ORNL-TM-3137.txt

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Fhis: r_;eport' was prepared as an account of work sponsored by the United-
 States . Goverament. Neither the United States nor the United States Afomic
Energy Commission, nor any of their empiayee’s, nor any - of their £ONlractors,
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assumes. any legal liability or respansibility for the accuracy, completeness or
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ORNL~TM~313T
Contract No. W-T4O5-eng-26

CHEMICAL TECHNOLOGY DIVISION

ENGINEERING DEVELOPMENT STUDIES FOR MOLTEN~SALT
BREEDER REACTOR PROCESSING NC. 2

I.. E. McNeese

FEBRUARY 1971

CAK RIDGE NATTIONAL LABORATORY
Oak Ridge, Tennassee
operated by
UNION CARBIDE CORPCRATICN
for the
U.S. ATOMIC ENERGY COMMISSION

A
B
i
|
4

JUERERRERY

3 445k 0135288 7
ii

Reports previously issued in this series include the following:

ORNL-413%9
ORNL -420k
ORNL-123Lk
ORNL-4235
ORNL-4364
ORNL-4365
ORNL -L366
ORNL-TM-30573

Period
Period
Period
Period
Period
Period
Period

Period

Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending

March 1967
June 1967
September 1967
December 1967
March 1968
June 1968
September 1968
December 1968
CONTENTS
Page
SUMMARIES . &+ ¢ v v v s o v s o e v o e e e e v
1. INTRODUCTION « v v ¢ 4 o v o o s o 2 s & o o o s o« o s o = 1
2. PROPOSED FLOWSHEET FOR PROCESSING A SINGLE-FLUID MSBR BY REDUCTIVE
EXT RA CT ION . . . . . a . . . . . . . . . - . . . . 1.
5. MSBR MATERIAL-BALANCE CALCULATIONS . . ¢« « ¢ ¢ ¢« & « « « & 5
5.1 Effect of Fuel~Salt Discard Cycle . « . « « + &« « + . & 5
3.2 FEffects of Removing Zirconium and Seminoble Metals 6
3.3 FEffect of Protactinium Removal Efficiency . . . . 8
3.4 Quantities of Heat and Mass in the MSBR Off-Gas System . 10
L. REMOVAL OF PROTACTINIUM FROM A SINGLE-FLUID MSBR . . . . . . 22
4.1 Mathematical Analysis of a Reductive Extraction System . 2h
4.2 Calculated System Performance 26
4.3 Stabilization of the Protactinium Isolation System . . 29
L. Transient Performance .+ « + « o o o o o o o o + o« . s L1
5. USE OF THE PROTACTINIUM ISOLATION SYSTEM FOR CONTROLLING THE
URANIUM CONCENTRATION IN THE BLANKET OF A SINGLE-FLUID MSBR . Ly
6. REMOVAL OF RARE EARTHS FROM A SINGLE~FLUID MSBR . . 52
6.1 Proposed Rare-Earth Removal Flowsheet 52
6.2 Effect of Rare~Farth--Thorium Separation Factor and Bismuth
Flow RAt@ « ¢ ¢ o o & o o o o & o o s oo o o + o« 5k
6.3 Effect of Location of Feed Point . . . .« . . « « « « . . 63
6.4 Effect of the Fraction of ThFu That Is Reduced in the
Electrolytic Cell . . . . . e e v e e e e e e 63
6.5 Effect of Wumber of Stages in the Extraction Column . . 66
6.6 Effect of Processing Cycle Time « + « o o ¢ + o o + 66
6.7 Rare-Earth Removal Times for Reference Conditions 7

jil
iv

CONTENTS (Continued)

Page
T. SEMICONTINUOUS ENGINEERING EXPERIMENTS ON REDUCTIVE EXTRACTION T8
T.1 Pressure Tests of the Feed-and-Catch Tanks . . . . . . . 78
7.2 TInitial Cleanup of Experimental System . . . . . . . . . 79
(.5 Loading and Treatment of Bismath . . . . . . . . . « . . 79
T.4 Operation of Gas Purification and Supply Systems . . . . 80
7.5 7Planned Experiments .+ « + o « + o « o o ¢ o « o 4 4 . . 80
8. ELECTROLYTIC CELL DEVELOPMENT . + « v « 4 v « ¢« « o o o « o . g2
8.1 Comparison of Cell Resistance with AC and DC Power . . . 82
8.2 Experiments in a Quartz Static Cell with a Graphite Anode 63
8.3 Studies of Frozen~Wall Corrosion Protection in an All-Metal
Cell o o v & v v i v s e e e e e e e e e e e e e e e 85
9. MSRE DISTILLATION EXPERIMENT . . . . « « « « « & o « o « + . 89
9.1 Instrument Panel . . .+ + « ¢« ¢ 4 4 ¢ i 4 e e e e e e e . 90
9.2 Main Process Vessels . « « « ¢ v v v v v 0 0 v e e e . 30
G.3 Valve BOX « v v v v v v v e e e e e e e e e e e e e e 9k
9.4 Condensate Sampler . « « « + « 4 4 4 4 e e e 4 e e e 96

10. REFERENCES . « ¢ « ¢« v v o v v « o v 4 4 v v o o o o o o o = 100
SUMMARIES

PROPOSED FLOWSHEET FOR PRCOCESSING A SINGLE-FLUID
MSBR BY REDUCTIVE EXTRACTION

A proposed flowsheet for processing a single-fluid MSBR is based
on reductive extraction, and routes the reactor salt through the prot-
actinium-isolation and the rare-earth~removal systems on 3- and 30-day
cycles, respectively. The flowsheet is scaled for a 1000-Mw (electrical)

MSBR.-

In addition to the two primary operations {protactinium isolation
and rare-earth vemoval), the flowsheet incorporates means for removing
materials such as zirconium and nickel. Other operations include
recovery of uranium from waste streams, UF6 collection, and fuel recon-

stitution and cleanup.
MSBR MATERIAL-~BALANCE CALCULATIONS

The MATADOR steady-state material-balance computer code has been
used to determine the effects of a number of processing-plant variables
on the nuclear performance of a 1000-Mw {electrical) MSBR. Variables
studied were fuel-salt discard rate, removal times for zirconium and

seminoble metals, and efficiency for the removal of protactinium.
REMOVAL OF PROTACTINIUM FROM A SINGLE-FLUID MSEBR

Isolation of protactinium by reductive extraction is proposad,
based on the fact that protactinium is intermediate in nobility between
uranium and thorium. By countercurrently contacting a salt stream con-
taining flucrides of uranium and protactinium with a bismuth stream
containing thorium and lithium, the uranium is transferred from the salt

to a downflowing metal stream and is carried out of the extraction column.
vi

The protactinium, which is trapped in the center of the column, will,
for the most part, be held up for decay in a tank through which wolten
salt is circulated. This protactinium isolation method was analyzed

mathematically; calculated results show it to be attractive.

USE OF THE PROTACTINIUM ISOLATION SYSTEM FOR CONTROLLING THE
URANTIUM CONCENTRATION IN THE BLANKET OF A SINGLE-FLUID MSER

The presently envisioned system for protactinium isolation produces
a salt stream that is free of uranium and protactinium at one point in
the system. This characteristic could be exploited to decrease the
uranium concentration {and hence the uranium inventory) in the blanket
region of a single-fluid MSBR. We concluded that the use of the prot-
actinium~isolation system for decreasing the uranium concentration in
the blanket region to 10% of that in the core region is restricted to
reactor designs having an exchange rate between the core and blanket

regions of less than 0.01% of the flow rate through the core.

REMOVAL OF RARE EARTHS FROM A SINGLE-FLUID MSBR

A reductive extraction method is proposed for removing rare earths
from a single-fluid MSBR. Operation of the system is dominated by the
low separation factors (1.2 to 3.5) between thorium and the rare earths.
Calculated results for the steady-state perforwmance of the system
indicate that an extraction column having approximately 24 stages and
a bismuth flow rate of 15 gpm will be required for removing the rare
earths on a 30-day processing cycle, which is roughly equivalent to

a 50-day removal time for all rare earths.

SEMICONTINUOUS ENGINEERING EXPERIMENTS ON REDUCTIVE EXTRACTION

Equipment for semicontinuous engineering experiments on reductive

extraction was installed, and the initial shakedown operation was com-
viil

pleted. The salt and bismuth feed tanks were pressuve tested. The
internal surfaces of the system were treated for removal of oxides by
contact with hydrogen at 600°C, and a 18L4-kg charge of bismuth was
added to the system and treated for oxide removal. Several experiments

planned with this system are described.

ELECTROLYTIC CELL DEVELOFPMENT

The proposed reductive extraction preocesses for protactinium
isnlation and rare-earth removal refuire electrolytic cells for re-
ducing lithium and thorium fluorides into a bismuth cathode to prepare
the metal streams fed to the extraction columns and for oxidizing
extracted components from the metal streams leaving the columns. Experi-
mental work is reported on (1) a comparison of cell resistances ob-
served with alternating and direct current, (2) the performance of a
graphite anode, and (3) the evaluation of an all-metal cell employing

frozen-wall corrogion protection.

MSRE DISTILLATION EXPERIMENT

Equipment ig being insztalled at the MSRE for demonstrating the
nigh-temperature, low-pressure distillation of molten salt as a means
for separating the lanthanide fission products from the components of
the M3RE fuel-carrier salt, which is a mixture of lithium, beryllium,
and zirconium fluorides. Modifications that were made after the system
was tested with nonradiocactive salt are discussed; they include a change
in the configuratioen of the feed lire to the still pot, installation of
electrical insulation between resistance heaters and process lines to
prevent the lines [rcm being damaged in the event of a heater failure,
installation of absolute filters in the vacuum lines from the feed tank
and the receiver to prevent the spread of radioactivity, completioca of
the secondary containment feor the valve box, and construction of a

condensate sampler suitable for remote operation with radicactive samples.
1. INTRODUCTION

A molten-salt breeder reactor (MSBR) will be fueled with a molten
fluoride mixture that will circulate through the blanket and core re-
gions of the reactor and through the primary heat exchanger. We are
developing processing methods for use in a close-coupled facility for
removing fission products, corrosion products, and fissile materials

from the molten fluoride mixture.

Several operations associated with MSBR processing are under study.
The remaining parts of this section describe (1) a proposed reductive-
extraction flowsheet for a single-fluid MSBR, (2) material-balance cal-
culations that show the effects of the removal time for zirconium, alkali
metals and alkaline earths, europium, and protactinium on reactor per-
formance and that indicate the magnitudes of the heat generation and
mass flows associated with the reactor off-gas, (5) calculated results
showing the steady-state performance of a protactinium isclation system,
(4) an evaluation of the use of Che protactinium isolation system to limit
the uranium concentration in the blanket of a single-fluid MSBR, (5) cal-
culations to predict the steady-state performance of a rare-earth removal
system based on reductive extraction, {6) preliminary testing of the semi-
continuous reductive-extraction facility, (7) experiments related to the
development of electrolytic cells for use with molten salt and bismuth,
and (8) installation of equipment at the MSRE for demonstrating low-
pressure distillation of moltea salt, using irradiated MSRE fuel carrier
salt. This work was carried out by the Chemical Technolegy Division during

the period January-March 1969.

2. PROPOSED FLOWSHEET FOR PROCESSING A SINGLE-FLUID MSBR
BY REDUCTIVE EXTRACTION

L. E. McNeese M. E. Whatley
The process flowsheet envisioned for a single-fluid MSBR is based

on reductive extraction, and routes the reactor salt volume through the

protactinium-isolation and the rare-earth removal systems on 3- and
30-day cycles, respectively. The present version of the process flow-
sheet (Fig. 1) is scaled for a 1000-Mw (electrical) MSBR. The protac-
tinium isolation and the rare-~earth removal systems will be described

in more detail in a later section.

The protactinium~isolation system exploits the fact that protac-
tinium is intermediate ia nobility between uranium and thorium. A
molten-salt stream is withdrawn from the reactor on a 3-day cycle (2.5
gpm) and is fed countercurrent to a 5.3-gpm stream of liquid bismuth in
a l2-stage extraction column. If the correct flow of reductant (thorium
plus lithium) in the bismuth stream entering the coutactor is used, the
uranium in the salt will transfer to the downflowing bismuth stream in
the lower part of the column. The protactinium, however, will concen-
trate midway up the cascade, where most of the protactinium in the
reactor system can be held up by diverting the salt through a suitably

large volume (200 ftj). At steady state, the 253 255U at

Pa decays to
the same rate that it enters the tank from the reactor. The concen-
trations of both protactinium and uranium in the salt leaving the column
are negligible; however, the concentration of rare earths at this point
is approximately the same as that in the reactor. Approximately 10% of
the salt stream leaving the protactinium-isolation column (i.e., 0.25
gpm) will be processed for the removal of rare earths. The remaining
salt passes through an electrolytic oxidizer-reducer, where lithium and
thorium are reduced into a flowing bismuth cathode to provide the metal
stream that is fed to the column. At the anode of the cell, bismuth is
oxidized to BiFf’ which is soluble in the molten salt. The salt stream,
containing BiFfi, is countercurrently contacted with the bismuth stream
exiting from the extraction column in order to oxidize uranium, prot-
actinium, and other materials, which are then transferred to the salt

stream and returned to the reactor.

The concentration of uranium or protactinium must be known at some
point in the column in order to control the concentration of reductant

in the bismuth stream [ed to the column. The uranium concentration is
SALT MAKE UP
043 1+ day
Li=0r2
Be.-—'2=0.€6
Thiy =042

 

0.25gpm
RE =0.00004%

F

3 STAGE
CONTACTOR

   
   
    

ELECTROLYTIC CELL
11 # 8/ ELECTRCDE

1
3 STACE
CONTACTOR

 

 

 

Bi
REMOVAL

 
 

 

 
 

 

Th=00016
gl |

.94 gpr = 00018

 

         

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

      
      

 

 

 
 
  

 

  
   
 

 

 

 

AND
REDUCTIGN
] 3 STAGE
EXTLFACTOR
Ha ] 12 STAGE
EXTR
REACTOR Régcs;”gfi TRACTOR
3 e
{461 > (EL FCTROLYTIC CELL
BR=1.06 (ExcESS 411 2/ELECTRODE -
{_Ufe Th=00016 ©.25 gpm
Limid 51 {30 day CYCLE)
- £E=0.0001i6
F,TO
f RECYCLE & STAGE &
NoF LB 2 UFg
400°C _ EXTRACTOR Ha 31.9 9 moles/doy
HF | z2
* {2 STAGE
EXTRACTOR
Fa £a DECAY
FLUOR 1 cemdval TANK ”
200 13, 0.08 5z
U= 2 2107% A5 g
£ iy Pu:O,OOI;u
28 kw/ft
25k 15 gpm B7
& 8 STAGE
C.i2& gpm U=37x10 ~ = EXTRACTOR LiF-Zrf,
{60 day CYCLED Pfl=‘l‘33_>‘10 TO WASTE
24,4 kw/Ht 3 5.2 moles /doy
0.53 ZcRy
FUEL SALT REPROCESSING 0.7 LiF
PLANT FEED : E=
25% gom TANK 5.2 gom Bi RE_) 190.90por§9
{3 doy CYCLE]} U=0.0018 -8 :
UF, = 0.003 Po = 5.6 410"
Paf,=1.54 110 Th=2.0% QJ
37.9 kw/H> Li=4xi0

Fig. 1.

