ORNL-4371.txt

 

 

 

Contract No. W-T7405-eng-026

 

ORNI~4371

CHEMICAL TECHNOLOGY DIVISION

Pilot Plant Section

PREPARATION OF ENRICHING SALT 7LiF-ESBUF‘ FCR REFUELING THE

MOLTEN SALT REACTCR

Jchn M. Chandler

S. E. Bolt¥

MARCH 1969

 

*Reactor Division.

 

OAK RIDGE NATTONAL LABORATORY

Cak Ridge, Tennessee

operated by

UNICN CARBIDE CCRPORATION

for the

U. 5. ATOMIC ENERGY COMMISSION

LEGAL NOTICE

This repori was prepared as an account of Government gponsored work, Neither the United
States, por the Commission, nor any person acting cn behaif of the Commissicn:

A, Makes any warranty or representation, expressed or implied, with respect to the accu-
racy, completeness, or usefuiness of the information contained In thig report, or that the use
of any information, apparatus, method, or precess disclosed in this report may not infringe
privately owned rights; or

B. Assutnes any liabilities with respect to the use of, or for damages resulting from the
use of any information, apparstus, method, or process disclosed in this report,

As used in the above, ‘‘person acting on behaif of the Commigeion® includes any era~
ployee or conmtractor of the Commission, or employee of such contractor, to the extent that
such employee or contractor of the Commission, or employee of such contractor prepares,
disseminates, or provides access to, wny information pursuant o his employment er contract
with the Commigsion, or his employment with such contractor.
 

 

 
 

 

iii

CONTENTS

ADSTYECT & ¢ o « o ¢ o s & © 5 & o e s s & s & o

15

2.

71

Tntroducticn o o ¢ ¢ o o ¢ 6 o s & & o s e & & s

o2
233 in the MSRE

233

Considerations for Substituting
2.1 Nuclear Characteristics of the U System
2.2  Chenistry of the Circulating Fuel . « . .

2.3 Fuel Preparation by Remote Means . . « . -«

Process Development .« ¢ « ¢ & « o ¢ & ¢ 5 o o
3.1 High~-Temperature, One-Step Process . . . .

342 Low-Temperature, Two=-Step Process .« « « o

Description of Salt Production Facility .
.1 Descripticn of the TURF . . . « o & « &
ll‘elcl Cell G @ s - @ @ - * e 3 - - & & ®

4,1.2 Redioactive Hot Drain — Hot Off-Cas
(RED-HOG) v v v v ¢ o o ¢ o« v o o &

k,1.3 Cell Ventilation . « « « ¢ « o & &
h.1.4 Shielding o « ¢ ¢ o ¢ o ¢ o s
h,1.5 Manipulators . « « « o o « « o o .
L.2  Process Flowsheets . + « o « ¢ « « ¢ ¢ «
k.2.1 High-Temperature Process
h,2.2 Low-Temperature Process . .
4.3  Process Equipment . . . « « ¢ « ¢ « o
4.3.1 Decanning Station . . . . « . . .
4.,3.2 Reaction Vessel « v« ¢ o« & o o o « &
L.,3.3 Salt Storage and Transfer Vessel .
4.3.4 Shipping Containers . - « « « + &
4.3.5 Scrubbers . . « « o ¢ o ¢ o &
4.3.6 Miscellaneous Equipment . . .

h.h  Operating Procedures . o o ¢ o o ¢ ¢ o o o

Cold Run with Depleted Uranium . . . ¢ o « ¢ ¢ &
5.1  Preoperational Activities . . . « . . .

5.2 Mechanical Operations . . « « ¢ ¢ + « &

a ®

& &

System

@ s

E__l

W W o o

= W

"N

10
11
11
13

LA W
O ON
8.
9.

1C.

5.3 Chemical Operations « ¢ o ¢« ¢ ¢ o o ¢ o o & ¢ o a
7LiF-233UFu Process Operations ¢ o o« s o ¢ o ¢ o o & o =
6.1 Feed MaterialS « o « v o 4 o o o o o o & 8 o o o
.2 Radiation Levels of the Oxide Feed . . o « « « o &
£.3 Oxide Feed Material Handling o« « « « o o ¢ « o
6.4  Reduction of Uranium OXide « + ¢ o s ¢ ¢ o o« o o a
6.5 Hydrofluorination of U02 e e e « o e+ 5 & e s s o e
6.6 Formation of the Eutectic Salt ¢ « ¢ o « & « o o &
6.7 Purification of the Eutectic Salt . . . . . . . .
6.8 Transfer of the FButectic Salt .+ « ¢ o « « o o &
6.9 Container Disassembly and Preparation for Shipment .
6.10 Transfer of |LiF-“3°UF, Salt Product to the MSRE .
6.11 Material Balance . « o« « « o o o o o o o ¢ o s & @

iv

Maintenance Engineering .« « o s o o s o o ¢ o o« o o

7.1
7.2
7.3
7.k

Spare Parts .+ v ¢ o s v s o 6 & s s s e ¢ o o o o
Redundant Fittings ¢ ¢« o « ¢ ¢ o « o o o s o ¢ s o
Tools and Work Tables o . « ¢ ¢ o o ¢ o o o o o

Maintenance ProceduresS « ¢« o o « o o s o s o o ¢ o

COnCluS ions ° @ ® € e o *® ¢ o @ ¢ ® ° @ - e o & L ° o *

Acknowledgments . o o ¢ o o o« © o a o ¢ o s o ¢ o & o o

Referenc eS ® o o € [ ® & . ¢ ® & © * ° & @ e o & € e s
 

PREPARATION OF ENRICHING SALT 7LiF—ggBUFLL FOR REFUELING
THE MOLTEN SALT REACTOR

 

- John M. Chandler 8. E. Rolt

ABSTRACT

The Molten Salt Reactor has been refueled with an
enriching salt concentrate, 7LiF=233UFq (73-27 mole %).
Sixty-three kilograms of this concentrate was prepared in
cell G of the Thorium-Uranium Recycle Facility at ORNL.
Its preparation in a shielded cell was required because of
the high 232U content (222 ppm) of the 233y,

In the shazkedown run with depleted uranium oxide, a
high-temperature, single-step process was used to reduce
the oxide and then convert it to UFL for use in making the
eutectic salt. Although this process ylelded a high-
quality product, severe damage to the equipment was ob-
served. Therefore, 1t was discarded in favor of a low-
temperature, two-step process in which the uranium oxide
was reduced to U0, by treatment with hydrogen, the UOQ
was converted to UFj, by hydrofluorination, LiF was added,
and the eutectic was formed by fusing the components. The
eutectic mixture, LiF-UF) (73-27 mole %), was purified by
treatment with hydrcgen, which reduced the corrosion prod-
ucts to metal and subsequently allowed their removal by
filtration. The quality cf the precduct was well within
the requirements established for the MSRE.

The fuel concentrate, containing 39.0 kg of uranium
(91.4% 233U), was packaged in nine variable-capacilty
(0.5 to T kg of ursnium) shipping containers for addition
to the reactor fuel drasin tank and in 45 enrichment cap-
sules, each containing 96 g of uranium, for addition to
the bowl of the fuel circulating pump. The fuel was shipped
in shielded carriers to the MSRE to accommodate the reactor
enrichment schedule.

1. INTRODUCTION

In July 1966, an ad hoc committee was appointed to study the
235

feasibility of substituting 233U for the U being used to fuel the

Molten Salt Reactor. It was the recommendationl of this committee that

233

 

a charge of eutectic salt containing U be prepared for enriching
235
U

The MSRE prior to removal of the from the system.
o

A single-step process, in which the fuel concentrate could be

prepared directly from 233UO and LiFF, in a single reaction vessel

3

located in a hot cell, was considered to be the simplest and most

economical appreoach. Bilological shielding would be reguired because

232 233

of the high U content (222 ppm) of the U feed material.

The reduired quantity and quality of the enriching concentrate

will permit operation of the MSRE st full power for at least one year.

This report summarizes the preliminary phases of the wcrk —

development, design, and construction — as well as the actual operation
'
e 2
7L e 33UF

of the process used to prepare the 1

salt.

i (73-27 mole %) eutectic

233

2. CONSIDERATIONS FOR SUBSTITUTING U IN THE MSRE

2 35, . .

The substitution of 33U for E“SU in the MSRE had tc be considered
from three standpoints: (1) the nuclear characteristics of the 233U j
system, (2) the chemistry of the circulating fuel, and (3) the preparation

233y 232y

and handling of & salt containing

233

2.1 DNuclear Characteristics of the U System

An assessment2 of the operation of the Molten Salt Reactor with

233U has indicated that the following could be learned by substituting

the new fuel charge:

(1) Critical losdilngs would provide a check on the available

233

riuclear data on U for predicting critical conditions in a
reactor with a neufron energy spectrum similar to that of the
proposed molten salt breeder reactor (MSBR).

(2) Measurement of aguU—tom£33U atom ratios over a periocd of sub-
stantial burnup might further evaluate available nuclear data

and calculationsl methods.

(3) Interesting differences would bhe observed in the dynamics cf

the reactor because of the smaller fraction of delayed neutrons G
L

 

233

2
available from U fission. Stable operation with 23“U in
the MSRE would tend to confirm tentative conclusions regarding

the dynamics of the MSBR.

;fig (4) EKnowledge of the fission product yields for 233

U, as compared
with those for 235U, would provide a means of more pesitively
identifying the resctivity transients that follow changes in

power levels.

(5) Changes 1In other nuclear characteristics, such as temperature
coefficients, neutron lifetime, and control recd worth, would

be observed.

2.2 Chemistry of the Circulating Fuel

The uranium concentration in the reactecr will be reduced to 0.2
mole %; however, nc significant increase in UF_ concentration or precipi-

3

tation of ursniuvm or U02 1s expected. Most of the properties of the new
fuel charge will be very similar to those of the 23?U fuel. No change
1s expected in the compatibllity cf the fuel with the graphite and the

Hastelloy K.

2.3 Fuel Preparation by Remote Means

A strong incentive exists for develcoping an economical method for
reprocessing irradiated fuel for reuse in a power reactor. The high

232 233

U content of the U will make remote processing of a 233U fuel charge
mandateory. Therefore, it 1s desirable to prove the feasibility of a
simple, direct fuel recycle process without the necessity of high-level
decontamination; this would emphasize the advantages of the fuel

preparatlon for molten salt reactors.
3. PROCESS DEVELOPMENT

The preparation of the enriching salt, 7LiF°233UFa (73-27 mole %),
7

from "LiF and 233UO3 by a direct high-temperature, one-step process, was

 

2
investigated” in the laboratory and Tound to yleld & satisfactory product.
However, during the cold run in Building 7930, the high temperatures
necessary for this process were found to promote corrosicn of the equip-
ment; thus, an alternative method — a low-temperature, two~step process —

was adopted. These processes are discussed in detail in Sect. k.2,

3.1 High-Temperature, One-Step Process

The chemical procedures that were developed for the production of
the fuel concentratie were similar to those used in the routine prepsara-
tion of UFLL from cxides. Necessary modifications included: (1) the use
of a vertically mounted, cylindrical reaction vessel rather than trays
or fluidized-bed reactors, (2) the addition of LiF to the initial charge
of material, and (3) the operation of the process at temperatures

sufficient to maintain the LiF in its molten state.

Iaboratory-scale experiments were conducted by the Reactor Chemistry
Division tc gain irnformation about the rates at which the reactions would
occur and alsc to examine possible process control techniques. The

reactions investigated wvere:

Sintering: Helium atmosphere at 900°C

U0, — U0, ¢ + 0.2 O,
Reduction: Hydrogen at 900°C

U0, ., + C.6 H, — U0, + 0.6 £
Eydrofluorinaticn: Ho-HF sparge at 900-550°C

uo, + LHF — UF) + 2H,0
Reduction of impurities: Hydrogen sparge at T700°C

MF2 + H2 — MC + 2HF

In the reduction step the progress of the reaction could be followed
by observing the generaticn of water vapor in the system. The temperature

of the gas effluent increased markedly during this periocd.

During the hydrofluorination trestment, hydrogen, along with anhy-
drous HF, was admitted to the reaction vessel to control the corrosion
of the vessel and also to ensure the complete reduction of UO3 to UOE’

The conversion step cf the process was carried out at temperatures

sufficiently high to keep all of the LiF in solution. Thus, as UFM was
 

<

 

produced, the liquidus temperature of the fluoride components decreased
from 845 to LO0°C, the melting point of the eutectic mixture. In the
equipment used in the development runs, the Iiguidus temperature could
be measured and the UF& concentration could be determined by comparison
with the phase disgram. A final treatment of the melt with hydrogen at
T00°C reduced the concentration of nickel and iron to acceptably low

ievels.

