ORNL-TM-2685.txt

   

OAK RIDGE NATIONAL LABORATORY
operated by

UNION CARBIDE CORPORATION
NUCLEAR DIVISION T

for the
U.S. ATOMIC ENERGY COMMISSION

ORNL- TM-2685

 

COPY NO. -

RECEIVED BY DTIE OCT 9 196¢

DATE - August 10, 1969

RECEIVED «.

INHERENT NEUTRON SOURCE IN MSRE WITH CLEAN 233U FUEL

R. C, Steffy, Jr.

ABSTRACT

After about three years of nuclear operation, the MSRE fuel, enriched
’ 235U, was replaced with a 233y fuel mixture. In this new mixture there
- are quantities of 232y, 233U, and 2%%U., Each of these, along with the 2329

decay chain, is a strong alpha emitter and interacts with fluorimne, beryllium,

and lithium to produce neutrons. This neutron source is time-dependent be-
cause of the buildup of 2329 daughters, and at the time of reaching critieality
with the 2%%U fuel, the neutron source in the MSRE core was about 4 x 108
neutron/sec, primarily from the reactions ’Be(a,n)!'?C and !°F(a,n)?2Na.

Alpha-n
reactions with lithium will produce <3 X 10° neutrons/sec. Spontaneous fission
will produce <102 neutromns/sec. ,

Keywords: Inherent neutron source, 233y fuel, (a,n) reactions, alpha
particles, particle sources, 232y decay chain, fluorine, beryllium, lithium.

NOTICE This document contains information of a preliminary nature
and was prepared primarily for interncl use at the Oak Ridge National

Laboratory. It is subject to revision or correction and therefore does
not represent o final report.

QETESUSESS U5 Do MOCUAISE i UBLALE
 

i — LEGAL NOTICE — —— —————

 

This report was prepared as an account of Government sponsored work., Neither the United States,

nor the Commission, nor any person acting on behalf of the Commission:

A. Makes any warranty or representation, expressed or implied, with respect to the accuracy,
completeness, or usefulness of the information contained in this report, or that the use of
any information, apparetus, method, or process disclosed in this report may not infringe
privately owned rights; or

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

As used in the above, “‘person acting on behalf of the Commission'’ includes any employee or

contractor 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, any information pursuant to his employment or contract with the Commission,

or his employment with such contractor,

 
Introduction . . .
Fuel Composition .
Alpha Source . . .

Neutron Source ., .

"Li(o,n)1°B .

»

Spontaneous Fissions.

CONTENTS

“Be(a,n) 2 and °F(upn)>7Na. .

Discussion . . . .

Appendix . . . . .

. -

Calculation of Inherent Neutron
for 23 Fuel Mixture . . . . . + v v v v v v v e e e .. 1D

Source from °Be(q,n)!*C and ®F(a,n)®®Ne . . . . . . . . 12

Source from "Li(a,n)?°B .

 

o
o
0

@CO*]*Q-\]O\L"J:‘}EQ

e e e e e e e e e e e 12

Source in MSRE

* . . . L e - - . ° - » - lT

LEGAL NOTICE

Thia report was prepared as an account of Goverament sponsored work. Neither the United
States, nor the Commission, neT ALy person acting on bshalf of the Commdssion:

A. Makes any warranty or representation, expressed or implied, with respect to the accu-
racy, completeness, or usefulneas of the information contained in this report, o¢ that the use
of any information, apparatus, method, or process disclosed {n this report may not infringe
privately owned rights; or

B. Assumes any liabilities with respect to the use of, or for gamages resulting from the
use of any information, apparatus, method, or process disclosed in this report.

