ML19220C707
| ML19220C707 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Crystal River, Rancho Seco, Crane  |
| Issue date: | 04/06/1979 |
| From: | Ornstein H NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19220C708 | List: |
| References | |
| TASK-TF, TASK-TMR NUDOCS 7905140055 | |
| Download: ML19220C707 (18) | |
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Have Feedwater transients occurred at other B&W facilities.
How many? Significant?
A.
Feedwater related and similar transients have occurred at TMI-2 and other plants with B&W reactors.
These are described in the attached memorandum.
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MEMORANDUM TO:
D. Eisenhut, Deputy Director, Division of Operatin9 Reactors FRCM:
H. Ornstein, Technical Specialist, TA EDO
SUBJECT:
FEED'3ATER EVENTS AT B&W PLANTS Enclosed is a typewritten copy of the material I providt.1 you on April 4,1979.
Also included are ccpies of relevant B&W and NRC memoranda, hk D
H. L. Ornstein Technical Specialist, TA EDO
Enclosures:
As stated above cc:
P. Check J. Watt 101 240
TMI-2 3/29/78
('ero Power Physics Testing 3 RC Pumps Operating)
RCS depressurization Vital bus 2-IV de-energized - caused opening of an electromagnetic relief val ve.
Reactortripped(duetoindicationthatAloopRCPumpswereoff-whi[6 8 loop RC pumps were on - actually 1 "A" loop pump was not in service at beginning of transient and failure of the vital bus 2-IV caused the RPS to sense that the remaining A loop pump was lost while both B loop RC pumps were opera ting).
The reactor depressurized, the operators closed the RCS letdown isolation valve. Temperature compensated pressurizer level indication was lost, as was RC; pressure indication powered from the Vital Bus 2-IV.
The operators did not know why the depressurization was occurring - as-they had no position indication of the electro-magnetic relief valve.
At 1 min 53 sec SFAS actuation for safety injection initiated.HPI pumps took suction from the BWST and the NA0H tank - at 2 min 23 see The Safety InjectionSignalSypasse4At4 min 13seethevitalbuswasreenergized through its alternate source, thereby closing the electromagnetic relief valve and returning all instrumentatior, to service.
The depressurization ended at a pressure of 1173 psig.
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101 241 5.
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TMI-2 4/23/78 (30% power - 3 RC Pumps Operating)
Excessive RCS cooldown and depressurization Reactor tripped due to a " Noise spike on NIS power range detector - The reactor tripped because RPS Channel C was already in the tripped state...
due to the inoperability of NI 7."
There was a turbine trip causing pressure increases in the OTSG's. The main steam relief valves on the SG's lifted and did not immediately reseat properly (finally reseated after 2-4 minutes with SG pressures at 550 and 600 psig).
I The operator took the proper immediate action in manually cutting back feedwater demand, shutting the letdown isolation valve, starting a second makeup pump, and opening the high pressure injection valves on the side of the j operating makeup pumps.
The operator failed to initially recognize that the J feed pump was in manual and did not run the feed pump speed back until approximately 1 minute and 20 seconds had elapsed.
The Integrated Control of the feedwater valves had not yet been initially tuned at the time of the event.
Integral vice proportional control was the dominating signal of the feedwater valves and although the valves responded in the proper direction, they responded much slower than the traditionally expected response.
Thus, the feedwater valves slowly going shut, rapidly 9g decreasing steam ger.erator pressure and a constant feed pump speed, too much water was fed into the steam generators.
a "A
The safety valves failing to reseat at the proper pressure coupled with over-O f
feeding the steam generators caused a rapid depressurization and cooldown of the reactor coolant system.
The reactor coolant temperature varied from Js 5830F to 4640F in 3 minutes. The RCS shrinkage frcm the cooldown caused the pressurizer volume to drop below the minimum indicated level range approximately o
one minute after the reactor trip.
g Due to the rapid depressurization of the y h By present design this injected Na0H into the reactor cool Ng the nigh pressure injection lines.
f Pressurizer level was restored two minutes into the event as a result of safety injection, the Turbine Bypass valve going shut and some of the B sidt Main Steam Relief Valves going shut.
Feedwater latch occurred 2'u minutes into the event and terminated feedwater flow to the steam generators.
Feedwater latch was the key event in terminating the transient.
