ML19257C081
| ML19257C081 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/03/1980 |
| From: | Villalva I Office of Nuclear Reactor Regulation |
| To: | Kane W Office of Nuclear Reactor Regulation |
| References | |
| TAC-42410, NUDOCS 8001240089 | |
| Download: ML19257C081 (39) | |
Text
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'o UNITED STATES
[ y > c q ',,g NUCLEAR REGULATORY COMMISSION g.
/
f WASHINGTON, D. C. 20555 0, ' ?r a
f JAN 3 iSSO MEMORAtlDUM FOR:
W. F. Kane, Acting Chief Standardization Branch, DPM FROM:
I. Villaiva, Project Manager Standardization Branch, DPM
SUBJECT:
AUDIT OF SMALL BREAK LOCA PROCEDURES FOR THE MILLST0flE 2 PLANT On December 18-19, 1979, members of the B&OTF visited the C-E simulator at Windsor, Connecticut and conducted an audit of the procedures being developed by Northeast Utilities for their Millstone 2 plant. The particulars regarding the visit to the simulator and our audit of the Millstone 2 plant are attached hereto as Enclosure 1.
Working copies of the procedures reviewed during the audit are attached as Enclosures 2 and 3 and a listing of those in attandance during the exit interview at Millstone 2 is contained in Enclosure 4.
Subsequent to our audit, we were provided with a listing of the instrument power supplies to certain instruments which we believe play a vital role in the small break LOCA procedures. We have not reviewed this listing. The efficacy of these power sources will be reviewed by the B&OTF Systems Group (G. Kelly), to determine whether changes are needed. The aforementioned instrument power supply listing is attached as Enclosure 5.
/
n 4Q cG I. Villalva, Project Manager Standardization Branch
Enclosures:
As stated cc: C-E Caners' Group Principal Contacts and C-E Owners'-Group Representatives (see attached lists)
Northeast Nuclear Energy Company P. O. Box 270 Hartford, CN 06101 ATTri: Mr. W. G. Counsil o
i' s em e4a 1797 277
ENCLOSURE 1
SUMMARY
OF THE BULLETINS AND ORDERS TASK FORCE'S AUDIT OF SMALL-BREAK LOCA EMERGENCY PROCEDURE AND CDERATOR RETRAINING AT MILLSTONE 2 On December 19, 1979, representatives of the Bulletins and Orders Task Force (B&OTF), accompanied by the cognizan: Division of Operating Reactors Project Manager and Office of Inspection and Enfcrcement Regional / Resident Inspectcr conducted an audit of the emergency ;recedures and operator retraining asso:iated with small break loss-of-coolant accidents (LOCAs) at the Millstone 2 plant.
The B&OTF representatives also visited Combustion Engineering Company's Pressurized Water Reactor Sinulator at Windsor, Connecticut, to observe its responses to a small break LOCA. A list of those individuals who attended the plant audit is enclosed.
BACKGROUND The purpose of the B&OTF audits was to review selected licensees' emergency prcccdurc:, c;; crc.tcr rat eining, operator awareness of the emergency procedares and their bases and systems consideratior.s associated with small break LOCAs.
The Millstone 2 plant was selected for the B&0TF audit based on its representation of C-E reactors and on the limited time available to conduct the audit.
Audits of the remaining C-E designed operating plants will be conducted by the Office of Inspection and Enfcrcement.
The licensees' small break LOCA emergency procedures and operator retraining were based on operator guidelines developed by the Combustion Engineering Owners Group which, as documented in our letter to the Owners Group dated November 14, 1979, have been approved by the B&OTF.
As set forth on Page 5 of Enclosure 6 to Darrell G. Eisenhut's letter to all operating plants dated September 13, 1979, the small break LOCA emergency procedures and operator retraining are to be implemented by December 31, 1979.
PLANT AUDITS The following matters were censidered during the plant audit:
Emergency Procedures The licensee's small break LOCA emergency procedure was compared to the approved operator guidelines.
The ciarity of the proccdure in terms of individual operator actions and cautions and the flow of the procedure with respect to the timely initiation of operator actions were considered.
Ooerator Retraining The retraining that the operators received or were to receive for the small break LOCA emergency procedure was reviewed.
Infomal training, formal class-room study and walk-throughs of the emergency procedure with their shift sucervisors or training coorainators were considered.
