ML20138L512
| ML20138L512 | |
| Person / Time | |
|---|---|
| Site: | University of Michigan |
| Issue date: | 06/12/1978 |
| From: | Kerr W MICHIGAN, UNIV. OF, ANN ARBOR, MI |
| To: | Reid R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20138L505 | List:
|
| References | |
| FOIA-85-587 NUDOCS 8512190290 | |
| Download: ML20138L512 (18) | |
Text
.
i
. c n;
%:. 2-
=-
- w o a :.x._..w ;O a.
.w:
l 9
!.?
?
b -lC ~)$
r% a 4 ;
THE UNIVERSITY OF MICHIGAN
[
(
'3 MICHIGAN MEMORI AL-PHOENIX f*ROJECT 3
8
,y f
CPP.C. OFT.Ot CTO.
P.co.NtX
.-O...L L..O..TO.7 y;
,,.c.
Po.
i i j; i
- ,e Docket No. 50-2 June 12,1978 j
License R-28 a
United States Nuclear Regulatory Commission ATTN: Mr. Robert W. Reid
?
Operating Reactors Branch #4 i j Division of Operating Reactors i
Washington, D. C.
20555 il i
.jj Gentlemen:
1.i ji This letter provides additional information concerning the University of Michigan's.
j reactor license amendment request to utilize aluminide and oxide fuels.
t,
y {
1.
The answers to 13 questions posed by the Nuclear Regulatory Commission concerning i !
aluminide and oxide fuel utilization are enclosed as R5PONSES TO QUESTIONS 3j CONTAINED IN INFORMATION ON THE UNIVERS'ITY OF MICHIGAN i !
REQUEST THAT THE FORD NUCLEAR REACTOR LICENSE NO. R-28 BE
', l AMENDED TO UTILIZE ALUMINIDE AND OXIDE FUELS.
j.
j j 2.
An amended SAFETY ANALYSIS is enclosed. Revisions are dated in the margins j
and underlined. None of the revisions are significant. Those dated 10/77 were j
made after a telephone discussion of the analysis with your consultant to correct
]'
minor mathematical errors. Revisions dated 4/77 were inserted after measurements ij were made and a peak / average flux ratio of 1.86 was calculated. A value of J
1.5 had been used in the original analysis.
1 y
3.
Insufficient data about UO was provided in the SAFETY ANALYSIS. Since 2
)
its use is not anticipated, the request to utiliza UOn is withdrawn. The proposed j,
revised section 5.2 of Technical Specifications shou'd read as shown below.
j Changes to the present section 5.2 are underlined, i.
5.2 REACTOR FUEL A!
ql a.
The fuel elements shall be of the MTR type, consisting of piates 4'
containing enriched uranium - aluminum alloy, uranium aluminide (UAl,UAl 2
1 aluminum. 3, UAl ),r uranium oxide (U 0 ) I"'I CI"d *I'h 4
38
]~
There shall be 18 fuel plates containing 140 (* 2%)
h grams of uranium-235 in standard fuel elements and nine fuel plates
]':
containing 70 (* 2%) grams of uranium-235 in control rod fuel elements.
Pertially loaded fuel elements in which some plates do not contain 3j uranium may be used. Elements containing uranium-aluminum j!
alloy, uranium aluminide, and uranium oxide may be intermixed i
within the core.
1 sj 4.
A fission density limit has not been set for FNR fuel in Technical Specifications, if :f
~
I) )
8512190290 851203 PDR FOIA AFTERGO85-587 PDR i
^.mmmm
,,_m.,--.n.
a
w w _/ a; whip-w.:, - ~ ~ ;_., m.,. z ag... _
3,;,:gg
~
.y n
..t l
3, J
J. i d.,.
. '3 1
j i If a fission density limit is required, it could be added as section 5.2.c. The i
a single value of 1.8 X 10 guested in the amendment have been reduc limits on fission densities r
'lf 2
fissions /cc for all types of fuel. This value is
[j -
equal to or below operational fission density limits already in use at other reactors.
p g
Operational Fission a c, Fuel Type Reactor Density (Fissions /cc) i) ]'
21 Uranium-Aluminum General Electric 2.0 X 10
[i Alloy Test Reactor (GETR) 21 i :
Uranium Aluminide Advance Test 1.8 X 10 1)
(UAl, UAl ' AI }
- *# I^
2 3
4 21 U!
