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COPY NO.
- 375
INHERENT NEUTFt3N SOURCES I N CLEAN MSRE FUEL SALT
\mile
Price $
/,
1
P . N . Haubenreich
P
ashington
25, D. C.
ABSTRACT Unirradiated MSRE fuel salt w i l l contain an appreciable neutron source due t o spontaneous f i s s i o n of t h e uranium, and (a,n ) r e a c t i o n s of . alpha p a r t i c l e s from t h e uranium with t h e f l u o r i n e and beryllium o f t h e s a l t . The spontaneous f i s s i o n source i n t h e core (25 rt3o f s a l t ) i s lo3 neutrons/sec or less, mostly from U238. The alpha-n source i s much larger, giving about 4 x lo5 neutrons/sec. i n t h e c o r e . Nearly a l l of t h i s l a t t e r source i s caused by alpha p a r t i c l e s from I
.
$34.
NOTICE P-
'c
i;
This document contains information of a preliminary nature and was prepared primwily for internal use at the Oak Ridge National Laboratory. It i s rub'ect t o n v i s i o n 01 corroctiun ond therefore doer not represent a final report. \he information i s not to ba abstracted, reprinted 01 otherwise givon public dirsominotion without the approval of the ORNL potent branch, Legal and Information Control Department.
Jj
P t
LEGAL NOTICE T h i s report was prepared as on account of Government sponsored work.
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3 CONTENTS Page Introduction---------------------------------------------------&el
Composition------------------------------------------------
Spontaneous Neutrons
Fission from
Alpha Alpha-n
Neutrons------------------------------------
Alpha-n &ission Yields
Reacticns--------------------------------by
Uraniu.m----------------------------------
in
Fuel
Salt--------------------------------
Discussion--------------------------------------------------------
Appendix - Calculation of Alpha-n Yields in MSRE Fuel Salt-----Dilution by Non-Productive Constituents of Fuel Salt------(a, n) Yield------------------------------------------Be9 ~19 (a, n) Yield------------------------------------------Li7 (a, n) Yield------------------------------------------References--------------------------------------------------------
5 5 6 7 8 9 10 12 12
14 14 14 16
V
5 Introduction When a r e a m o r i s s u b c r i t i c a l , t h e f i s s i o n r a t e and t h e neutron f l u x depend on t h e s t r e n g t h of t h e neutron source i n t h e r e a c t o r due t o various r e a c t i o n s and t h e m u l t i p l i c a t i o n of t h e s e source neutrons by f i s s i o n s i n the core.
By supporting t h e f i s s i o n r a t e i n t h e s u b c r i t i c a l r e a c t o r a t
s u f f i c i e n t l y high l e v e l s , a source performs s e v e r a l functions i n r e a c t o r operations.
The source s t r e n g t h r e q u i r e d f o r some functions i s higher
than f o r o t h e r s . I n a l l r e a c t o r f u e l s t h e r e i s always a source of neutrons due t o spontaneous f i s s i o n , b u t t h i s source i s r e l a t i v e l y weak, p a r t i c u l a r l y i f t h e f u e l i s h i g h l y enriched uranium.
I n many r e a c t o r cores t h e r e i s a l s o
an inherent photoneutron source produced by i n t e r a c t i o n of gamma r a y s with deuterium or beryllium i n t h e c o r e .
This source i s u s u a l l y not s i g n i f i c a n t ,
however, u n t i l a f t e r f i s s i o n product gamma sources have been b u i l t up by power o p e r a t i o n .
Therefore, i n n e a r l y a l l r e a c t o r s an extraneous neutron
source i s i n s e r t e d i n or near t h e c o r e .
The MSRE i s unusual i n t h a t t h e
f u e l i s a homogeneous fused s a l t i n which alpha-emitting uranium i s i n t i mately d i s p e r s e d with l a r g e q u a n t i t i e s of f l u o r i n e and beryllium,both of which r e a d i l y undergo alpha-n r e a c t i o n s .
Thus t h e r e i s a s t r o n g alpha-n
source inherent i n t h e MSRE f u e l s a l t even b e f o r e it has been i r r a d i a t e d .
It i s conceivable t h a t the source inherent i n t h e MSRE f u e l s a l t , which i s c e r t a i n t o be p r e s e n t whenever t h e r e i s any chance of c r i t i c a l i t y , i s s t r o n g enough t o s a t i s f y some, i f not a l l , of t h e requirements which
make an extraneous source necessary i n most r e a c t o r s .
