ORNL-TM-0080.txt

 

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X aw e AR R S

OAK RIDGE NAT'ONAL LABORATORY
operated by

   

UNION CARBIDE CORPORATION
for the
U.S. ATOMIC ENERGY COMMISSION
r""- "\‘ /( Z
ORNL- TM- & Ui ]
CoPYNO. - /7
DATE — December 6, 1961

LABORATORY-SCALE DEMONSTRATION OF THE FUSED SALT VOLATILITY
PROCESS

G. |, Cathers, R. L. Jolley, and E. C. Moncrief

ABSTRACT
The feasibility of processing enriched irradiated zirconium-uranium alloy fuel
by the fused salt-fluoride volatility procedure has been demonstrated in laboratory
tests with fuel having a burnup of over 10%. Uranium recoveries were good and
decontamination factors for radioactive fission products were 106 to 10”. The UF¢
product contained significant quantities of nonradicactive impurities, and additional

work in this area is needed.

For review by Nuclear Science and Engineering.

NOTICE

This document contains information of a preliminary nature and was prepared
primarily for internal use at the Oak Ridge National Loboratory. It is subject
to revision or correction and therefore does not represent a final report, The
information is not to be abstracted, reprinted or otherwise given public dis-
semination without the approval of the ORNL potent branch, Legal and infor-
mation Control Department.
INTRODUCTION
The fused salt~fluoride volatility process for zirconium=-uranium reactor fuel

consists of (1) hydrofluorination and dissolution of the fuel in molten salt, (2) fluori-

nation to volatilize UFy from the melt, and (3) complete decontamination of the UF,

in an absorption-desorption cycle (1,2). As a nonaqueous process it has not received

the large-scale development effort that has been expended on aqueous processing

methods. Its advantages include low waste volumes (<1 liter/kg Zr-U alloy), high

decontamination from fission product activities, a greatly decreased criticality prob-

lem with enriched fuel due to the absence of neutron moderators, and the form of
the product, UFg, which eliminates some of the chemical conversion steps needed
in the uranyl nitrate~uranium metal cycle. Some possible disadvantages are the
corrosion at high temperatures in a fluoride system, the necessity of a gas-tight

system, and the difficulty of manipulating molten salt.

The tests, carried out in a hot cell, included study of the hydrofluorination
and fluorination reactions, the behavior of various fission product activites, and
the degree of uranium recovery and decontamination. The tests were conducted
primarily in preparation for operation of the ORNL Volatility Pilot Plant, which
has been adapted to process Zr-U reactor fuel (3). If this operation is successful,
the pilot plant may be modified to test variations of the process with other types

of fuel.

PROCESS DESCRIPTION AND EXPERIMENTAL PROCEDURE
The major process steps of fuel dissolution, UF volatilization, absorption,

and desorption were used (see Fig. 1). In each of 12 tests, 650 g of 2- to 3-year-

 
Refrigerant
(-60°C)

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UNCLASSIFIED
ORNL-LR-DWG. 49010

 

 

 

 

2nd NaF Bed
400°C

 

 

 

 

 

 

UF6 Cold Trap

 

 

 

Cold Laboratory

Fig. 1 Schematic of laboratory process test equipment.
decayed zirconium alloy fuel, with gross B and y activity levels of 1.8 x 107 and
1.0 x 107 cpm/mg U, respectively, was dissolved at 500-700°C by hydrofluorination
in molten 57-43 mole % LiF-NaF with a liquidus temperature of 670°C. As hydro-
fluorination and dissolution proceeded, the composition of the salt was changed as
represented by the line shown in the phase equilibrium diagram (Fig. 2). Dissolu-
tion was completed at a salt composition of 31-24-45 mole % LiF-NaF-ZrFy, i.e.,
close to a eutectic composition melting at 449°C. The resuiting final UF4 concen-
tration in the melt was <1%. The system LiF-NaF-ZrF,4 is one of the few fluoride
systems known in which liquidus temperatures are so low for large concentrations

of ZrF4 (4). The composition of the initial dissolution salt was chosen so as to mini-
mize the liquidus temperature encountered in the 0-20 mole % ZrF 4 region of the

phase diagram.