Uponn successful completion of the laboratory-scale experiment, a
cold run, using equipment thet more clesely resembled the production
eguipment, was conducted by the Reactor Chemistry Division. This run
served to further confirm the feasibility of the one-step process and
also to train key personnel for the production operatilions at the

Thorium-Uranium Recycle Facility (TURF).

3.2 Low-Temperature, Two-Step Process

As an slternative to the high-temperature method, a two-step,
low-temperature process was evaluated in a labecratory experiment and
subsequently used in the salt production runs. In this process, which
is similar to that used to produce the original fuel salt for the MSRE,
the temperatures are generally much lower, and the LiF is not added
until the conversion of the oxide to the fluoride has been completed by

hydrofiucrination.
4. DESCRIPTION OF SALT PRODUCTICN FACILITY

The processing equipment was installed in cell G of the TURF, which
is located in the Melton Valley area of Czk Ridge National Iaboratory.
This faci_litylp contains shielded cells and process supporting systems

to permit remote fuel reprocessing.

4.1 Description of the TURF

The TURF, Building 7930, was constructed at Ozk Ridge National

Iaboratory to help develop and demonstrate economical remote methods
6

for reprocessing irradiated thorium-based fuel and for refabricating

the purified fertile and fissile material into fuel suitable for reuse
in a power reactor. The use of shielding and remote fabricaticn methods
will permit the use of simplified processes yielding only modest decon-

tamination factors.

The TURF has sufficient space to accommodate equipment for process-
ing and fabricating two types of fuel assemblies simultanecusly. The
facility is divided into four major areas: (1) an office area adjacent
to, but isolated from, areas that contain radioactivity, (2) an operating
areea with a development laboratory, chemical makeup area, and equipment
rooms for service equipment, (3) a maintenance operating area with ser-
vice areas for recelving and storing spent fuels, and (4) a cell complex

containing seven hot cells (six shielded and one unshielded).

Included in the facility are the services, ventilation systems,
crane and manipulator systems, viewing systems, and liguid and gaseous

waste disposal systems necessary to support fuel reprocessing.

L.1.1 Cell G

The 233U fuel charge for the MSRE was prepared in cell G. The

interior of this cell 1s 20 ft wide, 16 ft long, and 30 ft high. A false
floor was instslled to elevate the process equipment so that the through-
wall master-slave manipulators could bve used and maximum advantage could
be taken of the viewing capabilities of the shielding windows. This
decreased the effective height of the cell to 22 ft, which, in reality,
was Turther reduced to 14 ft of effective headroom because of the in-
cell electromechanical manipulator system space requirements at the top
of the cell. The walls, celling, and floor of the cell are lined with
stainless steel. Six cell operating modules (Fig. 1) are built into the
walls of the cell; four of these are equipped with viewing windows, and
two have window forms fthat are filled with removable shielding. There
are mcre than 100 penetrations into the cell for process services —

many more than are required for the salt preparation. The cell venti-

 

lation system normally maintains a pressure of =5 in. HEO and a Tlow

of 1033 ecfm of air through the cell; the wvessel off-gas system removes
 

ORNL DWG 68— 12283

1" SERVICE PIPE

 

 

 

8"‘0“ o
i
MODULE
fl ‘ SHIELDING

e ey e ot e e s e = PLATE ~ (-

 

 

i SECOND FL.
EL. 863°

LR B g A g @ s 6 ;” R

1 1

Mg c MANIPULATORS Talalt
Wak N % e MANIPULATOR o A
MODE L"A'+ h,  FMOREL™D" L SLEEVE " | hajlay
Y L - ! O : <+ ! =
N o A =

SMALLITEMS
ENTRY PORT

 

 

 

 

 

 

 

 

O

PERISCOPE

SLEEVE

 

ELECTRIC
SERVICE
SLEEVE

 

 

 

¢ WINDOW _ :

Mo 6" 8" SERVICE e Ay

| SLEEVE —\\,,;E-,_;_,;__:__- =

< A,

a DRAIN TROUGH — &~ = =~ —~ .,

, FIRST FL.EL.849%y L|° . <« . <
Fra

. . _
B Al Bl B g Ry 8 g i el Dl R ’ 4. .d'

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 1. Typical Cell Operating Module.
25 cfm of air from the cell and process vessels and routes it to a
system that is maintained at a pressure of =17 in. HgO° A small entry
port permits introduction of small tools and miscellaneous items through
a glove box and an alr lock. The rocf of the cell contzins a hatch with
a 10- by 6-ft opening that is sealed and shielded. The hatch provides
access to The cell with the 50-ton building crane. Eight master-slave
manipulators were installed in the cell: Tfour Central Resesrch Model A
and four Central Research Model D units. A Program and Remote Systems
Model 3000 electromechanical manipulator system is mounted on rails to

give complete coverage of the cell.

4.1.2 Radicactive Hot Drain — Hot Off-Gas System (RHD-KEOG)

The radicactive hot drain--hot off-gas system 1is a combination by-
product waste collection system and vessel off-gas system. It has inlet

connections in the hot cells and at verious points throughout the facility.

The network of stainless steel piping is designed to handle gas
and 1liguld in concurrent flow. Liquid waste 1s separated from the off-
gas and is collected in tank B-2-T in the belcw-grade and shielded waste
tank pit. The gaseous stream flows through a bank of absolute filters
and then into the cell exhaust system located upstream of the filter
units for that system. The liquid wastes may be pumped either into the
Melton Valley waste system or into cell G for recovery of vaiuahble

materials.

Two connections from cell G te the HOG system were made for the
salt preparation process. One connection was made to the capsule drill-
ing station in order to collect the particulate matter resulting from
the drilling operation. The other was made teo a multipurpose manifold
station. With the HOG system pressure controlled at —17 in. HEO’ the
flow of air from the cell was adjusted to 25 ¢fm by a manually cperated
valve. The gaseous effluent from the process scrubbing system was dis-
charged into this manifold station, thereby ensuring dilution, in the
case of hydrogen, to less than the explosive limit. In addition to the
scrubber dischsrge, the manifold station served the in-cell titration

station, the sampling station, the can-opening bcex, the scrubber system

 
\O

.......

and sirk liguid drsins, and all of the process vessels. Bome of these
connecticns were mede to the menifold station on the cell gide of the

manuval flow regulating valve, where the pressure was —5 in. H_ O; scme

2
were made on the HOG system side of the valve, where the pressure was
—15 in. H_.O. The location of each connection depended upon the pressure

requirenent for that particular use.

4.,1.3 Cell Ventilation

 

Approximately 1033 cfm of eir enters cell G from the cell G pump
room via a series of filters, a fire damper, a back-Tlow preventer, and
a cell pressure control valve. It discharges into the cell from dif-
fusers that are mounted on the cell ceiling and then flows through
roughing filters located at the false floor (2nd also at the cell floor
level) into the cell ventilation system marifcld in the north valve
pit. The flow is contrclled by a manually icaded valve in the exhaust
- duct. The exhaust is routed to the filter pit, where 1t passes through
two sets of filters in series; from there, it passes through a 30-in. duct
to the centrifugal blowers and into the 250-ft HFIR stack for dispersion
into the atmosphere. The controls on the ventilation system maintain
a pressure of -5 in. H O in the cell under normal conditiocns. Pressures

2
as low as —20 in. HEO or as high as =1 in. HEO can be maintained in the
cell under certain emergency conditions. Only a failure of all emer-
gency systems would allow the pressure in the cell to rise above —C.5

in. H20°

L.1.4 Shielding

The shielding of the TURF is designed in such a manner that, during
operaticn with radiocactive material having an intensifty of 105 r/hr,
the penetrating dose rates in normally occupled areas are no greater than
0.25 mrem/hr, with small hot spots no greater than 2.5 mrem/hr. Dose
rates greater than this are permitted in limited-access areas and for

short-term, non-routine operations.

 
10

To satisfy the allowable design radiation levels, the operating
cells have 5=l/2—ft—thick walls of normal concrete up to a height of
11 ft, 4-1/2 £t of normal concrete for the remaining portions of the
vertical walls, and 5-ft-thick concrete on the roof. This amount of
biological shielding was more than adequate to reduce exposure dose

rates to < 0.1 mr/hr.

The windows are essentially equivalent, in shielding thickness and
in their attenuation of penetrating radiation, to the concrete walls
in which they are installed. Each window consists of two major assem-
blies: the seal glass that is removable from inside the cell, and the
tank unit that is removable from the operating face of the cell. EFach
window is a composite unit consisting of 7 in. of glass and 58 in. of
zinc bromide solution; 1t is well sealed to minimize leakage of air

ground its periphery.

4.1.5 Manipulators

A Programmed and Remote Systems Model 3000 Manipulator System is
installed on a set of rails in cell G. The tube hoist on this manip-
ulator has a vertical travel cf 13-1/2 ft and a 1ifting capacity of 1000
1b. The trolley and bridge travel, along with the vertical travel of
the tube hoist, provides complete manipulator coverage of the cell down
toc the false flcor. This unit provides a1l the motions of the human
arm, plus wrist extensicn and ccntinuous rotation at the wrist and at
the shoulder. A grip force of 200 1b can be exerted with the fingers.
The hand is remotely removable and can be replaced by a hook fixture or

an impact wrench.

One Central Resezrch Model A master-slave manipulator and one
Model D master-slave manipulator are installed at each of the four
viewing windows 1In cell G. The Model A and the Model D manipulator have

maximun 1ift capabilities of 25 1b and 100 1b, respectively.

These master-slave units were installed to operate valves, to make

and break tubing disconnects and electrical and thermocouple disconnects,

and to conduct the hand operations required in the process. Conventional

hand tools were modified for use with these manipulators.
 

 

11

The PaR electromechanical manipulator was installed to perform the
heavy~duty maintenance, to convey heavy assemblies around the cell, to
reach some parts of the cell that were not accessible to the master-
slave manipulators, and to provide a "third" hand to simplify certain

operations.

.o Process Flowsheets

Two processes were developed for the preparation of the fuel
concentrate: (1) a high-temperature process, and (2) a low-temperature
process. The second process was adcpted after conclusion of the cold
run when it vecame evident that the origina’l process caused severe

corrosionr of the reaction vessel.

4.2.1 High-Temperature Process

 

The chemical flowsheet for the one-step, high-temperature process
as shown in block form (Fig. 2} includes the following major steps:

233

(1) Partial reduction of the uraniun oxide, Uté,by thermal

means .
(2) Further reduction to U’O2 by hydrogen treatment.

(3) Conversion of the oxide to UFu by hydrciluorination and con-
current dissolution of the UF& in the molten LiF to form the

evtectic mixture.

(L) Final purification of the eutectic mixture by trestment with
high-purity hydrogen.

'’ 233

UO3 in the initial cherge to the

reaction vessel. The reduction process was then conducted at tempera-

The 'LiF was combined with the
tures in excess of 845°C (LiF melting point) so that the oxide particles
were suspended in the molten fluoride. The moltern LiF served to keep
the oxide particles wet, thus reducing the possibility of entrainment

of particulate matter in the effluent gas stream. During the hydroc-
fluorination step, the progress of the reaction was estimated by

determining the liquidus temperature of the molten material, and then
 

 

LOADING

 

THERMAL
RECOMPOSITION

 

Load reaction vessel with
about 13 kg 233y (91.5%
enriched) as UO3 and
about 4 kg 7LiF.

 

REDUCTION TO
U0y

ORNL DWG, 67-11637A

 

 

 

Purge reaction vessel with
dry He at 1 liter per min
while heating to $00°C,

Partial reduction of UOg
to lower oxygen content.

 

 

 

 

 

 

 

Sparge with Ho~He

mixture through dip lire.

Total gas flows 2 to 10

liters per min, Melt temp:

?00°C,

Reduction of oxide to near
00% UQ9,

CONVERSION TO
UF4

 

 

 

 

Sparge with HF-Hg through
dip line. Hg at 10 liters per
min, HF at 1 to 2 {iters per
min, Periodically adjust melt
temp, to about 50°C above
tquidus but not below 700°C,

 

 

 

 

'

 

 

MELT ANALYSIS

 

 

REDUCTION OF
MF> IMPURITIES

 

When HFqyt equals HFjy
and liquidus tempy ap~
proaches 490°C, with=~
draw filtered sample of
melt and submit for chemi~
cal analysis,

 

 

HF STRIPPING

 

Increase melt temp, to
700°C and sparge with
H2 alone at 5 to 10 liters
per min until HF in gos
effluent becomes <0,01
milliequiv per Hter Ho.