I n determining
whether o r not a n extraneous source i s required, it i s necessary t o p r e d i c t t h e s t r e n g t h of t h e neutron source which i s inherent i n t h e clean
s l a t b e f o r e t h e photoneutron source becomes important. describes t h i s prediction.
The p r e s e n t repOrt
The question of source requirements w i l l be
considered l a t e r , i n a s e p a r a t e r e p o r t .
Fuel
Compos it ion
The s t r e n g t h o f t h e inherent neutron source depends on t h e composition I
of t h e f u e l s a l t and t h e iso-;opic composition of t h e uranium i n i t .
Three
6 f u e l s a l t s , with compositions shown i n Table 1, have been considered f o r t h e MSRE. Uranium concentrations shown are f o r t h e i n i t i a l c r i t i c a l experiment. For power o p e r a t i o n t h e uranium concentration w i l l be h i g h e r by about
17%
( t o compensate f o r c o n t r o l rods, xenon and o t h e r p o i s o n s ) . The i s o t o p i c compositions shown for t h e uranium are t h e values used i n the c r i t i c a l i t y calculations.
The U234
and U236
f r a c t i o n s a r e based on
t y p i c a l analyses o f uranium enriched i n t h e d i f f u s i o n p l a n t t o t h e i n d i c a t e d
u~~~ c o n t e n t . The l i t h i u m composition i s t h a t of l i t h i u m a c t u a l l y on hand f o r f u e l
s a l t manufacture.
Table 1. Fuel S a l t Compositions
A
Fuel Type
B
C
-
~~~
S a l t comp: (mole $)
65
70
1 0 -313
0
0
0.189
o .831
u238
3
5
64.4
144.3
134.5
142.7
Density a t 1200 OF (l b / f t 3 ) a
66.8
L i p BeF2 ZrF4 mF4 UF4
99.9926 5
23.7 3
29 4
29.2 5
Li7.
Spontaneous F i s s i o n Neutrons The r a t e o f neutron production by spontaneous f i s s i o n i s a s p e c i f i c p r o p e r t y o f t h e each n u c l i d e .
I n t h e c l e a n MSRE f u e l , U238 has t h e s h o r t e s t
h a l f - l i f e f o r spontaneous f i s s i 0 n . l
(See Table 2.)
-.
7
.
Table 2 . Neutron Production by Spo:Ttaneous F i s s i o n of 1Jranium Isotopes
w
S p e c i f i c Emission Rate sed (n/kg
-
Isotope u234
6.1
u235 ~236
0.31
5.1
15.2
~238
The e f f e c t i v e core of t h e E R E ( t h e graphite-containing region p l u s some of t h e f u e l i n t h e upper and lower heads ) contains 25 ft3 o f f u e l s a l t . The amounts o f each uranium isotope and t h e spontaneous f i s s i o n neutron source i n t h i s volume a r e gi\ren i n Table 3 for each of t h e t h r e e f u e l s described i n Table 1. Table 3 .
.
Spontaneous F i s s i o n Neutron Source i n Core Fuel B
Fuel A
u234 u23 5 u238 u238
0 -3 27 .o 0.3 1.5
Fuel C
2
0.2
1
0.2
1
14
16.5
8
26.4
13
2
0.2
1
0.9
1 13
0.2
22
47.5
722
-
4
23
40
-
737
Neutrons From Alpha-n Reactions Energetic alpha partic1Les can produce neutrons by nuclear i n t e r a c t ions with s e v e r a l d i f f e r e n t n u c l i d e s . upon t h e n u c l i d e .
Threshold energies vary widely, depending
Three nuc:Lides, Li7,
Be9 and F19, have a-n t h r e s h o l d s
below t h e maximum energy o f alphas from uranium.
The neutron y i e l d p e r
alpha p a r t i c l e i s a f u n c t i o n o f t h e i n i t i a l energy o f t h e alpha p a r t i c l e . I
and t h e composition o f t h e medium i n which it i s slowing down.
8 Alpha Emission by Uranium Among t h e uranium isotopes present i n f r e s h MSRE f u e l , U234 has by f a r t h e h i g h e s t s p e c i f i c alpha emission and a l s o emits t h e highest-energy The s p e c i f i c alpha sources a r e summarized i n Table 4 . l
alpha p a r t i c l e s .
Table 5 gives t h e t o t a l alpha source i n t h e e f f e c t i v e core of t h e MSRE during t h e i n i t i a l c r i t i c a l experiment (25 f't3 o f s a l t , containing t h e amounts of uranium shown i n Table 3 ) Table 4.