As used in the above, ‘‘person acting on behalf of the Commiasion’ includes any em-
ployee or of the C jnsl or empioyee of such , to the extent that
auch ployee or of the Ci izal or empl of such contractor prepares,
disseminates, or provides access to, any information pursuant to hia employment or contract
with the Commiasion, or his employment with auch contractor,

 

 

ENIPEDNCTS £ T3 TR

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i

2

\
INTRODUCTION

In the MSRE fuel mixtures, alpha-emitters are dispersed in large
quantities of fluorine, beryllium, and lithium, All three of these ele-
ments undergo alpha-n reactions to produce neutrons. Haubenreichl calcu-
lated the inherent neutron source for the initial fuel loading (~ 35% U
and ~ 65% ©8U), but the different concentrations of uranium isotopes for
the 233 fuel loading, all of which are strong alpha emitters, required
that another calculation be performed. The calculations reported herein
were performed before the ©° was loaded into the reactor.

This memorandum is divided into two parts. The first part presents
final results and general discussions; the second part (appendix),

intermediate results and details of the calculations.
FUEL COMPOSITION

The strength of the inherent neutron source is, of course, determined
by the composition of the fuel salt. Except for the change in uranium,
the major constituents of the fuel salt were essentially unchanged by the
chemical processing which removed the 235y~-rich uranium and replaced it
with & 223 fuel load. Thus, changes in the strength of the inherent
neutron source were basically a function of the change in the alpha pro-
duction rate and the associated alpha energies. Table 1 lists the
approximate weight percentages of the components in the £33 fuel loading.
These were estimated based on the predicted critical uranium concentration
of 15.8 grams of uranium per liter of salt (~ 1 1b/ft’). It was assumed
that all of the uranium in the fuel was loaded as part of the 233 fuel
loading, and that the small amount present from other sources was negli-
gible. The concentrations of the uranium isotopes which are present in
the 277U fuel loading are listed in Table 2 with a comparable listing for
the €U fuel loading.
Table 1

MSRE Fuel Salt Composition

 

 

_ Wt 9% in
Element 233 Mixture
F 684
*
Li 11.0
Be 8.8
r 11.0
U 0.73

 

*99.99% 7Li

Table 2

Uranium Isotopie Concentration in MSRE Fuels

 

 

 

- Atom % in Atom % in
Nuclide 233 Mixture Initial Loading
235y 0.022 ---
=33y 91.4g ---
=347 7.6 0.3
235U O-T ' 35
236y 0.05 0.3
23877 O.]_Lj. 6L . L

 

aData from Reference 2.

I
“Data from Reference 1.
ALPHA SOURCE

Each of the uranium isotopes in the 3-fuel mixture decays by
alpha emission, These all have long half-lives and emit alpha particles
at essentially a constant rate for the lifetime of the MSRE. With one
exception, the daughters of the uranium isotopes also have long half-lives
for alpha emission and never reach high enough concentrations to emit
alpha particles at a significant rate. The very important exception is
228Th, the daughter of 232U, which decays with a 1.91 year half-life by
alpha emission. This is followed by an entire string of alpha emitters
with very short half-lives; so short, in fact, that they are essentially
in equilibrium with the 223Th, An alpha decay by ©28Th is quickly fol-
lowed by a cascade of alpha emissions until the parent has finally decayed
to 2°8pp, which is stable,

Immediately after the 24 was purified (~ June 196L), the only alpha
emissions were from the decaying uranium atoms, but as the concentration
of ##8Th increased, the alpha activity obviously increased. So the alpha
activity in the fuel salt is composed of a time-independent contribution
from the decaying uranium atoms and a time-dependent contribution from the
272y decay chain. The time-dependent contribution is controlled by the
228Th half-life of 1.91 years and reaches 75% of the saturation value
about L years after the 27U purification.

The energies at which the alpha particles are emitted are also im-
portant., All of the significant alpha emissions from the uranium atoms
are between 4.1 and 5.3 Mev. The alphas from the lower members of the
233y decay chain are emitted with energies between 5.2 and 8.77 Mev.