The dissolution product salt containing UF4 was fluorinated at 500°C with
elemental fluorine, and the volatilized UF was absorbed on sodium fluoride at
100°C. The UF, vapor pressure over the UF 4-3NaF complex at this temperature
is ~2 x 1073 mm, and essentially all the UF . is absorbed out of the Fo-UF 4 gas
stream (5). Desorption consisted in heating the UF s~NaF complex bed from 100
to 400°C while passing F, through to a second NaF bed held at 400°C. The dis-
sociation pressure of the UF ,-3NaF complex exceeds 760 mm af 400°C. The final
UFé product was cold-trapped at -60°C, then hydrolyzed witha 1 M AI(NO3)3

solution for analysis.
UNCLASSIFIED
ORNL -LR-DWG 38115R

ZrF, 912

TEMPERATURE IN °C
COMPQSITION IN moile %

     
  
 
  
 

= INDICATES SOLID SOLUTION

 

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Fig. 2. LiF-NaF-ZrF4 phase diagram with process composition line.

_g_
Equipment for the laboratory tests was installed in a hot cell equipped with
Argonne Model 8 slave manipulators (see Fig. 1). It consisted of a dissolution
reactor, fluorination vessel, NaF absorption beds, cold traps, and the necessary

pneumatically operated valves for coupling the system together.

The Hastelloy N dissolver was 18 in. deep and 3 in. i.d., with a 250-mil-thick
wall, and had a loading chute. The L~nicke! fluorinator was also 18 in. deep, 3 in.
i.d., with a 250-mil wall. Both vesseis were heated by a 5-in.-dia 12-in.-long tube
furnace, supported vertically. The salt transfer lines of 3/8 in.-dia Inconel tubing
(30 mils wall thickness) were heated auto-resistively with high-amperage current.
The salt transfer line between the two salt reactors was also used as a common gas

inlet line for the two vessels.

In runs 1 through 7, U-tube nickel absorption reactors containing 200 g of
NaF, 12-20 mesh, were used. Since less than 10 g of uranium was involved, smaller
nickel vessels containing 50 g of NaF on a grade H sintered nickel filter were used
in runs 8 through 12. In these last runs the UF6 was desorbed in a “"cold" laboratory,
the absorption bed having a maximum activity of about 200 mr/hr at contact, with

a large part of this being due to external surface contamination,

The stainless steel cold traps for waste HF and product were cooled by trichlo-
roethylene and dry ice. The waste HF was jetted into ice water, warmed to ambient
temperature, sampled, and poured into a waste drain, The molten salt was sampled

by a dip rod-frozen salt technique before being disposed of in heavy iron cans,
sealed over with a high-melting wax while still warm,

RESULTS

Dissolution

The total dissolution time varied from 16 to 62 hr, the HF efficiency from 19
to 48%, and the average dissolution rate from 0.17 to 0.64 mg em=2min=) (Table 1).
These rates, although lower than in early laboratory work, are comparable to those
obtained in engineering studies. The variables in the dissolution tests were flow rate,
temperature, gas phase reaction, salt purification, zirconium hydriding, and HF con-
centration. Dissolution was most rapid in run 12, in which the salt had received some
prepurification, the zirconium was prehydrided, and a high HF flow rate was used.
Entrained or volatilized material carried over in the HF stream was usually <0.1%
(Table 11), and the maximum uranium loss, probably as UF 4, was 0.03%. Entrain-
ment of Na, Cs, Sr, and rare earth fluorides was similar in magnitude to that of

uranium. Volatilization combined with subsequent entrainment is indicated for

fluorides of Zr, Nb, and Ru.