Proposed Reductive Extraction Processing Flowsheet

ORM_—DWG 69-2353RA

3.6 63 SALT/DAY FOR 41 days
T TeTTTYTTTTTTT T

TRERE EARTH ACCUMULATION 4AND Pa DECAY]
29,613 23617 39.6i115l 396615 39,647

e e b L

7.2 f1¥day
(SEMF - CONTINUOUS)

Ufg

; 100°C 400°C Fi
; MNeF NaF
o
‘2

 

 

 

 

G493 fr¥ SALT/day
TO WASTE
RE=0.0069 _
Li=0Q,72

Bef, =0.16

Thig =0.1i3

  
   
 

ALL CONCENTRATIONS ARE GIVEN N MOLE FRACTIONS

for a Single-Fluid MSER.
determined by fluorinating approximately 5% of the salt entering the
protactinium decay tank and analyzing the resulting gas stream for UF6.
Means are provided for collecting the UF6 from this operation as well
as from other fluorination operations and for returning this material
to the fuel salt. The UF6 is simultaneously absorbed into the molten
salt and reduced to UFLL by a hydrogen sparge. A bismuth-removal step

will also be provided before the salt returns to the reactor.

Approximately 1.5% of the bismuth stream leaving the extraction
column (i.e., 0.08 gpm) will be hydrofluorinated in the presence of a
salt strzam for the removal of the seminoble metals (Ga, Ce, Cd, In, Sn,
and Sb), various corrosion products {Fe, Ni, and Cr), and fission prod-
uct zircenium. The salt is recycled between the hydrofluorinator and a
fluorinator, where uranium is removed. The principal components that

build up in this salt are fLiF and ZrF, ; the expected steady-state

composition is 47-53 mole % LiF—ZrFM, which has a liquidus of 520°C.

Salt that is free of uranium and protactinium but contains rare
earths is fed to the center of a 2hk-stage extraction column at the rate
of about 0.25 gpm, which is sufficient to process the reactor veolume in
30 days. The salt flows countercurrent to a bismuth stream containing
thorium and lithium. Typically, 60% of the rare ecarths are extracted
from the salt stream in the upper column {an effective removal time of
50 days), and the rare-earth concentration in the lower column is
increased to approximately 0.6G mole %. The bismuth flow rate through
the column is 15 gpm. Part of the salt leaving the column is returned
to the reactor, whereas the remainder is fed to the electrclytic cell
complex; this produces a net effect of reducing thorium and lithium into
the bismuth phase and transferring the extracted rare earths from the
bismuth phase to the returning salt. (Both the anode and the cathode are
flowing streams of bismuth) A Bi-Li stream generated at the cathode of
the cell is fed to the three-stage contactor, which removes most of the

ThFu from the incoming salt. The salt then picks up BiF5 as it passes
1

the anode. The salt stream containing BiF_ is passed countercurrent to

5
the bismuth stream entering the complex from the rare-earth-removal
column in order to oxidize the rare earths, thorium, and lithium from

the bismuth.

Salt containing rare earths at a concentration of 0.69 mole % is
4
withdrawn from the system, at the rate of 0.49 ffyday, through a set of

sequentially arranged 40~ft§

tanks. The active-metal fission products
(Sr, Cs, Ba, Rb, and Eu) are also present in the stream at a concen-
tration equal to that in the reactor, and are removed on a 3000-day
cycle. Use of this system limits the rate at which rare earths could
inadvertently return to the reactor. The salt is fluorinated for

uranium recovery when necessary.
5. MSBR MATERTIAL-BALANCE CALCULATIONS

M. J. Bell L. E. McNeese

The MATADOR computer code for calculating steady-state material
balances has been used to determine the effects of a number of proc-
essing-plant variables on the nuclear performance of a 1000-Mw
(electrical) MSBR, which has a thermal power of 2250 Mw. These var-
iables include fuel-salt discard rate, removal times for zirconium
and seminoble metals, and protactinium removal efficiency. The
reactor under consideration had a power of 2250-Mw (thermal) and
contained 1220 ;f:'t—.j of salt in the case of the studies of zirconium
and seminoble metal vemcval and 1463% ft° for studies of protactinium
removal and salt discard rate. The fuel-salt composition was 67.7-

20.0-12.0-0.3 mele % LiF -BeF,-ThF, -UF

2 o

5.1 Effect of Fuel-Salt Discard Cycle

The active metals {(Rb, Cs, Sr, and Ba) are not removed by reduc-

tive extraction, and their concentrations in the reactor system are
Lo

maintained at acceptable levels by the discard of salt. A study was
made of the effect of the rate of salt discard on the performance of
the reactor, as evidenced by neutron absorptions by the active metals.
The poisoning caused by these metals is shown in Table 1 for salt-
discard cycle times of 800, 1200, 2000, and 3000 days, and is compared
with the total poisoning for all fission products. It is evident that
the poisoning produced by the active metals is a small fraction of the
total at even the longest cycle times, and that low discard rates are
acceptable. However, the discard stream also contains concentrated
rare~earth fission products from the rare-earth reductive-extraction
system, and is the means for removing these materials from the system.
Cycle times longer than 3000 days are not possible because the solubil-
ity of the rare earths in the discard stream would be exceeded. A cycle

time of %000 days for a reactor containing 1220 ftj

of salt corresponds
to a salt discard rate of 0.4l fti/daya The approximate cost for the
fuel salt is $2000/ft5; thus the cost of discarding salt at the above
rate would be $250,000 annually (or 0.036 mill/kwhr), which indicates
that it might be economical to process this stream to partially recover

the carrier salt.

3.2 Effects of Removing Zirconium and Seminoble Metals

Zirconium is very similar in chemical behavior to uranium, and will
be extracted into the bismuth stream in the protactinium isolation sys-
tem. If means for removing zirconium from the bismuth are not employed,
the zivrconium will be returned to the reactor with the uranium and will
constitute a serious neutron peison. However, zirconium can be removed
from the system by hydrofluorination of part of the bismuth in the
presence of salt, followed by fluorination of the salt to recover the
uranium as UF6. This concentrates the zirconium in a salt stream as

ZrFu, which is then discarded.
Table 1.

of the Salt Discard Cycle

Neutron Absorptions by Active Metals as a Function

 

 

 

Discard Neutron Absorptions per 100 Fissile Absorptionsa

Cycle Total Total
Time Active Fission

(days) Rb Sr Cs Ba Metals Products
800 0.000059% 0.017L 0.00166 0.00672 0.0264 2.15
1200 0.00008588 0.0253% 0.00248 0.00925 0.0372 2.16
2000 0.00015 0.0398 0.00%9 0.0159 0.0598 2.18
3000 0.00022 0.0561 0.0055 0.0267 0.0885 2.21

 

e
aPoisoning x 10 .
A study was conducted to determine the effect of zirconium removal
time on neutron poisoning by fission-product zirconium. The results of
this investigation are summarized in Table 2, which shows the poisoning
due to zirconium isotopes for removal times of 50, 200, and 800 days.

The poisoning by zirconium is only 2% of the combined fission-product
poisoning for a 200-day removal time, and increases to 8% for an 800-

day removal time. On this basis, it was decided that the bismuth should
be processed on a 200-day cycle to keep zirconium poisoning at an accept-
able level. This requires the continuous hydrofluorination of 1.5% of
the circulating bismuth in the protactinium isolation system, or about

6.7 ft7 of bismuth per day.

Hydrofluorination of the bismuth stream to remove Zr will also
serve as a cleanup step to remove corrosion products (Fe and Ni) and
may remove fission products that are intermediate in nobility between
the noble metals and uranium (Zn, Ga, Ge, Cd, In, and Sn), which
are designated as seminoble metals. 1In the absence of the zirconium-
removal step, these materials would accumulate in the bismuth phase
until, eventually, saturation and precipitation occurred. Such an
event might impair the performance of the protactinium isolation system.
Table 3 shows the chemical composition of the mixture of seminoble
metals and the production rate of these materials, assuming that they
are removed on a 200-day cycle. This group of fission products, which
consists primarily of isotopes of tin, does not contain important neutron
poisons and is not a significant source of heat. The heat-generation
rate of the combined seminoble metals is about 0.02 Mw. This heat
generation rate is small compared to that of the zirconium isotopes

removed in this operation, which is 0.39 Mw.

5.3 Effect of Protactinium Removal Efficiency

Another parameter that was studied is the efficiency of the prot-
actinium isolation system, which has been defined as the ratio of a 5~

day removal time for protactinium to the actual removal time. These
k\f)

. a . ,
Table 2. Neutron Absorptions by Zirconium Isotopes
as a Function of Zirconium Removal Time

 

Removal Time (days)

 

 

 

Isotope 50 200 800
99, 0.0000035 0.000017 0.000062
o 0.00145 0.00608 0.0231
92 7y 0.00056 0.00148 0.00579
95y 0.00761 0.0305 0.1199
I 0.00012 0.00048 0.00196
96zr 0.000058 0.000234 0.000955

Total Zr 0.00954 0.0388 0.152
Total Fission

Products 2.18 2.21 2.32
aAbsorptions per 100 fissile absorptions. Poisoning x 102.

 

 

Table 3. Composition of Seminoble Metals Stream from 2250-Mw (thermal)
MSBR for a 200-day Removal Time
Production Rate
Element (mole/day) Mole Fraction
n 2.60 x lO‘r 9.27 % 10'6
ol -6
Ga 1.70 x 10 6.06 x 10
- -
Ge 6.65 x 107" 2.36 x 107"
cd 1.22 x 1072 L.35 x 107
=4 -3
In 1.76 x 10 6.27 x 10
Sn 2.60 x 107 9.27 x 107t
Total 2.81 x 107 1.00

 
10

studies were performed using a combination of the MATADOR material-
balance code and the ROD reactor optimization and design code. The
combinedVCode, called MODROD, uses the MODRIC nine-~group diffusion
calculation and two-dimensional flux synthesis to compute the fissile
inventories and a neutron balance for a given "lumped" fission-product
concentration obtained from MATADOR. The fissile nuclide reaction rates
computed by ROD are then used by MATADOR to obtain a new estimate of the
lumped fission-product concentrations. This procedure 1s repeated until
the lumped fission-~product concentrations converge. A few of the indi-
vidual points computed by the MODROD code were compared with the exact
ROD treatment of the same case and were found to be in excellent agree-

ment.

The proposed protactinium-isolation system will not perform with
100% efficiency; therefore, the effective protactinium removal time will
be greater than the 3-~day processing cycle time. Figure 2 presents the
effect of this longer cycle time on the fuel yield and the fuel-cycle
cost of the reference MSBR design. It is obvious from this figure that
the reactor performance varies only slightly for a protactinium removal

time of 3 to 5 days, which is the expected operating range.
3.4 Quantities of Heat and Mass in the MSBR 0ff-Gas System

A series of investigations conceraing the MSBR ocff-gas system has
been carried out. Significant quantities of noble metals and noble
gases and their daughters will be carried into this system by a helium
stream that has been contacted with the fuel salt. These materials
will be intensely radioactive and will generate about 30 Mw of heat.
The system must also be able to accommodate all volatile fission products
that are evolved throughout the processing plant. Table ! summarizes the
important volatile radionuclides that are generated in the processing
plant and designates their point of origin. This table shows that the

noble-metal group of fission products is an important source of volatile
FUEL YIELD (% of fissile inventory per annum}

ORNL—DWG 70—10033
Pa PROCESSING TIME (days)

 

 

 

 

 

 

 

 

 

QEL-CYCLE COST

emmemnemenli.

2.25 / \\ 0.0!
\

2.00 O
O 0.4 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0

Pa PROCESSING EFFICIENCY

o 30 15 10 75 & 5 375 3

3.50

£

3.25 E—_ 0.05 &

/ E

4—

3.00 YIELD £ 0.04 @

Q)

\ N

275 0.03 O

/ %

1

Lol

250 002 T

N Tz

L

&

=

<

T

o

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 2. MSBR Performance as a Function of Protactinium Processing Efficiency.

11
Table L.

Production of Volatile Radicactive Isotopes in an MSBR Continuous Processing Plant

 

Rate {gram-atoms/Mwc) at Which Voiatile Radiocactive Isotopes Are:

 

 

Recovered Total
Recovered from Evolved from from Fluori- Activity
Removed by Evolved from Fluorinator in Salt in Rare- nator in Rare~ Production
Sparging Separated Pa Isolation Earth Process- Earth Process- Produced Rate
Isotope Half-Life with Helium Noble Metals System ing Plant ing Plaat (Total) {curies/Mwd )
% . ~ -7 e -
“1 12.26 y 4.L8 x 10 ---- ---- L8 x 107! 1.6y x 1677
§7m, | -5 | -5 /. - -& ,
<K 1.86 h 1..8 x 107 £.90 x 10 7 .82 x 10 c 7.72 % 10 & ---- L.35 x 10 - 7.2% x 107
| “ - - =5 . el . -& - . !
Eomy b4 h 8.75 x 1077 1.66 x 1077 5.680 x 107 1.0% x 10 © - 1.00 x 1077 6.75 x 16
&+ . -6 . -5 ! - - -6 -
Tkt 10.76 y S .6k x 1070 5.61 x 107 8.75 x 10 Y 236 % 1077 545 x 1070 5.06 x 107
&7 . -k - ; -1 . -
Tkr 76 m 1.8% x 10 ---- 5.02 x 1077 9.60 x 1071 ---- 1,85 x 107 LA x 10
88 -k - - ] 3
Kr 2.80 h 2.21 x 10 ---- 1.0k x 10 7 6.9% x 10 7 --=- .ot ox 107 2041 x 107
o o -4 L . =& - -4 '

Total Kr® - 8.6L x 10 1.09 x 167" L0 % 10 5 vooh x 1077 ---- 5.9% x 10 2000 % 107
189I 1.7 % IOT ¥ - .35 x 107 1.18 x LO_b -——- 2 LhO x 10_T 849 x 1677 L.7h x 167" R3
131 - -5 = o -y L
5 I 8.05 d -——— 1.720 x 1C 1,19 x 10 “ ~--- 2413 x 10 7 .21 x 1o 1.91 x 1o’

2 -4 ‘ -& , -4 oo 5
15 T .26 h ———- 1.94% x 10 1.4% x 10 © ———— ---- 1.9k x 10 2257 x 107
133 ~ ., -4 | -7 L b ok

I 20.3 h _———- 2.30 x 10 G.ab o x 10 —_——— -— 2051 x 10 4.56 x 10

i3k 5.20 m S 2.20 x 10'L’ 3,07 x 10'8 S ———- 2,20 x ;(;“L’ 765 x LG

32 6.68 h ——-- 2.8 x 1077 1.32 x 10'6 S - 2.97 x 1077 1.25 x 10
: - - o -7 - -4 ,

Total Ia --- ~—-- ¢.7h x 10 L.hs x 10 6 - .90 = 10 : o.ou ox 10 107 x 10
131 - -& . =7 . -9 =8 -1li -6 i
> Xe it.6 d 1,89 x 10 © 9.60 x 10 ° 3.60 % 10 7 1.9% x 10 1.70 x 10 °° 1.12 % 10 1.21 x 17
%%y : -7 - -G -& ' e j -
Ly 2.26 d 750 x 107 .5l x 1070 Lok x 1070 5.01L x 107 ---- 620 x 107 Seobox LU
1%5. o o -5 o -L . 7 . b i N

Xe 5.27 @ Z.90 x 10 2.30 x 10 6.0 w1t 1.30 x 10 ---- 2.60 x 10 £.26 x 1C
Loomye 15.6 m L.66 x 107 8.51 x 1076 7.96 x 1071 £.0% x 107 ---- 5.61 x 107 £.59 x 107
L0 %e 3.1 n 1.64% x 107 2.8L x 1077 1.22 x 10'6 5,01 x 1077 —-- 1,96 x 10 &by x 10

- 3 .“’I = -6 = .;L - 5 -4 o T
Total Xea --- LOT7 = 10 5 7.18 x 10 ! 2.21 x 10 ¢ 2.5% x 10 1.70 x 10 L 1.7% x 10 7 2.8k x 10

 

a . . . .
Total preduction rates include stable isotopes not shown.