   

 

 

 

 

Sparge melt with He at §
liters per min for about 2
hours fo remove residual
HF and Ho.

SALT TRANSFER

 

 

 

 

 

 

Note: If results of chemical analysis {Step 5) do not meet specifications, continue HF~Hy treatment (Step 4).

Fig. 2.

 

Chemical Flowsheet for Preparation of the

Heat transfer line and filter

fo 600°C, Transfer salt to
storage vessel, Cool all
vessels and {ines to room temp.

 

 

233U Fuel Concentrate for the MSRE.

A

ol
13

 

the furnace temperature controller was adjusted to approximately 50°C

above this value. Frequent titrations of the reaction vessel effluent
gas stream yielded valuable informaticon cn the progress of the conver-
sion of the oxide to the fluoride since the utilization of HF remained

guite high and nearly constant until the conversicn was completed.

Chemical analyses of filtered samples of the melt were made to
determine whether the final process — the purification of the product
by hydrogen sparging = should be initiated or whether the hydroflucrin-

ation process should be continued.

h.2.2 Low-Temperature Process

 

The low-temperature process (see Fig. 3) was used to produce the

-

fuel concentrate. It differs in only one major respect from that used

in the production of the original 235U;bearing enriching salt for the
MSRE. In the earlier process, the starting material was UFA; in the

low-temperature process, the starting material is UO The oxide is

digested at 550°C and then reduced to U62 by treatmeit with hydrogen

at temperatures ranging from 400 to 550°C. Helium is used as a diluent
gas during the reduction step until the major portion of the excthermic
reaction is completed. The oxide is then converted to UFLL by hydro-
fluorination at temperatures ranging from 400 to 630°C and at HF (in
hydrogen) concentrations varying from 5 to MO%H The final steps of the
process involve mixing and fusing the fluoride salts, contactling with
HF and hydrogen to remcve residual oxides and corrosion product im-
purities, and filtering the molten salt during its transfer to the

storage containers.

The reduction and conversion processes were monitored by a thermo-
couple array that was inserted into the powder in the reaction vessel
and by measurements of hydrogen utilization during the reduction step
and of HF utilization during the conversion step. Unfiltered and fil-
tered samples of the melt were withdrawn for oxide, petrographic, and

metal impurity analyses.

.........
CHARGE UOss

HEAT TREAT U@3§
HYDROGEN REDUCTION:
U03 — UO2

HY DROFLUORINATION:

U02 — UF‘4

EUTECTIC FORMATION:
UF, + LIF —UF - LiF
EUTECTIC PURIFICATION:
MO + HF —= MF + HyO
MF + Hy — MO + HF

PRODUCT PURITY:

UO3+H

14

CRNL DWG 68~-9515R1

OVERALL REACTION
ot 4HF — UF4 + 3H2O‘I

UF4 + LiF — UF4- LiF

27% — 73% EUTECTIC COMPOSITION

~13.2 kg U AS UO,
3 TO 5 hr DIGESTION AT 550°C; COOL TO 400°C,

START 5% Hp AT 400°C AND INCREASE TO 50% Hg
TEMPERATURE RISES TO 490°C; TREAT AT 500-550°C
AT 100% USAGE OF Hg; COOL TO 400°C.

START 5% HF IN Ho AT 400°C; INCREASE TO 40%

HF IN Hp; TEMPERATURE INCREASES TO 450°C;

WHEN HF USE DECREASES BELOW 80%, INCREASE

THE TEMPERATURE TO 630°C STEPWISE UNTIL HF

LUSE BECOMES O; COOL TO 150°C,

ADD EXACT QUANTITY OF ?LEF; MELT UNDER 30% H2;
DIGEST AT 850°C FOR 3 TO 5 hr; COOL TO 700°C,

PURGE MELT 24 TO 30 hr AT 700°C WITH 20% HF
IN Hg; TREAT WITH Hy FOR 75 TO 150 hr.

UNFILTERED SAMPLE ANALYZED FOR OXIDE CONTENT,
FILTERED SAMPLE ANALYZED FOR METALLIC IMPURITIES,

Fig., 3. Chemical Flowsheet for the Low-Temperature Process for
Preparing the MSRE Fuel Concentrate.
The temperatures encountered in this process are generally 200°C

lower than those measured in the single-step, high-temperature process.

The adoption of this process was felt to be Justifiable without the

benefit of a fuli-scale cold run because of the:
(1) success of the laboratory-scale experiment,

(2) similarity of this process to the original concentrate prod-
uction process,

(3) improvements made in the vprocess monitoring ilrstruments and
technigues,

(L) +the time schedule involved.

4.3 Process Equipment

The equipment flcwsheet shown in Fig. 4 is a simplified presentation

of the msjor compeonents redquired in the process. These components are:
(1) the fuel decanning statiomn,
(2) the reaction or oxide treatment vessel,
(3) the salt storage and transfer vessel,
(4) wvariocus containers for shipping the preduct,
(5) the off-gas scrubbers.

In addition, the process requires many other smaller pleces of
equipment, such as: the oxlde can preparation equipment, the in-cell
titration assembly, the furnaces for the vessels, the enrichment capsule
driliing and weighing station, disconnect stations for electrical and
instrument lines and process gas lines, and work tables and tool racks.

Figures 5 and 6 show some of this equipment after installation.

All services and reagent scurces are located in the penthouse

- outside the cell.

 
16

 

ORNL DWG &7-il638 R1

   
 

 

 

  
 

 
  
   
   

 

 

 

 

 

 

 

 

 

 

 

 

 

   
   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1-2 g/min, HF
GAASNifiMéP[‘gE e - ‘ == [
Y : ‘ B RESTHOYOR
FUEL
CAN
ENTRY
CRUTE
1
Acggaus
ADDITlONI chLROD
GAS
FILTER
CELL_"_G_" U036 kgs —iiF4 kgs [~ 1~ p——He PURGES ~1/2 slm
CAN vOG
AT CUTTING i
TOOL
CAN CHUCK
AND DRIVE
CAN CPENING A T
DUMPING BOX °
. GELL
VENT
. % z
‘2ig‘§fm GAS SAMPLE
ANALYSIS
FLOW + SALT
7 OVERFLOW
AR B | ARARRA, ALARM
¢ CALROD | CALROD
SALT
FILTER
‘ F-2 F-3
— FURNACEZ| | FURNACE
AQUEQUS KOH |
SCRUBBING SYSTEM ;
T-1 T2 T-3
REACTION SALT ENRICHING CAPSULE
VESSEL STORAGE AND SALT CAN FILLING
VESSEL VESSEL

 

 

 

 

 

 

RHD
SYSTEM

Fig. 4. Simplified Chemical and Engineering Flowsheet for
Preparing 233U Fuel Salt,

 
7

 

BHORNL PHOTO NO. 90483 A

CAN OPENER AND
DUMPING BOX

SALT TRANSFER
LINE™

SALT SAMPLING
BOX =

REACTION VESSEL
FURNACE

 

. Fig. 5. Main Reaction Vessel Furnace and Auxiliary Equipment for
2J3U Fuel Salt Preparation - Cell G, Building 7930.

 
 

 

CRUBBER OFF GAS TITRATION
o rTiE EQUIPMENT

_ REACTION VESSEL
CAPSULE AND SALT FURNACE

RNACE SALT STORAGE
SEL FURNACE

233

Fig. 6. Auxiliary Furnaces and Equipment for
Building 7930.

U Fuel Salt Preparstion - Cell G,

ORNL PHOTO NO. 90464 A

QT
 

 

19

4.3.1 Decanning Station

 

The decanning station (Fig. 7) is a rather complicated work box
that is required for opening the double~container cans, extracting the
oxide powder, feeding the powder to the reaction vessel, and disposing
of the spent cans. To accomplish this, the station treated each can in

the following manner:
(1) Received the full can and fed it into a rotating chuck.

(2} Tightened the chuck to grip the can while the grocving tool
formed a groove deep enough to fix the inner and outer cans

together. This was done twice for each can.

(3} Pushed the can into the box, displacing the previously opened

car.
(L) Ejected the previcusly opened can and 1id.

(5) OCripped the can, supported it, ancd rotated it while the

cutting tocl cut it open.

(6) Removed and retained the cut-off 1id while the can was

emptied of cxide by vibration and brushing.
(7) Inspected the inside of the can and 1id for remaining pcwder.

(8) Transferred the powder from the box into the reaction vessel

by vibration and use of in-box tools.

The decanning station was designed for alpha containment. All
penetraticns fto it were sealed by O-rings and boots. kach spent can

acted as a sesal until it was displaced by the next can.

A ball valve and charging hopper situated on top of the box was

used to introduce the LiF to the system.

4.3.2 Reaction Vessel

 

All of the chemical reactions of the process were conducted in
the reaction vessel (Fig. 8). The cold-run vessel wag a right cylinder,

7-5/16 in. ID and 36 in. high, fabricated from type 30kL stainless
 

PHOTO 91345

 

 

Oc
21

ORNL DWG-67-11674

Material: Vessel & Pipe 3041 SST

Liner, Diffuser & Dip Line:
Nickel 20!

 

/—-Sheil 8" Sched 80 Pipe

36 -3/4"

 

 

 

h——Liner 7 5/16" 1.0.x.36" Lg.x1/8" Thk.

Diffuser & Dip Line

 

 

Fig. 8. Reaction Vessel T-1.

 
22

steel; it contained a free-standing liner, 1/8 in. thick, constructed
from type 201 nickel, and was designed, on the basis of stress rupture
data, for a specified service 1life. The vessel contained nozzles to
permift powder addition, salt sampling, gaseous effluent discharge, and
product transfer. Upon completion of the cold run, the design of the
vessel was modified to include a truncated conical bottom te provide
better contact of the reagent gases. The production runs were conducted
in this vessel. Design information for this vessel and the other process

vessels 1s listed in Table 1.

%.3.3 Salt Storage and Transfer Vessel

 

The transfer vessel (Fig. S) was a right cylinder, 4 in, in diameter
and 36-1/2 in. tall, constructed of type 201 nickel. It contained five
nozzles, one c¢f which was a spare. The function of this vessel was to
receive a filtered, purified batch of the enriching salt from the re-
action vessel and to dispense it to the various product shipping con=-
tainers. It was necessafy for this vessel to store salt for 24 hr or

more during the changeout of some of the shipping containers.

L.3.4 Shipping Containers

 

The shipping contaliners were arranged into three arrays for the
Tilling operstions. Later, upon completion of the filling operation
and freezing cf the salt, each array was disassembled into individual

ccntainers for shipment to the MSRE.

The first array (Fig. 10) consisted of 45 enrichment capsules
(see Fig. 1i), each of which was 3/L4 in. in diameter and 6 in. long
and designed to contain 96 g of uranium. The capsules were connected
in series and arranged in three 15-capsule decks. An overflow pot
containing ligquid level detection elements and thermocouples was the
last vessel in the series. The tubing connections into, and out of,
each capsule were precisely positioned so that uniform filling and

subsequent blowback of the overfill were possible. Two holes were

 

drilled in each capsule to permit the salt to flow out when the capsule
 

23

I

1

 

 

Table 1. Equipment Design Infofmation: MSRE 233U Fuel Salt Prepaiation
Reaction Reaction Salt Furnace Furnace Furnace
Vessel Vessel Salt Storage Addition  Enrichment Liner Liner Liner
Item or Requirement T-1 Outer Inner Liner Vessel T-2 Can® Capsule F-1 F-2 F-3
Type of material 304L S8 201 Ni 201 Ni 30LL SS 200 Ni 300 SS5 300 88 304 SS EIC
Maximum operating temp., °C 900 900 700 600 600 900+ 900+ 900+
Maximum 0pera£ing pressure, 56# at 9OO°Cb Free standing 20 20 20 - Open Open Open
psig ' 20# at T00°C” 1liner in reaction |
vessel; no pres- g
sure difference |
developed ?
Maximum operating vacuum, 15 - 15 15 15 - - -
psig
Design temperature, °C 900 - - - - - - -
Design pressure, psig 6# at 900°C - 20 - - - - -
_ (Limiting
condition)
Calculated maximum stfess, | T27 psi at Maximum stress is 2560 péi at - - - No load carried No load carried
psig - 6 psig that due to heed 20# gage by bottom plate by bottom plate
: of salt in liner or wall or wall
and is negligible
Allowable maximum stress 1150 psi at - 3000 psi at - - 870 psi at - -
at maximum operating - 900°C 700° CP 900°C
temp.
Overall height, in. 37 35-7/8 36-1/2 Variable 6-3/8 37 36-7/8 36-7/8
Bottom plate thickness, 1/2 1/8 1/h 5/16 - 1/2 - 1/4 1/8 1/8
in. (reinforced
with welded
beam on diam)
Flat head thickness, in. 3/h - 1/ 1/4 - - - -
Outer diameter, in. 8-5/8 T7-9/16 h-1/2 2-1/2 3/h 10-3/k 4-7/8 4-7/8
Wall thickness, in. 1/2 1/8 15/64 0.065 0.035 0.359 1/8 1/8
Inner diameter, in. 7-5/8 7-5/16 b-1/32 2.37 0.680 10.032 4-5/8 4-5/8
Top flange OD, in. - - - - - 16 4-5/8 L-/58
Top flange thickness, in. . - - - - - 3/8 1/4 1/4
Hemispherical bottom wall ‘
thickness, in. - - - - 0.035 - - -
Hemispherical top, in. - - - - 3/8 solid | - - -

 

®Designed by MSRE.

blO,OOO hr creep-rupture data;

design based on

service << 10,000 hr.