Some of the variables present in the test dissolutions have not been evaluated
adequately in cold laboratory work, but they were used in an effort to reduce the
dissolution time. The probable effect of some of the variations are noted in Table ]II.
The nature of the tests precluded, however, the drawing of firm conclusions about
optimum conditions for dissolution. The effect of impurities in the salt on dissolution

has not been definitely established although the piating-out of nickel on the zirconium
Table I. Typical Dissolution Results

 

 

 

Average HF Average
HF HF Utilization Dissolution Dissolution

Run Flow Rate, Concentration, Efficiency, Time, Rate,

No. ml/min % % hr mg cm™2-min~!
] 820 100 20.1 62 0.17

3d 1130 100 30.3 34 0.30

10° 1200 70 27.3 29 0.35

12° 1400 100 47.7 16 0.64

 

an. , e
Direct gas phase reaction occurred in first 6 hr.

bSdI’r prehydrogenated; fuel hydrided to ZrHo 33"

“Salt prehydrogenated; fuel hydrided to ZrH; .
Table I, Entrainment in Waste HF of Dissolution Step

 

 

Amount, % of total in feed

 

 

Run Gross Gross

No. U Zr Na B Yy Ruy Zry Nby Csy Srp TREB
1 0.007 - - 0.02 0.3 0,005 0.02 - 0.1 - -
3 0,007 - - 002 0.2 0.04 0.1 7 0.07r - -

10 0.01 0.3 0.04 0.03 0.2 0.3 0.04 0.9 0.05 0.02 0.004

12 0.007 0.1 0.05 0.08 0.2 2.0 0.4 1.2 0.04 0.02 0.006

 
~-10-~

Tabie I

 

Variation and Effect on Dissolution Rate

Remarks

 

Increasing HF flow rate increases rate
logarithmically

Increasing temperature increases rate

Nonsubmergence of fuel in salt promotes
direct HF-alloy fuel reaction

HF dilution with Hp decreases rate

Prehydriding possibly increases rate

Salt purification by H22reducfion of
Fe2t, Ni2t and Cr+ possibly

removes inhibition of reaction

Probably due to more chance of direct
gas-solid reaction

Generally true, but effect does not
involve much over a factor of 2
in available temperature range

Kinetic studies have shown that initial
rates of 100% HF with unconditioned
massive metal at 700°C results in
run-away reaction

Ho arises from reaction with metal or
metal hydride, or is added as
diluent

Indefinite at low H/Zr ratio, since
much surface cracking and swelling

generally occur only above a H/Zr
ratio of 1.5/1

Effect not proved although plating-out
of Ni, Fe, and other elements on Zr
surfaces has been noted

 
=11-

metal surface has been observed, and this presumably hinders dissolution of the zir-
conium. Other elements such as iron, molybdenum, tin, and chromium are also elec-
tronegative relative to zirconium and would presumably act similarly. All of these
elements are either fission products, arise from corrosion of the Inor-8 (71% Ni, 16%
Mo, 5%Fe, and 7% Cr), are present in the fuel alloy, or are impurities in the initial

salt. Two reactions involved in the nickel cycle are

2 Nify + Zr——> ZrF4+2Nij, AF°® = -130 kcal

Ni + 2HF ——> Hy + NiFg, 4F°® = +6 keal

Reduction of NiF5 to nickel metal at the zirconium fuel surface probably proceeds
mainly by the first reaction although the second reaction in reversal leads also to
reduction if the Hp/HF mole ratio is sufficiently high. At the end of zirconium fuel
dissolution the production of hydrogen becomes small and the second reaction leads

to total redissolution of the amorphous nickel metal formed in the earlier stage.

Uranium Recovery

 

Greater than 99% UF 4 volatilization was obtained only in the last four tests
because of inexperience with the equipment and the use of low fluorine flow rates
(Table 1V). Volatilization was repeatedly 99.8% or more in earlier work with simu-

lated process tests.