Total activity includes very short-lived

isotopes not shown.
radionuclides; thus the chemical behavior of the neoble metals in the
fuel salt will greatly influence the design of the processing plant.
Smaller amounts of volatile fission products will be evelved in the

fluorinaters and in the rare-earth decay tanks.

The noble-gas and noble-metal fission~product groups have been
examined in more detail using the ORIGEN isotope generation and decay
code. 1t was assumed that these materials migrated to circulating
helium bubbles with a 50~sec residence time in the fuel salt, and that
the helium bubbles were stripped from the salt on a 110-s2c cycle. 1In
this investigation, the MATADOR material~balance code was used to com-
pute the steady-state compositions of the noble-gas and noble-metal
fission-product streams leaving the fuel salt. These stream composi-
tions were then used as input to the ORIGEN code to compute the isotopic
compositions of each stream as a function of decay time in the off-gas
system. The results of these calculations are summarized in Tables 5-
8 and Figs. 3~6. Tables 5 and T and Figs. 3 and 5 give the flow rates
of the noble gases and the noble metals as a function of time after
removal from the gas bubbles. Tables 6 and & and Figs. 4 and 6 present
the heat-generation rates for these two fission~product streams as a
function of decay time. The tables are based on the production of
fission products during a one~day period, and the figures have been
normalized to obtain specific heat~generation rates. The heat-genera-
tion rates of both groups of fission products decrease considerably
after a holdup of about 10 hr. Large quantities of halogen and noble
gases are produced by the decay of the ncble metals. T1f a large frac-
tion of the noble wetals is held up in the reactor system, the same
quantity of material (but daughters in this case) will still require
processing; however, the heat load on the off-gas azystem will bhe greatly

reduced.
 

 

 

Table 5. Chemical Composition of MSBR 0ff-Gas Stream as a Function of Holdup Time
71 Flow Rate (g/day) After Holdup Period of:
ement 0.0 hr 0.0L hr 0.5 hr I hr L0 hr 100 hr
Kr 1.75x102 1.52x102 1.2Lx10° 7.12x101 b ihx10" u.h5x101
. 1 1 1 | 1 \ 1
Rb 0 1.71x10 2.26x10 3.95%1C L.81x10 4.81x10
St 0 %.52 2.59x101 6.20x101 7.91x101 7.a0x101
. =l -2 -1
Y 0 &.78x10 2.22x10 2.51x10 1.35 6.36
-G -7 -4 -2 . -1
Zr 0 2.40x10 8.79x10 1.51x10 1.02x10 1.42%10 _
o o o -
Xe 5.27x102 2.88x10° 2.16x102 1.75%10 1.29%10 1.20x10
1 1 . 1 a2 . 2
Cs 0 3.57x10 8.67x10 6.85%10 1.1Lkx10 1.23%x10
1 1 1 1
Ba 0 3 .25 2.30x10 T.22%10 T.08x10 6.96x10
-2 1 _ 1 1
La 0 1.49x%10 1.28 1.16x10 1.38%10 1.38x10
-5 - -2 -2
Ce 0 7.18x107" 2.03x10 > 2.45%10 9.78x1C 1.30
-8 5 - - _
Pr 0 1.92x1C 1.87x10 6.20x10™7 1.13%x10 2 8.6Lx10™
- - o = =)
Nd ® 4.82x10 3 5.50%x10 1o 1.53%10 [ 1.37x10 > 2.7hx10 *
g 2 o
Total 5.00x102 5 00x10° 5.00x10 5.00x102 5.00x102 5.00x10

 
Table 6. Total (B + 7) Power of Noble Gases and Their Daughters in the MSBR
O0ff-Gas System as a Function of Holdup Time

 

(B + v) Power (w) After Holdup Period of:

 

Element

 

0.0 hr 0.0L4 hr 0.4 hr 4 hr 4O hr 400 hr
Kt 9.2Tx108 2.29x108 1.99x10T 4.51x106 9.513102 2.98
Rb 0 1.27x1o8 5.51x10T 5.69x106 4.97x102 1.92x10-9

)i
Sr 0 2.05x10 2-981{10u 2.14;6}(1011L L.h9x105 2.55x103
v 0 6.88x10" 1.40x10° b .2hx10° b .06x107 6.61x10"
8 8 zeqnl 5 4 2 =

Xe 7.18x10 2.20x10 2.2%x10 4.09x10 2.84x10 2 .04x10
Cs 0 u.51x107 5.&5x107 7.’45x10b 7.20 7.19
Ba 0 1.u6xlo5 8.98x105 1.90x105 5.52x102 2.5ux102
La 0 9.52x10° 3. 1hx10” 1.90x10” 5.52x102 | 2 .5hx10°
Ce 0 1.54x10"" 1.1% 2.21 2.2% 1.12
Pr 0 2.76x10'6 2.58x10_h 5.60x10 4.10x107° 5.64x10-2

o - .
Total 1.65%107 6.20X108 1.12x108 9.5?x10t 3.66%10 L.21x10

 
16

 

 

 

Table 7. Chemical Composition of the Stream Containing Noble Metals and Daughters
as a Function of Holdup Time
£l . Flow Rate (g/day) After a Holdup Period of;
emen 0.0 nr 0.0L hr 0.4 br i hr 4G hr hOO hr
As 1.3% 1.13% 6.L1x10 7" 2.56x10'1 GL71x10 T 216wl
1 1 . ,
Se 2.76x10 1.99x10 1.13x10 .1 5.51 s
i 1 .
Br 0 £.95 1.06x10 505 5.00 5200
. . 1
Kr 0 1.82 6.34 1.42x10 1.55%10 1.55x10
S )
Rb 0 1.6hx10 ~ 1.17x10 1.12 2ok 74
2 | 2 . 1 o 1 e
Nb 1.50x10 1.44%10 9.13x10 L HOx10 5.01x10 2L EELO
o 2 - 2 o 2
Mo 1.oTx102 1.%9%10 1.60x10 1.91x10 1.88x13 1.62x10
Tc 1.58x101 1.T6x101 1.67x10 1.97 L.65x101 LB
Ru 2.02x101 3.1ux101 6.51x101 9.08x101 8.75xlol &.26x101
Rh 7.16x10"" 1.0% 1.%5 2.57 5.51 Y .0
Pd 1.77 1.92 2.94 %84 P52 AL
: -1 , . | -1 - -1 o -
Ag L.87x10 4.21x10 L4.07x10 L. 828x10 7.20x10 39T xS
-1 Vo -1 . -1 | N -
Cd 0 2.3%x10 h.65x10 L, TOx10 7.5ex10 bLshx10
L -4 o % s 1R -2 , -1
In 0 5.45x10 6.26x10 3.07x10 TLhexlo 1.60x10
. -7 ~ o wli -2 ey D 2
Sn 0 2.39x10 Z.32%10 1.85x10 6.7Tx1C {5uxlo
2 . e _ 2 2 . 2 o 1
Te 3. 17x10 3.01x10 2.45x10 1.26x10 1.02x1e L0x10
. 1 2 . ,
1 0 1.64x10" 6.67x10 1.20x10° T.Ofixlol %.28%10
o L E S oo | 2 .z
Xe o 6.85%10 3.52 6.7Ex10 1.32%10 1.55%10
Cs o 2.2%%10 77 t.01x10 ™ L .61x107h 1.53x101 £.81x10
- ~1i - . e - -8
Ba 0 £.57x10 18 1.30x10 - 1.L7x10 1 LAlxl077 &.a7x10 7
o : 2 - . 2 . 2 . .
Total 6.83x10 6.83x%10 6.8%%10 6.82x10 6.8%%L0 6.855%10

 
Table 8. Total [P + 7} Power of Noble Metals aad Their Daughters

as a Function of Holdup Time

 

Element

{B + y) Power {w) After a Holdup Period of:

 

 

0.0 hr 0.04 hr C.k hr I hr 50 hr 40G hr
- - = o : 2 -
As 1.k7x10" 2.25x107 1.30%x10” 2.51x10" 1.11x10° 1.75%10%
Se 2.48x10" 5.70x10" u.09x106 1.10x10° 2.08x107" 3.80x10'”
Br 0 1.3hx107 7.25x106 1.26x105 2.35 0
Ke Q 1.85x10h 6.20;;10L h.§2310u 1.21%10° h‘SYxlO_l
Nb 9-52x108 k. 20%10° 1.62x107 1-53x106 7. T5%10° 5.53%10°
Mo E-Ohxlob ,.98x107 1.71x107 9.963104 §.82x10" 1.6"u,x105
8 7 7 L ‘ 4 A e A2
Te 1.53x10 9.50x10 3.,20%10 1.07x10 1.21x10 2.G7x10
Ru 5.6hx106 3 95x106 5.u8x105 1.58x105 2.48%10" 1.55x105
Rh 6.16x10° 2. 67x10° b Tox10” 1.89x10" 3.59%10° 2.0hx10°
1 -
Pd 2. 77x10° 7.95%10° 7.08x10 " 2.55%10° 3 .G8x10° 5.57x10"
Ag u.o5x106 1.5Yx106 8.80:;10LL 6.6hx105 1.25x105 e.zhxlol
> 3 3 2
cd 0 5.47%10 6. 4710 2.53x10 1.97x10 3.13
. 1 2 N 2
In 0 1.5710 2.85x10 1.58x10 1.27x10 1.20
Te b 763100 9.58x10" 534107 1.82x10° 3.87x10" 1.52x10°
1 0 1o3xic® 1.01x10" k. .g0x10? 3.63%107 1.27x10"
z . I ]
Xe 0 2.21x10° 2.00x10" b lix10" 1.87x10" 1.91x10°
Cs o 9.52x10_12 1.67x10 72 1.7§x10'7 3.19x10'6 5.70x10'6
Total 2.10x10° T.T2x108 1.hex10° 8.58x106 5.17x10° E.SEXIOL

 

LT
(g/daoy)

FLOW RATE

ORNL -DWG-70-12475

 

17T

I

Xe

Kr

!'liii'!

 

j02

 

1 1 !

 

ITITT]

 

1 i iy

—

3

H

I!iil!] i T

=TT

L.l 11U

A

 

lil!lLl

 

 

 

10
| RARE
EARTHS
- +y _
Zr x 103
}— —
‘ !liilli J 1 1lil1ll L 4 Ll 12 1] 1 i illl}lz ) | L j 1 3 1]
10-¢ 101 s 10 (0 103
HOLD UP TIME (hr)
Fig. 3. Chemical Composition of MSBR Off-Gas Stream as a Function of Holdup Time for

Sparged Noble Gases.

Gl
SPEGIFIC HEAT GENERATION RATE {(w per gram of mixed fission products)

'—4
\O

ORNL-OWG 70-10034R1

 

108 103
-—Xe
< ACTI
METALS
105 102
104 10!
ACTIVE METALS
RARE EARTHS
+Y —
103 100
10-¢ o~ ! 100 10! 102 103

HOLDUP TIME {hr)

Fig. 4. Specific Heat Generation Rates for Noble Gases and Their
Daughters as a Function of Holdup Time in a 2250-Mw (thermal) MSBR
0ff-Gas System.
CRNL-DWG-70-12479

 

 

 

 

    
 

 

  

 

 

 

iosr_ T 1 yr T T T 1 !TT1TT! T ! ITTIIII 7 1 lll’!'!lI T 1 IT[T'!__
- NOBLE METALS B
-
= 102}
o —
o -
\ p—
oh
Lt -
= HALOGENS
o n
-
=
O
4
- SEMINOBLE
METALS x |0
- Xe ACTIVE .
METALS
0 i/i. Ilelii 1 jlilll] 1 1 llliill i i B L g
1G-2 10~ i 10 102 0

HOLD UP TIME (hr)

Fig. 5. Chemical Composition of Noble Metals and Their Daughters Produced ia a 2250 -Mw
J p >
(thermal) MSBR as a Function of Holdup Time.
ORNL-DWG 70-10035

 

10
NOBLE METALS

10

5 NOBLE GASES x 103

HOLOGENS

!

SEMINOBLE
METALS X 103

SPECIFIC HEAT GENERATION RATE (w per grom of mixed fission products)

3 ;
1072 2 5 o~ 2 5 10° 2 5 e} 2 5 102 2 5 103

HOLDUP TIME (hr)

10

Fig. 6. Specific Heat Generation Rates for Noble Metals and Their Daughters Produced in
a 2250-Mw (thermal) MSBR as a Function of Decay Time.

1
A

. REMOVAL OF PROTACTINIUM FROM A SINGLE-FLUID MSBR
L. E. McNeese M. E. Whatley

A method for isolating protactinium from a single~fluid MSEBR has
been described previously.l In the proposed flowsheet (Fig. 7), a salt
stream from the reactor enters the bottom of the extraction column and
flows countercurrently to a stream of bismuth containing reduced metals.
Ideally, the metal stream entering the top of the column contains suf-
ficient thorium and lithium to extract only the uranium entering the
column. The system exploits the fact that protactinium is less noble
than uranium but more noble than thorium. Hence, in the lower part of
the column, uranium is preferentially extracted from the incoming salt,
while the protactinium progresses farther up the column to the point
where it is reduced by thorium. In this manner, protactinium refluxes
in the center of the column, and relatively high protactinium concen-
trations result. A retention tank is provided at the center ol the
column, where the maximum protactinium concentration occurs in the salt,

to retain the protactinium until it decays to uranium {cf. Sect. 2).

