ML20058K899
| ML20058K899 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 02/13/1973 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 9104260262 | |
| Download: ML20058K899 (17) | |
Text
e
, N!dery F1b C, a
MSIs NORTHERN STATES POWER COMPANY M I N N E A ? O L.I S. M E N N E
- OTA 95401 N
/
February 13,1973 temo g,
3 USnLC y
-j FEB 141973
- V
-n
~
,f-.
-O it-I Mr. A Ciambusso arcutM MT g'
f4 Deputy Director for Reactor Projects O-hh fy h
'8 b g[(N 9,
g Directorate of Licensing j
United States Atomic Enerty Commission s
Washington, D C 20545 g
e?
4
Dear Mr. Giambusso:
MONTICELLO NUCLFAR GENERATING PIANT Docket No. 50-263 License No. LPR-22 Supplemental Report of a Change in the Transient Analysis as Described in the FSAR On August 14, 1972 ve informed you that results cf an analysis of reactor transients for the end of cycle differed from those presented in the FSAR.
We reported that the analysis assumed control rod ceram times faster than required in our present Technical Specifications.
In our letter we noted that recently measured scram times were approximately half of those used in the analysis; continued operation was therefore justified. We have since analyzed our control rod drive system and its past performance.
We see no reason for significant increases in the attainable scram times in the future; therefore, we can accomodate revisions to our Technical Specifications to reflect the faster scram times used in the analysis.
Attached is a report prepamd by General Electric entitled "Results of Transient Reanalysis for Ibnticello Naclear Generating Plant with End-of-Cycle Core Dynamic Cnaracteristics. " Refinements in modeling the transient conditions have shown a shift in the scram reactivity feedback curve for the exposed core having all rods out.
Since we do not plan to extend the present operating cycle to the al3-rod-out condition, and because of the relatively low averaEe core exposure, we do not expect the FSAR analysis results to be exceeded during the present cycle.
The cycle subsequent to the spring, 1973 reload will begin with more control rods in the core; a scram from this condition vill fall within the FFAR transient analysis results.
'lhe results of the transient analysis are, therefore, not representative until Inter in the second cycle.
t 1084 9104260262 730213 y CF ADOCK 05000263 C F _L
NORT.
.'R N OTATED POWER CON ANY g
Mr. A Giambusso February 13, 1973 r
Technical Specification changes are recom:nended by General Electric in Section IV of the nctached report. We will in the near future, submit a request for the recom:aended Technical Specification changes, but in a forest and wording consistent with our current Technical Specifications.
The changes will in-clude revised scram times, required operability of the fourth safety / relief valve, and appropriate wording changes in the Bases section. Analytical studies of the effects of increased exposure on future cycles represents a continuing effort. We wil.1 advise you should these studies reflect a re-sponse in future cycles different from that described in the PSAR and the attached report.
Yours very truly,
.O.
L O Ihyer, P.E.
Director of Nuclear Support Serzices LOM/lGV/br ec: B E Grier G Cnarnoff Minnesota Pollution Control Agency Attn. Een Ihugan i
_w e
RESULTS OF TRANSIENT REANALYSES FOR MONTICELLO IUCLEAR GENERATING PLANT WITH ENT)-OF-CYCLE CORE DYNAMIC CHARACTERISTICS I.
Introduction A recent reanalysis has shown that a significant change in the shape of the scram reactivity curve could occur by the end of a fuel cycle (see fig.1).
As can be seen, even though the total scram reactivity has increased somewhat, the insertion rate is slower at the beginning of the stroke where the tran-sient analyses described in the FSAR for single-event caused abnormal occurrences could be affected. This, in turn, could affect design, operational, and safety provisions derived from these analyses for such things as the relief and safety valve capacities, set points, and Technical Specification requirements. For this reason, all transient analyses previously performed for Monticello have been reviewed, those which affect relief and safety valve settings and capacity have been redone, and necessary Technical Specification revisions detemined.