Since the neutron yield increases sharply with incident alpha energy, the
higher~energy alphas in the 233j decay chain add weight to the importance
of the chain,

Tabulated information pertaining to the properties of the decay of
the uranium isotopes and the 233y decay-chain are given in the appendix
in addition to tables of the numbers of alphas/sec emitted by the various

isotopes.
NEUTRON SOURCE

Calculation of the neutron source is necessary primarily to insure
that, even when subcritical, there are enough neutrons available to meet
detection criteria and guard against a startup accident.® From this
standpoint, the graphite region in the core is the region of concern,
for only in this region of the MSRE is criticality feasible. The calcu-
lations presented herein are for 25 ft° of salt, the approximate volume
of salt in the effective core, whereas the total volume of salt is about
75 £t2. So multiplication of the source strength in the core by ~ 3 will
give the approximate source in a fuel drain tank when it contains the

entire fuel loading.

7Li{a,n) 198
Haubenreich? found lithium to be an insignificant neutron producer
when compared to beryllium and fluorine in the 35U fuel mixture. A cal-
culation (see appendix) of the neutron yield from the most energetic
alpha (8.77 Mev) in the 33U fuel showed that lithium produced less than
one neutron for every 110 produced by beryllium and fluorine. Alphas
starting at lower energies produce proportionately fewer neutrons from
lithium because of the rafiid decrease with decreasing alpha energy in the

4 (Threshold energy for

cross section of lithium for (a,n) reactions.
“Li{a,n)*®B is ~ 4.3 Mev.) The total neutron source from all (a,n) re-

actions with lithium is <3 x 10° n/sec.

Spontaneous Fissions

All of the uranium isotopes which are present in the £33 fuel loading
undergo spontaneous fission, with half-lives for this process ranging from
8 x 1012 yr for 233 to 3 x 107 yr for 23F. The total rate of spon-
taneous fissions in'fihe MSRE is simply the sum of the products of the
inventory of each isotope and its spontaneous fission time constant.

Less than 100 n/sec will be produced in the MSRE core by spontaneous

fission.
“Be(o, n) 120 and 19F(q, n) ©“Na

High neutron production rates result from alpha reactions with
beryllium and fluorine., For the alpha particles emitted from the uranium
isotopes (energies between k.1 and 5.3 Mev) the neutron production rates
from the beryllium and fluorine are approximately equal. The neutron

yield of fluorine®

increases more rapidly with increasing alpha energy
than does the yield of beryllium (see Fig. 1);%,7 therefore, for the
higher-energy alpha particles produced by the €3&j decay chain, most of
the neutrons result from (a,n) reactions with fluorine,

Figure 2 shows the total calculated neutron source for the MSRE core
region and the individual production rates for the most significant alpha-
emitters. The 233 decay chain obviously dominates the neutron production
with #12Po being the alpha emitter (alpha energy of 8.77 Mev) which causes

the most prolific neutron production,
DISCUSSION

When the #7%U mixture was placed in the MSRE fuel salt, the reactor
already had about three years of operating history. The long runs at high
powver insured that a large photoneutron source would be present for some
time after shutdown of the reactor. However, if sufficient time were al-
lowed to elapse, this source would weaken., More important as a reliable
neutron source in the MSRE is the inherent neutron production from (q,n)
reactions with the salt constituents. The (aq,n) source is essentially
independent of power history and actually increases asymptolically to a
limiting value. At the time the MSRE achieved criticality with 2°% as
the fissile material (the 23 mixture was the equivalent of ~ 4 years
01d),® the inherent neutron source in the core region was ~ & x 10%® n/sec,
about a factor of 1000 over the inherent source which was calculated for
the #75U fuel mixture.?

The accuracy of these calculations is dependent on the accuracy to
which the fuel isotopic composition and the various yield data are known.