An essential part of the flowsheet is an electrolytic oxidizer-
reducer, which serves the dual purpose of recovering the extracted
uranium from the metal stream leaving the extraction column and of
preparing the lithium-thorium-bismuth stream that is fed to the ex-
traction column. The metal phase containing the uranium extracted in
the column can serve as the cell anode, where urvanium and lithium will
be oxidized to UFM and LiF, respectively. Salt from the top of the
extraction column serves as the cell electrolyte and first passes over
a pool of liquid bismuth, which serves as the cathode into which

thorium and lithium are reduced.
ORNL DWG

68-9438

 

   

 

 

 

 

UF, IN SALT
ELECTROLYTIC
OXIDIZER -
REDUCER
COLUMN
SINGLE
FLUID
REACTOR
COLUMN
UF, AND PaF, IN SALT U IN Bi

Fig. 7. Protactinium Isolation for a Single~-Fluid MSBR.
2h

L.1 Mathematical Analysis of a Reductive Extraction System

Computations having to do with mass transfer in columns involving
the countervcurrent flow of two fluid phases can be carried out by one
of two standard techniques: a differential analysis (the HTU concept)
can be employed, or the column can be assumed to be equivalent to a
sequence of theoretical stages. TIn order to use the former, mass-
transfer coefficients must be known; for the later, the height of
column equivalent to a theoretical stage must be known. The theoretical-
stage approach was adopted for the present analysis. 1In this approach,
calculations proceed from one end of the column, where flow rates and
concentrations are known or assumed, to the opposite end of the column,
where an appropriate check is made on the calculated values (if the
starting concentrations and flow rates were assumed values). In either
case, the calculations involve the successive application of equilibrium

and material-balance relations for each of the theoretical stages.

In setting up the equilibrium relations for a system containing
N + 1 components that distribute between the molten salt and bismuth
phases, one component is conveniently chosen as the reference component.
The compositions of the two phases are then related by the following set

of expressions:

'XM ni/nr n.F
— i _,..i....., 1 - T T
i < %51 X exp g (Boy = Egp)y 1= 1.ooN
St
N+1
2 X
i=1

= X:MR,
where XSi"XMi = mole fraction of component 1 in salt and metal,
respectively,

mole fraction of reference component in salt and

1

XSr, XMr ‘
metal, respectively,
XMR = equivalents of transferrable components per mole of
bismuth,
n,, n_ = valence of component i and reference component in
salt, respectively,
F = Faraday's constant,
R = gas constant,
T = temperature, °K,
E' - Efi = difference between modified reduction potential of

ol or
component 1 and reference component.

A material balance arcund stage j of a column in which the stages

are numbered from the top yields the relation:

+ F = F X +
Sj Xsi}j I‘Mj XMiJj Sj+1 Sigj“l“l FMj"'l D{M:i.;j"l’

where I, = flow rate of salt leaving stage j,
= flow rate of metal leaving stage j,
= mole fraction of component i in salt leaving stage i,

XMi,j = mole fraction of compcnent i in metal leaving stage j.
For the present calculations, we assumed that
i~ Tejl T g
Fui = Mg~ Py

so that the above reduces to

FS(Xsi;j - Xsi,j*l) - FM(RMi}j—l - XMi,j)'
26

L.2 Calculated System Performance

The typical performance of the protactinium isolation system is
shown in Fig. 8§ for a case that will be taken as the reference case.
Conditions assumed for the calculations include the following: a
reactor salt volume of 1.5 x 106 g-moles (~1000 ftB), a decay-tank
volume of 0.3 x 106 g-moles (~200 ft5), a salt processing rate of
0.5 x 106 g-moles/day (~1.7 gpm), and six theoretical stages both above
and below the protactinium decay tank. The operating temperature was
550°C; the differences between the modified reduction potential of Th
and those for U, Pa, and Li were -0.18, -0.1h4, and +0.36 v, respec-
tively, and U was considered to be tetravalent in the extraction columns.
The protactinium production rate was assumed to be 10.7 g-moles/day for a

1000-Mw (electrical) reactor. The fuel carrier salt composition was 68-

20-12 mole % LiF -BeF,, “ThFu'

The minimum reactor protactinium concentration is obtained when the
bismuth flow rate is just sufficient to extract the uranium entering the
system. At slightly higher bismuth rates, protactinium will also be
extracted since it is the next component in order of decreasing nobility.
At bismuth rates slightly lower than the optimum rate, some uranium will
not be extracted; this uranium and most of the protactinium will flow
out of the top of the column. In either case, some protactinium is
allowed to return to Lhe reactor, and the effectiveness of the system
is diminished. The protactinium isolation system becomes ineffective
almost immediately for bismuth flow rates lower than the optimum rate;
for bismuth flow rates higher than the optimum, the reactor protactinium
concentration increases from the minimum value of 22 ppm, af the rate of
28 ppm for each percent increase in metal flow rate (for the conditions
of this particular case). Similar effects would be produced by varia-
tions in salt flow rate or in the total concentration of reduced metals
(equivalents of reduced metals per mole of bismuth) fed to the column.
Calculated concentration profiles in the extraction column are shown in
Fig. 9 for steady-state operation under optimum conditiong. The concen-

tration of uranijum in the salt decreases from the reactor concentration
ORNL-BWG 67-12907AR

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1400
|
i |
| CONDITIONS |
REACTOR VOLUME 15%x10% g- moles
1200 DECAY TANK VOLUME: 0.3x40% g~moles
\ SALT FLOW RATE: 0.5x10% g- moles /day
\ ; TOWER ABOVE TANK: 6 STAGES |
21000 DECAY TANK | TOWER BELOW TANK! 6 STAGES _|
3 i |
8 s
@ ? :
2 | j |
£ 800 ' | |
< |
= i f ' §
= ' | ;
< [ ] 5
= i : j
§ 600 ‘ j
= \ '
= |
)
QL
> |
O : |
O : i
o 400 f i i ! ;
& REACTOR AND ' ‘ ‘
/ DECAY TANK REACTOR
: | ‘il |
* i |
| |
O : ] E E ‘:
4.8 4.9 5.0 5.4 5.2 5.3 5.4 5.5 5.6 5.7 5.8

BISMUTH FLOW RATE x40~ % {g-motes / day)

Fig. 8. Variation of Protactinium Concentration in Reactor and
Protactinium Decay Tank with Bismuth Flow Rate.

Le
ORNL DWG 67-12883

 

 

 

 

 

 

 

oLIF T T f T T T ¥ T 1 10 " ©
®Bek;

% H10"!
elhF, diny -2
e o Thin Metal 110

I 0/\6"“ %“Si 41072

- \\\ 104
@—-v@-"-"‘@/ / / 6‘ p i}

roo <1072

g8l —0—o |} e

™ ‘8. 110

- \8
Lo
2
§ L
5t g
£ ¥ o =
8 -

- e &
2 o ey o
o J
o o 3
r | £ | s
U=.000063 | .000083
Pa=.00132 00130
1 L 1 L 1 1 I 1 '1\ j|0-19
| 2 3 4 5 66 7 8 292 [0 1l 12
STAGE NUMBER
Fig. §. Calculated Concentration Profiles in Reductive Extraction

Tower.
29

-% -5
of 3 x 107 mole fraction to 6.3 x 10 7 mole fraction at the inlet to the
protactinium decay tank, whereas the protactinium concentration increases

>

from the reactor concentration of 2.2 x 10 © mole fraction to 1.32 x 10.5
mole fraction at the inlet to the decay tank. The concentrations of prot-
actinium and uranium in the decay tank are 1.3 x 10.;j and 8.3 x 10_5 mole
fraction, respectively. Above the decay tank, the uranium and protac-

tinium concentrations decrease steadily to negligible values.

The flowsheet hkas several very desirable characteristics, which

25%

include a negligible holdup of fiszile U in the processing plant,

an almost immediate return of newly produced 2’2}5U to the reactor system,
and a closed system that precludes loss of protactinium, 255U, or other
components of the reactor fuel salt. However, the efficiency of protac-
tinjum removal is undesirably sensitive to minor variations in operating
conditions such as the salt or bismuth flow rate and the concentration
of reduced metals in the bismuth stream that is fed to the extraction
column. Methods for making system performance lesgs sensitive to minor
variations in operating conditions have been explored and are discussed

in the follewing section.

4.3 Stabilization of the Protactinium Tsolation System

Methods that were considered for making the protactinium system less
sensitive to changes in operating conditions included removal of uranium
from the center of the column by fluorination, withdrawal of salt con-
taining uranium and protactinium from the center of the column, relo-
cation of the protactinium decay tank, and partial oxidation of the metal
stream at the center of the column. The first tliree of these methods
were found to decrease the sensitivity of the system performance to
changes in flow rate; however, only uranium removal by fluorination
appears to be practical. 1In addition, consideration was given to using
a higher salt throughput with the original flowsheet in order to obtain
the desired time-averaged reactor protactinium concentration with a sgys~
tem allowing some variation in operating conditions. Effects of these

changes are discussed in the remainder of this section.
50

Effect of Uranium Removal by Fluorination. — The protactinium

 

isolation system caun be stabilized for bismuth flow rates lower than
the optimum rate if the uranium that is not carried out the bottom of
the column by the bismuth stream is removed from the center of the col-
umn by fluorination. The alternatives of fluorinating salt from the
decay tank or salt from the stream entering the decay tank were con-
sidered (Fig. 10). As will be shown later, fluorination of salt from

the stream entering the decay tank is the preferred operating method.

The effect of fluorinating the stream that enters the protactinium
decay tank is shown in Fig. 11 for a typical case in which 2% of the
uranium in the stream is removed. The protactinium concentration in
the reactor is also shown for the reference case (no fluorination) as
a dashed line. Removal of uranium from the inlet stream is seen to
have no effect on the reactor protactinium concentration for bismuth
flow rates higher than the optimum flow rate (5.024 x 105 g-moles/day
for the reference case), which will be designated as FMBO . As the

pt

bismuth flow rate is decreased below FMRB the reactor protactinium

opt’
concentration increases sharply from 22 pgm to 236 ppm with no fluor-
ination, but remains at 22 ppm until the bismuth flow vate has de~
creased Lo 4.972 x lO5 g-moles/day if 2% of the uranium flowing to the
decay tank is removed. As the bismuth flow rate is decreased further,
the reactor protactinium concentration increases at a slower rate than
in the reference case and reaches the maximum concentration (236 ppm)

>

at a flow rate of 4.7 x 10 g-moles/day. The minimum bismuth flow rate

that yields a reactor protactinium concentration of 22 ppm will be

designated as FMB . . The value of FMB ., 1is related to the fraction of
min min

uranium removed from the inlet stream as

FMB - FMB |
opt min

FMBopt

 

= 0.538 f,

where f is the fraction of uranium removed from the stream entering the

protactinium decay tank.
31

to
Reactor
UFs Pa
DECAY TANK
REACTOR FLUORINATOR| | 5

ORNL DWG 68-2083

 

 
   
  
    
 
 
 
      
   
 
 
  

OXIDIZER -
| REDUCER

  

EXTRACTION
COLUMN

 

 

 

 

 

 

OXIDIZER ~
to REDUCER

Reactor

  

Pa DECAY
UFg  TANK

EXTRACTION

REACTOR FLUORINATOR COLUMN

 

 

 

 

(b) URANIUM REMOVAL FROM SALT IN Pa DECAY TANK

Fig. 10. Uranium Removal from a Protactinium Isolation System by
Fluorination.
REACTOR PROTACTINIUM CONCENTRATION x 10° {mole fraction)

250

200

150

50

ORNL DWG €8-2084

Reference Case
{No Fluoringticn )

 

BISMUTH FLOW RATE x 107 ° (g moles/ day )

Fig. 11. Variation of Reactor Protactinium Concentration with Bismuth

q

Fiow Rate for Removal of 2% of Uranium in Inlet Stream by Fluorination and

for No Uranium Removal.
(NN
W

The value of FMB0 is fixed when system parameters other than the

C
fraction of uranium reioved by fluorination are specified; thus, for the
ref@{ence case, the bismuth flow rate interval in which the minimum reac-
tor protactinium concentration is maintained increases at the rate of
0.538% {based on FMBOpt) for each percent of uranium removed from the
salt stream entering the protactinium decay tank. For bismuth flow

rates below FMBOpt,,the peint at which the maximum reactor protactinium

concentration is reached (for the reference case) is given by

FMB@R mZBMBfl,

FMBflR

= 0.982 f,

where FMBER is the bismuth flow rate yielding the maximum reactor prot-
actinium concentration without flucripation {5.016 x 105 g-moles/day),
and FMBE ig the bismuth flow rate yielding the maximum concentration

with fluorination.

The variation of the UF6 concentration in the fluorinater off-gas
with the bismuth flow rate is shown in Fig. 12 for several values of f.
The rate at which flucrine is fed to the fluorinator and the fraction of
uranium that is removed from the salt passing through the fluorimator
were assumed to he constant. For bismuth flow rates greater than FMBopt’
the UF6 concentration is seen te be independent of the fraction of
uranium removed from the steam entering the decay tank. As the bismuth
flow rate decreases_below iMBopt’ the UF6 concentration rafiio increases
sharply from about 5 x 1077 at FMBopt to about 0.54 at FMBmin, Thus,
the concentration of UF6 in the fluovinateor cff-gas is extremely sensi-
tive to minor variations in bismuth flow rate in the desired operating
range (FMBmin E_FMB'S_FMBopt)e It is anticipated that the optimum sys~-
tem for controlling the bismuth flow rate (or more likely, controlling

the oxidizer-reducer amperage) will be based on sensing the UF6 concen-

tration in the flucrinator off~gas.
2] ,'4

ORNL DWG 68-2085

Moximum Uranium Removai=6%—

UFg CONCENTRATION FROM FLUORINATOR/MAXIMUM UF, CONCENTRATION

. FMBpin - FMB i 1 FMB qin L FMBopi.
(6% ) (2%) (0.6%]) ‘

 

340 344 348 3.52 356 36 364 368
BISMUTH FLOW RATE (ga!/ min)
Fig. 12. Variation of Ratio of UFg Concentration in the Fluorinator

Off-Gas to Maximum UF6 Concentration in ¥luorinator with Bismuth Flow Rate
and Fraction of Uranium Removed from Stream Eantering Decay Tank.
A

L

Fluorination of salt that is withdrawn from the protactinium decay
tank [Fig. 10(b)] produces an effect similar to fluecrination of part of
the stream entering the tank, but is less desirable for several reasons.
As shown in Fig. 13, the UF6 concentration in the fluorinator off~-gas is
less sensitive to variations in bismuth flow rate for this mode of oper-
ation than for fluorinaticon of part of the inlet stream, since the
uranium concentration in the decay tank is appreciably higher {for FMB
E'FMBopt) than in the inlet stream. Fluovrination of salt withdrawn from
the decay tank is also less desirable since a long time interval is
required for the uranium concentration in the decay tank {and hence the
measgured UF6 concentration) to reflect minor variations in operating
conditions (tank volume/flow = 0.6 day), whereas the change in uranium
concentration in the stream entering the tank wilil be almost instanta-

neous .

Effect of Withdrawing Salt from Center of Extraction Column. — The

 

protactinium isolation system can alsc be stabilized with respect to low
bismuth flow rates by withdrawing salt containing protactinium and ura-
nium from the center of the caluhn, as shown in Fig. 14. The stream
passes through the protactinium decay tank before its return to the
reactor system. The effect of withdrawing %% of the salt flowing to the
protactinium~igsolation system is shown in Fig. 15. The minimum protac-
tinium concentration in the reactor is 75 ppm for a decay-tank volume
equal to that of the reference cése. The concentration of 22 ppm is
approached as a limit only as the volume of the decay tank becomes
infinite; therefore, withdrawal of salt from the center of the column

is not considered to be a practical method of stabilization.