- 11. Conclusions and Recommendations j
When the analyses were completed, all the design guidelines, margins, cri-teria, etc. were met with the exception of the relief valve adequacy tran-l sient (turbine trip without bypass).
In this case, the margin from the peak of the pressure transient to the setting of the first safety valve was not sufficient to meet the General Electric recommended design guideline minimum of 25 psi.
In order to meet this guideline, a slight improvement in the control rod drive scram time was assumed, to that being used on the '67 and later GE-EKR Product Line plants (see Fig. 2). With the improved scram time assumption, the csiculated margin was 27 psi, which is acceptable.
In conjunction with this new assumption, the corresponding Technical Specification change is recommended and included later in this document.
Other appropriate proposed Technical Specification changes based on the results of the reanalysis are included at the end of this document, but do not involve any major changes in safety philosophy.
III. Discussion A.
Basis for Changes It has been recognized in the past that there could be substantial changes in axial reactivity characteristics with increased exposure which could affect the shape of the scram reactivity curve. However, it had been assumed that during most of the fuel cycle, enough stubbed (partially in-serted) rods were availabic to effect a fast scram reactivity rate at the beginning of rod stroke. On the other hand, even though all rods could be out at the end of the fuel cycle, flux would be peaked at the bottom of the core and this, combined with other exposed core characteristics was expected to le
=
J sufficient to still obtain a relatively fast scram reactivity rate at the beginning of rod stroke. Thus, the old curve was previously judged to be adequate to cover.the worst of these cases without being so extreme as to unduly penalize the plant.
Later generic development work was based on following the ibling principle I
which establishes an axial flux distribution which peaks flux toward the bottom of core early in the fuel cycle to reduce reactivity at that loca-tion while the control rods were still there to control it. This procedure prevents limiting flux peaks from developing towards the bottom of the
.i core when all rods are out and voids are reducing reactivity at the top of the core. This principle has been successfully applied to core manage-ment of earlier plants and reload applications.
With the rods all the way out (or the resulting tendency to all rods out/
in patterns) and a reduced flux peak at the bottom of the core, there was concern that the early part of the scram stroke might not be as effective, i.e., the scram reactivity insertion might be slower at the beginning of the rod stroke.
Information from operating plants has confirmed this ten-dency. Further, improved analytical capability allowed a more refined cal-culation of scram reactivity characteristics for expcsed cores.
B.
Assumptions Used in Reanalyses Because the new scram reactivity curve represents an end-of-cycle con-dition, the new analyses were also done with other inputs at end-of-cycle conditions for consistency and to ensure that a realistic worst case would be found between the two sets of conditions. For example, the void co-efficient is reduced at the end of the cycle end this will tend to reduce the peak of the pressurization transients.
Other conservative assumptions used in the original transient analyses, such as a multiplier og the void coefficient, and average control rod scram times equivalent to the Technical Specification limit, were also used in the reanalysis. Ibwever, when the reanalyses were completed, the mini-mum acceptable design margin from the peak of the pressure transient to the setting of the first safety valve could not be met on the turbine trip with-
~out bypass transient, (i.e., the General Electric recomended minimum de-sign margin for such transients is 25 psi). To meet this margin, it is pro-posed that slightly improved control rod drive scram times be assumed, the same as those used for later plants, such as Vermont Yankee and Browns Ferry.
The control rod drive equipment is the same at Monticello as at later plants and is easily capable of meeting the improved scram time requirement, as illustrated in Figure 2.
As can be seen, the improvement is in the early part of the scram stroke where it can be of most benefit to the results of the transient analyses.
Because it was undesirable to await another complete set of transient analyses, only the transients of most concern were redone. These were the "Ibrbine trip without bypass transient (identical to instantaneous loss of condenser
~
vacuum transient) for checking relief valve adequacy and the MSLIV valve closure with indirect scram for checking safety valve adequacy.) The resulting set of curves based on the different assumptions have been appropriately labeled 4
and are included herein.