The fuel composition is thought to be known to well within #* 5% and is
ORNL—-DWG 69— 40028A
10

Be (DATA FROM REF 6)
N=0152 £3-85

n

— == EXTRAPOLATED

NEUTRONS PER MILLION ALPHAS

N’

F (DATA FROM REF 5)
S N=1.02x10"*F683

 

0
0.5 1 2 5 10
£, ALPHA ENERGY (MeV)

10

Fig. 1. Neutron Yield from Thick Beryllium and Flucrine Targeis as
a Function of Incident Alpha-Particle Energy.
INHERENT NEUTRON SOURCE IN THE MSRE CORE

10

ORNL—DWG 69-100304A

 

10
5
TOTAL NEUTRON SOURCE
2
212p, (877 Mev)
\ALPHA ENERGY
108
216py(6.77)
TOTAL URANIUM
5
220Rn (6.27)
233y (4.7-4.8)
224pq (5.2~ 5.7) 232\ (5.1 -5.3)
2
2281 (5 2-5.4)
(OCT 1968 - MSRE CRITICAL
; WITH 233)
10

0 { 2 3 4 5 6 7
TIME AFTER PURIFICATION OF THE 233U FUEL (years)

Fig. 2. Total Inherent Neutron Source and the Neutron Source
Resulting from Individual Alpha Emitters in the 233U-Fueled-MSRE Core.
11

probably not an appreciable source of error. The area of this calculation
which contains the most possibility of error is the yield data for the
reaction, °F(q,n)®?Na. Due to insufficient data, we had to extrapolate
the known yield data (see Fig. 1), which only went to alpha energies of
5.3 Mev, to get the yield from the higher-energied alphas from the Z22U
decay chain, We made the assumption that the yield continued increasing
with the same functional relationship to the higher energies. While it

is reasonable to expect the neutron production rate to increase with alpha
energy, it is not known whether the yield will increase according to the
same power function which could be used to describe its behavior for lower
energied alphas. Yield data for beryllium (which is known from experiment
for alphas to >6 Mev and is extrapolated to ~ 8 Mev by therexperimenters)6
is also shown in Fig. 1 and does indeed increase as a power function with
increasing alpha energy. Since the data for beryllium increases uniformly
to higher alpha energies, one would tend to expect the assumption about
fluorine to be good, but even if the assumption were not good and the
yield data for the 8.77-Mev alpha were lower by a factor of 2 than the
extrapolated curve, the calculated total neutron source would be in error
by only ~ 25%. We feel that it is fair to assign a probable error band

of + 25% to these calculations with the expectation that the error is much
smaller but with the realization that more experimental data on high-energy
alpha interactions with fluorine is necessary before stronger confidence

in the calculated neutron yield is warranted,
12

APPENDIX

Calculation of Inherent Neutron Source in MSRE for

233 Fuel Salt Mixture

 

Source from ZBe(a,n)13C and !°F(q,n)Z=Na

For both reactions %Be(q,n)'2C and '°F(q,n)%®Na, the general equation

used to determine the neutron source was

N = (Ax/10°)(N__ Y(N/N__) (1)

where
Ax = Alpha activity (a/sec) in MSRE core (25 £t of salt)
N = Neutron source (n/sec)

N = Maximum number of neutrons produced if target were

max . .
composed completely of neutron-producing nuclei
(°Be or °F as the case may be). N,,, is given in
units of neutrons per million alphas.

(N/Nmax)= Fractional yield of neutrons produced in fuel
mixture to number that would be produced in pure
target material.

Each of the three terms enclosed in parentheses in this equation is
found independently. Each term is discussed in the following paragraphs.
For the life of the MSRE, each of the uranium isotopes present in
the fuel will be a near-constant alpha source due to its long half-life
(see Table 3). Each of the isotopes, except 74U, decays to a nuclide
which alsc has a long decay half-life, The second, and subsequent nuclides
in each of these decay chains are thus eliminated as significant alphs
sources.,
While 273y produces a constant supply of alphas, it also produces
the same number of “*8Th atoms. Thorium-228 has a 1.91 year half-life
and starts a chain of short half-lived elements (see Table 4). Since the
fuel mixture was about four years old when loaded into the reactor, we may
assume that ®28Th and its decay chain are in equilibrium (i.e., the activity
(alphas/sec) from #25Th and each member of its decay chain will be the

same) .
13

Table 3

Alpha Source in MSRE Core from Uranium Isotopes

 