 

 
 

24

ORNL DWG-67-11675 T

Material: Nozzlgs & Vessel: Nickel

Nozzles: 3/8" Sched 40 Pipe

 

36

172"

 

 

_— Vessel : 4" Sched 40 Pipe

 

 

 

Fig. 9. Transfer Vessel.

AN
b‘
25

 

PHOTO 91466

 

 

Fig. 10. Filling Array Consisting of 45 Capsules.

 
5 5/8"

ORNL DWG-67-11676 RI
v Material : Nickel 200
3/4

«— 3/16"x0.025" Wail Tubing

/8"

 

T

‘/———m 3/4" x 0.035" Wall Tubing

5.375"

 

 

 

 

 

Fig. 11. Typical Enrichment Capsule.
 

 

27

is lowered into the MSRE pump bowl. A cable-and-latch assembly was
sttached to each capsule for use iIn the sampler-enriched operation at

the reactor.

The second array (Fig. 12) consisted of four 2-1/2-in.-diam oy
34-in.~-long cans connected in series for filling operations. Each was
designed to contain 7 kg of uranium and had an instrumented overflow
pot. The cans were later cdisassembled, and the top and bottom plugs
were removed; then the cans were fitted with 1ifting bails and bottom
"stopper-type" plugs and weighed. They are being shipped, one at a
time, to the MSRE, where the bottom plug of each can is removed before
the can is lowered into the drain tank. In this tank, the salt melts

and runs out of the container.

The third array (Fig. 13) contained a group of six 2-1/2-in.-diam
variable-length cans (see Fig. 14) that were similar in design and
arrangement to those in the second array. One of these cans was used
to store excess product material that was blown back from the other
five cans after they had been filled to overflow. The latter cans were
designed to contain 0.5 te 3 kg of uranium (2 cans, 0.5 kg each; one

can, 1 kg; one can, 2 kg; and one can, 3 kg) .

4.3.5 Scrubbers

©

A caustic scrubber system was installed to neutralize the HF
effluent from the reaction vessel before it was discharged to the TURF
HOG system. The scrubbers, namely, four 13-gal polyethylene bottles
filled with 10% KCH, were connected in series and fitted with the
necessary fill-and-drain connections. The first bottle in the series
was kept empty, and the dip lines in the subsequent bottles were posi-
ticned at an exact depth. Thus the first bottle had an adegquate volume
to retain any liguid that might flow back toward the reaction vessel
(in the event that the pressure distribution in the system became
reversed). Also, a vacuum bresker was installed to prevent the occur-

rence of a vacuum, which could cause collapse of the bottles.
28

 

PHOTO 91470

 

Fig. 12. Second Filling Array: Four Cans, Each Containing 7 kg
of Uranium.

 
 

PHOTO 91471

 

 

 

Fig. 13. Third Filling Array Containing Six Shipping Containers
of Miscellaneous Sizes.

 
30

ORNL DOWG-67—11677 i

Material: 304L SST

 

 

 

 

 

 

 

 

 

Nozzie Salt
i Capacity
Lo DIMENSION &' Kg
ugs
’ m\\fl H 33.592" 7.0
' [4.802 3.0
| 10.102 2.0
5.402 1.0
3.052 0.5

  

 

 

 

 

 

 

 

 

2 2-1/2" x0 065" Wall
o -_fl/////m__-Tumng
Uy
£
bl
= il
O 5" 0D, |
Orain
.‘-‘-NOZZ|G

 

 

 

Fig. 14%. Salt Addition Can.

 
 

 

31

Raschig rings had been installed to prevent ccllapse of the
bottles in the scrubber system used in the cold run; however, these
rings prevented uniform mixing and were subsequently replaced by a

vecuum breaker.

4.3.6 Miscellaneous Equipment

 

Oxide Can Lengthening Equipment. — The 23jUO decanning station

 

3

was designed by using can drawings and sample cans from the Savannah
River Laboratory. After the decanner was fabricated, it was discovered
that the actual cans were 1-3/& in. shorter than denoted on the drawings.
(The measurements of the actual cans had not been checked closely be-
cause of the high radiation ievels of the oxide.) Since modification
of the decanner was impractical, we decided to lengthen each can 1-1/2
in. by cementing an extension on it. Three items of equipment, as
shown in Fig. 15, were required to accomplish this: (1) 2 machining
fixture to bevel the end of the can, (2} a press tc force the extensicn
onto the can, and (3} a fixture to cure the epoxy resin used in the

cemented Jjoint.

In-Cell Titration Assembly. — The off-gas titration assembly

 

consisted of a train of three 3-in.-diam by 6-in.-tall glass vessels,
a wet tegt meter for measuring gas flows, and the necessary tubing and

valves to permit reagents to be added and flushed from outside the cell.

Furnaces and Heaters. — Three high-temperature, electrical-

 

resistance-heated furnaces were installed in cell G for temperature
control of vessels used in the treatment, storage, and transfer of the
salt. A 24-kw, 12-in.-cavity Turnace was used for the reaction vessel;
a T-1/2-kw, 4-7/8-in.-cavity unit was used for the transfer vessel;
and a 10-kw, Q-?/Snin.—cavity unit was used for the shipping container
assemblies. Locally fabricated clamshell heaters using tubular
resistance-heating elements provided heat to the nozzles located cn
the tops of the vessels. The salt transfer lines were heated by at-
taching tubular heating elements to the tubing and then applying thermal
insulation. The off-gas line (and filter) was heated with electrical,
resistance-heating tape to prevent condensation of HF or water vapor

in the line.
 

Can Lengthening

Equipment.

PHOTO 91472

 

 

2t
 

 

33

Enrichment Capsule Drilling and Weighing Fixture. — Two holes were

 

drilled in each of 45 enrichment capsules to permit the salt to flow

out when the capsule is immersed in the salt in the reactor pump bowl.
Upon disassembly of the capsule filling array, each capsule was inserted
into the work box (Fig. 16); here, a bottom outlet heole and a top vent
nole were drilled in the capsule, the capsule was weighed, the lifting
bail was tested for integrity, =snd the capsule was inserted intc the
shipping carrousel. The transparent box served as containment during
the driiling operation. Off-gas from the capsule drilling box passed
through an absolute filter and was then discharged to the plant vessel

cff-gas system.

Disconnects. — The electrical and thermocouple disconnects used
in the cell were standard commercisl units. Generally, one-hzlf of the
disconnect was mounted rigidly to facilitate making and bresking cpera-

tions with the master-slave manipulators.

Several types of pipe and tubing disconnects were used. In genersl,
the gas service lines utilized ball-check quick disconnects. Compression
fittings were used in the salt transfer lines and In the main off-gas
lines. Conventional screwed pipe connections were used in areas where
remote operaticon of the Jjoint was not considered necessary. Locally
designed and fabricated connections were used in several large-diameter

Joints that would possibly reguire remote maintenance.

Work Tables and Tool Racks. — The prccess equipment was pecsitioned

 

in the cell to obtain maximum advantage of the capabilities of the eight .
available master-slave manipulators. Extensive use was made of support-
ing stands and mounting framework. Tocl racks for the required hsnd
tcols were leocated at the major work centers. Where possible, the vecid
spaces between the process equipment and the cell walls were filled with
trays te provide extra work areas, space for temporary tool storage, and
a surface on which to catch tools and small equipment items that werer
drcpped. A large work table (Fig. 17) was installed to accommodate the
product shipping container disassembly, the shipment preparation opera-

tions, and the capsule drilling and weighing fixture.
 

Fig. 16.

Capsule Drilling and Weighing

Station.

 

PHOTO 92787

 

 
 

-

Fig. 17.

In-Cell Work Table.

 

 

Y -84919
36
The tables, racks, and supports were fabricated from carbon steel;

the trays were fabricated from stainless steel.

L. L Operating Procedures

Operating procedures were developed for each of the 25 separate
operations required for the fuel production. (These procedures were
written before the cold run was made.) Brevity and conciseness were
stressed. Complete revision of most of them was necessary as & result
of the experience gained in the cold run. Review and revision, on a

233

day-to-day rasis, continued throughout the U production runs; the
minor equipment failures and malfunctions that occurred during these
runs led to extensive procedural changes to ensure continued production

of a high-quality fuel concentrate.

Tables 2 lists the runsheets that were prepared.
5. COLD RUN WITH DEPLETED URANIUM

The run with depleted uranium provided full-scale testing of the
chemical process, operational procedures, and of most of the equipment
in the absence of a penetrating radiation field. In this run, 14.6 kg
of uranium oxide (11.6 kg of 238U) pilus 3.4 kg of LiF were charged to
the system to produce 18.7 kg of 1’_,:3'.,16''U"E‘LL (73-27 mole %) eutectic.5

5.1 Preoperational Activities

The period from September 15, 1967, through January 15, 1968, was
devoted to facility testing; process eguipment fabrication, installation,
and testing; operator training; preparation of procedures and manuals;
and preparation of the process safety review, the criticality review,

and the radiochemical plant safety analysis.

5.2 Mechanical Operations

The run with depleted uranium began Januvary 15, 1968, with the

a38U0 -filled cans into cell G from the shielded carrier. e

loading of 3
 

 

37

2
Table 2. List of Runsheets Prepared for 2"‘O)U

Fuel Salt Prcduction

 

1he
15.
16.
17.
18,

25,
26,
27
28.
29,

Packaging of LiF.

Removing Empty Fuel Carrier from Building 793C Penthouse Pedestal.

Preparing and Transferring Empty Fuel Carrier from Building 7930
Penthouse to Building 3C19 Penthouse.

Preparing and Transferring of Loaded Fuel Carrier from Bullding
3019 Penthouse to Building 7930 Penthouse.

Checkout of Cell G.

Preparation of KOH Scrubber Systen.

Preparation of Inert and Reagent Gas System.

Pressure Test of Helium Header.

Pressure Test of H2 Manifold.

Pressure Test of HF Manifold (before any run).

Pressure Test of T-1 and T-2 Systems.

Preparation for Operation: Valve and Rotameter Settings.
Discharge oif Fuel Cans from Carrier to Cell G.

Capping Fuel Cans.

Decanning and Charging Fuel to T-1.

Charging LiF to Reacticn Vessel.

Transfer of 0dd-Lot LiF tc Cell G.

Eutectic Formation.

Hydrofluorination Reaction Progress.

HF Titration (Inlet).

HE Titration (Outlet).

oample Procedure.

Salt Transiter from Reaction Vessel T-1 to Transfer Vessel T=-2.

Salt Transfer from T-2 to T-3, Run No. 1 (3/L-in. Capsule
Assembly) .

Salt Transfer from T-2 to T-3 (Partial Filling of T-3).
Replacing Salt Vessel Assemblies (T-3).

Capsule Drilling and Weighing.

Weighing and Removal of Salt Shipping Vessels from Cell G.

Transfer of Salt Shipping Vessels to MSRE.

 
38

 

The UO3 had been processed at the Y-12 Plant, and the empty inner and
outer cans had been obtained from the Savannah River Laboratcry. The
cans had been filled with the oxide and sealed by a magnetic forming

technique to simulate the product cans that would later be used in the
production runs. The can opening box did not perform satisfactorily;

therefore, it was removed from the cell and extensively modified. The
alignment tolerances were relaxed, components were strengthened, addi-

tional equipment was installed to aid in transferring the powder from

the box, and the viewing capablilities were increased.