Ibwever, the discussion only includes a statement t
t.
t abodtthemarginstheydemonstrate.
C.
Transients Not Reanalyzed The FSAR included about 20 analyses of wrst case abnormal transients i
in six categories of events. These categories are primary system pressure increases, moderator temperature decreases, reactivity insertions, core l
coolant inventory decreases, core coolant flow increase, and core coolant flow decreases. These were all reviewed to determine those which might be significantly affected by the new end-of-cycle core characteristics assumo-tions. The breakdown of categories, events, and logic for those in which only a review was deemed to be adequate, is sho n below.
Category Event Reason Reanalysis Not Needed i
Nuclear System All except those These events are less severe Pressure Increase identified in than those analyzed.
II,IB bbderator Temperature All All these events are less j
severe than those analyzed.
Decrease The only event of signifi-cance is the feedwater con-troller failure, maximum demand, which is terminated by the high level turbine trip resulting in a pressu-rization transient which is less severe than the turbine trip without bypass tran-sient analyzed herein.
Reactivity Rod Withdrawal These transients are temi-Insertion error nated by the rod block moni-tor and not a scram, so i
they will not be affected by the scram reactivity 1
curve change. Other core characteristics will not change sufficiently to sig-nificantly affect the out-come of the analysis or the l
block setpoint.
Decrease of All These transien'ts result in a Coolant Inventory RPV depressurization and, in some cases, a low level scram.
Power level drops due to void formation before the scram, and FDER effects are minimal.
A mild repressurization on MSLIV. closure at 850 psig occurs on some. RPV tempera-ture transients are the only concern on scrne..
I 1
I Category Event Reasons Reanalysis Not Needed Core Coolant All These transients are not severe Flow Increase and are affected by the reacti-vity increase due to sweeping out voids from the core and the addition of colder water to the core. The lower void coefficient l
at the end of life conditions makes these transients less severe than as previously analyzed.
Core Coolant All A scram does not occur as a direct Flow Decrease result of this transient, so the void coefficient change will be the principal effect changing the results of these transients.
The change in void coefficient is not sufficient to significantly affect the the results. Start up test results where actual recirculation pump trips were con-ductpd_, demonstrate that these transients
.are of a mild nature.
Others Any other transient analyses conducted ancillary to the standard ones or for other special purposes, such as a DC power interruption, do not include con-siderations pertinent to this discussion and are therefore not included.
D.
Results of Transient Reanalyses 1.
Scope of Reanalyses
. The following transients were reanalyzed in order to detemine the specific changes that might occur to the previous analytical results:
Turbine trip without bypass (Relici valve adequacy check)
Main Steam Isolation Valve Closure, (includes delayed scram case for safety valve adequacy check)
Specific write-ups for these analyses, together with the transient curves obtained, are included herein. Curves are shown for the turbin6 trip without bypass, relief valve adequacy transient and the MSLIV closure transient, safety valve adequacy transient. This corresponds with the extra analysis done with the improved control rod drive scram time assumption to show ade-quate design margins.
~
It should be noted that the original FSAR analysis used for the safety valve sizing transient was the turbine trip without bypass with flux scram. Ibw-ever, it was detemined with later plants that the main steam line isolation 1 !
f with flux scram could be more severe. During the reanalysis work reported herein, this possibility was checked by perfoming both analyses and the results showed a somewhat higher peak pressure with main steam isolation valve closure. Hence, this analysis is used for checking safety valve ade-quacy in this report.
The dwell time is a special Technical Specification requirement to take account of an AEC concern about the possibility of a delayed scram.
It was arbitrarily assumed that one of the more probable of the transients,
- he turbine trip with Sypass, occurred and a neutron flux initiated scram was delayed until the J,afety limit was reached. The amount of the delay, 0.95 secs. for !bntice;10, is a Technical Specification. Analyses performed for all other plants (e.g. Millstone) have shown that the neutron flux peaks are lower and broder than previous analyses have shown (Turbine Trip with bypass transient).