The NaF absorption-desorption cycle also operated effectively, resulting in
little uranium loss and duplicating the behavior observed in earlier laboratory and

pilot plant work. The total uranium retention on NaF beds wos less than 0.1%, and
-12-

Table V. Typical Uranium Volatilization in Fluorination Step

 

 

 

Fluorination Fo U

Run Temp, Time, Flow Rate, Volatilized,®

No. °C hr ml/min % of total inventory
8 500 4 300 97.9

9 520 4 300 99.4
10 500 3 340 99.2

11 520 3 340 99.9
12 500 3 430 99.8

 

a . . .ys . .
Based on analysis of fluorinated salt. Initial salt contained ~0.5% uranium.
- 13-

this probably represents an upper limit with reuse of the beds and the tendency for

refluorination of any retained residue.

PRODUCT DECONTAMINATION

Overall

Overall gross B and y decontamination factors ranged from 106 to 107 (Table V).
In all but a few cases the amount of activity in the product UF, was less than 10-
fold the "natural™ activity of U-235 (Table V1). In run 11 the ruthenium y activity
was high because an accidental pressure buildup and release entrained dust from
the NaF into the UFg cold trap. In runs 1 and 6 the product UF 4 was trapped on
NaF instead of in a cold trap, with the result that subsequent hydrolysis prior to
analysis gave an excessively dilute solution and a high and uncertain background
correction led to more apparent activity than in later runs. Runs 2 through 5 are
not included since they were not complete flowsheet tests. In runs 7 through 12
the services of a special low-activity-level analytical laboratory were used, which,
in conjunction with an improved product hydrolysis method, gave o more accurate

picture of the activity in the product.

The principal chemical impurities in the UF, product were molybdenum, techne-
tium, neptunium, and chromium, in order of decreasing concentration (Table VII).
Molybdenum, the end product of several fission product decay chains and a hydro-
fluorinator vessel corrosion product, varied in concentration from 1200 to 10,100 ppm.

Volatile MoF 4 formed in the flucrination step complexes with NaF, similarly to the
~14-

Table V. Overall Decontamination Factors in Complete Process Tests

 

 

Run Decontamination Factors

 

No. Gross B Grossy Ruy Zry Nby GCsy Sr B TRE B

 

8 2¢00% >1x10° 2108 8x105  4x107 3x10°  3x108  5x10°

o  3x10° 2108 4x107  2x10°  >2x108  1x107  33x107  >1x10'°

10 9x10®  5x107  2x106 3x10°  8x107  1x10°  >1x10'0 37107

4 5

11 1x10°  210°  5x10% 1x10° 36x107 2x1010 >9¢10%  3x107

12 2%10°  6x107  ix107  ox10%  8x10®  2x10°  »5x10°  >1x10'!

 
_]5..

Table VI. Ratio of Product Uranium Activity to Activity of Unirradiated U-235

 

Basis: 232 B cpm/mg U, 11 y cpm/mg U for 90% U-235 in equilibrium with

 

 

 

daughters
Run Activity Ratio, product/unjrradiated U-235
No. Ruy Zry Nb y Csy Sr B TRE P
1 <10 <10 <10 <10 1 <]
6 <10 <10 <10 <10 <1 <1
7 <] <1 <] <1 <1 <1
8 <<1 <1 <<] <<] <<1 <1
9 <1 <1 <<1 <<} <1 <<1
10 <10 <] << <] <] <<}
11 <100 <10 <<] <<] <] <1

12 <1 <10 <1 <<1 <] <<]

 
-16-

Table VII. Impurities in UF, Products

 

 

 

 

Run Amount, ppm of U
No. Mo Np Tc Crd

8 10,100 260 1,020 200

9 5,200 240 490 290
10 2,200 58 260 150
1 2,500 310 240 80
12 1,200 290 60 <100

 

oProbcbly from corrosion of the cold trap during the hydrolysis of the product to

obtain a representative aqueous sample.
-17-

behavior of UF,. Investigation of the dissociation pressure of MoF,-NaF complex
showed that it is approximately 1 atmosphere at 225°C, compared to 360°C for the
UF 4-NaF complex (_(3) Some separation of MoF , and UF is thus achieved in the

absorption step, dependent on the conditions of temperature, time, and gas flow.