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qOL L
TMI-2 11-7-78 (92% power)
LER 78-65-99X Reactor Trip and Safety Injection During a power runback due to loss of 1 feedwa'.er pump.
On November 7, 1978, TMI-2 experienced a reactor trip during a power runback from 92% rated thermal power.
Prior to the reactor trip, testing per TP 800/05, Reactivity Coefficients at Power, was in progress. All operatfng garameters were normal except for RC Tave which had been elevated to 588 (6 F above normal) for temperature coefficient measurement.
At 0523:37, a heater drain tank low level alarm was received.
This automatically tripped operating Heater Drain Pumps HDPlA and 1B which normally supply approximately 30% of the total feedwater flow to the suction
(
of the feedwater pumps.
The feedwater pumps tried to meet the increased feedwater demand, however, Condensate Booster Pump COP 2C tripped on low h
suction pressure. This automatically tripped Feedwater Pump FWP13.
The Integrated Control System (ICS) began a power runback to 55% rated thermal power based on the loss of one feedwater pump.
However, due to the elevated Reactor Coolant System (RCS) temperature required by the testing in progress, L
the reactor tripped at 64% power.
This trip occurred prior to completion y
of the pcwer runback. as all four Reactor Protection System (RPS) channels received a variable temperature pressure trip signal.
At this point, the cr.erator secured the letdown flow (closed Makeup Valve MUV376). A second makeup pump was then started prior to the safety injection.
Ng RCS p" essure continued to decrease and safety injection was automatically initiated at 1640 psig thus limiting the pressure decrease to 1550 psig at 25 seconds after the reactor trip.
The decreased RCS volume due to the cooldown and depressurization caused the pressurizer volume to decrease below zero indicated level for approximately 30 seconds. However, calculations show that the pressurizer was not emptied during the transic-t.
Approximately Zh g minutes after the reactor trip, RCS pressure increased above 1600 psig.
101 2a3 e
TMI-2 LER 78-69-99X 12/2/78 (22% power)
Reactor Trip with Safety Injection due to pinned open main Feedwater regulating valve.
r-On December 2, 1978, TMI-2 experienced a reactor trip from 22%
rated thermal power, while switching from the startup to the Main Feedwater Regulating Valves.
Prior to the reactgr trip, all operating parameters were normal except A
for RC Tave of 584 F.
Tave was higher than normal due to Feedwater Heaters D
fbeingplacedinservice.
hg Due to the changing FW flow, the startup feedwater valves (FW-Vc5A/B) reached 80% open. As the main feedwater block valves (FW-V10A/B) opened, FW valve d/p decreased to zero, prcmpting the operator to increase feedwater pump k
speed.
It was later determined that the Main Feedwater Regulating Valves L
(FW-V30A/B) were full open by manual hand wheel with Instrument Air isolated.
The increased feedaater flow resulted in a rapid RCS cool'!own. At this point the operator secured letdown flow (closed MU-V376), started a second makeup pump, reduced feedpump speed, and closed the main feedwater block valves.
j The overfeeding of the OTSG's resulted in the reactor trip on low RCS pressure, g followed by Safety Injection 215 minutes later.
Due to the above operator g action RCS pressure recovered to above the Safety Injection setpoint within 17 seconds.
101 244
Rancho Seco 78-1 3/20/78 Excessive cooldown transient 70% power (Frem I&E report 50-312/78-03) lost non-nuclear instruments - including SG and pressurizer levels - and all RCS temperature 5, Loss of RCS hat leg temp input to the ICS caused termination of feedwater flow.
Reduced heat removal in steam generators causedRCS temperature and pressure to increase.
The reactor tripped on high RCS pressure folicwed by a turbine trip.
The secondary sides of both steam generators emptied due to operation of condenser bypass valves, atmospheric dump valves and auxiliary steam loads.
Pressurizer level was maintained (using computer indication) by manual operation of a high pressure injection pump.
"A" steam generator level control (actually lost at time zero - but the channel drifted slowly downward - while "B" SG channel drifted slowly upward) initiated emergency feedwater injection (the turbine driven auxiliary feedwater pump had started on loss of feed-water flow)
RCS cooldown started as a result of emergency feedwater #10w to A steam generator and possibly also due to main feedwater pum manually operated).
Decreasing RCS pressure (1600 psig) actuated all safeguards pumps and the motor driven auxiliary feedwater pump.