Difficulty was encountered in transferring the LiF from the de-
canning box to the reaction vessel. The avallable vibration was in-
adequate, and no provision had been made to sweep or "rod" the powder
out and down through the powder addition line. The vibrators had been
located on the equipment in such a way that most of their force was
imparted to the fixtures and mechanisms in the box. This resulted in
some damage to the equipment. Because of this, the vibrators were used

sparingly and with cautiocn.

5.3 Chemical Operations

Chemical operations began on January 2k, 1968, with the sintering
operation, which was conducted at 900°C with a helium flow of 1 std
1iter/min through the bed. The sintering treatment was immediately
followed by the hydrogen reduction step at the same temperature. A
2-liter/min ficw of gas, consisting initially of 0.1 liter of hydrogen
and 1.9 liters of helium per minute, was attsinea initially. The com-
pcsition of this gas was adjusted periodically until a flow of pure
hydrogen was obtained. Then the flow rate was increased, in increments,
to 10 1iters/min, where it was maintained for 5 hr. During this treat-
ment, the sintered-metal off-gas filter became plugged. ©Subsequent
examination of the filter unit revealed the presence of a hard filter
cake. Water wvapor, which is evolved during the reduction proccess,
coculd have condensed on the filter in the event that the filter heater

was 1mproperly operated; this would have permitted the filter temperature

to drop below the dew point of the effluent.
 

39

The conversicn of the oxide to the fluoride began January 27, 1968,
Anhydrous hydrogen fluoride mixed with hydrogen (10 meq of HF per liter
of hydrogen) was injected through a dip line to the bottom of the UO2
bed at the rate of 2.2 liters/min. This reaction began at 900°C, and
the temperature was pericdically decreasged to maintain the melt at
about 50°C above the liquidus temperature. The in-cell titration appa-
ratus was used to monitor the HF in the effluent, but the results were
inconclusive; thus another titration unit that permitted precise addi-
tion of the equilibrating reagents from outside the cell was designed
and installed. The data obtained with the new equipment showed that
the utilization of HF was essentially zerc (which indicested that the
conversion of the oxide was complete) after 295 hr of hydrofiuorination.
An attempt to sample the melt was unsuccessful because of the presence
of a hard top layer (or crust). No noise was evident from a listening
rod installed to detect bubbling from the dip lire. It was concluded
that the dip line had broken, by corrosicn or melting, above the liquid
interface. This would account for the lack of HF utilization and the
slow progress In the conversion of the oxide. Subsequent examination of
the dip tube showed a large hole approximately at the interface. A new
dip tube, zlong with a sensitive sound detector, was installed, and the
HF treatment was resumed. Bubbling in the melt was easily detected;
and, for the first time, reascnable values for the HF utilization were
obtained. After 250 hr of additional HF treatment, the dip line be-
came plugged at a location about 1m3/h in. from its open end. Upcn
exemination, the plug material appeared to be uranium oxide. Four sub-
sequent lines plugged approximately at the gas-liguid interface within
the tube; in these instances, the plug materizl also appeared to be
uranium coxide. It was obvicus, then, that excessive moisture was being
introduced into the melt by the gases used in the process. This problem
was sclved by installing a column filled with magnesium perchlorate in
the helium system and by installing a catalytic recombiner; a molecular-
sieve drying column (with regenerative capabilities), and a moisture

analyzer in the helium system.
e

The effluent gases from the reaction vessel were directed to a
caustic scrubbing system, which consisted of polyethylene bottles filled
with 20% KOH solution and Raschig rings (to prevent collapse). At an
early stage in the process, the dip lines in the scrubber botties
plugged, and a white precipitate appeared. Frequent sparging to promote
mixing, frequent recharging cof the caustic solutiocn, rinsing with hot
water, and a decrease (from 20% to 10%) in KOH concentration permitted
use of the system for the remainder of the cold run. It was felt that
the Raschig rings were a major contributor to this problem in that they
promoted stratification of the solution; this was confirmed by the
colored interface that appeared in the bottles when phenolphthalein
indicator was present. At the conclusion of the cold run, the scrubber
system was completely replaced by one that contained spare lines and a
vacuu breaker (to prevent possible bottle collapse), and permitted

remote replacement and flushing of all lines.

Although it gave satisfactory results in the small-scale development

runs, the technique of measuring the liguidus temperature of the melt

and subsequently determining, from the binary UFu—LiF phase diagram,

the progress of the conversion reaction did not prove successful in

this system. The heat capacities and thermal insulation of the systemn,
along with poorly located thermccouples, obscured the inflection in the
temperature decay curve of the melt upon cocldown. However, the HF
utilization data, alcng with material balances for the system, provided
sufficient reaction progress information. When the reactlon was Jjudged
to be complete, an unfiltered sample of the melt was withdrawn and ana-

lyzed, chemically and petrographically, for cxide content.

The purification cf the melt, by the reduction of the metal fluo-
rides to metal, was conducted without incident. Titration of the HF in
the effluent gas proved to be a reliable means of determining the end |
point of this reaction. Filtered and unfiltered samples were taken at
the conclusion of the purification step. The Tiltered samples were
considered to be representative of the packaged product since the melt
was filtered in the transfer operation. The cold run product analyzed

as follows:
 

b1

 

Concentration
(wt %)

U O Ni Fe Cr AT S

 

61.0 0.02 0.C065 C.007 C.26 0.1 0.002

 

The product wag filtered into the intermediate storage vessel;
then it was transferred to the capsules and shipping containers by
subjecting the salt to = helium overpressure of approximately 12 to
15 psig and forcing it out of the vessel through & dip line. This in-
volved several operations because the capsules were, first, filled and,
then, replaced in the system by the shipping can container assembly for
filling. The compression-type tubing fittings used for this operation
were very difficult to open after thermal cycling and after salt had
contacted them. (Also} it was difficult to make up these fittings
properly and to obtaein leak-tight joints, using the master-slave manip-
uwlators.) The approach toc a solution of these difficulties consisted

of:

(1) modification of the intermediste transfer vessel by additicn

of spare nozzles,
(2) reorientation of the vessel to improve accessibility,
(3) use of redundant fittings when necessary,
(H) use of a high-temperature thread lubricant on all fittings,

(5) modification of the operating procedure to reduce the number

of times fittings had to be opened, and

(6) training operators in use of the manipulators on a simulated
piping arrangement.
In summary, the cold run with depleted uranium was invaluable in
that it:

(1) 1led to a process that was more simple to operate and that

was compatible with the equipment,
Lo

(2) pointed out design and fabrication deficiencies in the

equipment,

(3) provided on-the-job training for the operators and supporting

personnel,

(4) tested the operational procedures, showing the need of exten-

sive changes in them, and

(5) showed that comprehensive remote maintenance procedures would
probably be mandatory to ensure successful completion of the

233

U fuel concentrate production runs.
6. (1a7-%3UF, PROCESS OPERATIONS

Three production batches, using the low-temperature flowsheet,
were required in order to prepare 63.4 kg of the fuel-enriching concen-
trate, TLiF—EngFh. This concentrate contained 3¢ kg of uranium (35.6
kg cf 233U), The first run began May 9, 1968, and the third run was
completed July 30, 1968. The 45 enrichment capsules were filled with
a portion of the first-run product; the four T7-kg shipping containers
were Tilled with the remainder of the first-run product, all cf the
second-run product, and a portion of the third-run product; the mis-
cellaneous shipping container assembly was filled with the remainder
of the third-run product. The ten salt shipping contsiners and the 45
enrichment capsules were delivered to the Molten Salt Reactor as required

in the reactor enrichment schedule.

6.1 TFeed Materials

The 233UO3 had been prepared in batches by using 7 M NHAOH (in

excess) to precipitate hydrous uranium oxide from solutions that con-
and NH, NC,.

3 43

The uranium in the feed solution had been purified and isolated in 1964

tained 10 to 40 g of uranium per liter and were 1 M in HNO

and 1965 by solvent extraction followed by ion exchange. These treat-

ments decreased the concentrations of plutonium, thorium, fission prod-

 

ucts, and corrosion products (iron, nickel, and chromium) to acceptably

low levels. Table 3 shows typical analyses of the oxide feed material.
 

 

 

 

 

Table 3. Compositionsa cf Teeds for Preparing
MSRE Fuel Selt
Run No. RU-33-1 RU=-33-2 RU-33-3
Oxide, g 15,929.4 15,901.7 16,376.7
Total U, g 13,082.4 13,105.8 13,533.0
233y, o 11,966.9 11, 969.7 12,377.8
Impurities, ppm
Chromium 50 &7 Ll
Tron 372 588 271
Nickel < 16 L7 < 13
232 £D5 218 ool
Isctopic Analysis of Uranium, at. %
233 A
33y 01.L47 91.35 91.46
)
23 'y 747 7.4 7.46
232y 0.809 0.948 0.816
236U 0.058 0.0701 0.059
238U 0.191 0.245 0.202

 

Data furnished by manufacturer.
Ly

The LiF feed (99.97% TLi) had previously been densified +to 1.2

g/cm3 by hydrogen treatment st elevated temperatures.

6.2 Radiation Levels of the Oxide Feed

The high radiation levels of the oxide feed material (see Table &)
result from the daughter activity of the 232y (222 ppm) in the 233y,
Alpha, beta, and gemma radiation is emitted in the transitions (see
Fig. 18). From the shielding standpoint, the 2.6-Mev gamma radiation
from the QOBTl and the 2.2-Mev gamma radiation from the 21251 are the

most important.

Table 4. Radiation Levels™ of Storage Cans
Containing Oxide Feed Material®

 

 

 

 

Time After Rediation Level {(r/hr)
Purification Distance from Source
(months) 0 1 ft T m 10 ft 20 ft
o 1
o6 10 1.1 0.120 0.028
36 250 25 0.120
Lo 300 25 0.200

 

"Measured with 2 paper shell cutle ple.

Prach can contained 450 g of 233U.

6.3 Oxide Feed Material Handling

The 233U-’O3 had been packaged in double-walled alurinum containers,
containing approximately 450 g of uranium. The cans, 2.8 in. 0D by 8 in.
long, had been sesled on one end by welding, and on the other end by a
magnetic forming technique. They had been stored in Building 3019 at
ORNL.

233

To prepare the U fuel concentrate, we removed 89 cans from the
facility and transported them, in six loads, to the TURF in a shielded
carrier. The carrier was positiconed on a pedestal atop celli G for

unloading the cans intoe the cell.
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

*
232%  (n,2n) 233
* Y232 |y RECYCLE FUEL Ui U
~80% FROM ngzz B13d
~ 0% FROM U a T4y
pg®>2 Gamma :
Energy
(n, ¥ Nuclide ' (Mev) |Yield
’ ! Pb2!2 | 0.250 |0.40 §
pg23! TH228 0720 0.19 §
0830 0.19 §
B25h .03 006 |
24 Bi2!2 | {34 005 |
Th™  6.i3h al9y 181 007
22 00358
v 0510 025 §
232 228 224 :
Th Ac Ra 1208 | 0582 080 |
0859 | 0.15 §
262 | 1.00 §
a 1.39x10'§ BTy a 3.64d
v ¥
RQZZB RnZZO Po2ié
B
64 %
@ 54.5s 3.04x1077s
4 ¥
p02|6 BEZIZ prOB
606 m (stable)
a
b 36%
a 0.158s " B3.Im
h
Y ¥
pb2!2 TIZOB
Fig. 18. Decay Chain and Gamma Activity in %06-233[}02 Fuels.
Le

No spread of contamination or excessive radiation exposure to
perscnnel occurred during the removal of the oxide cans from the storage
facilit during their transfer to TURF, or during their discharge from

s g 3 g

the carrier to cell G.

In cell G, the cans were gaged, and any burrs or excess weld metal
were removed by filing. Three separate operations were required to
elongate the cans to the 9—1/2—in. length required for proper operation
of the can opening box. First, the cans were machined on one end tc
"true” them. A premachined end cap (with epoxy cement applied) was
inserted into the cell through the small-items entry port. This cap
was installed on the can in a pressing fixture, and then the can (with
cap) was placed in a curing fixture, where five cans at a time were
cured under longitudinal pressure for 1 hr at a temperature of 110 to
120°C. The can preparation equipment functioned satisfactorily, except
on one occasion in which the drive motor on the can machining rixture
burned out. The remaining cans for that run were dressed with a file,
and the caps were installed without Incident. ILater, a new motor was

installed remotely in the fixture.

The elongated cans of oxide were opened, one at a time, in the
decanning staticn. This was a time-consuming operation involving the
use of equipment that had been recognized during the cold run to be
marginal. The decanner had a rugged chuck assembly that held and ro-
tated the cans while a cutting wheel opened tThem inside an alpha-sealed
box. Most of the cother attachments in the box were of little value,

and in some instances, actually hindered the operation.