Uranium Oxide High Flux isotope 1.9 X 10
(') j (U3 8}
h{
The reduction has been made to expedite approv I of the amendment and because S
It is not anticipated that a level above 1.8 X 1 fissions /cc will ever be j
required at the Ford Nuclear Reactor, I
c.
The fision density limit shall be 1.8 X 10 fissions /ce.
.' l Your prompt attention to the amendment request is requested in order that we
.l i may continue reactor operation. The reactor has seven fuel elements remaining, j,
enough to last through September,1978. The original amendment was submitted m June, 1977.
j The amendment is requested to utilize fuel types which were developed to provide q
an improvement over uranium - aluminum alloy ~ fuels and which have been operationally 5
proven in reactors which operate at much higher thermal power densities, fission 4.
densities, heat fluxes, flow rates,. and temperatures than exist in the Ford Nuclear Reactor.
3 j;
Sincerely, g
ji NI&m/6 i
William Kerr l
1-Director d<
i, Enclosures
(
WK/RRB/at Subscribed to and sworn to before me this /5 day of 19 ff g;
..e,.
7
,-i
,.., e ' ' ' ' / at Ann Arbor, Michigan, County of Washtenaw Y
lll
. l.
- 8) j
\\.,; e,
< l seta,x Fubii.
y 3n 2.a._
~ -
(
c!
'.. j., g W i
- .:.. m"-... :~...u =
.p b
k r
_- t a.,,:; e:x a.a.
c,_.___
1 b
.i
- l t
1 4
t i.i.
1 i
${1 i
i j i RESPONSES TO QUESTIONS CONTAINED
.1
-5
~i IN
'.1 i
!i. 4 INFORMATION ON THE UNIVERSITY OF MICHIGAN e
i REQU5T THAT THE FORD NUCLEAR REACTOR c
i LICENSE NO. R-28 BE AMENDED TO UTILIZE i
1 ALUMINIDE AND OXIDE FUELS t
t 3
t
+
k
.,3 s
-i
't ll DOCKET NO.: 50-2
- t
.t t
i s
r
.i.
4
- 1 1
t t
4 1
1 I
a 5
i
- s
- 1 e
.11 g
June,1978 i!,
i-Ti t
?
?
t
+
L d
aux.
w,., =, _..,e
\\,_2...;. _.,_c.h.5 g 3_A J ~.
di.\\
d;
,:Jj '
1.
What maximum stress level will the FNR elements be operated ot? Compare jj this with the aluminide and oxide core stress limit at which plastic deformation /
6 j l failure occurs and show that the strength reduction of the aluminide and oxide 3,. j core is not 3 serious analytical consideration.
1 Qb; j
RESPONSE
- . i a4 Maximum stress levels in the FNR fuel elements are not known. However, FNR elements ore operated under minimal thermal and hydraulic stress levels. Peak fuel plate temperature is approximately 156 F. Coolant velocity is approximately 2.5 ft/sec which is almost
- i laminor.
Fission product and thermal swelling do not cause odiacent elements to come into contact d!
so stresses due to these phenomena are not produced. The minimum radial clearance
"' {
available to each element in the reactor core grid configuration is.029 inches which
)!
represents 0.9% of the corresponding radial dimension. Tests of 19 Advance Test Reactor 20 j
(ATR) aluminide fuel element plates at fission fissions /cc resulted in average radial swellings of 0.13%qensities up to 11.0 X 10 j
This represents only 14% of the minimum j
clearance available in the FNR core. FNR elements are unrestricted in the axial -
direction.
The ATR o ed buckling in fuel plates which had been operated to fission densities of 2.0 X 1 fissions /cc and s+ ding core swelling levels of 13%AV/V, measured j
as on increase in plate t,icknen. The plate length change corresponding to 13% AV/V 1
is approximately 0.35 ' Figure 1 in the amendment SAFETY ANAL.YSIS shows that 1
core swelling is I than 13% AV/V for alloy fuel up to the requested fission density limit of 1.8 X 1 fissions /cc. Since alloy fuel has no inherent voids, it provides the worst fission density-swelling combination possible. It is our contention that any rj voids will tend to reduce swelling.
<j-In the ATR, the buckling which was observed was in pe axial direction along the plate 2
j length, it occurred at a fission density of 2.0 X 10 fissions /ce. Up to the point j
of buckling, no plate failures were observed.