Full auxiliary feedwater was initiated to both steam generators.* RCS hit a minimum of 1475~ psig which was then increased and maintained at 2000 psig by manual control of an HPI punn.
Restoration of tne NNI restored all lost indications and controls - operating personnel secured the auxiliary feedwater pumps and started RCS pressure reduction via pressurizer spray.
NM
- maximum flow on both aux feedwater pumpg autcmatically provided to both steam generators.At Davis Besse 1 and Crystal River 3 safety features Actuator signal does not initiate maximum AFA flow rate to both steam generators.
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Rancho Seco LER 79-01 1/5/79 (100% powe*) Loss of ICS - Excessive cooldown rn" Short on ICS - loss of logic power sm3the feedwater valves back to the 50% posi' ion - caused RCS pressure to increase resulting jn a high pressure trip./6 ~4 /?CS depressurization to 1600 psig actuatM HPI and auxiliary feedwater.
ICS was res+ored after 5 minutes, and feedwater flow increased.
The operator then terminated most of the feedwater flow - 2 minutes later the main feedwater pumps were tripped - thereby allowing aux feedwater to supply the OTSG's.
During the transient the gbTSG was filled to the top of the operating range, and it stayed at that level for 10-15 miautes.
SMUD believes that the excessive feedwater to the "B" OTSG from the AFS was "the single mos t significant cause of the excessive cooldown rate."
No info on pressurizer level - SMUD will review (by 4/30/79) ar. evaluation of the necessity of aux feed on safety features actuation.
B&W is reviewing the transient and will forward reccmmendations to SMUD for corrective acticn.
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O 10i' 246
NP 32-77-16 Davis-Besse 1 9/2k/7/
9% power Loss of RCS pressure due to failure of pressurizer power operated relief valve 9% power -bypass system operational, spurious signal resulted in closure of feedwater control valves to 1 steam generator. Upon a icw-low steam generator level. or main steam and feedwater isolation valves closec and 2 aux feedwater pumps started. A problem (don't know what) developed in 1 AFW pump, and it was shutdown by the operator.
The plant operated at 9% power on i SG with 1 aux FW pump.
Feedback to the RCS resulted in increasing pressurizer level and pressure. The operator indicated a manual scram. (T=1 min. 47 sec)
RCS pressure increased to the setpoint of the pressurizer relief valve.
It cycled 9 times and then stuck open. When pressure dropped to 1600 psig ECCS was initiated (2 min 51 sec).
Full high pressure injection flow was established, and started to raise pressurizer level. AT T= 6 min 14 sec the operator stopped the high pressure injection pumps (The operators had b/.en
" heavily involved before this time in regaining seal injection flow to the reactor coolant pumps which had been stopped by the SFAS actuation." By T=5 min 20 sec the appropriate SFAS signals had been overridden and the normal flows restored to the seals of the pumps).
RCS pressure continued to decrease until saturation pressuro was reached, steam formed in the RCS (T=8 min) - causing an upsurge of water into the pressurizer and the pressurizer went off scale.
During the level increase the operator saw the RCS avg te.nperature and pressurizer level increase - he then stopped one reactor coolant pump on each loop (T=9 min) to reduce the heat input to the RCS.
At approximately T = 21 min., it was determined that the power relief valve was remaining open and the block valve was closed, isolating the power relief valve on the pressurizer and stopping the venting of the reactor coolant system to the quench tank. At T = 31 min., pressurizer level came back on scale. At T = 41 min. the operator started a second makeup pump to try and stop the pressurizer level decrease. This additional cold water started the 4
reactor coolant system on a slow decreasing zemcerature transient. At k
T = 43 min., pressurizer level reached the low level interlock and cut off the pressurizer heaters. At T = 49 min. the operator started a high Q.
pressure injection pump to try and stop the decreasing pressurizer level.
N With pressurizer level well on its way to recovering, the operator stopped b
the high pressura injection pump (T = 53 min. 24 sec.).
At T = 57 min.
y he restored reactor coolant makeup flow to normal.
This stopped the slow s
decreasing reactor coolant temperature transient which started at T = 41 min.
g All plant parameters were now fully under control and the plant was brought to a steady state condition, and a normal plant cooldown started.
101 20
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~11/29/78 - Unit 3 One reactor building pressure sensor channel vas intentionally tripped, spurious trip on another channel 0 4 psi, caused E.S. actuation, operators took manual centrol and secured E.S. cquipment.