Effect of the Location of the Protactinium Decay Tank.~ In the

 

reference flowsheet (Fig. 7) and all modificaticns considered thus far,
the protactinium decay tank is located at the center of the extraction
column, which consists of six stages above aund six stages below the tank.

The variation cf the protactinium ceoncentration in the reactor with
N
O™

ORNL DWG 68-2086

Flugorination of
Salt From
Decay Tank

Fluorination of
Salt From
inlet to Decay
- Tank

UFg; CONCENTRATION FROM FLUORINATOR/MAXIMUM UF, CONCENTRATION

Magzimum Uranium Removal= 6%

 

4.75 4.8 4.85 4.9 4,85 5.0 5.05 5.1
BISMUTH FLOW RATE x10 (g moles / day)
Fig. 12. Comparison of UFg Concentration in Fluorinator Off-Gas for

Salt Withdrawn from Protactinium Decay Tank with UF6 Concentration for
Salt Taken from Inlet to Tank.
ORNL DWG 68-2087

 

 

2
N

REACTOR

 

 

OXIDIZER -
REDUCER

 

EXTRACTION
COLUMN

 

 

o

Fig. 14. Stabilization of Protactinium Isolation System by With-

drawal of Salt from Center of Column.

)e
€

REACTOR PROTACTINIUM CONCENTRATION x 10 (mole fraction)

ORNL DWG 68-2088

250 o L 5 ; ' Conditions
i : &
: o : Ll Reactor Volume 1.9 x% e g moies
Sait Fiow Rate 0.5x% {0 g moles/day

Tower Above Withdrawol Point 6 Stages
ST T | : T Tower Below Withdrowai Point © Stages
200 Frr =+ e e Fraction of Salf Withdrawn 0.05

 

180
.
o
= T _ - Decay Tank Voiume
100 e e = : —r = =1 =3x10° g moles
56
- Decay Tank
Pt Volumes ©
U T z . T . . T T y 0T LA I
4.7 4.8 4.9 5.0 8.

BISMUTH FLOW RATE x 107 ° {g moles/day)

Fig. 15. Variation of the Protactinium Concentration in the Reactor
with Bismuth Flow Rate When Salt Is Withdrawn at Center of Column.
bismuth flow rate is shown in Fig. 16 for six or less stages below the
decay tank (total of 12 stages in the column). Although the system
becomes less sensitive to variations in bismuth flow rate as the number
of stages below the decay tank decreases, this effect is more than off-~
set by a continual increase in the minimum protactinium concentration

in the reactor.

Effect of Partial Oxidation of the Metal Stream. — Partial oxida-
tion of the metal stream adjacent to the protactinium decay tank was
found to digplace the minimum concentration of protactinium in the
reactor in the direction of highér metal flow rates without changing
the general behavior of the system. It produced no tendency toward a

mare nearly stable system.

Effect of Higher Salt Throughput with Initial System. = A specified
protactinium concentration in the reactor can be cobtained even when

bigmuth flow rates are slightly higher than FMBO if the cycle time for

pt

reactor processing is appropriately decreased. 1In the reference case,
the concentration of protactinium in the reactor at steady state in-~
creases from a minimum of 22 ppm, at the rate of 28 ppm for each percent

increase in bismuth flow rate above FMBD that is,

Pt;

C = 22 + 2800 W, 0 <W<0.09%

reactor protactinium concentration, ppm,

where C =
FMB -TFMBO ¢
W = Tn ' , the fractional increase in bismuth flow rate
above TMB .
opt

If a salit throughput equivalent to a one-day cycle time were uged in-
stead of the three-day cycle time for the reference case, the steady-
state concentration of protactinium in the reactor would increase from

T.76 ppm at the rate of 25.1 ppm for each percent increase in flow rate.

That is,

C = T.76 + 2510 W 0 <W<0.1093%.
REACTOR PROTACTINIUM CONCENTRATION x iOS(moie;fmction)

250

200

G0

50

4.9

ORNL DWG 68-2089

3 Below

vy .

|

8 or 4 Stoges Below —=

4.95 5.0 5.05 5.1 5.15 5.2
BISMUTH FLOW RATE x 10~ { g moles/day)

Fig. 16. Variation of Protactinium Concentration in the Reactor
with Bismuth Flow Rate and Location of Protactinium Decay Tank.

5.25

 

5.3
b.4  Transient Performance

Calculations were made to determine the transient behavior of the
protactinium~-isclation system. Since the concentration of uranium in
the salt entering the decay tank had been shown to be very sensitive to
operating conditions when these conditions were near optimum, measure-
ment of this concentration was chesen as the means of controlling the
system {Fig. 17). We assumed that the uranium concentration in the salt
entering the decay tank could be measured by fluorinating approximately
5% of the stream. The measured uranium concentration would then be used
by a controller having proportional, integral, and derivative actions to
conicrel the flow rate of bismuth through the columns. For purposes of
simulating the actual system, a random error distributed normally about
the controller output and having a specified standard deviation (usually
5%) was imposed on the control system. The system was assumed to operate
for a specified time interval (0.06 day) at each bismuth flow rate
selected by the control system. During this period, the extracticon col-
umns were assumed to operate at.steady state; however, rhe reactor and
decay tank were treated as perfectly mixed vessels having inlet concen-

trations equal to the column effluent concentrations.

The calculated system response is shown in Fig. 18 for an initial
praotactinium concentration of 10mh mole fraction in the reactor. The
reactor volume used for the transient calculations was 1000 ftB)&Hd
the other conditions were identical to those used for the steady-state
calculations. Resdlts show that the pretactinium concentration decreases
to approximately 4 x 1077 mole fraction, which is acceptably low. (This
value would be expected to be even lower for a larger reactor volume.)

Therefore, contrel of the protactinium isolation system in the manner

suggested is believed to be practical.
UF4IN SALT

ORNL DWG $8-4090R

 

 

SINGLE

FLUID
REACTOR

 

 

  
    

ELECTRCLYTIC
OXIDIZER—
REDUCER

COLUMN

 

 

CONTROLLER

 

 

  
 

COLUMN

I

I
550
o =203

 

 

UF, AND PaF, IN SALT

Fig. 17.

U iN Bi

Schematic Diagram Showing Method for Controlling Metal

Flow Rate by Measurement of Uranium Concentration in Salt Entering Prot-

actinium Decay Tank.
ENTERING TANK x10®

U

{mole fraction)

BISMUTH FLOW RATL {gpm]

"I

 

 

 

 

PROP BAND = 50,000 %

- a =005
= Aot
Treser = 1.0 day
TnERly = C.05 cay
X
1 .
r - *
.
*an * »
. .
.
.
Ao
— .-. .'.
-
l...
.
.l. .............. -
...
.
.,
.I
...'I‘
neg
.
.
..,
.,
fmn. lp".
»
-,
n.-.-.“,
s an
.
e
. *,
."--._..,.' .'.. u-..‘a‘_”q.._’.“ 0""". e,
Yo, ., o » s, wte
"-..“‘-,." - eo
-

ORNL-CWG 68 -410TR2

 

 

TIMZ {deys)

Control of Protactinium Isolation Process Through

Fig. 18.
Measurement of Uranium Concentration in Salt Entering Decay Tank.

100

an

7o

ENYRATION »10°

{moie fractien )

TOR PROIACTINIUM CONC

REAT

e
L

5. USE OF THE PROTACTINIUM -ISOLATION SYSTEM FOR CONTROLLING THE
URANIUM CONCENTRATION IN THE BLANKET OF A SINGLE-FLUID MSBR

L. E. McNeese

A characteristic of the system for protactinium isolation, as
presently envisioned, is that it produces salt that is free of uranium
and protactinium at one part of the system. This characteristic could
be exploited to decrease the uranium concentration in the blanket of a
single-fluid MSBR, and thereby decrease the inventory of uranium in the

reactor system. A possible operating method is shown in Fig. 19.

Assume that the core and blanket regions are separated by a mem-
brane that allows limited exchange of salt in the two regions and that
the core and blanket volumes have separate heat exchangers and separate
salt-~circulation systems. The salt flow rate through the core to the
processing plant is small as compared with the flow rate through the
core heat exchanger; likewise, the flow rate through the blanket to the
processing system is small as compared with the flow rate through the
blanket heat exchanger. Salt containing UFA and PaFLL is withdrawn from
the core for removal of protactinium; salt is also withdrawn from the
blanket. These two streams are fed to the bottom of an extraction col-
umin that is a part of the conventional protactinium-~-isolation system-.

In the lower part of this column, the uranium is extracted iato the
downflowing metal stream, and the protactinium is isolated in the prot-
actinium decay tank. Part of the salt stream flowing out of the top of
the extraction column (i.e., a stream that contains essentially no prot-
actinium or uranium) is returned to the blanket. The remaining salt
flows to the cathode side of the electrolytic cell to provide lithium and
thorium fluorides, which are reduced into the bismuth cathode to form the
metal stream that is fed to the column. Uranium flowing out the bottom
of the column is oxidized at the cell anode and transfervred to the salt

stream flowing through the cell.
ORNL DWG 68-12267

 

 

 

 

 

 

 

 

 
    

 

 

 

 

 

 

 

 

 

 

 

  

 

&
OXIDIZER
REDUCER
F
KF COLUMN
Y
- L L L
CORE BLANKET
0 ‘-—L———- p)———
CORE HEAT BLANKET
EXCHANGER HEAT Po DECAY
EXCHANGER TANK COLUMN
F {14+K)F

 

 

Fig. 19. Use of the Protactinium Isolation System for Decreasing the Uranium Concentration

in the Blanket of a Single-Fluid MSBR.

in the text.

Flow rates of various streams are denoted as discussed

o
Because of leakage between the core and blanket volumes, uranium
tends to flow from the core to the blanket. The proposed system would
limit the concentration of uranium in the blanket by purging this vol-
ume with salt containing no uranium or protactinium. Uranium extracted
from both the blanket and core streams would be returned exclusively to
the core region in salt flowing out of the cell anode. A study was made
to examine the extent of exchange that could be tolerated between the
core and blanket volumes without unduly affecting the protactinium-

isolation system. The results are given below.

Since the uranium and protactinium concentrations in the salt leav-
ing the upper column are assumed to be negligible, a uranium material

balance around the blanket volume yields:

LU_ = LU+ KkFU, (1)
where L = rate of exchange of core and blanket fluids,
o = uranium concentration in core,

Ub = uranium concentration in blanket,
F = flow rate through core from protactinium-removal system,

k = ratio of blanket processing rate to core processing rate.

Solving for L, one obtains:

 

 

U

b

kF UC

= — 2
L " (2)

1~ =
U

C

The ratio of the exchange rate to the total flow rate through the core
can then be obtained by dividing through by FC, the flow rate through

the core heat exchanger, to give:

T

L c
= =k G o (3)

c U

U
b

Values for the ratioihh% are given in Fig. 20 as a function of k
and the ratio Ub/Uc for an assumed core heat exchanger flgw rate of
128 ftB/sec and an assumed core processing‘rate of 487 ftj/day. Thisg
figure shows that, if the uranium concentration in the blanket is to
be 10% of that in the core, then approximately 0.05% and 0.5% of the
flow through the core can exchange between the core and blanket when
the values of k are 10 and 100 respectively. Also, the allowable

exchange rates are less than 1% per pass through the core for k values

of 100 or less.

From a material balance on uranium in the core, one obtains the

relation:

U+ LU, = LU+ FU, (1)

where Up is the uranium concentration in the salt stream veturning to
the core. Combining Eqs. (1) and (4) yields the following relation for
UP/UC;

UP/Uc = 1 + k{17 /U_), (5)

which also represents the ratio of the rate at which uranium is reduced
in the present system to the rate of uranium reduction in the conventicnal
protactinium-isclation system. If the concentration of uranium in the
blanket is 10% of the concentration of uranium in the core, the value of
the ratio, Up/Uc’ is 2 for k = 10 and 11 for k = 100. That is, in the
first case, twice as much uranium must be reduced in the protactinium-
isclation system as compatred with the conventional pretactinium-isolation
system, and the stream returning to the core will have a uranium concen-
tration twice that of the core. 1In the second case, 11 times as much
uranium must be reduced in the protactinium-isolation system, and the
uranivm concentration in the stream returning o the core is 11 times

the concentration in the core.
10'3 ORNL DWG 68-12262
1

 

 

 

 

 

Fig. 20. Variation of Ratio of Allowable Core-Blanket Exchange Rate
to Core Flow Rate for Given Blanket~-to-Core Uranium Concentration Ratios
and Blanket-to-Core Processing Rate Ratios.
Of primary interest is the effect of the increased salt flow rate
(and the attendant requirements for increased uranium reduction) on the
protactinium-isolation system. We will assume that the permissible flow
rates of salt and metal in the packed-column extractor are given by the

relation

1/2
. + v, / = 21.6, (6)
where V. is the superficial velocity of the continuous (salt) phase and

V4 is the superficial velocity of the discontinuous (metal) phase, both
expressed in feet per hour. We will further assume that the metal stream
entering the column contains 0.0016 mole fraction thorium, and that the
concentration of uranium in the core is 0.00% mole fraction. Since all
the uranium is removed from the salt, the required quantities of uranium
and reductant are stoichiometrically related; thus, the ratio of the super-~
ficial velocity in the discontinuous phase to that in the continuocus

phase (after inserting a conversion factor to convert molar rates to
linear velocities) is:

e
v 1.98F 3
4. ¢
v, (1 + k)¥F

o1
= 1.98 G o (7)

Substituting Eq. (7) into Eq. (6) and solving for the square root of the

superficial velocity of the continuous phase yields:

. IVERS 21.6
C

 

' (8)
1+ [1'98(Up/uc)/(1 . k)]l/e
The required column diameter is related to the superficial velocity of

the continuous phase by

_ u1/2
D = %. fil;i?jihi ] (9)

c

Substitution of Eqs. (3), (5), and (8) into this relation yields the
required column diameter, D, as a function of the relative exchange rate
between the core and blanket volumes (L/FC) and the ratio of rhe uranium

concentration in the blanket to that in the core:

" 1/z

1/2 I

}__,_13.;_
U

C

L
. C
1 * 7 - l: + 1.98 1 + 7

Fr
’TJ!‘L—i

C
21.6
(10)

Figure 21 shows the required column diameters as a function of L/FC and
the ratio UbZUC- In this case, the rate of salt withdrawal from the core,
F, is 487 ftj/day (i.e., a three-day cycle with respect to the combined
core and blanket volumes) and the core flow rate, FC, is 128 fts/sec.

The figure shows that, for a Ub/Uc rat}o of O.l,‘the required column
diameters for L/FC values of 10'5, 10—4, and 1Omj are approximately

9.3, 20.0, and 58.9 in., respectively.

The practical range of column diameter values is believed to be
20 in. or less, for which the acceptable exchange rates are quite low if
the uranium concentration in the blanket is to be 10% of that in the
core. Hence, we conclude that the use of the protactinium-isolation
system for decreasing the uranium concentration in the blanket of a
single-fluid MSBR is practicable only for reactor designs having an ex-
change rate between the core and blanket regions of less than about 0.01%

of the core flow rate.
56 ORNL DWG TO-412473R2

 

F= 487 13/ day
Fo = 128 3/ sec

48 |-

REQUIRED COLUMN DIAMETER (in.)