 

 

 

 

 

 

Decay o
Half Life Energy Emission a Source
Isotope (yr) (Mev) Percentage (a/sec)
233y 7l 5.31 68 1.29 x 1012
- 5.26 31 0.59 x 1012
5,13 0.3 0.06 x 1012
23 1.62 x 10° 4,82 83 3.01 x 1012
4.78 15 0.5k x 1012
h.73 2 0.07 x 101®
234y 2,48 x 10° 4,76 Th 1.45 x 10t?
%.70 23 0.45 x 101t
4.60 3 0.06 x 101%
238y 7.13 x 108 4,39 86 5.30 x 10°
h,57 10 0.63 x 10°
4,18 b 0.25 x 10°
2386y 2.39 x 107 4,50 T3 9.80 x 108
b, L5 o7 3.61 x 10°
238y 4,51 x 10° 4,19 T 1.53 x 10°

4,15 23 0.46 x 10°

 
ik

Table &

Information Related to Members of <y Chain

 

 

 

Decay Decay Alpha Energy Emission
Nuclide Mode Half-Life (Mev) Percentage
£28 o 1.91 yr 5.42 71
5.33 28
5.20 0.6
224py o 3.64 day 5.68 95
5. hh 4.6
5.19 0.4
“2CRn a 52 sec 6.25 100
£18pg o 0.16 sec 6.77 100
“1Zph R” 10.64 h none 0
£12py B~ (66.3%)  60.5 min rionie 0
a(33.7%) 60.5 min 6.09 27
6.05 70
“18pg o , 0.30k ysec 8.77 100

 

Activity of each of the uranium nuclides is simply the product AN,
where )\ 1is the decay constant and N is the number of atoms of the par-
ticular isotope. Activity of #23Th (and other members of decay chain)

may be calculated using the relation

N At -A ot
(Aa)g = KQNE = %‘2&)\"{'%1 (e k':L - e )\'2 ) (’))
15

where
(AQ) 2 = ©28Th activity (a/sec)
Ay = “°%U decay constant
Ao = 228Th decay constant
Number of 233U atoms

Number of ©28Th atoms

= =
no»
noo

Results of this calculation are shown in Tables 3 and 5 and completes
the calculation of the alpha activity.
Runnalls and Boucher6 give the neutron yield for alphas incident

upon a beryllium target as a function of alpha energy, E (in Mev),

N oy, = 0.152 E>:®5 neutrons per million alphas, (3)
This relation is expected to be good over the range of alpha energies
encountered in the MSRE fuel.

Apparently little work has been performed on the neutron source from
(a,n) reactions with fluorine. The only neutron yield data we could lo-
cate was in an article by Segre and Wiegand.® The yield data was only
given for alpha energies up to 5.3 Mev., Up to this energy, the yield was
increasing as a power'function of energy. Assuming that this held true
up to 8.77 Mev (as was the apparent case with beryllium), the data was
extrapolated, The analytic expression for the neutron production of

fluorine as a function of alpha energy was found to be

N ooy = 1.02 % 10™% E®-83 neutrons per million alphas. (4)
Equations (3) and (4) analytically define the expressions to be
used for the N term in Equation (1)
The term (N/Nfiax) in the general equation converts the neutron yield
from the yield in a pure medium to yield from a mixture. An expression,

which agrees well with observed data,6 for calculating this term is
16

 

(—) - R (5)

where n, is the number density of a nuclide and Si is the 'relative atomic
stopping power". Subscript "p" refers to the particular nuclide for which
the calculation is being performed., Number values for "S" were obtained
from an article by Livingstone and Bethe® where data is given for a variety
of elements as a function of incident alpha energy. Table G gives values
for "S" for MSRE fuel salt, obtained by interpolation of the Livingstone-
Bethe data. |

Table 9

Relative Atomic Stopping Powers, Sj

 