This is as a result of the lower void coefficient which reduces the rate ci reactivity insertion and slows the progress of the transient.
The dril time for the Monticello plant calculated for end of life conditions will merefore be longer than that calculated by previous analysis. Because the previous analysis represents a worst case condition, those results are retained as the technical specification requirement.
2.
'Ibrbine Trip Without Bypass - Relief Valve Adequacy Transient A scram signal is initiated at the same time a turbine trip occurs by position switches on the turbine stop valves. This transient causes a rapid pressure increase in the reactor pressure vessel. Primary system relief valves are provided to remove sufficient energy from the reactor to prevent safety valves from lif ting. The initial reanalysis showed that peak pressure in the stear.i line at,the safety valve location _did _ not meet the,GE margin of 25 psi to the safety valve set point. Hence, the transient was reanalyzed usine the u' nnroved control
' rod drive scram time assumption discussed previously and illustrated in Figure 2.
The results are shown in Figure 3.
The peak pressure in the steam line at the safety valve location was1183 psig,which provided an adequate margin of 27 psi to the first safety valve set point. Thus, the adequacy of the four relief valves was confirmed for these conditions. Ilsing the parameters associated with the end of life conditions, four relief / safety valves are required to operate to prevent this pressure transient from exceeding the safety valve set point. The rapid pressure rise Le to rapid closure (0.10 sec.) of the turbine stop valve without bypass operation causes core voids to collapse and neutron flux reaches 241 percent of design (Figure 3) before the scram shuts down the reactor. Peak surface heat flux is less than 110 percent (Figure 3) thus adequate themal unrgins are maintained.
3.
Closure of All Main Steam Line Isolation Valves i
(Flux Scram) - Safety Valve Adequacy Transient The AS4E Nuc1 car Boiler and Pressure Vessel Code requires that each vessel designed to meet Section III be protected from the consequence of pressure and temperature in excess of design conditions. "Ihe ASA Code for Pressure Piping also requires overpressure protection. The set points of the safety valves comply with the AS4E pressure vessel code taking into account static heads and dynamic losses.
5-
-~
l l
To detemine the required flow capacity of the safety valves, it is assumed that:
a.
The reactor is at 1670 MWt, b.
The reactor experiences its worst main steam isolation transient, Direct reactor scram is neglected (based on isolation valve position c.
switches),
d.
The backup scram due to high neutron flux shuts down the reactor, The Target Rock relief valves act as safety valves with low set points.
l e.
i Both a turbine trip without bypass and closure of all main steam line isolation valves produce severe overpressure transients. Analyses for these two events have shown that the 3 second closure of the isolation valves is slightly more severe for the final plant configuration when direct reactor scram is neglected. This results because the longer steam lines, allowing more volume for steam compression, more than compensates for the faster acting turbine stop valves in the fomer transient, when com-pared with MSL1V closure. The latter transient is therefore provided here as the basis for determining the adequacy of the safety valves.
Pressure increases follow this rer. tor isolation until limited by the opening of the safety valves. The peak allowable pressure is 1375 psig (according to ASE Section III, equal to 110 percent of the vessel design pressure of 1250 psig). The Target Rock set points are < 1080 psig and the spring safety valve. set points are at 1210 psig (2 valves) and 1220 psig (2 valves). Thus the ASME code specifications that the Icwest safety valve be set at or below vessel design pressure, and the highest safety valve be set to open at or below 105 percent of vessel. design pressure are satisfied. The four spring valves together have nameplate capacity greater than 35 percent of turbine design flow.
Figure 4 shows the resulting transient assuming the capacity of only 3 of the 4 relief / safety valves (35% of main steam generation rate) and only 2 of the 4 safety valves (18% of main steam generation rate). An abrupt pressure and power rise occur as soon as the isolation becomes effective.
Heutron flux reaches scram at approximately 1.8 seconds initiating reactor shutdown.