The major problems in the decanning operation can be sumarized

as follows:

(l) Alignment and clearance tolerances were tco close for hot-cell

operation.
(2) Visibility in the decanning box was limited.

(3) The powder failed tc flow ocut of the box easily.

 
 

 

W7

(k) A vibrator shaft seal failed, leading to loss of oxide powder
from the box. This caused rather extensive contaminaticn of
the cell. Modifications were made remotely during the fuel
concentrate production to increase visibility, repair the seal,
and to provide tools inside the decanner to assist in trans-

ferring the powder to the reaction vessel.

6.4 Reduction of Uranium Oxide

!‘)
The expanded bed of ;33UO3 was heated for 2 hr at 550°C in a
helium atmosphere to remove, by pyrolysis, any traces of ammonium com-

pounds or cother volatiles remaining from the chemical processing.

The bed was then cooled to U00°C before hydrogen treatment was
started. This temperature was sufficiently low to accommodate the
temperature rise expected from the exothermic reaction of hydrogen with
U03:

U0, + By = U0, + H,0 + 72,000 cal/g-mole.
The concentration of hydrogen (in helium) was adjusted initially tc
5 vol % and was gradually increased to 50 vol % during the first 4 hr
of treatment. The temperatures within the bed were monitored by an
internal probe with 12 thermocouples that were placed at 2-in. intervals
along the vertical axis of the bed. These temperatures rose in response
to an increase in hydrogen concentration and then became constant after
the initial excursion. The procedure of incrementally increasing the

hydrogen concentration was repeated until the hydrogen:helium volume

ratio was 1:1; the gaseous flow rate was 2 liters/min.

The location of the reaction zone and the zone movement 1nside
the bed were clearly defined by the temperature profile. As the
reaction progressed, the reaction zone rose, in the form of a band,

up through the powder bed.

Figures 19 and 20 show plots of the temperature at 6 in. and
16 in., respectively, from the bottom of the 2Lk-in.-high bed during the

uranium recuction phase cf the three production runs. The rise in the
   

880
560
S 540
g
B 520
-
e
&
& 500
&
=
[P
= 480
Bad
=
2 460
®
5
Z 440
|
420
400
380

Fig. 19.

ORNL DWG 68-9477

 

 

 

 

 

 

 

 

"RUYK no. | He = 0.4 sim purge
RUN po. 2 ———— ) Ha = 1.0 sim
RUN no. 3 mn—rm- S e He = 1.0 sim
\
\!
\
\l
\ \‘
\ \
\ \
4
\ \
\ %
Vo
\ \
\ \
\ \
I e e e
4 8 i2 6 20 24 28 32 36 40 44 48 52 56 60
TIME (hrs)

Temperature 6 in. from Bottom of Oxide Powder Bed During Uranium Reduction.

8N
 

580

560
~ 540
[ 8]

L
& 520
>
}—
=
£ 500
O.
F
i
= 480
[FY]
=
9 460
x
[T1]
Q. 440
Q.
2

420

400

380

Fig. 20.

ORNL DWG 68-3476

 

  
 
 

He = 0.4 sim purge
Hz"‘ 3.0 SIm
He = 1.0 sim

 

 

 

 

 

8 12 ie 20 24 28 22 36
TIME (hrs}

40 44 48 52 56

60

Temperature 16 in. from Bottom of Oxide Powder Bed During Uranium Reduction.

 

ot
50

temperature of the lower zone during the first 8 hr of treatment is due
{to the exothermic reaction (the reaction vessel furnace was set to con-
trol at 400°C). After the temperature excursion resulting from the
initial increases in hydrogen flow had subsided (approximately 12 hr),
the temperature of the furnace was increased, at the rate of 30°C/hr,
until the bed temperature was 525 £ 25°C. The reduction operation was
continued at this temperature, with 50 vol % hydrogen=--50 vol % helium,
until 50 to 100% excess over the stoichiometric amount of hydrogen had

been passed through the bed.

Hydrogen utilizetion within the bed was 100% until the reaction
zone approached the top of the powder bed; then a slight decrease was
observed. Hydrcgen usage was determined by a material balance of the
gas Tlowing into the reaction vessel, as measured by rotameters, and

the gas outflow, as measured by the in-cell wet test meter.

6.5 Hydroflucrination of o,

Upon completion of the oxide reducticn step, the bed was allowed
to cool to LO0°C. The coaversion of Uo, to UFH by hydrcfluorination,
using HF gas diluted with hydrogen, began at LOO°C and was completed

at 625°C. TFive to seven days were required for the conversion.

The EF gas wes supplied to the process by withdrawing it from the
vapor space of a heated 100-1b HF cylinder. The cylinder heater was
thermostatically controlled to provide a vapor pressure to 12 to 14 psig.
A differential-pressure Transmitter across a capillary restrictor in
the HF gas supply line was used to monitor the flow. The gas was passed
through =& maze of tightly packed nickel wire in a 2-in.-diam nickel tube
that was maintained at 625°C to remove sulfur from the stream. It was
then mixed with 2 metered amount of dry hydrogen, filtered, and intro-
duced to the reaction vessel through a dip tube that extended to the

bottom of the U02 bed.

At the beginning of the hydrofluorination step, the composition of

the gas was 95 vol % hydrogen--5 vol % HF; the flow rate of the mixture

was 2 std liters/min. Over a period of 3 to 4 hr, the HF concentration

33
 

 

was incrementally increased to 40 vol % as the exothermic reaction
caused the bed temperature to rise from 400 to 450°C. During these
initial hours, the temperature within the bed was constantly monitored
to determine when the temperature excursion resulting from each HF flow

adjustment had ceased and when another adjustment could be made.

ATter the HF concentration of the hydrofluorinsting mixture had
reached 40 vol %, the reasction zone traveled, in the Torm of a band,
up through the bed (in a manner similar to that observed during the

hydrogen reduction) as the U0, was converted to UFH' The reaction
uo, -+ LHF — UF) + 2E,0 + 144,000 cal/g-mole

is more exothermlic than the reducticn reactlon, but it does not have as
great a tendency to cause thermal excursions. Probably, this is the

result of differences between UO UOE’ and UFLL with respect to bed

3)
permeabllity and thermal properties. The reaction-zone temperatures
for the three production runs are plotted as a function of time in

Fig. 21. In runs 1 and 2, several furnace temperature adjustments were
made during the first five days of treatment in an eiffort tc increase
the bed temperature to L75-500°C (vecause the HF gas is less corrosive
to nickel at the higher temperature). In run 3, the furnace operatec

satisfactorily and did not reguire adjustment.

The vrogress of the reaction was followed by observing the reaction
zone travel through the bed and also by cobtaining a material balsnce of
the HF in the system. The HF utilization was essentially 100% for the
first five days and then decreased sharply as the reactlion zone reached
the top of the bed. Then the temperature of the bed was increased to
625°C, where it was held for two days to ensure completeness of the
reaction. The HF utilization did not increase at the higher temperature;
instead, it continued to decrease, suggesting that the rezction had been

complete at the end of the fifth day of hydrofluorination.

A total of 13.5 kg of wranium, as U0, was converted to UFM in
o ‘
each of the three runs; only very minor differences in the runs were

noted.
5e

ORNL DWG 68-9478

 

680

660 |-

640
. He = 0.4 sim purge

g20 | He = 1.2 sim -] \}{
|/

B

!

|

|

!

|

|

|

L HF
600 -

40% (C.8sim} in Hy

580 -

560 —

540 —

520

500 —

480 |-

460 —

REACTION ZONE TEMPERATURE (°C)

440

42Q

 

400 -

 

380 | | 1 ! 1 | |

0 24 48 72 96 120 144 ieg 192
TIME (hrs)

 

 

Fig. 21. Maximum Temperature Within Powder Bed During
Hydrofluorination of UOQ..

c
53

 

£.6 TFormation of the Fubectic Salt

The eutectic mixture UF)-LiF (27-73 mole %) (Fig. 22) was formed
by adding the stoichoimetric quantity of 1ithium fluoride powder to the

uranium tetrafluoride powder and fusing the mixture.

Dry LiF powder was added to the reaction vessel by the fellowing
series of operations: (1) it was poured into a hopper located on top
of the decanning station, (2) it was dumped from the hopper intec the
decanner box, and (3) it wae transferred from the box to the vessel by

vibration and brushing.

The temperature of the reaction vessel containing the stratified
UFLL and LiF powders was then increased to 855°C in order to melt the
1ithiun fluoride (mp, 835°C). The melt was digested at this tempera-
ture for 3 hr while it was sparged with hydrogen (at a flow rate of 0.2
1iter/min) to reduce any extraneous compounds that might have been

introduced during the LiF addition.

A sound detector was atliached to the reaction vessel dip line to
sense the bubbling of the hydrogen sparge that would occur in the
presence of a liguid in the bottom of the reaction vessel. This was
an excellent device for determining when the meltdown had started. The
internal thermccouple probe confirmed that, at meltdown, a very wide
range ol temperatures existed along the axis of the melt. A rapid
cooldown to 65C°C was observed in the lower region of the vessel where
ligquid existed. At the same time, temperatures as high as 850°C were
measured in the upper regions of the vessel where fusing of the com-
pounds had not yet taken place. As more liguid was formed, a gradusl
shift in the temperature profile occurred, indicating a rising liquid
level in the vessel. The temperature of the pool eventually reached
850°C, the set point of the furnace temperature controller. The initial
presence of the ligquid and the rising level were also indicated by
changes in the dip line pressure as The back pressure on this line

increased to overcome the rising liquid level.

Differences, with regard to conditions during initial meltdown,

were noted in the runs. In runs 1 and 3, the UFA and LiF had to be
5k

ORNL-LR-DWG 4745TA

 

$40C

//

900 P

7

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o
c.
& 80O \
> \ /
<
o ! /
& 700 E
! ;
- z
t
600 N
<&
W <
500 S W
v W
4LiF -UF, | - -
400 | f
LiF i0 20 30 40 50 60 70 80 90 UF,

UF, (mole %)

Fig. 22. LiF-Tfl‘)_P Phase Diagram.
 

 

52

heated to about 850°C before melting occurred. In run 2, initial

meliting occurred at 650°C, nearly 200°C below the melting point of the
lithium fluoride. The low-temperature initial melting must have resulted
from the presence of a sizable heel of salt (mp} hg0°C) that remained

in the reaction vessel from run 1. Probably, this heel acted as =a

"seed" to permit fusing of the UFlL and LiF powders at the lower tempera-

ture.

The G=-in.-deep pool of eutectic salt (mp, LO0°C) was next treated
with 20 vol % HF--80 vol % hydrogen (flow rate, 2.4 liters/min) for 24
hr at a temperature of 700°C to remove the last traces of oxide from
the salt prior to the hydrogen purification procedure. At the conciu-
sion of this ftreatment, an unfiltered sample of the salt was withdrawn
=nd analyzed for oxide content. (A 1/2-in.-0D x 2-1/L-in.-long nickel
cup was immersed in the salt to withdraw a 25-g sample.) After the
sample had solidified, a l-in.-long section was cut from the center and
analyzed petrographically and chemically for the presence of er' The
remaining portion of the sample was submitted for = complete chemical

analysis.

Hydroger. treatment of the salt was started prior to obtaining the
results of the analyses. The more rapldly obtained petrographic results
were used to determine whether hydroflucrination should be resumed or
whether hydrogen purification of the melt should be continued. In each

of the three runs, the UC. content was reported to be less than the

2
lower limit of accuracy (200 ppm) for the petrographic appraisal; thus,
subsequent HF treatment was unnecessary. Chemical analyses of the same
samples showed oxide contents (in the product sslts) of 62, 34, and 32

ppm for runs 1, 2, and 3 respectively.

£.7 Purification of the Eutectic Salt

The eutectic salt was purified by bubbling pure hydrogen gas
(3 to 10 ppm HOO), at & flow rate of 2 std liters/min, through the
10=in.-deep meit, The temperature of the molten salt during this
reaction,

G
MF + 1/2 H, — HF + M,
56

is 700°C. The progress of the reaction was followed by titrating the
effluent gas with the in-cell titration assembly. These data are
plotted in Fig. 23. The high rates observed initially can be attributed
to the evoiution of the soluble HF in the melt. The first inflection
and plateau of each curve correspond to the reduction of the nickel
fluoride; this is followed by a second inflection and plateau, which
represent the reduction of the iron fluoride. The end point of the
purification was evident from the leveling off of the HF evolution rate
at 0.025 meq per liter of hydrogen. The hydrogen flow rate was in-
creased on several occasions during the process, and the reaction rate

seemed to be almost independent of the hydrogen concentration.