-i Thus, even though strength is somewhat reduced in aluminide and oxide cores, it is j
not a serious analytical consideration at fission densities up to the limit requested, c.]
1.8 X 1021 fissions /ce, and under FNR thermal / hydraulic conditions.
rt q
(a y
- }
a 3,l A,
I J
l
!i e
1.2. avl;.;.. '..;..,.,a.m -
- a..
_2:
e.,
..C...
.,_,_y..
- _~
..==s--.
..n: ~.
- j.,,, g,,.
i t
4 4
3-2-
1 2.
Does the phrase "... the some temperature..." in the FNR statement, M
it "In any case, it is the disruption of the fuel clad which could result in the
)~.:
release of fission products, which would occur at about the some temperature for any type of fuel.", refer to 1200 F which is the melting point of the j
Al clad (as well as the Al in the matrix mixture)?
3 Ni
RESPONSE
d, Yes, "the some temperature" refers to the 1200 F mel' ting point of aluminum.
o i'
I!
d 5
f 6
g
.4
't
[.ti ri t!
1
-t b
a
.0 1
.s o
9
.]
N 1
} :5y%=.;a..;3e,4 i.
J A.
- u. :
y g.7 y-, +
e 3
.g g
- i I il 4
A j
3.
Provide the basis for concluding that the " curved geometry" of the FNR j
j fuel is more stable, and; therefore, a 7% AV/V would not result in any i
failure mode.
4j-..:
3
RESPONSE
in flat plates, when buckling occun, it is possible for the clad and the fuel core to gj l
separate and for the clad to bow in one direction while the core buckles in the ij opposite direction.
sij In curved plates, the clad and fuel core will tend to bend in the direction of curvature. The probability of clad-core separation and subsequent warping due to d
fission product swelling is minimized. In this sense, curved geometry is more stable than flat geometry.
f,.
The conclusion concerning curved geometry refers to section 5.1 of the amendment
'j '
safety analysis. A 7% AV/V swelling limit was set for flat plate Engineering Test j
Reactor (ETR) fuel because some warping, though not failure, was periodically.
j observed in flat test plates irradiated to the 7% AV/V swelling level. ATR, which has curved fuel plates, has operated up to 13% AV/V swelling before buckling was observed. The buckling observed was along the plate length rather than along the plate width, which is the ' irection of curvature. This operational experience seems d
,j,
to confirm the increased stability of curved geometry.
j s
')
i
-j Sa 1l'l.
Q 9
b Y
k I
H 3
Y
'J q,
i L:
.i 1, aI
'~ y g4. 4.....,.
- 3 h,_,,
.a,,L,'3b 2 2 -.:.&:,h; %.a;;_.q.;". M 2?L.
.i i
t
,;)
- a ti
~r.
1 4.
Provide a comparison of the thermal-hydraulic parameters for the cases with j
and without a 20% change in the fuel thickness (.004 inches) and show Lj that there is no opemtionally significant change, et3i 1;:
RESPONSE
11 a
r A comparison of thermal-hydmulic parameters for cases with no swelling and a 20%
4 change in fuel core thickness follows. The format is in the some order as APPENDIX A to the safety analysis. Those parameters which change are denoted by on asterisk.
4 1
The peale/ average flux ratio in item A.3 was changed to a value of 1.86 basert on d;l measurements made in April,1978.
l The maximum fuel plate clad temperature actually decreases from 156.2 F to 154.6 F.
]!
The reason is that for an assumed constant flow rate of 980 gpm, coolant velocity L1 increases slightly as the not core flow area decreases. The subsequent heat transfer L]
coefficient increases, thermal resistance decreases, and clad temperature decreases.
i <
5:j; in any case, clad temperature change and thermal-hydmulic parameter changes are 1
very small for a 20% change in fuel core thickness.
1
(
I t
4 1'),
.,1 6
i, I
O s
t e
>~
ii.
1 1
1 j
o I
- I c.
I i i
,,... h.... f.- 1 1 3
d
"" ~ rz =:.:::..
.uxt;td;mac. int.c..anwn.
Y.
h 9
1 n
]
5-d 4
APPENDIX A a
l!