It peaks at a value of 610 percent. Peak fuel surface heat flux is slower, reaching a peak of 127 percent at about 27 seconds. The assumed safety valve capacity (Target Rock plus spring safety capacities) keeps the peak vessel pressure 92 psi below the peak allowable A9E overpressure of 1375 psig. Therefore, the relief valves plus the spring safety valves provide adequate protection against excessive overpressurization of the nuclear system process barrier with a large margin, because of the reduced capacities assumed for this analysis.
6-
l IV Technical Specification Changes A.
Scope of Changes
- Ihe principle changes of interest concern the slightly improved control rod scram times. This is needed to be consistent with the new assumptions used in the transient reanalyses and is discussed in detail in Section IIIB. Other changes are those associated with the results of the transient reanalyses discussed in Section IIID. None of these are of a crucial safety nature and mostly affect statements about margins for various pressurization transients.
B.
Specific Changes ITB4 IOCATION GRNGE REASON Basis statement Pg.22 in last sentence-105%
This reflects the re-for 2.3.F to 110% and 1.9 to 1.8 sults of this new analysis.
Basis statement last para.
in the second sentence This reflects the re -
for 2.2 Pg.24 change " turbine trip" sults of the new to closure of all the analysis
~
main steam line isola-tion valves.
(MSLIV).
~
in the third sentence This reflects the re -
change 1187 psig to-sults of the new 1183 psig analysis in the last sentence This reflects the re-change " turbine trip -
sults of the new valve" to-MSLIV analysis position Basis statement Pg. 25 change 1293 psig This reflects the re-for 2.2 to-1283 psig sults of this new analysis i
Basis statement second para.
in the sixth line change This reflects..the re-for 2.4 Pg. 26
" turbine trip" to4tSLIV suits of this new position.
analysis in the ninth line change The adequacy of the 3 293 psig to-1283 psig.
safety valves is veri-Delete "Section 4.4.3 fied and presented in FSAR" and reference this this analysis analysis. Delete" in the FSAR" and reference this analysis.
, r
4, ITEM LOCATION CIRNGE REASON Basis statement Pg.39 in the sixth and This transient has for 3.1 first para, seventh line delete been reanaly ed and reference to the the results are pre-FSM1 and reference sented in this this analysis analysis Pg.39 in the last line This transient has Third para.
delete reference to been reanalyzed and the FSAR and refe-the results are pre-rence this analysis sented in this analysis Spec Table On Change to The transient rea-3.3.C.1 Pg. 79 the following:
nalyses were done with these new scram time requirements
% Inserted From Average Scram Insertion Fully Withdrawn Time (secs) 5 0.375 20 0.900 50 2.00 90 5.00
'f"3b.2 Tblgg Change to on The transient reanalyse Pg the following:
were done with these ne scram time requirements
% Inserted From iAverageScramInsertion Fully Withdrawn Time (secs) 5 0.398 20 0'.954 50 2.120 90 5.300 Basis statement Pg 82, A.1 in the eleventh line, This is in conjunction for 3.3 and 4.3 First para, change the reference with the new scram to refer to this reactivity curve used analysis in this analysis Pg 85 starting in thecighth This is consistent para. C line, change " turbine with the results of stop valve closure" this new analysis to-closure of the main steam isolation valves in the eleventh line, This is consistent witi:
change 1.9 to 1.8 the results of this net analysis j
t 4-3 IDCATION OiANGE PEASON ITEb!
in the twelfth line, Thisisinconjunctiog change 390 to 290 with the changed scrar time requirement and is consistent with current design practie on Transient analyses completed on other plants in the thirteenth line This is in conjunctio2 delete all remaining with the changes in after.
"This is----
the scram time and to through the end of the be consistent with paragraph on pg.86 and this analysis.
replace with - This is adequate and conserva-tive when compared with the typical time delay of about 210 milli-
~
seconds estimated from scram test results.