12/14/78 - Unit 1 Feeduster pumps tripped, stes= 3eneratorsvvent dry..s? Inj ection actuated, FORv's operated satisf actory.
C 7/6/73 - Unit 1 E.S., actuation due to operator error, EPI inj ection ra tiad pro levet $have level instrument range, but pressure remained less than 2215 est o.; _pperator action properly avertec W ing at pressurizer relief valves.
11/22/74 - Unit 2 During cooldown subsequent to a resetor trip resulting frc= a RCP meal l eak the corn fleed tanks were depressurize by bleeding. nitrogen to the quench tank.
This caused the quench tank rupture disk' to burst, and the resultant jet severed the i= pulse, line on the pressurir.er level instru=cntation and dansged pressurizer insulation.
Mor=all,, C7 tank bleed would be to the vaste gas filter or vent h ead e r.
However, the valves required to operated to acconplish this were inaccessible due to the seal leak.
AO-270/74-2.
5/30/74 - Unit 2 operator error resulted in EFI inj ection into RCS vhen valves were creencously opened (no ES actuation signal).
Reactor trio resultad c" nressurizer'hi'gh level.
NOTE:
They carly found problens with the PORV block' valves and by october 1976 all three units had replaced the valve with an 1=pr.oved design.
They have installed a deflector plate between the quench tank and the pressurizer to keep stes=/ vater mixture from hitting the pressurizer.
Cc=pleted as follevs Unit 1-5/76, Unit 2 - 6/77 and Unit 3 - 10/.76
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l01 248 e
100% power Full Load 'urbine Trip Test + Failure of 1 trEn of Aux Feedwater Turcine Trip, main feedwater pump trip, automatic actuation of emergency feedwater pump. A MOY required for feeding t e "A" steam generator through the AFW nozzles failed to open.
Feedwater to steam generator continued satisfactorily.
Subsequent operator action re, stored feedwater to both steam generators.
101 249
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Oconee 3 6/13/75 A0 287/75-7 Excessive Cooldown Rate Routine maintenance shutdown 100% -915LYnile at 15 % power there was a mismatch between power generation (ll5 MW)and unit load demand.
(65 MW).
ICS attemptdto match load with power. The main steam bypass valves opened, and then closed when the main steam pressure decreased, feedwater flow and steam generator level oscillated, as did RCS temperature and pressure.
The power operated relief valve opened at 2255 psi but failed to close when the pressure dropped. Decreasing RCS pressure caused a reactor trip and HP! actuation.
The operator closed the block valve after the reactor trip to terminate depressurization.
The clock valve was later reopened because of rising pressurizer level, and was again closed when the pressure dropped to 800 psi - terminating the transient.
\\o\\
250
12/14/78 Oconee I-R0-269/78-77 98% power A short ca; sed ICS T ave recorder error ICS withdrew control erroneously-reactor trip on high temp / press.
Both Normal FW pumps tripped on high discharge pressure.
Emergency Pd pump started then stopped when normal FW pumps were reset and started.
2 rs. later OTSG levels dropped to 6 and 0 inches (VS 110 inches n S
level was restored within 3 hrs whereas it took about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> until S filled through the emergency P4 header. Apparently malfunctioning of I w
valve:
t e normal and emergency feedwater paths led to the long fill 1.,
time for S 3.
The HPI was actuated on low RCS pressure during the event - however from the existing information it is not clear when this occurred.
10\\ 25\\
Crystal River 3 3/2/77 40% power Loss of AC bus-excessive cooldown rate "B" vital AC bus was lost due to a failed output diode in the inverter.
Power was lost to ICS causing reactor trip, turbine trip and opening of the steam dump valves (50%).
Main FW pumps tripped (due to loss of vacuun) and SG feed continued from the emergency FW pumps.
RCS temperature 0
dropped 164 F wir.nin 15 minutes.
Plant normalization was begun immediately, with vacuum restored within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
101 252
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CLtN CLLvN tLLthets 60337 Janut.ry 8, 1979 Locke: Nc. 50-500/501 50-329/330 MD 0?.ANOUM 7CR:
J. F. Streeter, Chief, Nuclear Supper: Section 1 T ECP. :
J. S. Cresvell, Reac:or :nspec:c:
SU3 JECT:
CONVEYING N k' INFORMATION TO LICI::SI.'O 30A?. S-OAVIS-33SSE UNITS 2 & 3 AND MIDLCD UNITS 1 6 2 Durins :he ccurse of =y inspecticns a: Davis-3es se, ce r:ain is sues have cc:e a.....'.c...+...ch 7
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.c F c:edure 1530A (Neve b er 16, 1978), s:ep 3 and inic-a:i:n s;;p'.ied to me per step 1..
ine issues :c censidera:.,cn are:
1.