.
Alpha Emission by Uranium
Half-Life f o r a-decay ( y )
Decay Rate (dis/sec-kg)
u234
2.48 x 105
2.28 x 1011
u23 5
7.13 x
Isotope
E, (Mev)
f a Source (all00 d i s . ) (a/sec-kg)
~~~~~
lo8 io7
u236
2.39 x
u238
4.51 x 109
Note:
E
a
7.90 x lo7
2.35
109
1.23 x
io7
4.77 4.72 4.58 4.47 4.40 4.20 4.y 4.45 4.19 4.15
1.64 0.64 0.79 0.24 6.56
72 28 10
3 83 4 73 27 77
x 10l1
x 10l1 x io7 x io7 x io7
io7
0.32
1.72 o .63 0.95 o .28
23
x 109 x 109 x lo6 x io6
i s t h e i n i t i a l energy o f t h e alpha p a r t i c l e and f i s
t h e percentage y i e l d of alphas of t h a t energy i n t h e n a t u r a l alpha decay of t h e n u c l i d e . Table 5.
Alpha Source i n MSRE Core Source S t r e n g t h (a/sec )
Isotope u234 u23 5
~236
u238
Ea (Mev)
4.77 4.72 4.9 4 -47 4.40 4.20 4.Y 4.45 4.19 4.15
Fuel A
Fuel B
Fuel C
lolo io10 2.13 x io8 0.65 x lo8 17.7 x lo8 0.86 x lo8 5.01 x io8
2.90 x 1O1O 1.13 x i o 1 O
3.71 x l o l o 1.45 x ioxo 2.09 x io8 0.63 x lo8 17.3 x lo8 0.85 x 1 0 ' 3.92 x lo8 1.44 x io8 0.45 x lo8 0.13 x 1 0 '
4.74 x 1.85 x
1.83 x io8 1.40 x lo6 0.41 x lo6
1.30 x
io8
0.40 x 10' 10.8 x lo8 0.53 x lo8 3.06 x lo8 1.12 x io8 0.86 x lo6 0.25 x lo6
a
9 w
Alpha-n Yields i n Fuel S a l t
F19, and L i 7 vary with t h e energy of
The y i e l d s of neutrons from Be', 6
t h e a l p h a - p a r t i c l e , g e n e r a l l y increasing with energy.
Yields f o r 4.77-Mev
a l p h a - p a r t i c l e s i n t h i c k t a r g e t s of pure m a t e r i a l a r e 40,
6 and 0 . 1 neutrons
per m i l l i o n a l p h a - p a r t i c l e s i n beryllium, f l u o r i n e , and lithium-7, respectively.
I n t h e MSRE f i e 1 sa:Lt, t h e productive nuclides comprise only a
f r a c t i o n of t h e t o t a l , and t h e y i e l d is a f f e c t e d by t h e d i l u t i o n with o t h e r elements.
Yields f o r a l p h a - p a r t i c l e s o f each energy i n Table 5, i n each
of t h r e e f i e 1 s a l t s , were c a l c u l a t e d by procedures described i n t h e Appendix. Table 6 i l l u s t r a t e s how beryUium, f l u o r i n e , and l i t h i u m c o n t r i b u t e t o t h e t o t a l y i e l d for t h e most numerous and highest-energy group of alpha p a r t i cles. Table 6. Neutron Yields f o r 4.77-Mev Alpha P a r t i c l e s i n MSRE Fuel S a l t Yield (n/lo6/cx) Constituent Fuel A
Fuel B
Fuel C
Be F Li
2.65
3.32
3.20
4.36
4.44
4.40
0.02
0.02
0.02
Total
7 -03
7.78
7.62
Table 7 g i v e s t h e neutron source i n t h e e f f e c t i v e core o f t h e MSRF: when t h e uranium concentration i s a t i t s i n i t i a l , clean, c r i t i c a l v a l u e .
r-
10
7.