THERMAL-HYDRAULIC PARAMETERS
. 3. ;<
l i
20% Change in Fuel
~,j No Swelling Core Thickness (.004 in)
'r f
A.1 FUEL Pl. ATE PARANETERS 2
A.1.1 Surface Area (ft )
345 345 c,
A.1.2 Fuel Meat Volume (ft )
.290
.347*
3 Li 9
Core Meat Thickness (ft) 00167
.00200*
3 A. I.3 Core Volume (ft ) 2 3.125 3.125 1
A.1.4 Core Flow Area (ft )
884
.848*
Channel Thickness (ft)'
0098
.0094*
m
[}
A.2 REACTOR THERMAL POWER i
ff DENSITY (MW/l) 0226
.0226 A.3 REACTOR FISSION 20 20,
,jI DENSITY (fissions /cc) 5.44 X 10 4.55 X 10 j
Peck / Average Flux Ratio I.86 *
- 1.86* *
.],
A.4 FISSil.E MATERIAL DENSITY (ggcc)
.395
.395 (Fresh Fuel) j A.5 REACTOR HEAT FLUX (BTy/hr ft )
3.68 X 10' 3.68 X 10' a
j, A.6 COOL. ANT VELOCITY (ft/hr) 8,919 9,298*
3 3
3 A.7 REYNOLDS NUMBER 9J5 X 10 9.03 X 10,
t Li ;
Characteristic Dimersion (ft) 01 %
.0188*
A.8 PRANDTL NUMBER 3.15 3.15
', y '
A.9 NUSSELT NUMBER 53 53 E'
A.10 HEAT TRANSFER COEFFICIENT 1022 1063*
}
(BTly hr ft4 "F)
-4
-4
.4 A.11 THERMAL RESISTANCE 9.8 X 10 9.4 X 10 2
[]
(hr-ftz *F/ BTU) y A.12 MAXIMUM FUE Pl. ATE 156.0 15 4.6*
1 CLAD TEMPERATURE (oF) s,1
$'.d.
h.
- Parameterchangescaused by 20% change in fuel core thickness.
i3
- Measured value, April,1978.
g:.
L Ii i
i.
i
,..._f. (. ify;;; k, 3
._.=.,g.-_..
u
.n,wo. m.agu.,.,q., n 3,.,_,,,
.. a 4,..
J, a
- l
~t3
- iiIlj,
5.
On page 10, FNR stated that 19 ATR irradiated fuel plates had an average length ii.7 change of.22% caused by fission product swelling. For a 2-foot long fuel plate, I
this change would amount to on increased plate length of approximately.050".
f.3.i Provide the following information: (a) provide an analysis which would demonstrate il '
the obsence of any plate buckling. (b) if this analysis demonstrates there is buckling fi !
what is the maximum expected chan~nel closure for a plate length increase of lj :
.050"? (c) what would the thermal-hydraulic consequences be in such an eventuality?
g }i (d) dat would be the significant operational considerations for such a change?
li :l3j
RESPONSE
(a)
Experimental test r its at ATR showed that buckling did not occur below fission Q~l: j densities of 2.0 X 1 I fissions /cc and corresponding swelling levels of 13%
y) :;
AV/V. More recent operating experience that no buckling has occurred during to a swelling level of opproximately 14% AV/V.gons/cc which corresp operation up to fission densities of 2.2 X 1 fis f.
.i 1
i ll
't 4,.:
ll *
- f q,
s1;
.; +
..j r.
9}1
- f i
xl
.t 7:
1 3
N q:.,
.4 ijs !
M' Vi !
Nt
>? l
.f q n- =
..-.- ~
,; y y. -
\\
d'L.hitw%m :. 2"& a;z.;a n c.w ~ m '
- r-
- ~
'~
~.l. ;.T..
~
ro i
i
' i 6.
It is noted that the fission densi D) at which the.22% length change was
' j observed was in the 7 to 11 X 1 fiss/cc range. FNR is proposing a FD limit of 20 X 1020 fiss/ce. Provide the following information: (a) what would the i
expected length change be for this FD? (b) how would this amplify the buckling l
concern? (c) What would the thermal-hydraulic and operational consequences
- j be from this amplified condition?
i.
- c
' i'
RESPONSE
(a)
Similar to swelling, length change is a linear function of fission density. Assuming that the average length change of 0.22% corresponds to a fission density of 9.0 X 1020, a fission density of 2.0 X 1021 would produce a length change
}
of 0.49%.
(b)
Based upon information contained in the response to question 5, buckling would
[
j not be expected.