The effect ¢f the hydrogen treatment on the concentration of the
corrosion products is shown in Table 5. The chromium concentration in
the melt was not affected by the hydrogen treatment. Higher initisl
values of nickel in the third run, and of iron in the second and third
runs, were the result of residual material left in the bottom of the
reaction vessel from previous runs. Apparently, the metals had pre-
cipitated and thus had not been filtered cut during the transfer

operation.

The levels to wnich the iron and nickel concentrations were lowered
by the hydrogen reduction of the metal fluorides are believed to be
neay the 1imit of accuracy of the sampling system and the laboratory
analyses. The actual concentrations of these Impurities in the en-
riching concentrate, as delivered to the MSRE, should be somewhat lower
than the values reported because an additional 24 hr of hydrogen treat-
ment was provided in each run after the filtered samples were withdrawn
for analysis. Table 6 gives the chemical analyses of the fuel shipments

to the MSRE; Table T gives the isotepic analysis of the uranium.

During each of the three runs, the reactiocn vessel was exposed for
2C days to 40 vol % HF--60 vol % hydrogen at temperatures ranging from
400 to 850°C. Approximately 5 g of nickel (from the nickel liner of
the reaction vessel) was lost to each melt. This corresponds to =
uniform corrcosion rate of slightly less than 0.001 in./year -~ g 1low

rate for this type of process.
HF OUTLET (meq HF/I Hp)

0.20

018

o.oe

0.06

0.04

.02

0.00

ORNL DWG 68-948

 

 

 

-. Ho = 2.0 sim
\ \ He = 0.4 sim
\; | ‘\ TEMPERATURE = 700°C
A .‘
o RUN no. |
‘ \\" RUN no, 2 —— — —
L ] \
N RUN no. 3 ——-—-—

 

 

24

48

Fig. 23.

72 96 120 144 168 192

TIME (hrs}

Purification of the Eutectic Mixture.

2is

 

LS
Table 5. Concentrations of Corrosion Products in the Eutectic Salt

 

 

 

 

 

 

 

Nickel
(%)
‘Type of Run
Sample 1 2 3

 

Unfiltered:
Before H? Treatment 0.0191 0.0038 0.430

Filtered:

After H_  Treatment 0.0066 0.0050 0.0160

2

0.0295 0.0750

Chromium
(%)
Run
1 2 3
M
o

 

 
 

3

Table 6. Analyses™ of the 2 jUF)_L-?LiF (27-63 mole %) Fuel Concentrate

 

 

Cans B, C, D, and B

 

Capsules (45)

 

Cans ¥, H, J, L, N, P

 

 

Chemical Spectrographic Chemical Spectrographic Chemical Spectrographic

Element Analysis Analysis Analysis Analysis Analysis Annlysis

Al

B < .01 < 0,01

ca < 0.004 < .00k

Ce < 0.033 < 0.021 < 0.038

Cr 0.0057 < 0.002 $.0055 < 0.002 0.0045 < 0.01

Cu

Fe 0.0104 0.047 0.0051 < 0.004 0.01k45 < 0.09

Gd < 0.003 < 0,002 < 0,00k

Li 4.83 4.91 L.8

Mn < 0.0004 < 0.C00k

Mo < 0.004 < 0.C04

Ni 0.0082 < 0.04 0.,0066 < 0.002 0.01E0 < 0.09

0, b ppm 62 ppm 32 ppm

Pb

Pr < 0.030 < £.010 < ¢.038

8i

s < 20 ppm < 20 ppm < 20 ppm

Sm < 0.007 < 0,004 < 0.008

Sn

Ti < 0.002 < 0.002

. i

Zn < 0.002 < 0.002

U 61.38 61..8 61.19

 

a . .
In wt % unless given in ppm.

65
60

Table 7. Isotopic Analysis of Uranium in MSRE
Fuel Concentrate

 

 

Capsules (U45) Cans Cans
B, C, D, E F, H, J,

L, N, P

Isctope

233y 91.465 01.378 91.466

23%; 7.185 7.463 7.466

235y 0.803 0.87h 0.810
236 0.058 0.065 0.0585

23'8U 0.186 0.218 0.196
Total uranium, kg 4.300 28,059 6.583
Average at. wt 233.100 233.150 £33.090
Total uranium, wt % 61.800 61.380 61.190

 

6.8 Transfer of the Eutectic Salt

Eight transfer operations were necessary to convey the three
13.5=kg batches of eutectic =ssl1t mixture from the reaction vessel to
the intermediate transfer vessel and then to the variocus shipping con=-
tainer assemblies. The transfers were made by applying helium gas
pressure cover a pool of molten salt having a temperature of 600°C. The
salt was forced out of the dip line and thrcugh the transfer line and
salt fillter tc the receiving vessel, which was vented. The transfer
lines were heated with tubular resistance heating elements, insulated
toc conserve heat and avoid cold spots, and equipped with thermocouples.
The transfers ranged in size from the 13.4 kg of uranium (4.7 liters
of salt) in the production batches to 4.3 kg of uranium (1.5 liters of

salt) that was reguired to fill the 45 enrichment capsule array.

The primary transfer line (3/8-in.-0D x ©.065-in. wall type 200

nickel tubing), which connected the reaction vessel to the lk-in.-diam

storage vessel, alsoc served as the treatment gas supply line to the dip
61

 

tube in the reaction vessel. During the processing cperations, this
line was ccnnected to the ftreatment gas supply line instead of the
transfer vessel. In each of the feed batch transfers; a compression-
type Titting had to be broken and the transTer line had tc be connected
to & new salt filter that had previously been installed in a nozzle
atop the storage vessel. The sintered nickel filfer was 3 in. in
diameter, and the pore size was 20 p. Transfers through tnhis line re-
gquired an overpressure of 15 psig to overcome the static head (18 £t
HEO) and the restriction to flow of the filter. Time required to com-
plete the transfer varied from 78 to 140 min. Initial movement of the
salt through the line was obvious from the rapid response of the attached
thermocouples to the salt tempersture. The progress of the transfer
was followed by observing the rising back pressure on the helium purge
to the storage vessel dip line; this indicated a rising liquid level.
The rush of gas through the line at the end of the transfer was audible
from the sound detector. These conditions were verified by a sharp
drop in the reaction vessel pressure and a rapid increase in gas flow

through the salt transfer line.

Five transfers were required to fill the three-product container
assemblies. They were made via a 1/4-in.-OD nickel transfer line that
connected the assembly to a dip line in the storage vessel. Since
filters were not involved in these transfers, the transfers were com-
pleted in approximately 10 min. Thermccouples cn the transfer line and
on the shipping containers indicated the movement of salt. An overflow
pot at the end of the containers (in series) was egquipped with liquid
level detection probes and internal thermocouples to indicate the preg-
ence of sglt. A blowback of gas then ensured that the contasiners were

filled to the proper level.

The transfer operations were conducted at a salt temperature of
600°C, (This temperature had been arbitrarily selected, and the con-
: tainers had been fabricated to contain the desired quantity of fuel at
this temperature.) The filled containers were allowed to cocl, and the

salt was allowed to solidify before the disassembly operaticn began.
62

6.9 Container Disassembly and Preparation for Shipment

The shipping container assemblies were stripped of the thermo-
couples, overflow pots, heaters, and supporting hardware. The containers

were separated from each other by cutting the interconnecting tubing.

The 45 enriching capsules were clipped and trimmed, were tested
(i.e., their lifting bails were tested) with & 15-1b pull, were drilled
for draining and venting, and were ldentified and weighed. Then they
were packaged, in groups of six, in carrousel shipping containers. The
equipment, which was designed to do these operations remotely, functioned

satisfactorily.

The 2-1/2-in.—diam product cans were easlily removed from the filling
arrays; nhowever, removal of the top and bottom plugs from each can was
guite difficult. An inordinate amount of force was necessary to break
the bottom plug locse from the thin salt Tilm that had formed in the
annulus between the plug and the bottom nozzle. Several fixtures de-
signed for this task proved to be inadequate and had to be replaced with
sturdier units. A lifting bail was installed on each can. Also, a
stopper was inserted in the bottom of the can to minimize the loss of

salt during handling and shipment.

The cans were welighed by using a beam balance that was located in
a glove box mounted cn top of the processing cell. The glove box be-
came a part of the cell contaimment for this operation. A long cable,
with the 1ifting bail of the can attached, was suspended from the
balance into the cell. The gross weight of the cans ranged from 1.6

to 15 kg.

All of the shipping containers were then stored in cell G to await

shipment tc the MSRE.

6.10 Transfer of TLiF-233UFLL Salt Product to the MSRE

Six of the 2-1/2-in.-diam product salt containers (total uranium
content, 33 kg) were delivered, individually, to the MSRE in the 10-ton

Pu-Al carrier. The Pu-Al carrier was slightly modified for this opera-

 

tion. The three-barrel magazine was removed, and a winch-and-cable
 

 

63

assembly was installed on top of the carrier in a sealed housing &o

that it became a part of the carrier containment.

The carrier was placed on top of the pedestal over a 6-in.-diam
port to cell G, and beeame.a part of the cell containment when the
slide drawers in both the carrier and the pedestal were opened. The
cable hook was lowered into the cell to receive the product can and
then raised to withdraw the can from the cell into the carrier. The
pedestal drawer and the carrier drawver were closed, isolating the car-
rier and cell from each other and from the ambiIent atmosphere. In this
manner, containment of the fuel was maintained at all times. The car-
rier was then remcved from the pedestal, sealed by the installation of

the cover plates, and cleaned of any surface contamination.

The fuel carrier was then transported tc the MSRE, where it was
positioned over a turntable (Fig. 24) that was designed for charging

the enriching concentrate to the reactor fuel draln tank.

Measurements, at the surface of the Pu-Al carrier, of the radiaticn
emitted by one T-kXg uvranium fuel can are given in Table 8. The neutrons
are proauced by (a,n) reactions on lithium and flucrine (of the eutectic)
by the high~energy alpha particles emitted by several of the daughters

of the =37,

The enrichment capsules were shipped tc the MSRE in a smaller
carrier that was designed to accommodate six capsules in a carrousel
fixture. The carrcusel served as a support Tixture during trsnsport
and as a holder for six bottom plug assemblies that became a part of
the sampler ~ enricher system at the MSRE as the capsules were removed

Trom the carrier through a l-l/2=in.-diam hole in the top.

A small glove box was fabricated for use in removing the carrousel
assemblies from cell G. This box was mounted on top of the small car-
rier, which, in turn, was mounted on the pedestal above the cell. Then
the pedestal slide-valve drawer was opened, and the glove box and the
carrier became a part of the cell containment. A cable from a winch
inside the box was lowered into the cell to receive the carrousel. The

carrousel was withdrawn into the carrier, the pedestal drawer was closed,
Tank.

6l

ORNL-DWG 68-967

  
 
    
       
      

TURF CARRIER

-GRAPHITE SAMPLING
s SHIELD

PURGE GAS
SUPPLY—

e e
e

o . P
MAINTENANCE SHIELDz S
e T ~

 
 

 

CONTAINMENT ENCLOSURE  /
AND STANDPIPE ASSEMBLY .

~

Fig. 24. Arrangement for Adding 233U Enriching Salt to Fuel Drain

 

 
65

Table 8. Radiation lLevels, Measured at Surface of Pu-Al Carrier, of
TLiF—233UFlL F'uel Containing 7 kg of Uranium

 

 

 

 

Instrument

Loecation Along Radiation Level (mrem/hr) at Specified

Vertical Axis Type of Distances from Cerrier Surface

of Carrier Radiation 0 12 in. 5 £t

Gamma, <2 < 1 < 1
Ny 5 50 12.5
12 in. below T
top of carrier I\T C:3 0.3 0.2
T I
N, o+ 25 15 5
Gamna <7
NF 200 75 10
Center of carrier NT C.6 0.3 0.3
- _
N ¢ 75 Lo 7.5
Gamma. < 1 <1 < 1
L \

18 in. =bove NF 2 o0 =
bottom of N, 0.3 C.6 0.2
carrier ~ = v

Nint 25 17.5 6.5
aN?, NT, Nint = Neutron radiation (fast, thermal, and intermediate).

and then the carrier was removed from the pedestal and placed on 1its
base. The glove box was removed from the carrier, and shield plugs
were inserted in the l/Emin.mdiam and l—l/z-in.—diam access openings
in the top of the carrier. The carrier contained a S5-in.-diam cavity

with 4 in. of lead shielding around it.