The low activity levels of the UF4 products in runs 8 through 12 and the fact
that the usual individual fission product B activity contributors did not total up
to the gross B activity indicated an unknown B contributor. This was found to be
technetium by both chemical and radiochemical analyses. it was calculated that
the feed contained >3900 ppm based on the uranium, assuming >10% burnup. The
volatile compound TcFg is presumably formed in fluorination and possibly behaves

in the same way as MoF ; and UF ¢ in absorption.

Neptunium hexafluoride also appeared to follow UF¢ through the absorption
step. Chromium may have been introduced by hydrolysis of the UF, product in a

stainless steel vessel.

Decontamination factors for the separate process steps are given for run 12

in Table VIII.

Dissolution Step

There was considerable but highly erratic disappearance of ruthenium and
niobium activity during dissolution. The ruthenium y d.f.'s were 5 to 120, and the
removal of Ru is believed due primarily to adsorption on the dissolver wall. Some

ruthenium was volatilized with the excess HF. The niobium d.f.'s were 3 to 620.
- 8-

Table VIll. Step Decontamination Factors in Run 12

 

Decontamination Factors

Step GrB Gry Ruy Zry Nby GCsy Sr P TREB

 

Dissolution 2 ] 40 2 30 ] ] ]

Fluorination 4x105 3x10°  4x10°  5x10%  4x10°  4x10°  1x107  2x107

3

Absorption- 2x103 2x10%  1x102 1 6  3x10°  >5x102 >5¢10°

Desorption

7

Overall 2107 6x107  1x107 9107 8x10°  2x10°  >5x10° >1x10']

 
~-19-

The removal of niobium is believed primarily due to volatilization rather than ad-
sorption on metal surfaces. Material balances for both ruthenium and niobium were
low, possibly as the result of condensation of some NbFg (f.p. 225°C) in the top
zone of the dissolver, and of adsorption of the ruthenium. The nondisappearance
of cesium, strontium, and rare earth activities was expected since these elements

form fluorides that are nonvolatile and difficult to reduce.

Fluorination Step

Decontamination from the most important volatile fission product activities,
Ru y and Nb y (probably in the form of RuF 5 or RuF4 and NbFz) in the fluorination
process was much higher at a fluorination temperature of ~500°C than in previous
work at 600-650°C (1). In the first six tests conducted in the hot cell work, the
absorption off-gas was trapped in caustic. From analyses of these solutions and of
the first NaF beds, the gross B, gross y, Ru y, and Nb y activity decontamination
factors were calculated for the fluorination step (Table IX). At 600-650°C the
gross B and y d.f.'s were usually ~10° and the Ru y and Nb y d.f.'s ~5-10.

Uranium is thus largely decontaminated from fission product activities in the

fluorination step if the temperature is kept as low as possible.

Absorption-Desorption

 

Decontamination factors in the absorption-desorption step were in the range
10-100 for the more volatile NbF5 and RuFg or RuF .. The amount of Ru y activity

in the off-gas stream passing through the first NaF absorption bed was highly variable,
-20-

Table 1X. Measured Decontamination Factors Obtained in Fluorination

 

 

 

I:\;J:. GrB Gry Ruy Zr y Nb y Cs B Sr B TRE B
1 3x10°  1x10°  4x102  7x10% 5108 2¢10°  4x10®  1x107
2 2:0°  sxi0* 210 1x10° w0 - ; 6x10°
3 1x10° 3x10%  6x10%  sx10* 1x10° - 8x10° -

4 1x10°  4x10% 1x10° 20%  2x10° - x10° -
5 8x10%  3x10* 2«10 3x10f 03 - - -
6  3x10%  2a0* 2107 1x0% 2103 4x10*  3x10°  5x10°

 
-21-

due principally to the small amount volatilized in the 500°C fluorination step: 82,
11, 8, 13, 11, and &% of the total volatilized Ru y activity in runs 1 through 6, re~
spectively. In previously reported work relatively more Ru y activity passed through
the first NaF bed and resulted in much higher d.f.'s, since a larger amount of Ru y

activity was volatilized in the fluorination step at 600-650°C.