(c)
No significant thermal-hydraulic consequences would be expected.
R y:
E i
(,-
Lt
]
nu I
y G,
'b.
J4$
}'
a 1
l4 a
di si!
l:
i 1
t-
. v.?t
Q(._
.w r:...ac = - --,
_w_..,...----
3:v...
7 M
- j,
y
] \\j..
U[l 7.
What is the void fraction of the aluminide fuel FNR intends to use? How does this compare with the void fraction of fuel whose irradiation test data base FNR l
uses to support their analysis?
}b I
J j
RESPONSE
t
.I.
Fuel specifications were prepared in co-operation with ATR management. Void fraction 1
is not specified. ATR management does not specify void fraction for its fuel. A normal manufacturing void fraction of 3-4% occurs. As fuel loading increases, void fracthan increases.
1 The fuel whose irradiation tests form the basis for the FNR amendment request was 1
manufactured without a specified void content.
@n-j lt should be pointed out that alloy fuel has no voids so any voids in aluminide or oxide 1
i fuels are a bonus, i
v
-3
.0 l
- .1 4
.{ i q.
3 -i t
.I j11 b
a!
$I
,Il -
.1 J 4 i
,U
,$ E:
- -j
-l
,y
,t.
w
- 1. l Yk
,n.
"h
/;
24 i f
y,/
i 3
41Il 1,
il \\t I
-. ~
. g., :., - -- -..
_- ;-.; 3 ;3:
, 3, 3
l) y d,
8.
On page 10 pertaining to the XA 003F plate swelling data that is plotted q
in Figure 6, what is the explanation for sample 3-3's swelling exceeding
(;.j the theoretically predicted value of 6.38% AV/V per 1021 fissior/cc ?
,,.j, d$
RESPONSE
u Ti Tt.e report, reference 16 in the amendment safety analysis, does not provide a hj ~
statements, "It is evident, by comparing the theoretical potential for solid specific explanation for the behavior of sample 3-3. It does enake the general J
fission product growth with the results, that the accomodation of solid fission p
products by fabrication voids has not occurred to a great extent. This fact is
'i further evidenced by copious, large voids observed metallographically during post
]j Irrodiotion examination. The small extent of voidage utilization is probably the P
result of a relatively low opemting temperature."I el
.h it has been previously stated that FNR has not specified void content.
i r
.i I
a 4
4 e
1.
r
- k s
4 s
m 4
e
.l M
?*'
d-j>q.
Q E.
s
[8 y
- ! 9 G4
+j g
t l
-m
4
\\..
a..
..-..;...,,,...<_z t
a a
~:s
'd 9.
What are the differences in chemical environment between the FNR and those fj reactors in which the irradiation tests were performed that FNR uses in their
'i analysis? If there are differences, explain why these differences would not Q
enhance conditions that could cause fiol.ures of those plates containing aluminide tj or oxide fuels under the FNR operating conditions, w,
dd'
RESPONSE
??)
. 11 il Chemical conditions in the reactors in which the irradiation tests utilized in the m
. safety analysis were performed are more severe than FNR chen;icci conditions due
]G to higher tempwatures and flow rates.
Id In general, each reactor utilizes domineralized water as coolant. pH is maintained 4
in the 5-7 range to minimize corrosion. No additional chemical controls are maintained.
Table 1 of the safety analysis provides a comparison of parameters such as temperatures 9
and flow rates.
- j
.:l,
.}
. 'l d
..k.
.t.
l 14
- d
.t
- t ;
3:
i b
'i.
7.1 t
M f
em Q
3
.y,
't 1] e 11-
., -.x;.
- -.-...~.....,
..,.n. ~.,a, y,,.-. w.=..:..., _ _ _ _. ca.;.a.a s p-; y..
w- -
=;
ui a
d J
\\
. 'l l.
P 10 Since the operation of FNR, have there been any fuel plate failures related to reactor operation? If so, what was the cause of the failures?
.3)
.1 L4
'7 1)
RESPONSE
...i.;
There has been one pinhole look in one FNR fuel plate. The pinhole was a
[J manufacturing defect.
3.<
.i
- 1.,
4..
.i] :
y t _o.
j
~
g I
1 4
i.
f g
4 8
e 9
4 OIt u
t t!
d p
e
~8 3
4 n,
g
.~
64 h
j 9 h
n
.a
'1 4
f
., y.:
g.._. 9 v.,.;... m_. _..,.. 7,
-- - -. - ~
.. a. _.... -.
w...