During a recen: inspec:icn a: Davis-3 esse Uni-1 i.fernatica has been a::ained which indicates that at cer:ain condi:icns cf reac:c ccclant visecsi:y (as a f unction of tempera:ure) dere liftin; =ay occur.
The licensee infer =ed the inspecter :ha-this issue involves o:her B&W facilities.
The Davis-3 esse FSAR sta:es in Sec:icn 4.4.2.7:
The hydraulic f e:ce en the fuel asse=biv. receiving :he res:
fiev is shevn as a f unction of system fice in Figure.4-39.
Addi:icnal forces actina en the f uel assembiv. are :he assembly 5
veigh: and a held devn spring f orce, which resulted in a ne:
dernvard f e:ce a: all ti=es during nc::a1 sta:icn Opera:icn.
c The licensee states tha: there.is a 500 ? interlock f er the starting cf -he fourth reac:c: ccclant pu=p.
Hevever, nc Technical Specifi-cation requires tha: the pe=p be started a: cr above this t e pera-ture.
A cencern regarding this ca::e vould be if asse:blies ' cved upward into a posi:ica such that cen rol cd =cve=en: veuld be hindered.
2.
Inspec:ien Repor: 50-346/78-06, paragraph 4, reper:ed reac:ivity -
pcver escillations in the Davis-3 esse cere.
These es:illaticas have also occurred a: Ocence and are a ribu:ed :e staa: sanera:e level escillations.
35W reper: 3AW-10027 sta es in A9.2:
- fm A y
'?
e w~,
a*-
101 255 1
' ' J. 'T. 5:rceter 2
January 5, 1979 The OTSG laboratory model tes results indica:ed' :ha periodic oscillations in steam pressure, s:ca fica, and s:can genera:::
pri=ary ou:let temperatures could occur under cer:afn c:..di:1cns.
It was shcun tha: the oscillations were of the type associa:ed with the relationships be:veen f eedvater hea ting chamber pres-sure d:cp and :ube nes: pr e s sur e d rop, whi ch ar e e.,;.; na:e., c:
reduced to levels of no consequence (no f eedback to rea _:c; system) by adjustment of the tube ncs; inle: resis:ance.
As a resul: of the tests, an adjus table crifice has been ins:alled in the devnce=er section of the steam generate:s to provide for adjustment of the tube nes; in'e: resis:ance and :o provide
- the means for elimination of oscil.ations if : hey should develep during :he operating lif eti=e of the genera: ors.
The ini:ial crifice setting is chesen conservatively to minimi:e the need fc: f urther adjus:=en: during :he star:up :es: prestan.
Ve also note tha: the'effect en the incere de:2c:c: sys:e= fer
=eni:cting core para =e:ers during :he cecilla:icns is ne: c lea r.
3.
Inspec:ico and Inforcement Reper: 50-346/75-06 documented tha: pres-suri:er level had gone of f scale f or moprcxima:ely fi.e -inu:es dur-ing :he November 20, 1977 less of offsi:e pcVer even:.
Anere.are sc=e indications that other 3&W plants may have problers main:sining pressurize level indica:icns during transien:s.
'In addi:ica, under m
cer:ain conditions such as less of feeduater a: 100% pcue vi:h :he reactor ecclant pu=ps running the pressurizer ta'y veid cc:pletely.
A speciil analysis has been pe rf er=ed concerning this event.
This analysis is attached as sne3.esure 1.
3ecause of pressuri er
.3 eve.3 saintenance prcblems the sizing of the pressuri:er may require f urther review.
Also noted during the even: vas the fact that Tcold'ven: o f f scale 0
(less than 520 ?).
In addi: ion, it was noted tha: the =akeup flov cenitoring is li ited to less than.50 sp: and that makeup flev
=ay be substan:ially greate than this value.
This infer =a:1on should be examined in ligh: of the require =ents of GDC 13.
4.
A =e=o' frca 36W regarding control red drive systen trip breaker
=aintenance is attached as Inclosure 2.