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2-Operator Awareness of tne Emergency Procedures and Their Bases The operators' understanding of small-break LOCAs as well as the differences between small-break LOCAs and other depressurization events, and the bases for the approved guidelines and their familiarity with the small-break LOCA emergency procedare were reviewed. The effectiveness with which the emergency procedure can be carried out was also considered.
Systems Considerations The systems-related aspects of the emergency procedure were reviewed to assure that the necessary operator actions can be performed.
FINDINGS The more significant findings resulting from the plant audit are summarized below:
Emergency Procedure The licensee's small-break LOCA procedure contained most of the elements required by the approved generic small-break LOCA guidelines which were prepared by C-E for the Combustion Engineering Owners Group. The staff identified certain discrepancies that ext-tcd Latoaar. the new procedure and the C-E guidelines as well as those between the new procedure and a related procedure and a licensing requirement.
The more salient recommendations made to the licensee regarding the procedure are as follows:
1.
The " Symptoms" section of the procedure included the terms " rapidly changing pressurizer level" and " rapidly decreasing pressurizer pressure". Since for a small-breck LOCA, these parameters would not change rapidly, it was recommended that the word " rapidly" be deleted.
2.
Since many of the identified symptoms may appear singularly or in conjunction with other symptoms during certain small-break LOCAs, it was recommended that the procedure should make it clear to the operator that the symptoms may not all appear simultaneously.
In brief, each sumptom statement should be considered to be an "0R" statement.
3.
The "Immediate Action" section of the LOCA procedure includes a statement to the effect that the " Emergency Shutdown" procedure should be carried out.
In reality, only the "Immediate Action" items of slid procedure need be implemented; therefore, it was recomended that such a notation be made.
4.
The "Immediate Action" section also included a requirement to monitor the pressurizer pressure.
Since this information would be of marginal value in the short time period associated with immediate action items, it was recommended that this action be deleted from the "Immediate Action" section.
5.
The procedure did not include a statement to the effect that HPSI should be initiated manually if it failed to be automatically initiated; therefore, it was recommended that such an action be included in the "Immediate Action" section.
6.
It was recommended that the "Subsecuent Actions" section be.estructured so as to give a higher priority to the isolation of leak paths.
1797 279
7.
It was recomended that a cautien s'.atement to the effect that pressurizer level, under the dynamic conditior.s associated with a small-break LOCA or otner depressurizing event, may nct be a reliable indicator of core 1e.el.
In addition to the above recommendatic.s a specific conflict regarding feedwater ficw to the steam generators was discussed. On the one hand, the new LOCA pro-cedure, in keeping with the bases stipulated in the approved guidelines, states that a minimum flow of 300 gpm of auxiliary feecwater flow should be delivered to each steam generator.
On the other hand, the existing "Erergency Shutccwn" procedure, in conformance with a licer. sing restriction related to water-ha mer, restricts the auxiliary feedwater flow to each steam generator to 168 gpm under certain prescribed conditions.
This ciscrepency must be resolved.
Operator Retraining At the time of the audit, the operator retraining program for the small-break LOCA emergency procedures had not beer. initiated.
lle were informed, however, that the retraining program would include a fcur-hour lecture session as well as a pruc.e 'un al naik-uiruugh in the centrol room.
We were further informed tnat this training would be given to eacn licensed cperator orior to Decem-ber 31, 1979.
Subject to fulfilling this commitment, we consider this to be adequate training for the small-break LOCA procedure.
Operator Awareness of the Emergency Procedure ar.d its Bases Two Senior Reactor Operators and one Reactor Operator were interviewed to deter-mine the degree to which they understood the small-break LOCA procedure and its bases. As a result of these interviews, the staff determined that the operators' knowledge of the procedures and their bases were adequate.
However, the following general weakness were revealed which ray reflect certain deficiencies in past operator training programs:
1.
The operators were confusec en what a saturation curve represents.
For example, they did not know how the plant would respond on the saturation curve to an open FORV.
In adcition, they were unsure of the reasons for subccoled, saturated, and superheated plant condi-tions.
2.
The operators did not have a clear understanding of the reasons for tripping the reactor coolant pumps subsequent to the automatic ini-tiation of HPSI due to low RCS pressure.
3.
The operators did not have a complete understanding of the relation between heat removal capability, natural circulation flow, and the delta-T associated with natural circulation.
Further, they did not understand the heat transfer relations across the steam generator.
179.7 280
Systems Considerations For the most part, the instrt. 4.tation and controls associated witr sys ems required to mitigate the effe:ts of small-break LOCAs were located sucP that the necessary operator actior.s :an be performed without undue diffi:ul-In brief, all the required acticns can be performed within the control ro:.