....:..._:..~
.w=_
4 I
9 N4,..,.
1
'j i 11.
The FNR analysis is primarily based on fuel performance data from irradiation tests and reactor operation. Describe how FNR will assure that the fuel to be used in their reactor will have specifications comparable to the fuel used
}'
in the tests and other operating reactors.
!j i a;
RESPONSE
.,'j FNR is presently in the process of having aluminide fuel manufactured by Atomics International (Al). Atomics Intemational manufactures ATR fuel. FNR fuel drawings q.
and specifications were prepared by EG and G under DOE authority at the National q l' Reactor Test Site (NRTS), Idaho Falls, Idaho. The specifications were written by j
engineers and quality assurance personnel associated with ATR.
- 1 8,'
AI, EG and G, and FNR penonnel met for the final approval of the drawings and g -;
specifications. All revisions must be approved by EG and G.
a.
i l if oxide fuel were used at some future time, a similar situation would be set up with
^!
personnel from Brookhaven National Labo,oksy.
i s..
9*
L
.'4
- g. t.
-.'A i A'
M 0h '
. s T
v.'
t 5
h3
[.'
14 1 I t,
4 1
?55 l5 i i!
r.1 i
.i 1:
'. l
.y
- e7_sg.m{-.
me
.j,., - g. g, p.g, y.e.
- s.,. y m
...t..
,j
+
i
- A?;?
12.
Provide the basis from which it con be reasonably assumed that powder
~
metallurgy UO w uld have nearly the same swelling characteristics as U 0
- 2 38 u i.
i
",~ !
RESPONSE
c, lJJ,
Sufficient data on the swelling characteristics of UO w s not provided, in addition, 2
j j the hear transfer characteristics of UO are much poorer than those of UAi and U 0
- 2 x
38
/x.
Since the utilization of UO fuel is not anticipated, the request for its use is 2
<r withdrawn.
A
'i.t
.' i Ni 1
I i d
.i l
4 4
I
- i.
a S
'.i i
,I i
k f
,E,,
'y a
N so I
'; s 73
,- e
!?
1 1
3 :
.1.s
$}
/1 i 1
,! l i
i
- -c
- T ? h_ w ' dyb.:,I y.;. n. n -.
n
.W-i.7. t ~ < fR N O e J T.m_'W ' * <* f N"~ '.%,
-+ w m
,: g.,;
c;;w.: C-f7 v
m_.
8 t
e e
I y; r,.
'3 13 What is the fission density limit for alloy (U-AI) at the FNR?
Os
- .0 1 i d,
- 3
RESPONSE
N
.i
.j A fission density limit for alloy fuel at the FNR is not specified.
1 I.I $
.u...
fJ f s
~
1 l<
.4
."t.*. I.
- eg3 i
,,, ' 2
,q1
. i ],
a,
~.l
. I
,r s
s 4
i
.i s
%.t
'L
.,J, i
y e.
l A
1 t
f.
.=
.o:1,.
'gi N
b
^.
1 l
i t
'h f
I 'i
- ,a 6
k
- r 3
s_-__-,__
- ja -y _,.
- .e.
. n.nr.
,..,,v.; a a w. - <,,.,. -.., y.., ::-g: _:~ _< r ~
~
~, o Y:
.;o
- J.
d 1
^
r 41 l.-
REFERENCES i
A !
)
'.'! [
1.
M. Grober and R. Hobbins, " Irradiation Testing of Sample Fuel Plates to Very q;
High Burnups", INC-16-1, U. S. Atomic Energy Commission Report ANCR-1016, Lj l October,1971.
e
..,d q
j 2.
J. H..Crawford and M. C. Wittels, " Radiation Stability of Nonmetals and
[j
.i Ceramics", Proceedings of the Second United Nations international Conference on the Peaceful Uses of Atomic Energy, Geneva,1958, Vol. 5, United Nations, Geneva,1958, pp. 300-310.
- ,4 i
j i 3
Telecon with W. C. Francis, E. G. and G., NRTS, Idaho Falls, Idaho,
+ l May,1978.
M.
3 a
e 1
?
4 e
t I
,1 I
l i
.i l bl a.
'}
1 3
j t
I 1
- s si.
>}
ji 1l
/'4 ul; 1(i >
o d'
f
[$
.