When the small carrier housed six capsules, each containing % g
of uranium in approximately 150 g of fuel concentrate, the gamms radi-
ation level was 200 mrem/hr at the surface and decreased to 20 mrem/hr

at 6 f£. The neutron radiation levels at the surface and at 6 ft were

 

105 mrem/hr and 10 mrem/hr, respectively; measurement made at the open
66

bottom of the carrier during removal from the pedestal indicated s

radiation level of 50,000 mrem/hr.

The removal of the fuel from cell G and its subseguent transfer
to the MSRE were characterized by very low radiation exposures to
personnel and essentially no spread of contamination. In fact, only
the top surface of the pedestal and a small area immediately adjacent

to it had to be decontaminated.

£.11 Meterisl Balance

A total of 39.72 kg of uranium, in the form of uranium oxide
powder, was transferred to cell G of the TURF. Of this amcunt, 38.95 kg

of uranium, as LiF-UFh, was packaged as the finished product.

The can-opening operation accounted for a loss of 245 g of uranium
as uranium oxlde. DMost of this, 200 g, was lost through the empty can
discharge chute because excessive powder had accumulated in the box. A
lezking sezl on the vibratory screen drive shaft accounted for 45 g.
Remote modification to the decanning station and changes in operating
procedures eliminated these sources of loss after the first batch (29

oxide cans) had been processed.

Sampling accounted for a loss of 100 g of uranium. Two 15-g and
two 5-g salt samples were required in each of three production runs.

In addition, several extrs samples were required for special purposes.

The presence of heels in certain vessels and the holdup in transfer
lines and filters were responsible for the majority of the loss. The
heel in the reaction vessel amounted to 27 g. However, the storage
vessel contained 346 g of uranium as heel as a result of a dip line
replacement. The dip line was installed improperly and, as a result,

about 1/3 in. of unrecoverable salt heel was left in the vessel.

A summary of the uranium accountability is given in Table 9.
 

er

Table 9. Uranium Material Balance

 

 

 

Uranium
(g) (Wt %)

Received as oxide 39,721 100
Finished product, LiFmUFA 39,952 98.1
Measured loss tc heels,

semples, etc. Li6 1.1
Estimated loss to dusting,

lines, filters, etc. 322 0.8

 

7. MAINTENANCE ENGINEERING

Early in the run with depleted uranium, it became obvious that
some of the equipment in cell G could not be considered reliable and
would have to be repaired or replaced before the conclusion of the
three production runs. Many of the deficiencies were corrected in
the interval between the depleted-uranium run and the 233U’ run. Since
it was impractical to consider major equipment modifications at this
time becsuse of the fuel delivery schedule reguirements at the MSRE,
emphasis was placed on: (1) storage of spare parts in the cell,

(2) installation of redundant fittings, (3) procurement and modifica-
tion of tcols and work tables in cell G, and (4) formulation of main-

tenance procedures.

7.1l Spare Parts

Because of the difficulties involved in obtaining access tc the
hot cell, a generocus supply of replacement parts was stored inside the
cell and cataloged Tor easy retrieval. Where applicable, the parts
were prefitted to ensure compatibility with the system. Threads on

the fittings and nuts and bolts were treated with a lubricant that is

suitable for use in high-temperature environments.
68

 

7.2 Redundant Fittings

The compression-type fittings used on the salt transfer lines and
on the gas supply lines at the vessel nozzles were extremely difficult
to make and bresk after they had been thermally cycled from room tempers-
ture to 600-700°C. The passage of molten salt through one of these
fittings made the operation even more difficult beczuse the salt formed
a seal around the protrusion of the tubing into the fitting. BSince
several of these fittings had to be routinely operated in the product
transfer operations, there were several lines in the system for which
no alternate had been provided. In these cases, redundant fittings
were installed in such a manner that a second Titting was available 1if
the primary fitting could not be operated. Although complex fitting
arrangements are normally conducive to leakage, no such prcblems were
encountered here. On two occasions, use of the standvy fitting was
necessary in order to replace the sinbtered metal filter in the salt

transfer line.

7.3 Tools and Work Tables

A large assortment of hand tools, medified for use with the manip-
ulators, were stored on tool racks in the cell. Usually the medification
consisted of welding grips to the tool to aid in handling by the master-
slave manipulators. A few special-purpose tecols, such as air=-operated
snips (for cutting the interconnecting tuking of the capsules) and an
impact wrench (for use with the electromechanical manipulator), were
also stored for later use in the cell. End wrenches and other fixed-

size tools were color coded by sizes for ease in identification.

Work tables and trays were located between the viewing windows
and the major pieces of equipment. They proved to be invaluable In the
many mechanicael and maintenance operations conducted in the three sslt
production runs because of the limited reach ©i the master-slave manip-
ulators and also because of the limited field of wision available through
the windows. A secondary, but very important, function of the trays
involved retrieval of tools and miscellaneous items that were dropped S

because of wear of manipulator parts or because of aswkwardness cn the
 

 

part of the operator. When an item fell to the Tloor, rather than
onto a tray, it usuzslly had to be replaced since retrieval from the

floor in many locations cof the cell was impossible.

The large heavy-dulty work table (Fig. 17) designed for the product
container and capsule array disassembly operations served well in these

functicns and was also used extensively as a general-purpose work bench.
T -

7.4 Maintenance Procedures

Remote maintenance of the process equipment was recognized, during
the run with depleted uranium, to be a formidable task that could not
be left to the operator’s discreticn when the need arose. Thus an
engineer wac assigned, on a full-time basis, to develcop detailed main-
tenance procedures that would be applicablie during any conceivable

failure as well as during the planned maintenance operstions.

A comprehensivé set of close-up photographs was taken of the
equipment in cell G before the cell was sealed to process the radioactive
material. These were later sugmented by photograpins taken through the
chielding windows and the monocular viewers. Altogether, more than 300

photographs were made.

Fifty-five detailed, step-by-step maintenance procedures were
written by using the photographs and field sketches {more than 300)
that had been used in the fabrication and installstion of the equip-
ment. A complete list of tocls and materials necessary for the job was
included in each procedure. Where appropriate, the maintenance proce-
dure was referred to, cor inecluded in, an operating procedure to ensure
continuity of a process cperation. Although it was not necessary to
use all of these procedures in the fuel production, they were instru-

mental in ensuring a safe and orderly completlon of this task.
&. CONCLUSIONS

1. Reactor-grade eutectic salt concentrate, 7LiF-

Jur, (73-27
2 27

233

UO2 or UFA

23
mole %), can be prepared by remote means from Li¥ and 33

by the process described in this report.
70
S
2. Hydrogen reduction and hydrofluorination reaction rates of
oxide beds can be controlled satisfactorily by adjusting hydrogen and
HF flows to avoid excessive temperature excursions from the exothermic
resctions involved. Temperatures can be determined by placing a number

of thermocouples within the beds.

3. Equipment for a remote process of this type should be designed
with a minimum number of close-tolerance fits. It should alsc be de=
cigned on & modular concept sco that components can easily be replaced,

thus eliminating, if possible, remote repair operations.

L. 1In general, commercial pipe, tubing, and electrical fittings
are satisfactory in this type of operation. However, they are not
satisfactory in instances where high temperatures or exposure to molten
salt are expected. In these cases, special-purpose fittings should be

developed.

5. Concentrations of the corrosion products nickel and iron can
be reduced to satisfactorily low levels by hydrogen treatment of molten

salt at 550°C. Chromium was not reduced by the process described. .

6. A binary mixture of UFu-LiF having a eutectic composition can
be fused at a temperature of 650°C in the presence of a small amount of
ligquid eutectic heel remaining in a vessel. In the absence of the lig-
uld heel, it is necessary to heat the materials to the melting point

of the LiF (835°C) to achieve liquefaction.

7. Tc chbtain a product containing less than 40 ppm of oxygen and
to avold plugging of dip lines by the formation of oxides within them,
it was necessary to install & catalytic recombiner and a molecular-
gleve drying column in the hydrogen supply system. The use of Linde
type L1k 1/16-in.-diam sieve material in a regenerative drying column
reduced the moisture content of commercial-grade hydrogen to less then

3 ppm.

8. Maintenance procedures are necessary to ensure orderly com-

pletion of a process of this nature. They are vital when entry into a

 

contaminated cell 1s necessary and when equipment and the prcduct must

be removed from the cell.
 

 

Ig!
9. ACKNOWLEDGMENTS

The authors express their appreciation to the following percons
whose assistance was essential to the successful preparation of the
enriching 233U Tuel concentrate for the MSRE: G. I. Cathers and
J. H. Shaffer (Reactor Chemistry Division) for process development;

E. L. Nichclson and W. F. Schaffer, Jr., for process equipment design;
S. Mann for assistance in operations and data analysis; J. P. Jarvis
(on loan to the Metals and Ceramics Division from the General Engineer-
ing Division) for process equipment installation and modification, and
for development of maintensnce procedures; and J. W. Anderson and

D. M. Shepherd (both of whom are on loan to the Metals and Ceramics
Division from the General Engineering Division) for facility engineering

and supporting services.
10. REFERENCES

1. P. N. Haubenreich, private communication, December 1966.

2. J. R. Engel, MoRE Design and Operations Report, Part XI-A,
ORNI~TM-2304 (September 1968).

 

3. Chem. Tech. Div. Ann. Progr. Rept., May 31, 1968, ORNI-U272,
pp. 28«33,

 

L. J. W. Anderson and J. M. Chandler, Safety Analysis for the Thorium-
Uranium Recycle Facility (TURF), Building T793C, ORNI-4278 (%o be
published).

 

5. C. J. Barton et al., "Phase Bquilibria in the Alkali Fluoride-
Uranium Tetrafluoride Fused Salt Systems,” J. Am. Ceram. Soc.,

h(2), 63-69 (1958).

 

6. P. N. Heubenreich et al., MSEE Design and Operaticns Report,
Part V-A, Bafety Analysis of Operations with 233U, ORNL-TM-2211
(February 1968), p. 37.

 

 
 

 
 

 

T3

ORNL-4371
UC-80 — Reactor Technology

INTERNAL DISTRIBUTIOR

1. Biology Library 76, C. E. Larson
2-l4, Central Research Library 77T. X. H. Lin
5-6. ORNL = Y-12 Technical Library 78. R. B. Lindauer
Document Reference Section 79, A. P, Litman
T-41, Laboratory Records Department 80. A. L. Lotts
42, Laboratory Records, ORNL R.C, 81. M. Lundin
L3, R. G. Affel 82, H. G. MacPherson
Wi, J. L. Anderson 83. R. E. MacPherson
45, J. W. Anderson 84, 5. Mann
46, C. F. Baes 85. H. E. McCoy
47, 8. E. Beall 86. J. R. McWherter
L8, M. Bender 87. R. L. Moore
49, E, S. Bettis 88. J. P. Nichols
50, E, G. Bohlman 89. E. L. Nichoclson
>1. 5, E. Bolt 90, A. S. Meyer
52. G. E. Boyd 91. A, M. Perry
53. R. B. Briggs 92-93, M. W. Rosenthal
54, R. E. Brooksbank 9L, Duniap Scott
55. W. D. Burch 95. W. F. Schaffer
56, W, 4. Carr 96. J. E. Shaffer
57. G. I. Cathers 97. M. J. Skinner
58-59, J, M, Chandler 98, J. W. Snider
60. F. L. Culler, Jr. 99. D. A. Sundberg
61. 8. J. Ditto 100. R, E. Thoma
62, W. P, Eatherly 101, W. E. Unger
63. J. R. Engel 102. J. E. Van Cleve, Jr.
64. D. E. Ferguson 103. A. M. Weinberg
65, L. M. Ferris 104, J. R. Weir
66. J. H Frye, Jr. 105. M. E. Whatley
67. R. E. Gehlbach 106, G. M, Watson
68. W. R. Grimes 7 107. J. C. White
69, A. G. Grindell 108. R. G. Wymer
70, R. G. Guymon 1069. Gale Young
T1. P. N. Haubenreich 110. J. L. Youngblood
T2. W. H. Jordan 111. P. H. Frmett (consultant)
73, P. R. Kasten 112. J. J. Katz (consultant)
Th, J. W. Koger 113. J. L. Margrave (consultant)
75. C. E. Lamb 11k, E. A. Mason (consultant)
115. R. B. Richards (consultant)

EXTERNAL DISTRIBUTION

116, D, E. Bloomfield, Battelle-Northwest, Richland, Washington
117. J. A. Bwartout, Union Carbide Corporaticn, New York
118, Laboratory and University Division, AEC, ORO
119-338, Given distribution as shown in TID-4500 under Reactor Technology
category (25 copies — CFSTI)