At the conclusion of the plant audit, -he licensee's representatives were aske: if the fforts involved in developing tr.e e ergency procedure and the retrair.'ng :f operators have enhanced the capabilitias of the plant and tha operators tc cc:e with small-break LOCAs and why.
The answers given by the licensee's representatives were generally affirmative; however, tney nad reservations regarding the crash ty;s pro-gram associated with developing and italerenting the new procedures.
The affi mative aspects of their answers included the :enefit derived from the collective contribu-tions of the Owners Group in developing the guidelines, and the increased attention given by the Owners Group and the licensensees' management in developing the :ro-cedures and the specialized operator retraining.
SIMULATOR The B&OTF representatives that particisated in the audit of the Millstone 2 plant also visited the Combus< tion Engineering sinulator to observe its response to a small-break LOCA.
Since the ability cf the simulator to adequately represent a small-break LOCA is somewhat limited, :ertain modifications had to be made.
T.9e actual simulation was accomplished by sinulating the rupture of several steam generator tubes. Hence, the break could be classified as a "large small-break'.
Nevertheless, the systems response was observed, and since the simulation repre-sented the case where offsite power was not lost, system parameters were watchid closely, and when it was detemined that a reactor trip had taken place and tr'it HPSI had been initiated by low RCS pressure, the reactor coolant pumps were tr'pped.
In brief, although this break does not represent tne " worst-case" break fc-C-E designed reactors, the system parare e-s could be reacily conitored durine the course of the event.
C0tlCLUSIONS On the bases of the findings resulting from the B&OTF audit of the Millstone 2 plant, it is concluded that licensee's emergency procedure and operator retraining associated with small-break LOCAs can be implemented by December 31,1979, as e-quired.
It is further cont'uded that -he licensee's emergency procedure for s all-break LOCAs, subject to i'- aeing modi #ied per the staff's recommencations, art operator retraining provide added assu-a.ce that PWR plants and their operators can mitigate the effects of a small-:raak LOCA in an acceptable manner.
3 n7 i / //
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DATE:
7-1-77 Plant Superiritendent EilCLOSURE 2 STATIO:! PROCEDURE COVER SMEET A.
IDEtiIIFICATIO!1 flu:aber OP 2506 Rev.
8 Ti tle LOSS-OF-COOLAtlT It:CIDErlT Prepared By J. Kelley, Jr.
B.
REVIEl1 I have revicaed the above procedure and have found it to be sa tisfactory.
TITLE SIG!!ATURE DATE DEPART!iE!1T 11EAD C.
U:;REVIE',iED SAFFTY QUESTIO:1 EVALUATIO:1 D3CU:'Ef!TATIO:1 REQUIRED:
(Significant change in procedure method or scope as described in FSAR)
YES [ ]
110 [ ]
(If yes, deco:aent in PORC/50RC meeting minutes)
Ef; VIRO:,iE !TAL Il'.?ACT (Adverse environmental impact)
YES [ ]
I:0 [ ]
(If yes, document in PORC/SCRC meeting rainutes)
D.
PORC/50RC APPROVAL PORC/SORC i'eetir.g f; umber E.
APPROVAL A!!D ItsPLEl'EtiTATIOt1 The attached procedure is hereby approved, and effective on the dates below:
Plant Superintendent / Unit Superintendent hpproved Date i.ffective Gate SF-301 F.e v. 2 7^' 78?
OP 2506 Page 1 Rev. 8 Date:
l\\
q LOSS-0F-C00LA!1T IriCIDEf1T PAGE NO.
EFFECTIVE REVISION DATE 1
8 2
8 3
8 4
8 5
8 6
8 7
8 8
8 9
8 10 E
11 8
12 8
13 8
14 8
15 8
16 8
1797 283
OP 2506 Page 2 Rev. 8 Date:
1.
OBJECTIVE To provide a prccedure to ensure that the core is adequately cooled following a loss-of-coolant incident of any size, up to and including the double-ended rupture of the largest reactor coolant pipe; thereby maintaining the release of radioactive products within the limits specified in 10 CFR 100.
2.
DISCUSSI0i1 A loss-of-coolant incident has. occurred whenever primary system leakage is in excess of the makeup capabilities of the charging pumps and pressurizer level cannot be maintained. The LOCA's can be classified into two categories, small and large break.