 

 

 

0 | { ! { Pt § 41 | ] i { P
107> 2 3 4 5 6 7 8 9 5% 2 3 4 5 & 7 8 953

RATIO OF EXCHANGE RATE BETWEEN CORE AND BLANKET TO CORE FLOW RATE, L/Fe

 

Fig. 21. Variation of Required Extraction Column Diameter with Ratio of Uranium Concen-
tration in Blanket to That in Core and Ratio of Exchange Rate Between Core and Blanket to
Core Flow Rate.
52

6. REMCVAL OF RARE FARTHS FROM A SINGLE-FLUID MSBR

L. E. McNeese

The rare-earth fission products are among the more important neu-
tron absorbers in an MSBR, and successful operation of this type of
reactor requires the removal of such materials on a cycle of approxi-
mately 50 days. Removal of rare earths from a single-fluid MSBR is
complicated by the need for separating the rare earths from thorium,

a major component of the salt, since thorium fluoride and the rare-~

earth fluorides are chemically similar.

6.1 Proposed Rare-Earth Removal Flowsheet

The proposed method for removing rare earths from a single-fluid
MSBR is based on differences in the extent to which the rare earths and
thorium distribute between molten salt and liquid bismuth containing a
reductant. The removal system is shown in its simplest form in Fig. 22.
A molten-salt stream that consists of fluorides of lithium, beryllium,
and thorium as well as rare~earth fluorides is fed to an extraction
column. The salt flows countercurrent to a stream of liquid bismuth
containing thorium and lithium. In the upper part of the column, a
large fraction of the rare earths is reduced and transfers to the down-
flowing metal stream. Below the feed point, the rare-earth countent of
the salt and metal streams is increased in order to produce a concen-

tration suitably high for disposal.

Molten salt leaving the top of the column containg rare earths at
a low concentration. Part of this salt is returned to the reactor, and
the remainder is sent to an electrolytic-cell complex. The complex is
used to add thorium and lithium to bismuth for use as extractant and to
return the extracted rare earths, which enter the complex with bismuth

from the bottom of the cascade, to the cascade as reflux by oxidizing
ORNL DWG 68-9443

 

 

»

ELECTROLYTIC
OXIDIZER -
REDUCER

PROCESSED SALT
gl

EXTRACTION

COLUMN SALT CONTAINING
v LnFy TO WASTE
e e

SALT CONTAINING LnF,

 

EXTRACTION
COLUMN

 

Bi CONTAINING Ln >

 

SALT CONTAINING LnFy ¢

 

Fig. 22. Removal of Rare Earths from a Single-Fluid Reactor by
Reductive Extraction.
N
L
£

them out of the bismuth and transferring them to the returning salt
stream. The electrolytic-cell complex consists of an electrolytic

cell with contactors above and below it. Both the anode and the
cathode of the cell are pools of flowing bismuth, and the electrolyte
is salt that does not contain large amounts of rare earths or thorium.
The cathode adds lithium to the bismuth stream that flows into the
lower contactor and extracts essentially all (about 99%) of the thorium
out of the entering salt. The anode releases BiF5 into the salt that
flows into the upper contactor and oxidizes essentially all of the rare
earths out of the entering bismuth.

Efficiencies for removal of rare earths were calculated for a
range of operating conditions to establish the importance of the rare-
earth-~thorium separation factor, the fraction of ThFu that is reduced
in the electrolytic cell, the location of the feed point, the number
of stages in the extraction column, the metal/salt flow ratio, and the
concentration of rare earths in the discard stream. A rveactor volume
of 1461 ftB, an operating temperature of 600°C, and a reactor power
of 1000 Mw (electrical) were assumed. The thorium and lithium con-
centrations in the bismuth stream that was fed to the extraction col-
umn were each 0.0016 mole fraction. Within the range of interest, the
concentration of the rare-earth fluorides in the withdrawn salt stream
had no effect, and a concentration of 0.0069 mole fraction was assumed
in all cases. This is less than 50% of the rare-earth fluoride solu-

bility in the fuel carrier salt at 600°C.

6.2 Effect of Rare-Farth--Thorium Separation Factor and Bismuth Flow Rate

Calculated results showing the effects of rare~earth-~thorium
separation factor, bismuth flow rate, and feed-point location are given
in Figs. 25-%30 for a 30-day processing cycle and a total of 24 stages
in the extraction column. It was assumed that 99% of the thorium was

reduced from the salt as it passed through the electrolytic cell, and
30

C
o

4
(o

o
QO W

REF REMOVAL TIME, DAYS

B3 @
o O
|

160
200
300

Fig. 2%.

   
 
 
  

ORNL -DWG-70~1472

1.0

0.9

0.8

o.7

0.6

0.5

c.4

0.3

FRACTION REF REMOVED

0.2

C.1

e . —— el o z “ - el . e . . E L . St - - ALY
1.0 15 2 3 5 7.5 10 15 20 30 40
BISMUTH FLOW RATE, gpm

 

Variation of Rare-Earth Removal Time with Bismuth Flow Rate and Rare-~Earth--

Thorium Separation Factor for 2 Stages Above the Feed Point and 22 Stages Below the Feed

Point.

GS
ORML-DWG-70-1471

30

 

35
g 490
2 45 @
Wi >
= 50 =
- ]
3 o
< 60 i
> m
o @
=
Ll r =
C 39 2
ka.
& 100 a
=
150
200
300
1.0 1.5 2 3 5 7.5 10 15 20 30 40
BISMUTH FLOW RATE, gpm
Fig. 2L. Variation of Rare-Earth Removal Time with Bismuth Fiow Rate and Rare-Earth--

Thorium Separztion Factor for 4 Stages Above the Feed Point and 20 Stages Below the Feed Point.
ORNL—DWG-70-1470

30

35

40

45
50

60

80

REF REMOVAL TIME, DAYS
FRACTION REF REMOVED

100

150
200
300

 

1.0 1.2 2 3 5 7.5 10 15 20 30 40
BISMUTH FLOW RATE, gpm

~ Fig. 25. Variation of Rare-Earth Removal Time with Bismuth Flow Rate and Rare-Earth-- -
Thorium Separation Factor for 5 Stages Above the Feed Point and 19 Stages Below the Feed Point.

.S
DAYS

REF REMOVAL TIME,

10 15 2 3 5 75 10
BISMUTH FLOW RATE, aom

 

ORNL—DWG-70-1469

FRACTION REF REMOVED

Fig. 26. Variation of Rare-Earth Removal Time with Bismuth Flow Rate and Rare-Earth--

Thorium Separation Factor for © Stages Above the Feed Point and

-~

1o Stages Below the Feed Point.
QRNL-OWG-70-1468

30

35

40

45
50

60

FRACTION REF REMOVED

80

REF REMOVAL T!ME, DAYS

160

150
200~
300~

 

1.0 15 2 3 5 7.5 10 18 20 30 40
BISMUTH FLOW RATE, gpm

Fig. 27. Variation of Rare-Earth Removal Time with Bismuth Flow Rate and Rare-Earth--
Thorium Separation Factor for J Stages Above the Feed Point and 15 Stages Below the Feed Point.
ORNL-DWG-T0-1473

REF REMOVAL TIME, DAYS
FRACTION REF REMOVED

 

1.0 1.5 2 3 5 75 10 15 20 30 40
BISMUTH FLOW RATE, gpm

Fig. 28. Variation of Rare-Earth Removal Time with Bismuth Flow Rate and Rare-Earth--
Thorium Separation Factor for 12 Stages Above tné Feed Point and 12 Stages Below the Feed Point.

09
 

 

 

 

wgd.ufic.Jq>ozmm 43y

 

e
S

.

 

 

 

 

 

 

il

T

oy

2

 

Rare—Earth-~

e

.

10n of Rar

Varigs

int,

the Feeg Po

Stageg Below
REF REMOVAL TIME  DAYS

ORNL - DWG- 70-9]

30 0
09

35
,,,,,, 08
S L ; LT | 07

45 ; . Lo IS A |

5C -
50 0.5
C.4

80
IGO0 3
SO T T A T T e 2

200C
306 ol
0

 

0 1.5 2 3 5 75 1G 15 20 30 a0

BISMUTH FLOW RATE,gpm

FRACTION REF REMOVED

Fig. 50. Variation of Rare-Farth Removal Time with Bismuth Flow Rate and Rare-Farth--

Thorium Separation Factor for 18 Stages Above the Feed Point and & Stages Below the Feed Point.

<9
63

that the concentration of rare-earth fluorides in the withdrawn salt
stream was 0.69 mole %. The system is seen to be relatively efficient
for separation factors of 2 or greater but relatively inefficient for

. separation factors lower than 1.5.

6.3 Effect of Location of Feed Point

Figure 31 shows the effect of feed-point location for operating
conditions that include a bismuth flow rate of 15 gpm, a total of 24
stages, 99% reduction of thorium in the electrolytic cell, a rare-
earth fluoride conceéentration of 0.0069 mole fraction in the withdrawn
salt stream, and a processing cycle time of 30 days. It is seen that
the optimum feed-point location depends on the rare-earth--thorium
separation factor and shifts from the center of the column for a
separation factor of 2.0 to near the top of the column for separation
factors near 1.0. The metal/salt flow ratios are relatively high for
the conditions given, that is, 34 and 85.4 for the upper and lower

columns, respectively.

6.4 Effect of the Fraction of ThF, That Is Reduced in the
Electrolytic Cell

Figure 32 shows the effect of the fraction of ThFLL that is reduced
in the electrolytic cell for a rare-earth--thorium separation factor of
2, a 24 -stage celumn, and a rare-earth concentration in the withdrawal
stream of 0.0069 mole fraction. It should be noted that the importance
of this parameter is equal to, if not greater than, that of the other
parameters considered. If the fraction of ThFu that is reduced is
decreased from 99 to 90%, a significant decrease in removal efficiency
results; the system becomes ineffective if less than 50% of the ThF

N

is reducead.
&l

ORML-DWG-70~92

RARE EARTH REMOVAL TIME (DAYS)

 

0 4 8 12 16 20 24 28
NUMBER OF STAGES IN LOWER COLUMN

Fig. 31. Variation of Rare-Earth Removal Time with Feed-Point
Location and Rare-Earth--Thorium Separation Factor.
REF REMOVAL TIME, DAYS

I‘hFM

ORNL—DWG-T0-83

30

35

40

45
50

60

FRACTION REF REMOVED

80
100
150

200
300

 

1.0 1.5 2 3 5 7.5 10 15 20 30 40

BISMUTH FLOW RATE, gpm

Fig. 32. Variation of Rare-Earth Removal Time with Bismuth Flow Rate aund the Fraction of
That Is Reduced in the Electrolytic Cell.
6.5 Effect of Number of Stages in the Extraction Column

Calculated rare-earth removal times for systems having a total of
2h, 18, and 12 stages are shown in Figs. 28, 33, and 34 for a feed point
at the center of the column. Assumed operating conditions include a
50~day processing cycle time, a concentration of rare-~earth fluorides of
0.0069 mole fraction in the withdrawn salt stream, and 99% reduction of

the ThFh in the electrolytic cell.

From the previous sections, it is apparent that, for a rare-earth--
thorium separation factor of about 1.2, about 20 stages are required
below the feed point in order to achieve a marginally acceptable rare-
earth removal time. The effects of the bismuth flow rate and the number
of stages in the upper column are shown in Figs. 24 and 35-37. The
effect of the number of stages in the upper column for a bismuth flow

rate of 15 gpm is shown in Fig. 38.

6.6 Effect of Processing Cycle Time

Figure 39 shows the effects of the processing cycle time and the
rare-earth--thorium separation factor for a bismuth flow rate of 15 gpm
and 5 stages above and 19 stages below the feed point. As expected, the
rare~earth removal times were insensitive to processing cycle time for

low separation factors.

The effects of rare-earth--thorium separation factor and bismuth
flow rate on the rare-earth removal time are shown in Figs. 25 and
LO-k2 for processing cycle times of &, 7.5, 19, and 30 days in the

case of a system having 5 stages above and 19 stages below the feed

point.
ORNL-DWG-70-924

 

30 1.0
- 08 _
35 0l
=
> = 08
g 40—_ b
0.7
S 45 i
=
= 50 06 »
_‘ 9
F.
T 60 06 ©
< <
= o e T T e A T T A T T T T A e v
L e A T A e T T e e N i
i 0.4
L 8O
wl i .
100+ 0.3
150 0.2
200
300 0.1
0

1.0 1.5 2 3 5 7.5 10 15 20 30 40
" BISMUTH FLOW RATE, gpm

Fig. %3%. Variation of Rare-Earth Removal Time with Bismuth Flow Rate and Rare-Earth--
Thorium Separation Factor for Nine Stages Above the Feed Point and Nine Stages Below the
Feed Point.
QRNL-DWG-T0-95

Laipeigad

30 1.0
0.9
35
g 08
< 40
- o7
45 ;
- 50 0.6
I
3 60 —05
E .
W
04
w 80 -
i
x !
100 0.3
150 0.2
200

300 CA

  
   

1.0 1.5 2 3 5 7.5 10 15 20 30 40

ISMUTH FLOW RATE, gpm

Fig. 3k. Variation of Rare-Earth Removal Time with Bismuth Flow Rate and Rare-Earth--
Thorium Separation Factor for Six Stages Above the Feed Point and Six Stages Below the Feed

FRACTION REF REMOVED

Point.

89
30

35

45
50

60

80

REF REMOVAL TIME, DAYS

100

180

300

Fig. 35.

 

ORNL-DWG-70-96R!

G.9

0.8

0.6
0.5

04

FRACTION REF REMOVED

-0.3
0.2
0.1
10 15 2 3 5 7.5 10 15 20 30 40
BISMUTH FLOW RATE, gpm

Variation of Rare-Farth Removal Time with Bismuth Flow Rate and Rare-farth--

Thorium Separation Factor for 6 Stages Above the Feed Point and 20 Stages Below the Feed Point.

69
ORNL-DWG -70-927

w O

)—

< S

o o

w 5

= o

=

. o

<

3 o

O

=

l s

o o
'..-
Q

[N |

1 =

o '

 

e 1.5 2 3 5 7.5 10 15 20 30 40

BISMUTH FLOW RATE, gpm

Fig. 36. Variation of Rare-Farth Removal Time with Bismuth Flow Rate and Rare-Earth--
Thorium Separation Factor for § Stages Above the Feed Point and 20 Stages Below the Feed Point.

0l
DRNL - DWG-7C-98

P
™
2 w
: 3
< s
o Lat
- T
-3 ,
v
g G
e o
2
z
* 5
u —
ul o
o =1
«
[

 

1O 5 2 3 5 75 10 5 20 30 40

BISMUTH FLOW RATE, gpm

Fig. 37. Variation of Rare-Earth Removal Time with Bismuth Flow Rate and Rare-Earth--
Thorium Separation Factor for 10 Stages Above the Feed Point and 20 Stages Below the Feed Point.