Approximately the first 90 milliseconds of the
- time interval results from the sensor and circuit delays; at this point the pilot scram solenoid deener-gized. Approximately 120 milliseconds later-control rod motion is estimated to begin. Ibw-ever, to be conservative, control rod motion is not assurned to start until 200 milliseconds later. This value was included in the tran-sient analyses and is in-cluded in the allowable scram insertion times of l
specification 3.3.C.1 and 3.3.C.2.
Basis statement Pg 134 In the second line of the All changes on the for 3.6 and 4.6 Para E.
third paragraph delete page are to be consis;
" turbine trip initiated" tent with this analysf and replace with-MSLIV closure.
t j I
IThi LOCATION OWE REASON-In the third line delete "no steam bypass system flow and change "tur-bine" to-MSLIV and
" trip" to-position In the fifth line change (35.4%) to
-(35%) and in the sixth line change (18-5%) to -(18%).
f a
4 e
p 4
e i
1 i
9.
-40 i
t t
M i
)
I YEW ANAtysis FOR I
END-OF-CYCLE I
.so j
i i
\\
s 5
i 1
PRgvjoW t
S u st vsiS
?
-20 r
u 4
ad s
L s
i 4
\\
ao i
i i
)
r i
t ao i
r 9
0 l
\\
)
i t
o t
o 2
4 s
s t
TIME (sect i
t FIGURE 1. SCRAfA REACTIVITY CURVES - IDNTICELID e
e 4
.m a
i 9
l 90 C.
71
. i.
i;*
80
<.:s/: *.-
- a g
W,.$!
k x!
,1:
pr O'
10 if3/ b fl3f
~~n
,1.. f
^
60 J
I R ANCE OF TYPICAL, j[
f RECENT EXPERIENCE
'l
, s0
,,7$,.
i u
,.p.[':
I y
.,Y' h
l 5 40
.rr-gc 49 d
!V-l 3
.y.'"
i 30 t, *-
c-a' CURRENT.
g 20 TECHNICAL SPECIFICATION I
i
- PROPOSED NEW 10 I
TECHNICAL SPECIFICATION I
I i
t 0
0.2 0.4 0.6 0.8 1.0 2.0 3.0 4.0 50 O
t I
ELAPSED TIME AFTER SCRAM SIGN AL (ml FIGURE 2. CONTROL ROD DRIVE SCRAM TIMES.WhTICELLD G
5
$O f
t t
6 i
.____.__m--.,-.
.----.~_.---._.,-m......~_-.,....--.-~.-~.r.~.-----,-..-.-.....w,..-~
s.,
-. ~.
._,__~.._,---...-...--.,----..-4 I
9.
v Y
[
J
= _ _ s 00p3 6
~
t ad I
\\
\\
g e}
l I
N gyg$
- 7 x
,s
[EEv ug u kb55b
/
p g_
=_
O i $
E e
e u
w
\\
B w
/
c c
i q
y t
u
/
i k
ll e
I t
i g
y
....i....
=
Ng 4
g i
y g
i i
9 8
8 8
g
~
to21w aa um u3a M
F V
k g-Eph a h
t 4 M!
c 4
~E bd y r
y
,,IEh I
_g k(3
__.h s' E'.~W b,u A
3 e_
3 N
w m
7 E
.j-N P
a
/.
s1, I
CE055 f~
~
c4hr
)
N a F
%.d6 4
R JLvMSE f
~
~
e
[{r5
- p E
E r-r g5551 3
a 3.a
~
f
)
~G m
a 2
d nc w
l d
-s I
- l 5
S I
vs Bii f
WW N
d 3
i s
x he d
-~
~.5, e
[.
y
,,,,I,,,j
., u !,, i, 'j m
6 8
E E
N E
a-(031W 3D IN3] U3d1 g
5d y
y I
l g
i E bc l.5 t
?E a
i o
e 4'E s lg phh
-\\
d g
8
~
ggi$
[
[
~
dly eghw
..de K
w ch
- N cohr
^
f s?