 

 

 

 

 

b 5 6 9
Li 0.55 0.55 0.55 0.55
Be 0.63 0.63 0.63 0.63
F 1.1 1.1 1.1 1.2
Zr 2.8 2.9 3.1 3.2
U 4.0 L.5 4.8 5.0

 

Shown in Table 10 are the calculated values for the fractional yield for
beryllium and fluorine as a function of incident alpha energy. In brief,
to calculate the neutron source from a particular alpha emitter, Equation (1)
is used. Alpha activity (Ax) may be obtained from Table 3 or 5 or calcu-

lated using Equation (2). Using Equations (3) and (L), N, . may be calcu-

ax
lated for both beryllium and fluorine., The appropriate (N/N, ,,) may be

found in Table 10. Multiplying these quantities, one gets the source from
beryllium and fluorine. Addition of these two gives the total source from

a particular alpha.
17

Table 10

Fractional Yield Data for Be and F

 

 

 

 

 

(N/N ..)
¢ Energy
Mev) h > 6 9
Element
Be 0.079 0.079 0.078 O.QTh
F 0.695 0.692 0.688 0.64T

 

Source from ‘Li(a,n)*°B

 

Haubenreich® found that for lower energy alphas (<5 Mev) lithium was
not a significant neutron producer in the MSRE. To be sure that it did
not become significant at higher alpha energies, the neutron source from

the 8.77-Mev alphas was calculated.

Neutron yield from lithium may be calculated using the relation®
‘ 0
N = Aozf N, o(E) (-dx/dE) 4E, (6)
4,3
Where
N = Neutron Yield (n/sec)
Ao = alpha activity (alphas/sec of energy E)
Eq = Energy of incident alpha (Mev)
Np = Number density of lithium
o(E) = Cross section for (m,n) reaction
4.3 = Threshold energy for "Li (a,n)?°B reaction.

The term (-dx/dR) was obtained from an article by Harris'l in which
he plots (-dx/AdE) as a function of alpha energy for a number of elements,

For the MSRE fuel mixture this was found by the relation
18

(-dx/dE)mix = }4 wi(-dx/dE)i (7}
i

where w, corresponds to the weight fraction of component i and (-dx/dE),

is the value taken from Harris for the 1°0 component, Units on (-dx/dE)

are gm.cm~Z:Mev™!, Table 11 gives the values of (-dx/dE) for the components
of the MSRE fuel mixture.

Table 11

(-dx/dE) for MSRE Fuel Components

 

{-dx/dE) x 10° (gm-cm”™2-Mev~1)

 

Wy Incident Alpha Energy (Mev)
Element (vt %) L 5 6 9
Li 11 1.1 1.3 1.k 1.9
Be 8.8 . 1.2 1.4 1.5 1.9
F 68.L 1.3 1.5 1.8 2.4
Zr 11 2.5 2.8 3.1 3.9
U 0.73 L. L 5.0 5.5 7.3

 

(ndx/dE)mix =

Z wi(-dx/dE) 4 1.4 1.6 1.9 2.5

 
19

Lithium's microscopic cross section for (a,n) reaction* reaches a
maximum of ~ 250 mb for T.l-Mev alphas, and drops exponentially with de-
creasing energy. Between 7.5 and 8.8 Mev the cross section is constant
at ~ 150 mb. |

By approximating the cross section by straight lines and carrying
out the integration, the neutron scurce from 8.77-Mev alpha reactions
with lithium is found to be <1.1 x 10° n/sec. This is seen to be negli-
gible when compared with the ~ 108 n/sec from fluorine and beryllium,
The lower energy alphas will produce proportionately less due to the

exponential decay of the cross section at lower energies.,
10.

11.

20

REFERENCES

P. N. Haubenreich, Inherent Neuiron Source in Clean MSRE Fuel Salt,
USAEC Report ORNL-TM-611, Oak Ridge National Laboratory, August 27,
1963. '

Oak Ridge National Iaboratory, MSRP Semiann., Progr. Rept. Aug. 31,
1967, USAEC Report ORNL-4191, p. 50.