Table
Alpha-nNeutron
Source i n Core D
Alpha Source u234 ~ 2 3 5
~ 2 3 6
u238
Neutron Source S t r e n g t h (n/sec )
Ea (MeV) Fuel A
4 -77 4.72 4.3 4.47 4.40 4.20 4.Y 4.45 4.19 4.15
Total
3.33 x 1.21 x
1.14 x 0.30 x 7 3x
o .28
x
2.45 x o .83 x 4.36
io5 io5 io3 io3 io3 io3 io3 io3
1.21
4.67
105
Fuel B
Re1 C
2.26 x 105 0.82 x 0.78 x 0.21 x
io5
io3
io3 io3 0.20 x io3 1.68 io3 5.20 x
0.57 io3 3 -07 o .85 3.17 x
io5
x io5 x io5 x io3 x io3 x io3 x io3 2.10 x io3 0 . 7 2 x io3
2.83 1.03 1.23 0.32 8.13 0.31
157 43 3.99 x
io5
Discussion The c a l c u l a t i o n s i n d i c a t e t h a t t h e bulk of t h e neutron source i n h e r e n t i n t h e clean MSRF: f i e 1 i s due t o alpha-n r e a c t i o n s , with spontaneous f i s sion contributing r e l a t i v e l y l i t t l e .
Furthermore, about 97 p e r c e n t o f t h e
neutron source i s caused by alpha p a r t i c l e s from a s i n g l e i s o t o p e , U234, which comprises a v e r y small f r a c t i o n o f t h e t o t a l uranium.
Therefore, t h e
neutron source w i l l be xrery closel:; p r o p o r t i o n a l t o t h e U234
content of t h e
fuel salt. I n n a t u r a l uranium, t h e abundance of U234 i s only 0.0057%, or 0.0079 of t h e U235 abundance.
I n a gaseous d i f f u s i o n p l a n t , however, t h e U234/U235
r a t i o i s increased, so t h a t i n uranium containing over U234
/U235
9%U235
the
r a t i o i s 0.010 or above.
The U234 f r a c t i o n s which were used i n t h e c a l c u l a t i o n s are based on
t y p i c a l analyses of enriched uranium, and t h u s are o n l y estimates o f what w i l l appear i n uranium which w i l l be used i n making up t h e MSRE f u e l s a l t .
The estimate i s probably good t o w i t h i n +20% i n t h e case o f Fuels A and B,
which use h i g h l y enriched uranium.
I n t h e case of Fuel C it was assumed
t h a t t h e uranium would be taken from t h e d i f f u s i o n p l a n t a t about 35% U235,
-*
11
'br
and t h a t t h e U234 content would be only 0.30%. It now appears t h a t t h e uranium may be added t o t h e MSRE f i e 1 s a l t i n two batches:
the first of
n a t u r a l o r d e p l e t e d uranium; t h e second, h i g h l y enriched.
If t h i s course
i s followed, t h e U234 content of Fuel C would probably be higher, perhaps by as much as a f a c t o r of
1 . 4 . The neutron source for Fuel C would t h e n
be higher by t h e same f a c t o r .
12 APPENDIX
d
Calculation o f Alpha-n Yields i n MSRE Fuel S a l t Information on alpha-n y i e l d s from various nuclides u s u a l l y appears i n one of two forms:
1) t h e microscopic c r o s s s e c t i o n o f t h e nuclide f o r
t h e a-n r e a c t i o n as a function o f alpha energy, or 2 ) t h e y i e l d of neutrons p e r m i l l i o n alpha p a r t i c l e s o f a given i n i t i a l energy emitted i n an i n f i If t h e alpha p a r t i c l e s a r e emitted i n a
n i t e medium of t h e pure n u c l i d e .
mixture, it i s necessary t o t a k e i n t o account t h e d i l u t i o n of t h e product i v e nuclides by o t h e r s which only slow down t h e alpha p a r t i c l e s . Dilution bv Non-Productive Constituents of Fuel S a l t The c o r r e c t i o n f o r t h e d i l u t i o n o f a productive nuclide i n a mixture i s e s s e n t i a l l y t h e f r a c t i o n o f t h e alpha energy l o s s which i s a t t r i b u t a b l e
t o t h e productive c o n s t i t u e n t .
L e t nmax be t h e y i e l d of neutrons f o r alpha p a r t i c l e s emitted i n an i n f i n i t e medium c o n s i s t i n g e n t i r e l y of a productive n u c l i d e .
L e t n be t h e
y i e l d f o r t h a t nuclide i n a mixture.
that a fairly
It has been
good approximation i s n n max
-
N S P P EN. S i
i
i
where S i s t h e " r e l a t i v e atomic stopping power", N i s t h e number d e n s i t y of a nuclide, and p r e f e r s t o t h e productive c o n s t i t u e n t . The b e s t information on r e l a t i v e stopping powers i s s t i l l a a r t i c l e by Livingston and Bethe.*
1937
They g i v e S r e l a t i v e t o a i r f o r 16
elements f o r 6-Mev a l p h a - p a r t i c l e s and f o r 6 elements a t 7 o t h e r energies from 2 t o 52 Mev.