- sot, lpt adu e a
~
/
9 9
p, sis
~
v) vi zn I
y-.
W i
=
=
l
AEC DX"UBUTION FOR PART 50 DOCr22 Fr 'IAL (TMORARX FORX)
CONTROL ;;0:
1054 FIII-FROM:
DIG OF DOC:
COED Northern States Power Company Minneapolis, Minn.
55kOl T_.
O. Mr er 2-13-73 2-lk-73 1
TO:
~
Mr. G N 1;sso 1 signed CIASS:
,(UfniOPINFO IhW2
- iO CYS REC'D DOCn;T h0
ho 50-263 DESCRIP2105:
E!iCLO!TURES:
Ltr trans the following:
SUPPL REPORT: "Results of Teansient Reanalysis for Monticello Nuclear Generatin6 Plant with End-of-Cycle Core Dynamic Characteristics".
NorE: *PLFASE CIRCUIATE-INSUFFICIDC CCPIES PICEIVED FCE FULL DISTRIBUTICN p
( LO cys ree'd) uQ NO[ Rg PLA;i? NAMES: Ponticello FOR ACTION /I;iFolCGTIOS p.3 s. n rn EUILER(L)
SCFa'ENCER(L)
SCE2GL(L)
}2;IGICOI;(E)
W/ Ccpies W/ Ccpies W/ Copies W/ Copies CLARK (L)
STOI2(L)
V ZIEMANN(L)
YOUNGEIDOD(E)
W/ Ccpies W/ Ccpies W/ 9 Copies W/ Copies GCLTER(L)
VASSALID(L)
CEITWOOD(IM)
REGAN(E)
W/ Copies W/ Copies W/ Copies W/ Copies
' XNIEL(L)
H. DECCN DICEER(E)
W/ Copies W/ Copies W/ Copies W/ Copies IICER 'AL DISTRIEUTION (PR5 ~FIM TECH REVIEW VOLUCR H/JiIISS WADE E
AEC PDR HEruRIE DECON SHAFER F&M P OGC, ROOM P-506A SCHROEDER Ui1MES F&M ERDVN E
V MUICZING/ STAFF p MACCARY*
G14MLL Mf G. VII11AMS E CASE 13;IGHI(2)
EAST!;ER NUSSEAU G E. GOULBOUPJ;E L GI/J3USSO PAWLICTI EAILARD A/T IND e BOYD-L(En'R)
SHAO SPfd;3IIR LIC ASST.
LEAIT!%N DEYOU';G-L( Fa'R)
/ 13rJiH
- SutVICE L SALTZMAN y SEOVHOII:'-L STF1Tn ErrR0 MASON L
P. COLLINS FOORE R~D na WILSON L
PI/J;S HOUSTON DICEER 1%IGIII L MCDJNALD REG CPR
/ TEDESCO' 13iIGITION SMITH L
DUEE y FILE & PSOICN (2)
IDNG YOUN3EID0D GEA1GN L
10ERIS LAINAS PROJ I?LtDER DIGGS L
I!TO STEELE EENAROYA TUTfS L
C. MIIIS b
PIGA'i ID~TEP!AL DISTPIF570N y 1-IDCAL PDE vin m ylir_ vi nn V 1-ITfIE(AEERNATHY)
(1)(5)(g)-NATIONAL IA3'S FDR-SAN /LA/NY
/ 1-NSIC(EUCH/J.!sii) 1-R. C/JiROLL-0C, GT-E" -
1-GERALD LELIDUCEE l ASL3-YCII/SAYP3 1-R. CATLIN, E-256-GT Eh0Cft:EAVEN NAT. IAB W30DWARD[H.ST.
1-CONSUIAIC 'S 1-AG'cD(WALTER r.0 ESTER, y 16-CYS ACES >!9ueTx3 Sac TO C C ASST.
NE,yAFK/LUfe/AGAEI;c; E= C L27, CT )
F. DIGGS CN 2-15-73 1.pD...MUILER...F-37/GT