Actual absorption-desorption decontamination factors could be calculated
only in runs 1 through é where the absorption off-gas activity was trapped and
measured. The absorption-desorption d.f.'s presented in Table VIII for run 12 are
minimum values since the absorption off-gas was not measured. Actually, inclusion
of the off-gas activity in the calculation would affect only the Ru y and Nb y d.f.'s
and those only slightly. The d.f.'s given for the separate process steps in run 12
(Table VIil) were calculated from prorated activities measured in the first absorp-

tion bed since the same material was used in runs 8 through 12.

CONCLUSIONS
The fused salt-fluoride volatility process gave satisfactory decontamination
of uranium from fission product activities at a burnup of >10%. The main impurities
in the product UF4 were molybdenum, technetium, and neptunium. Some separation
from molybdenum and technetium evidently occurs in fhhe absorption~desorption step,
and further work to optimize this effect appears warranted. The results also indicate
that use of a fluorination temperature of 500°C rather than 600-650°C minimizes
to a large extent the carryover of Ru y and Nb y activities along with the UF¢ gas

stream in the fused salt fluorination step. Evaluation of the absorption-desorption
-22-

step was only tentative due to the small carryover of all activities from the fluori-

nation step.

The overall results appear to confirm the chemical feasibility of the process
with irradiated Zircaloy-2-U alloy fuel. Areas of chemical uncertainty exist, par-
ticularly in the dissolution step, but these are apparently not serious. The dissolu-
tion step involves gas-liquid-solid contact and is necessarily quite dependent on
geometry. |t appears desirable, however, to study further the zirconium metal re-
duction of NiF; and analogous impurities in the fused salt to determine how this
affects the dissolution rate and the concurrent volatilization or deposition of rela-

tively noble materials such as ruthenium and molybdenum.

ACKNOWLEDGMENTS
The assistance of T. E. Crabtree and C. J. Shipman in performing the labo-
ratory work is gratefully acknowledged. The authors also express their appreciation
for the work of personnel in the Analytical Chemistry Division of ORNL under the
supervision of C. L. Burros, J. H. Cooper, W. R. Laing, C. E. Lamb, H. A. Parker,

and G. R. Wilson.
w

REFERENCES

. G. |. Cathers, Nucl. Sci. and Eng., 2, 768-777 (1957).

. W. H. Carr, Jr., Chem. Eng. Symposium Series, 56, 57-61 (1960).

R. P. Milford, S. Mann, J. B. Ruch, and W. H. Carr, Jr., Ind. and Eng.
Chem., 53, 357-362 (1961).

F. F. Blankenship et al.,, ORNL-2548, p. 60, Oak Ridge National Laboratory
(1959).

G. I. Cathers, M. R. Bennett, and R. L. Jolley, Ind. and Eng. Chem., 50,
1709-10 (1958). T

G. |. Cathers, "Dissociation Pressure of MoF 4~NaF Complex and the Inter-
action of Other Hexafluorides with NaF," paper presented at American
Chemical Society Meeting, Sept. 3-8, 1961.
24~

DISTRIBUTION

Nuclear Science and Engineering
E. J. Murphy

F. L. Culler

D. E. Ferguson

H. B. Graham

F. R. Dowling, Wash. AEC

E. L. Anderson, Jr., Wash. AEC

. J. Vanderryn ORO AEC
. G. |. Cathers
. R. L. Jolley

E. C. Moncrief
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