14
ORNL-DWG-70-99

100
90

700
600
500

400

300

200

100
90

70
60

50

RARE EARTH REMOVAL TIME { DAYS)

 

2 4 8 8 10

NUMBER OF STAGES IN UPPER COLUMN

Fig. 38. Variation of Rare-Earth Removal Time with Rare-Earth--
Thorium Separation Factor and Number of Stages in Upper Column When the
Lower Column Coutains 20 Stages.
ORNL-DWG-7T0-100

SALT FEED RATE, (gpm)
0 0.2 0.4 0.6 08 1.0 .2 1.4

RARE EARTH REMOVAL TIME, { DAYS)

 

30 15 1Q 75 6

PROCESSING CYCLE TIME, (DAYS}

Fig. 39. Variation of Rare-Earth Removal Time with Process Cycle
Time and Rare-~Earth--Thorium Separation Factor for a Bismuth Flow Rate
of 15 gpm and a Total of 2I Stages.
ORNL-DWG-70-1465

8_1

REF REMOVAL TIME, DAYS

20

30
40—z

-~

50

{.0 1.9 2 3 5 7.5

<
o

BISMUTH FLOW RATE, ¢opm

Fig. 40. Variation of Rare-Earth Removal Time with Bismuth Flow Ra
Thorium Separation Factor for a Process Cycle Time of 6 Days and a Total

 

FRACTION REF REMOVED

20 30

te zné Rare-Farth--
of 24 Stages.
ORNL-DWG-70-1466

REF REMOVAL TIME, DAYS
FRACTION REF REMOVED

 

1.0 1.5 2 3 5 75 10 15 20 30 40
BISMUTH FLOW RATE, gpm

Fig. 41. Variation of Rare-Farth Removal Time with Bismuth Flow Rate and Rare-Earth--
Thorium Separation Factor for a Process Cycle Time of 7.5 Days and a Total of 2l Stages.
ORNL-DWG-70-1467

1
=1

w
fa)
2 20 tal
G 3
- =
wd
L
‘E o
= 25
.
- w
< o
-
& 30 =
= Q
= o
& =
40 2
W : <
w i
£ 50
i00
150

 

1.0 5 2 3 5 75 10 i 20 30 4c

BISMUTH FLCW RATE, gom

‘ Fig. 42. Variation of Rare-Earth Removal Time with Bismuth Flow Rate and Rare-Earth--
Thorium Separation Factor for a Process Cycle Time of 15 Days and a Total of 2L Stages.

2L
6.7 Rare-Earth Removal Times for Reference Conditions

Reference operating conditions, which have been selected on the
basis of the previous data, include a bismuth flow rate of 15 gpm, a
total of 24 stages (5 above the feed point), a 30-day processing cycle,
99% reduction of ThE) in the electrolytic cell, and lithjum and thorium
concentrations in the metal fed to the column of 0.0016 mole fraction.
The thorium concentration in the bismuth in the extraction columns is

about 90% of the thorium solubility at the operating temperature of 600°C.

The separation factors for several of the rare earths have been
determined for thorium-saturated bismuth in contact with T72-16-12 mole
Z
% LiF-BeF, 7

Pm, La, Sm, Nd, and Ce at the reference conditions are given in Table 9.

-ThFu. The separation factors” and the removal times for Eu,

The removal times range from 155 days for Eu to approximately 33 days

for Nd and Ce, and are considered adequate.

Table 9. Removal Times for Various Rare Earths at Reference Conditions”

 

Removal Time

 

Rare Earth Separation Factor (days)
Pm 1.7 oh
Nd 3.0 3% .8
Sm 2.0 h8.2
La 1. 60
Eu 1. 155
Ce 5.5 A2

 

pismuth flow rate, 15 gpm; processing cycle, 30 days; number of stages in
extraction column, 24 (5 above the feed point); fraction of ThF), reduced
in the electrolytic cell, 99%; lithium and thorium concentrations in the
metal fed to the column, 0.0016 mole fraction.
78

7. SEMICONTINUQUS ENGINEERING EXPERIMENTS ON REDUCTIVE EXTRACTION

H. D. Cochran, Jr. B. A. Hannaford
L. E. McNeese

Equi_pment+ has been installed for semicontinuous engineering exper-
iments on reductive extraction. The installation was treated to remove
oxides, and inspected to the maximum feasible degree. Before loading
the bismuth and salt into the treatment vessel, the salt and bismuth
feed tanks were stress-relieved, heated to normal operating temperature
(about 600°C), and pressurized to rated pressures to test their integ-
rity. The process system was heated and evacuated to remove air and
degas the graphite crucible. A final hydrogen treatment reduced most
of the rust on carbon-steel surfaces. Then, bismuth (184 kg) was charged
to the treatment vessel, where the molten bismuth was treated with hy-

drogen for oxide removal. Planned experiments are described in Sect.

T.5.

7.1 Pressure Tests of the Feed-and~-Catch Tanks

The mild-steel concentric tanks used for feeding and receiving the
bismuth and salt were stress-relieved at 650°C and 1 atm pressure.
Following stress relief, the inner and outer shells were proof-tested
twice at 50 psi and 25 psi, respectively, at 600°C. Subsequent dimen-
sional measurcments of the vessels were somewhat ambiguous. The ap-
parent changes were both positive and negative, and amounted to less
than 0.5% strain in each case. These changes were ascribed to the
effect of stress relief. Additional measurements were made on the more
accessible parts of the vessels to serve as a base line for future

measurements which will be made to detect and measure creep.
9

T.2 1Initial Cleanup of Experimental System

In order to reduce oxygen contamination of the process system to a
low level, the system was thoroughly flushed with argon and held at an
absolute pressure of 0.35 mm Hg at room temperature for a period of about
100 hr. This treatment was followed by a period of approximately 100 hr
at 600°C, during which the pressure rose briefly (probably due to out-

gassing of the graphite crucible) and then fell to 0.35 mm Hg again.

The system was treated with hydrogen in order te reduce the rust
that was present on the interior surfaces. The quantity of rust was
estimated to be approximately 400 g, based on an assumed 0.5-mil layer

of Fe_ 0, on the vessels and tubing. The flow path of hydrogen was
2

2
nominally from the treatment vessel, through the process system, and
into the off-gas header. A continuous sample drawn from the off-gas
header through a hygrometer provided a means for following the rate of
FeBO5 reduction. A 16-scfh flow of 30% hydrogen in argon produced a 39
water content in the off-gas, corresponding to the reduction of about 32
g of Fe, 0, per hour. Measurement of the moisture level during the entire

23

12-hr treatment was not possible because of the failure of the hygrometer.

T.3 Loading and Treatment of Bismuth

Following the hydrogen treatment of the process system, the temper-
ature of the treatment vessel was reduced to 350°C. Bismuth was intro-
duced into the graphite crucible in 12.5-kg batches, which provided a
volume calibration of the crucible. The total charge was 184 kg, of
which about 145 kg (15 liters) can be transferred to the feed tank.

The bismuth was sparged with hydrogen at the rate eof 10 scfh for approx-
imately 12 hr at a temperature of about 625°C. Because the hygrometer
was incoperative, the progress of the treatment was followed by visual
observation through a temporary 2-in.-diam sight glass. Hydrogen treat-

5

ment was interrupted {after 125 std ft” had been fed) by a steadily
increasing back pressure in the off-gas system, which was caused by

the collection of several grams of solids (probably carbon) in the small
off-gas filter. The Teflon cloth filter was replaced with a fritted
Inconel bayonet filter having an area of about 40 in.g. Further hydrogen
treatment will be followed by addition of thorium metal to the bismuth
and circulation of the solution through parts of the process system for

further oxide removal prior to charging salt to the system.

(.4 Operation of Gas Purification and Supply Systems

Operation of the Serfass* hydrogen purification unit was not com-
pletely satisfactory. The actual capacity of the purifier was less than
5% of its rated capacity of 15 scfh. After about 50 hr of operation,
the unit was replaced with a spare because of a flow obstruction in the

line used for removing impurities.

Operation of the argon purification system was generally good in
terms of water removal. The moisture level reached a steady value of
1 to 2 ppm at a usage rate of about 1.5 to 2 scfh. The instrument for

reading oxygen concentration is not yet in use.

7.5 Planned Experiments

Two types of experiment are planned with the facilities described
above. The first experiments will consist of hydrodynamic studies in
which the salt-phase pressure drop through the column will be measured
under various flow conditions. A range of flow rates from 0.05 to 0.5
liter/min is available for each phase. It is hoped that limiting flow
rates (flooding) and bismuth holdup may be inferred from the pressure
drop by analogy to experiments with a water-mercury system. Preliminary
calculations using a modified Ergun equation indicate that a pressure

drop as high as 3 to 6 ft of salt may be observed at maximum flow rates.

 

.X_
Serfass Hydrogen Purifier; product of Milton Roy Co., St. Petersburg, Fla.
31

After the first few hydrodynamic experiments, uranium tetrafluoride
will be added to the salt phase, and thorium metal will be added to the
bismuth, so that mass-transfer performance may be observed. It is
expected that the height equivalent of a theoretical stage will be 1 to
2 ft in this system. Since the packed column is 2 ft long, we expect to

observe the performance of the equivalent of perhaps one to three stages.

Stage~to-stage calculations were made for one, two, and three stages,
assuming various experimental conditions, in order to predict the masg-
transfer performance of the system and to choose the best experimental
conditicns for observing mass-transfer performance. Metal/salt volu-
metric flow ratios from 10 to 0.1 were studied with UFM concentrations
from 0.0001 to 0.003 mole fraction in the salt and thorium concentra-~

tions from 0.0002 to 0.002 mole fraction ino the bismuth.

The most obvious result of these calculations is that it will be
very difficult to distinguish between one-, two-, or three-stage perform-
ance. In fact, it will probably be possible to make this distinction
only under carefully selected conditions. The most suitable experimental
conditions are metal/salt flow ratios from 5 to 1 with 0.0005 to 0.001
mole fraction of UFM in the salt and similar concentrations of thorium
in the bismuth. Under the best conditions, there will be about a factox
of 2 difference in the UFM concentration in the effluent salt betwzen one
and two stages and less than 20% difference between two and three stages.
Concentrations of UF, in the effluent salt between 10 and 200 molar ppm
are expected under tfiese conditions. Further calculations to determine
the sensitivity of these results to small errors in the input variables

and constants are now under way.
82

8. ELECTROLYTIC CELL DEVELOPMENT

M. S. Lin .. E. McNeese

The proposed reductive extraction processges for protactinium iso-
lation and rare=~earth removal require the use of electrolytic cells for
reducing lithium and thorium fluorides into a bismuth cathode to prepare
the metal streams that are fed to the extraction columns and for oxidizing
extracted components from the metal streawms leaving the columns. To date,
results of experiments with static cel.lslL have indicated the need for an
electrically insulating material that can withstand the corrosive condi-
tions at the cell anode and the need for studying heat generation auad
removal in cells. The use of a layer of frozen salt, which is known to
be a good electrical insulator and will not be attacked by the salt or by
molten bismuth, has been proposed as a means of corrosion protection.

The heat-removal problem could be studied without the complications of
electrochemical reactions if an ac source were used. A comparison of cell
performance with ac and dc power will reveal any significant differences.

As an alternative to a bismuth anode, graphite could be used.

Experimental work is reported on (1) a comparison of cell resist-
ance with alternating and direct current, (2) the performance of a
graphite anode, and () protection of an all-metal cell from corrosion

by a frozen layer of salt.

6.1 Comparison of Cell Resistance with AC and DC Power

We performed an experiment that was essentially a duplication of
the static-cell experiments reported previously; however, ac rather than
dc power was used during most of the experiment. During the ac portion
of the run, both the voltage and the current were recorded; the results

obtained during this part of the experiment can be summarized as follows:
&3

1. The quartz electrode divider was attacked, but at a much lower
rate than in the corresponding dc portion of the run. There

was ne visible gas evolution.

2. Both of the electrode surfaces lost their metallic lugter as
goon as the power was turned on. At first, they turned a
brownish color; later, dark-colored material was observed on

the surface.

3. The calculated cell resistance was 0.16 ohm at 500°C for the ac
portion of the run, as compared with 0.15 ohm at 515°C for the

subsequent dc portion of the run in the same cell.

L. A residual dc voltage of 0.2 v, which indicated that some
rectification had occurred, was measured after the ac portion
of the run. This may have resulted from sparging of one of

the two electrodes.

5. When one of the two electrodes was sparged with argoen, currents
2 i 2
of 87.5 amp (2.6 amp/em™) and 100 amp (3.2 amp/cm” ) were required
to maintain cell temperatures of 660°C and T50°C, respectively.
The cell was operated at each temperature for about 2 hr. The

. . . 2
highest average current density achieved was 5.2 amp/cu’.

6. The amount of black suspended material was less than that in
the comparable dc portion of the run. The wmaterial disappeared
when the cell was maintained at 660°C or higher and reappeared
when the cell temperature was lowered to about 500°C; this
suggests that a change in the solubility of the black material

in the salt with temperature was being observed.

8.2 Fxperiments in a Quartz Static Cell with a Graphite Anode

An experiment was carried out in a simple, [lat-bottomed quartz tube,

L in. OD and 20 in. long, of the same design as the quartz static cell
8l

described previously. The cell was charged with 6.7 kg of bismuth and

1.5 kg of molten salt (66-%4 mole % LiFwBeFE). The bismuth pool, which
had been treated with hydrogen at T00°C overnight prior to charging the
cell with molten salt, served as the cathode. The anode was a l-~in.~diam
graphite rod with the lower end cut at an angle of 10° from the horizontal
to promote disengagement of gas that formed on the end of the rod. The
area of the cathode was 71 cmg, whereas the area of the anede was 5 cm
(considering the bottom of the rod only). The distance between the two

electrodes was about 1/4 in.

A Hasting mass flowmeter with a nominal full-scale flow rate of 1
liter/min was installed in the off-gas line downstream from the dibutyl
phthalate bubbler. A wet-test meter with a capacity of 0.05 ft5 per
revolution was connected to the mass flowmeter for monitoring the rate

of gas evolution.
The results of the experiment can be summarized as follows:

1. As soon as the cell was charged with salt, an amber color was
observed in the salt phase. The color darkened and finally
became black. The black material, which collected overnight
at the salt-metal interface, could easily be dispersed through-
out the salt by an argon sparge. The material settled quickly
when the sparge gas was turned off, which indicates that the

material had a relatively high density.

2. During dc operation, a constant voltage was applied to the
cell. The current surged to a maximum value and then decayed
rapidly, showing smaller intermittent peaks. 1In a typical case,
15 v was applied across the electrodes; the maximum current was
L5 amp (9 amp/cm2 at the anode). The current quickly decreased
to 22.95 amp after 5 sec and to 4.5 amp after 17 sec; it reached
a near-steady value between 0.2 and 0.4 amp, with intermittent

increases.
5. TImmediately after the power was turned off, the cell potential

wasg about 1.3 v; the potential then increased to 1.6 v in about

5 sec and to about 1.7 v in about 30 sec. The potential remained

steady for a while and then decreased very slowly. This behavior

f

and the maximum cell potential (1.7 vs 2.2 v) are quite different

from data obtained with a bismuth anode.

L. The gas that evolved from the anode produced only a momentary

surge on the flowmeter, on the order of 10 to 20 ce/min.