This te=o shculd be evaluated in : erns of shutdevn ca: Sin =ain:enance and ATWS censidera:icns par-ticularly in ligh: of large positive =ederator coefficien:s -allevable
. a.. n o =,.ea c.<.,.a. < e s.
e /
O%O
\\
<..J.' 'T. 5:reete:
3 Ja.uary 5, 1979
.5.
Inspe::ica and Inforcemen: Reper: 30-346/75-17, pars.; aph c refers te inspec:icn fir;diugs regarding :he capabill:y of the incere de:e:-
to syste: te de: ermine vers: case thermal condi:: ens.
...e rasc:c:
can be operated per the Technical Specificatiens vi:h the cen:er incere string cut of service.
If :he peak pouct loca:icns is in the center of the ccre (this has been the ca se a: Davis-Sesse),
fac:crs are not applied te censervatively =enitor values such as TQ and F delta R.
6.
n e.3csure 3 c.escribes an even: :ha: cccurred a:
sca
- ac
,:7 vnic.a 1.
resulted in a severe ther=al ::ansien: and ex: ere dif ficulty in enntrolling the plan:.
The af eremen:iened f acilities shculd be revieved in light of this inferna:ica fc pcssible saf e:y implica-
- tons.
!")
(,-
)
. & ~ '.
J. 5. Cresvell EeaC:C InspeC:c!
Enclosu:~s:
As stated e
cc v/c enciesures:
G. Fiorelli
..C.
Ancp T. N. Ta=bling 4
101 257 en@enGe + e
PRELIMINARY NOTIFICATION April 23,1979 PRELIMINARY NOTIFICATICN CF EVENT CR UNUSUAL OCCURRENCE--PNO-79-91 This oreliminary noti fication constitutes EARI V notice of event of POSSIBLE safetiv or ouolic interest __ significance.
The information presented is as_ initially received without verification or evaluation and is basically all that is known by IE staf f on this date.
Facility:
Northeast Nuclear Energy Company Millstone Point Unit 2 (DN S0-336)
Waterford, Connecticut
Subject:
VAPOR BINDING OF SHUTDOWN COOLING SYSTEM PUMP Reportable Occurrence 79-08, Loss of Shutdown Ccoling (SDC) and subsequent RCS heatup to Mode 4, describes operation of the SDC system in a degraded mode on March 14, 1979 (30 Day LER). While in the SDC mode the LPSI pump became air bound resulting in a loss of SDC ficw.
Subsequent RCS heatup up to 208 degrees F resulted in entry into Mode 4 (Hot Shutdown).
Compliance with Mcde 4 LCO's was verified. To resture LPSI suction the licensee took safety measures including establishing containment integrity and having all personnel leave the containment. Suction to the LPSI pump was then shifted to the RWST and SDC was restored by RCS floodup.
Full suction was restored by vacuum priming the SDC suction loop seal.
These planned restoration actions resulted in the spillover of approximately 15,C00 gallons of water through an open Steam Generator Manway to the containment. There was no release of radioactive material to the environs.
This PN is issued for information only due to media interest.
The State of Connecticut has been informed.
A press release is not planned by the licensee nor NRC.
NRC, Region I, (Philadelphia) received notification of this occurre.:ce by telephone frcm the Resident Inspector at 2:15 FM on March 14, 1979.
Contact:
GKlingler, :E x28019 FNolan, IE x28019 SE3ryan, IE x23019 Distribution:
Transmitted H St 2: 2C1 Chairman Hendrie Ccmmissioner 3radford S. J. Chilk, SECY Commissioner Kennedy Commissiccar Ahearne C. C. Kammerer, CA Ccmaissioner Gilinsky (For Distribution)
Transmitted: MN3B 7 31 P. Bldg J. G. Davi s, 9 TM)
IE L. V. Gossick, EDO H. R. Denton, NRR Region ;;;_ _
H. L. Ornstein, ECO R. C. DeYoung, NRR J. J. Fouchard, PA R. J. Mattson, NRR N. M. Haller, MPA V. Stello, NRR
( MAIL _)
R. G. Ryan, CSP R. S. Boyd, NRR J. J. Cummings, CIA H. K. Ghapar, ELD SS Sidg _
R. Minegue, SD W. J. Dircks, NMSS 101 258 S. Levine: RES m.---
PRELIMINARY NOTIFICATION 7 90514 0Chly