J. R. Engel, P. N. Haubenreich, and B. E. Prince, MSRE Neutron Source
Requirements, USAEC Report ORNL-TM-935, Oak Ridge National Laboratory,
September 11, 1964,

J. H. Gibbons and R. L. Macklin, Charged Particle Cross Sections,
Phys. Rev., 114, p. 571, 1959.

E. Segre and C. Wiegand, Thick-Target Excitation Functions for Alpha
Particles, USAEC Report LA-136, Los Alamos Scientific Laboratory,
September 1944 (also issued as USAEC Report MDDC-185).

O. J. C. Runnals and R. R. Boucher, Neutron Yields from Actinide-
Beryllium Alloys, Can. J. Phys., 3k:94g (1956).

J. H. Gabbons and R. L. Macklin, Total Cross Section for ZBe(a,n),
The Physical Review, 137, No. 6B. B1508 - B1509, 22 March 1965.

J. M. Chandler, personal communication to P. N. Haubenreich, Oak
Ridge National Laboratory, October 1967.

M. S. Livingstone and H., A. Bethe, Nuclear Dynamics, Experimental,
Revs. Mod. Phys., 9:272 (1937)

W. N. Hess, Nuetrons from (o,n) Sources, Annals of Phys., 2: 115-133,
(1959).

D. R, Harris, Calculation of the Background Neutron Source in New,
Uranium-Fueled Reactors, USAEC Report WAPD-TM-220, Bettis Atomic
Power Laboratory, March 1960.
o .
-

O o= 0w EFw o+

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MomHoOHEAPa-gT Q0GOS s s sEHGyDdOHDTENLDNOQ QT

. G. Affel

I.. Anderson
F. Baes

. J. Ball

F. Bauman
E. Beall

S. Bettis
Blumberg

G. Bohlmann
Boyd
Briggs
Chandler
Compere
Cook
Cottrell
Crowley
Culler
Ditto
Eatherly
Engel
Ferguson
Ferris
Fraas
Franzreb
Fry
Gabbard
Gallaher
Grimes
Grindell
Guymon
Harley
Haubenreich
. Hightower
outzeel
Hudson

. Kasten
Kedl
Kerlin
Kerr
Kirslis

- - .

-

gnf-az*.q:n.r!:r::uzmmg‘);owmzm*uzw:ufum_tflpw;np.zwm

21

Internal Distribution

b1,
42,
L3,
Ly,
45,
L6,
L.
48,
49,
50-51.
52,
53.
5k,
55.
56.
5T
58.

60.

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>z.zc4§)w>mm>c+bmraommwwz

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UQEURYNGYIHO >R

ORNL-TM-2685

A. I. Krakoviak
T. S. Kress
R. B. Lindauer

Tundin

Lyon
MacPherson
. McCoy
McCurdy
McGlothlan
. McIntosh, AEC Wash,
McIlain
McNeese
McWherter
Miller
Moore
Nicholson
Perry
Prince
Ragan

. Redford
Richardson

- -

PFHEHERECOESOTEHES OQOEHEZ

. W. Rosenthal
. W, Savolainen

nlap Scott

J. Skinner

N. Smith

L. Smith
Spiewak

C. Steffy, Jr.
A, Sundberg

. R. Tallackson

E. Thoma
. Trauger
. West

Whatley
. Whitman

Q
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s
d.
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Gale Young
22

Internal Distribution (continued)

91-92, Central Research Library (CRL)
93-94.  Y-12 Document Reference Section (DRS)
94-97. lLaboratory Records Department (LRD)
98. Laboratory Records Department -Record Copy (LRD-RC)
99, ORNL Patent Office.
100.  Nuclear Safety Information Center (NSIC)

External Distribution

101-115. Division of Technical Information Extension (DTIE)
116. lLaboratory and University Division, ORO