Table
8 gives values of
S f o r the constituents of the
MSRE f u e l s a l t , obtained by i n t e r p o l a t i o n i n energy and atomic number of t h e Livingston
-
Bethe d a t a .
The r e l a t i v e stopping powers i n t h i s t a b l e
a r e evaluated a t 4.5 MeV, because t h i s i s approximately t h e energy o f t h e uranium a l p h a - p a r t i c l e s
. P
,
Table 8 . Relative Stopping Power of Constituents of MSRE Fuel S a l t f o r 4.5-Mev Alpha P a r t i c l e s
Li Be F Zr Th U a
N S / ~ N si ~
sa
Constituent
Fuel A
Fuel B
Fuel C
0.159 0.085
1.19
0.163 0.068 0.692
2.8
0.077
0.707 0.047
3.9
0.016 0.005
0
0.149 0.082 0.699 0.056 0
0.003
0.014
0.57 0.70
4.2
Atomic stopping power r e l a t i v e t o a i r .
If t h e microscopic cross s e c t i o n of a nuclide f o r t h e alpha-n r e a c t i o n i s known, then t h e number o f neutrons produced by an alpha p a r t i c l e can be
found from5 EO
(2 1 0
Harris’ has presented d i f f e r e n t substances.
dE’-l
as a function o f alpha energy f o r s e v e r a l
For a mixture one may assume t h a t
- -dE1
e E)
i s t h e weight f r a c t i o n of c o n s t i t u e n t i i n t h e mixture. Table 9 i gives values o f - - f o r t h e c’onstituents o f t h e MSRE taken from r e f e r -
where u
ence 1, and t h e products o f t h i s q u a n t i t y and t h e weight f r a c t i o n s f o r the three different fuel salts.
--’mix
E ‘
dx
f o r each s a l t .
The sum a t t h e bottom of each column i s
14 9.
Table
Slowing-Down Parameters for Alpha P a r t i c l e s
i
Fuel B
Fuel A
Fuel C
5 Mev
4 Mev
4 Mev 5 MeV 4 Mev 5 Mev 4 Mev 5 Mev ~~~
~
Th
885 840 730 385 228
781 741 645 351 208
U
222
202
Li Be
F zr
Be9
~~
(a, n )
91 38 419 38
108
95
43 474 42
57 513 37
Y
1-3
12
0
4 679
4
-
3
-
602
718
103
453 33
96 53 489 42
85 47 432 39
0 2
0 10
0
633
69
9 612
Yield
For alpha p a r t i c l e s with an i n i t i a l energy Eo, emitted i n pure b e r y l 1ium2 , nmax
=
0 .l52
For libels A, B, and C , n/nmu (See Table
8).
The product
is
neutrons/lO'
(41
a
0.068,0.085 and 0.082r e s p e c t i v e l y .
i s t h e y i e l d i n t h e f i e 1 s a l t which i s tabu-
l a t e d i n Table 10.
F19 (a, n ) Yield Segre and Wiegand' measured neutron y i e l d s f o r alpha p a r t i c l e s of begins t o be various energies i n t h i c k t a r g e t s o f F. The y i e l d , n max ' measurable a t 3 Mev and r i s e s t o 10 neutrons/lO' alphas a t 5.3 MeV. From Table
8, n/n-
f o r Fuels A, B and C a r e
duct of t h i s n/nm
and n
maX
0.692,0.705 and 0.699.
The pro-
from t h e d a t a of Segre and Wiegand i s given
i n Table 10. Li7
( a , n ) Yield
f o r L i 7 . The This w r i t e r knows o f no d i r e c t measurements o f n max c r o s s - s e c t i o n f o r t h e L i 7 ( a , n ) BIO r e a c t i o n as a f'unction of alpha energy
-.
was c a l c u l a t e d and reported by Hess cross-section r i s e s t o higher e n e r g i e s .
8 mb
iat
.
Above a t h r e s h o l d a t
4.8 MeV,
4.36 MeV,
the
t h e n decreases t o about 6 mb a t
Hess' c r o s s - s e c t i o n w a s used t o compute y i e l d s from L i 7
i n t h e f i e 1 s a l t from equations (2) and
(3). The i n t e g r a l i n E q . (2) was
evaluated by r e p r e s e n t i n g t h e cross s e c t i o n curve by s t r a i g h t - l i n e segments and by approximating points at
4
v s . E by l i n e a r r e l a t i o n s f i t t e d t o 'mix and 5 Mev given in Table 9. Results appear i n Table 10.