Toward

the end of the run, the off-gas was diverted to a starch-KI

solution bubbler. The color of the solution in the bubbler

turned faint yellow after a few minutes of contact.

5. During ac operation, the current varied linearly from 12

to

120 amp for applied voltages of 1 to 8 v; at the end of the run,

the cell resistance was 0.065 ohm as compared with about 68 ohms

in the case of dc operation. At higher voltages (8 to 20 v),

the cell temperature increased, resulting in a lower cell

resistance and hence a higher current (from 120 to 260 amp).

6. After the run was completed, the graphite electrode was removed

and inspected. No sign of attack was apparent.

&.3 Studies of Frozen-Wall Corrosion Protection in an All-Metal

Cell

The first attempt to use a layer of frozen salt as a means of pro-

tection for the anode was carried out in an all-metal cell. The main

body of the cell was .a 6-in. sched 40 mild-steel pipe that was 186

long and had a flat, 1/b-in.-thick bottom. The cell body was the

container. The upper part of the cell was flanged, and the upper

was electrically insulated from the cell body by a Teflon gasket.

in.
cathode
flange

The

anode of the cell, before and after assembly, is shown in Figs. 4% and

b respectively. Tt consisted of a double-walled, fluid-cooled cup having
Fig. L

7

).

Anode Cup Before Assembly.

PHOTO 94547

 

o3
PHOTO 94867

 

Fig. 44. Double-Walled, Fluid-Cooled Anode Cup Ready for Installation.
83

an inner diameter of 1.75 in. (a gross anode area of 15.5 cmg) and an
outer diameter of 2.75 in. The cup was suspended from the top flange
by two 1/L-in. tubes, which served as the coolant (N2 and water) intet
and outlet as well as the anode lead. Provision was made for raising
and lowering the anode cup. Viewing ports were provided on the top
flange, and provision was made for sampling the salt or bismuth without
introducing air into the system. A tube was installed to sparge the

center of the anode.

The cell was charged with 16.% kg of bismuth, which was subsequently
treated with hydrogen at T00°C for 16 hr. The cell was then charged with
molten salt (4.5 kg of 66-34 mole % LiF-BeFE) that had been purified by

hydrofluorination, hydrogen reduction, and filtration.

The first step was the formation of a layer of frozen salt over the
anode cup. The cup was then lowered into the bismuth pool to be filled
with molten bismuth. Finally, the anode was raised so that the bismuth
level in the cup would be about 1/4 in. above the bismuth level of the
cathode. (It was essential to keep the salt frozen in all of these and
subsequent operations.) After a few trials and some modification of the
furnace, we could maintain a frozen salt layer on the anode in the oper-
ating position. However, the thickness of the salt layer could not be
easily controlled. It was either too thick, which left no room for bis-

muth, or it was too thin and melted away when power was applied to the

cell.

During our attempts to form frozen salt layers, the salt temperature
was controlled at about 490°C, and the salt sparge tube was located just
above the cathode-salt interface. The anode was sparged by argon; the
lower heating section of the furnace, which surrounded the cathode, was
turned off; and, the cooling-water rate was adjusted go that the water

exit temperature was about 400°C.
89

Direct current was applied to the cell three times. 1In the first
period, about 2.6 v dc was applied to the cell and a back EMF of about
1.5 v was observed; this should be compared with 2.2 v in the quartz
cell. The cell resistance was 0.1 ohm. The c¢ell shorted scoon after
the test began. In the following two runs, the measured cell resistances
were 1.35 and 6 ohms, respectively. This variation in resistance was due
mainly to changes in the effective area of the anode (the cross section
of the inner cup, which is not covered by frozen salt). The back EMF was
about 1.5 v for the second runm, and only 0.2 v for the third run. The
high resistance and the low cell potential in the last run could mean
that the bismuth was fully covered by frozen salt and that only a small
unprotected mild-steel portion of the anode was exposed in this instance.
In the last run, cooling water was noted ian the cell after about 1 hr;
therefore, the run was terminated. Examination revealed that the cooling-
water exit tube, which was covered by a relatively thin layer of frozen

salt, had corroded (probably by anodic oxidation) and had ruptured.

9. MSRE DISTILILATION EXPERIMENT

J. R. Hightower, Jr. L. E. McNeese

The MSRE distillation experiment will consist of a demonstration of
the high~temperature, low-pressure distillation of molten salt as a means
for sepavating the lanthanide fission products from the ccmponents of the
MSRE fuel=~carrier salt, which is a mixture of lithium, beryllium, and
zirconium fluorides. Originally, the experiment was planned as a demon-
stration of the lanthanide removal step in the fuel-stream processing of
a two-fluid MSBR. Although the two-fluid MSBER concept has been deempha-
sized in favor of a one-fluid concept, the distillation of fluoride salts
still has potential application as a feed-adjustment step for reductive
extraction processes and as a means for partially recovering lithium and

beryllium fluorides from waste streams.
Q0

The experimental work is divided into two phases: (1)} nonradioactive
testing of the equipment using salt that has the MSRE fuel-carrier compo-
sition and contains NdF5, and (2) demonstration, at the MSRE, of the dis-

235

tillation of irradiated fuel-carrier salt from which the U has been

removed.

The first phase of the experiment has been successfully completed.
Equipment has been installed at the MSRE for the second phase of the

experiment.

9.1 Instrument Panel

The instrument panel, from which the process is operated, contains
all pressure, temperature, and level recorders and controllers, valve-
operator switches, electrical power-supply controls, and various temper-
ature and pressure alarms (Fig. 45). The instrument panel was installed
at the MSRE in the high bay area along the east wall, south of the spare

cell where the still was installed.

9.2 Main Process Vessels

The main process vessels5 — the feed tank, the still pot, the con-
denser, and the receiver are mounted in an angle-iron frame to facilitate
their transfer between Building 3541 and the MSRE, as well as installation
at the MSRE. Since the equipment is installed in a cell that is not much
larger than the equipment frame, all extra piping, thermocouples, heaters,
and insulation were installed on the equipment before its placement in the
cell. Fig. 46 shows the equipment after piping, thermocouples, and heaters
were installed, but before the thermal insulation was added. A stainless
steel pan was placed under the equipment to catch the molten salt in the
event that a vessel should rupture. The equipment is shown in Fig. 47

after installation in the cell.
ORNL PHOTO 89322A

  
   
    
    
  
    

   

ALARMS

PRESSURE
RECORDERS

e B

 

 

  

= -
TEMPERATURE
RECORDERS

 

 

TEMPERATURE ' i
CONTROLLERS L

 ;w | c;:caavaiy

 

 

VALVE doss
OPERATOR
SWITCHES ——

 

        
     
    
 

 

ST
} {?*lei'f.k L
STILL PRESSURE
MEASUREMENT

 
 

   

 

. {;

. . -:

STILL-POT
LIQUID- LEVEL
RECORDER
... CONTROLLER

;i:‘ . CJ;’.

   

 

   

STILL-POT
LIQUID-LEVEL
RECORDER

 

Fig. 45. Control Panel for MSRE Distillation Experiment.
VACUUM
LINE
FILTERS

FEED TANK
FILL LINE

 

CATCH
PAN
FOR

MOLTEN
SALT

e

 

Fig. L6. View of Still and Piping Before Insulation and Placement
in Cell.
ORNL PHOTO 94840A

FILTER IN RECEIVER
VACUUM LINE

   
 
  
  
  
 
 
  
  
 
 
  
  
   
   
 
  
 
     
  
 
 
 
 
 
   

FILTER IN
FEED TANK = ¥ » TO HCV-4 IN
VACUUM LINE (R T -~ VALVE BOX

SAMPLE LINE

SAMPLE LINE
EXPANSION
JOINT

 

SAMPLER
VACUUM

BLOWER AND F|L
FOR SAMPLER :
VENTILATIQN

2

ONDUITS FROM
UEL PROCESSING

DENSATE
EIVER

STILL POT
DRAIN LINE

TION BOXES FOR

JUNCTION BOXES L POT LIQ.

THERMOCOUP]
POWER LEAD:!

| j FILTER IN
L FEED TANK

| FILL LINE

B (LINE 112)

 
  

LINE 442 ¢ENETRATION L '
(FROM FUEL STORAGE ?AN%(} e |

Fig. 47. Equipment Installed in Spare Cell at MSRE.
ol

Some minor alterations were made on the basis of experience from
the nonradiocactive tests. The configuration of the feed line to the still
pot was changed to reduce the length of line that would be heated to
nearly 1000°C by the still-pot heaters and to provide a longer horizontal
section just ahead of the still pot. Both conditions should reduce the
probability of plugging the line with nickel and iron deposits. With
more of the line at a lower temperature, the rate of deposition may be
less; and, with the longer horizontal section, more material must be
deposited before the line can be obstructed (assuming that the material
will spread out along the complete horizontal section). In the non-
radioactive runs, a Calrod heater on an argon line failed and damaged the
tubing. 1In order to prevent a similar failure during the radiocactive
experiment, electrical insulation (glass tape) was installed between all

Calrods and the lines that they heat.

Absolute filteré*were installed in the vacuum lines from the feed
tank and from the receiver in order to prevent fission products such as
952r and 95Nb from entering the valve box. On testing the filters, we
found that they removed 99.997 % of 0.3-u DOP particles. These filters

can be seen in Figs. L6 and L47.

9.% Valve Box

The valve box (Fig. 48) contains all differential and absolute-
pressure transmitters, all valves that handle potentially radioactive
gaseous material, and two vacuum pumps — one to evacuate the reference
side of a differential-pressure transmitter and the other to evacuate

the distillation process vessels.

Before the valve box was installed at the MSRE, we added a bypass
valve to the differential-pressure transmitter associated with the liquid
level in the condensate receiver. The handle for this valve was extended

through the wall of the valve box to allow its operation when the valve

 

*
Flanders High Purity filters; product of Flanders Filters, Inc.,
Washington, N.C.
ORNL PHOTO 92845A

ABSOLUTE -PRESSURE § ;
TRANSMITTER ! ENCIESEURE

SECONDARY
CONTAINMENT B

ELECTRICAL
CONNECTIONS
FOR
VALVE -
POSITION
INDICATORS

 

(4

Fig. 4. Valve Box Before Being Sealed.
96

box was sealed. Figure 48 does not show the metal plates that were later
bolted to the front and the back of the box to complete the secondary con-
tainment. After the connections were made between the valve box and the
still, these covers were put in place and the box was sealed. With the
box under a pressure of 15 in. HéO, the leak rate was 0.1 scfh. During

operation, the pressure in the box did not exceed 0.5 in. H_.O; the leak

2
rate was negligible at this pressure.

9.4 Condensate Sampler

The condensate sampler (Fig. 49) is the most important piece of
equipment for obtaining information from the distillation experiment.

253

This sampler is patterned after the equipment that was used to add
to the fuel drain tanks and to take salt samples from the drain tanks.
Modifications were made to allow the sampler to be evacuated to approxi-
mately 0.5 mm Hg so that condensate samples can be withdrawn without

interrupting the run.

Figure 50 is a cutaway diagram of the sampler. The main components
of the sampler are: (1) the containment vessel in which the samples are
stored, (2) the turntable, which allows the sample capsules to be aligned
with the handling tool and also with the removal tool, (%) the capsule-
handling tool, with which empty capsules are attached to the cable in
order to be lowered into the sample reservoir, and (4) the reel assembly,

with which empty capsules are lowered and filled capsules are raised.

The following sequence was followed in withdrawing a condensate
sample. With HV-62 (the valve between the containment vessel and the
still) closed and the containment vessel at atmospheric pressure, the
sample-handling tool was raised to its highest position. The cable was
attached about 20 in. from the top of the tool so that, with the cable

reeled to the highest position, the top end of the tool protruded through
g1

ORNL-FHOTO- 94943 A

 

REEL
ASSEMBLY

MECHANISM FOR
ROTATING TURNTABLE

REMOVAL
TUBE

CONTAINMENT
VESSEL

WIRING FOR
INTERNAL
LIGHTING

-

Fig. 49. Condensate Sampler.
98

ORNL DWG 70-12477

 

  

 

 

 

 

 

 

 

 

) g:Eh
T~ i P
™ S
nMoo" | . Vedlny,
f 1 /\ru SELE"T
! Ny
1 ‘
‘/\"0 alt
of ] | 340
w0 ¢ H ! c:""sp Eg Fop
| | o
surl'l:.?‘al’z \
9ATuR
N
— FILTER | FILTER —
Q S A\VE 3 S AT VE Q
BLOWER HV-83 HV-64 VACUUM
¢ PUMP
X v
Ly
Wy,
82

Fig. 50. Cutaway Diagram of Condensate Sampler.
HV-66; with this arrangement, the samples on the turntable could pass
under the lower end of the tool. With the tool in its highest position,
an empty capsule was rotated underneath the tool's lower end. The tool
was then lowered onto the stem of the capsule and locked in place by an
adjustment at the end of the tool that protruded through HV-66. The tool
(with capsule attached) was raised again, the turntable was rotated until
the sampling notch was under the tool, and the tool was lowered below
valve HV-66. This valve was then closed. The vacuum pump was turned on
at this time, and the containment vessel was evacuated. When the pressure
in the containment vessel was 0.5 mm Hg, valve HV-62 was opened, aund the
sample-~handling tool and the empty capsule were lowered until the sample
capsule rested on the bottom of the sample reservoir at the end of the
condenser. The teool and the capsule were then raised above valve HV-62,
which was subsequently closed. The containment vessel was pressurized

to atmospheric pressure with argen. Valve HV-66 was opened, and the
sample~handling tool was raised to its highest position. The empty
sample holder was rotated underneath the sample-handling tool: the

sample was lowered into its holder and released from the tool. The
gample tool was again raised, another sample capsule was rotated under-

neath it, and the process was repeated.

The turntable had provision for 11 sample capsules. The samples
were stored in the containment vesggel until the end of the experiment,

at which time they were removed and sent for analysis.

A blower that induced a flow of air into the top of the line at
the reel assembly was provided to prevent contamination of the high bay
area when HV-66 was open. The air that was handled by the blower was

fiitered and exhausted into the cell.
2.

e

1CC

10. REFERENCES

L. E. McNeese and M. E. Whatley, Unit Operations Section Quarterly
Progress Report, January-March 1968, ORNL-L36lL.

 

J. S. Watson, QOak Ridge National Laboratory, personal communication.

M. W. Rosenthal, MSR Program Semiann. Progr. Rept. Feb. 28, 1969,
ORNL-4396 p. 28k.

L. E. McNeese, Engineering Development Studies for Molten-Salt
Breeder Reactor Processing. No. 1, ORNL-TM-3053 {November 1970).

W. L. Carter, R. B. Lindauer, and L. E. McNeese, Design of an
Engineering Scale Vacuum Distillation Experiment for Molten-Salt
Reactor Fuel, ORNL-TM-2213% (November 1968).
CoO—3] OV BN O

€S -
70

71-
12.

75 -
Th-

5.

101

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R. E. Blanco %29. J. Both

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P. N. Haubenreich 57. R. G. Wymer

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C. W. Kee 59-60. Central Research Library
R. B. Lindauer 61-62. Document Reference Section
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