Table 10.
.
4.77 4.72 4.58 4.50 4.47 4.45 4.40 4.20 4.19 4.15
Neutron Yields f o r Alpha P a r t i c l e s i n E R E f i t s 1 S a l t (n/@ a)
Fuel A
E,
(MeV )
E)'
(-
Be
F
Fuel B
Li
2.65 4.36 0.0153 2.60 3.94 0.016 2.31 3.04 0.008 2.18 2.70 0.00:s 2.12 2.56 o.oo:{ 2.10 2.42 0.002
Be
F
Fuel C
Li
Be
F
Li
3.32 4.44 0.019 3.20 4.40 o .017 3 -25 4.02 o .016 3.13 3.98 0.014 2.89 3 .io o .008 2.79 3.08 o .007
2.72 2.65 2.63 2.01 2.25 0.00:- 2.52 1.70 1.52 o 2.13 2.og 1.67 1.45 0 1.31 o 1.63 2.Ob
0.004 2.61 o ,003 2.47 0.002 2.29 0.001 1-55 0 1.48 0 1.34 0 2-75
2.62 2.56 2.53 2.43
2.73 2.59 2.45 2.27
0.004
o .003 0.002 0 .ooo 0 2.05 1.9 2.02 1.47 0 1.97 1.33
0
16
Re f e r enc e s
1. D . R . Harris, "Calculation of t h e Background Neutron Source i n New, Uranium-Fueled Reactors," USAEC Report WAPD-TM-220, Power Laboratory, March
B e t t i s Atomic
1960.
2. A. 0 . Hansen, "Radioactive Neutron Sources," p . 3 i n Fast Neutron Physics, P a r t 1, ed. by J . B. Marion and J . L. Fowler, I n t e r s c i e n c e , New
York, 1960.
3.
0 . J . C . Runnalls and R . R . Boucher, "Neutron Yields f'rom
Actinide -Beryllium Alloys
4. M. S. Livingston Revs. Mod. Phys., 9:
,
It
, 34 :
Can. J . Phys .
and H. S . Bethe,
949 (1956 ) .
"Nuclear Dynamics, Experimental,"
272 (1937).
5. W. N . Hess, "Neutrons f'rom (a, n ) S o u r c e s J f fAnnals o f Phys., 2: 117-133 0959 1* 6 . E. Segre and C . Wiegand, "Thick-Target E x c i t a t i o n Functions for Alpha P a r t i c l e s , " USAEC Report LA-136, Los Alamos S c i e n t i f i c Laboratory, September
1944
( a l s o issued as USAEC Report
MDDC-185).
Internal Distribution
1. MSRP D i r e c t o r ' s Office, 2.
3.
4. 5. 6. 7. 8. 910* 11.
12.
13 *
14. 15-19. 20. 21. 22.
Rm. 219, 9204-1 S . E . Beall M. Bender E. S. Bettis A . L. Boch F. F. Blankenship H . C . Claiborne S. J. Ditto J . R . Engel E . P . Epler A. P . Fraas W. R. Grimes R . H. Guymon S . H . Hanauer P . N . Haubenreich R. B. Lindauer R . N . Lyon H . G . MacPherson
W . B. McDonald H . F. McDuffie 25* A. J . Miller 26. R . L. Moore 27* A. M . P e r r y 28. B. E . P r i n c e 29* H. W . Savage 30. D . S c o t t 31. J . H . S h a f f e r 32. M. J . Skinner 33. I . Spiewak 34. A . Taboada 35 J . R . Tallackson 36. D . B . Trauger 37-38. C e n t r a l Research Library 39-40. Document Reference Seet ion 41-42. Reactor Division Library 43745. Laboratory Records 46. ORNL-RC
23 *
24.
-
Ekt e r n a l D i s t r i b u t i o n -
47-48. 49.
D . F. Cope, Reactor DS-vision, AEC, OR0
52. 53-67.
M . J . Whitman, Division o f Reactor Development, AEC, Washington Division o f Technical Information Extension, AEC (DTIE)
R . L. Philippone, Reactor Division, AEC, OR0 p . H. M. Roth, Division o f Research and Development, AEC, OR0 51. W . L. Smalley, Reactoy Division, AEC, OR0