A small break LOCA is characterized by:
A slow loss of RCS pressure during the short term (10 to 30 minutes) and equilibrium pressure above 300 psia in the long term (30 to 480 minutes) resulting from matching safety injection flow and flow from the break.
A loss of PC5 inventory during the short term followed by a refilling of the RCS during the long term.
Core cooling is by the steam generator (s) and the shutdowa cooling system. The break does.not provide the necessary heat removal yet depletes RCS inventory.
The steam generators provide cooling for natural circulation and, if natural circulation is lost in a boiloff and reflux mode.
The shutdown cooling system is used after the RCS has been refilled and pressure control is provided by the HPSI pumps and the charging pumps.
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OP 2506 Page 3 Rev. 8 bate:
The large break LOCA is characterized by:
A rapid loss of RCS pressure in 10 to 30 seconds, with equili-brium pressures below 300 psia and, in the case of the largest breaks, the RCS pressure nearly equal to containment pressure.
Core cooling is provided for by large flow from the injection system due to low RCS pressure.
The flow from the break provides sufficient heat removal.
Baron precipitation control is required to prevent possible boric acid accumulation in the core.
Reactor coolant system pressure is used to differentiate between small and large break LOCA's.
However, the delineation between small and large breaks does not need to be precise since there is a range of intermediate breaks for which either response will produce satisfactory results. The procedure takes this into account with the decisions to be made after eight hours.
A reactor trip is initiated by thermal margin /lc./ pressure or high containment pressure and the engineered safety systems are actuated as follows:
a.
Safety injection system, containment air recirculation, enclosure building filtration and containment isolation are automatically actuated by low low pressurizer pressure (1600 psia) and/or high containment pressure (5 psig).
b.
Cor.2ainment spray is actuated by high high containment pressure (27 psig).
c.
Hydrogen control is manually actuated by operator action.
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OP 2506 Page 4 Rev. 8 Date:
During a LOCA elevated temperatures inside containment will cause erroneous pressurizer and steam generator level indication.
Procedure steps specifying control peints for level include the necessary corrections.
For containment temperatures up to 300 F pressurizer water level will be between the taps for indicated levels of 19S to 87%.
For steam generators the indicated level must be between 155 and 30%.
In the unlikely event of loss of normal power occurring simultaneously with loss of coolant, ther.e would be no aeditional consequences.
All engineered safety system components receive power from emergency buses 24C and 24D or th-ir associated 480 volt load centers and/or motor control centers. Refer to Emergency Procedure 2503 (Electrical Emergency).
To ensure the diesel generators are not overloaded, the losas are sequentially energized ovec a 20-second time span, rather than energized immediately as they are when normal power is available.
The actuation of the engineered safety systems, containment isolation and natural circulatica for small break LOCA's ensures that fuel damage will be minimized and radioactive releases will not exceed release levels set forth in 10 CFR 100.
Due to the similarity of symptoms for a loss-of-coolant incident and a steam line rupture, it is difficult to differentiate cetween the two events. There are certain parameters which can be used to accurately distinguish which of the following has occurred.
a.
Loss-of-coolant incident.
b.
Steam line rupture downstream of MSIV's.
c.
Steam line rupture upstream of MSIV's/feedwater lire rupture downstream of air assisted check valve and outside the containment.
i777 286
OP 2506 Page 5 Rev. 8 Date:
d.
Steam l'ine/feedwater line rupture inside the containment.
The block diagram on Figure 7.1 provides the operator with a means of establishing which emergency is occurring. The immediate actions of this procedure and Emergency Procedure 2509, Steam Line Rupture, are identical, thus allowing this determination to be made in an orderly panner.
3.
SYMPTOMS'0F LOSS-OF-C00LAllT 3.1 Major Symotons
.3.1.1 Rapidly' changing pressurizer level (C03).
3.1.2 Rapidly decreasing pressurizer pressure (C03).
3.1.3 Increasing containment pressure (C01).
3.1.4 High containment radiation (RC-14).
3.1.5 Auto start of all standby charging pumps (r';2/3).
3.2 Other Symptoms 3.2.1 Safety injection actuation signal (C01).
3.2.2 Containment isolation (C01).
3.2.3 Enclosure building filtration actuation (C01).
3.2.4 Thermal margin / low pressure reactor trip (C04).
3.2.5 Auto start of diesel generators (C08).
3.2.6 PORV/ Safety valve open (C02).
l 3.2.7 High Quench Tank level, temperature and/or pressure (C03).
4.
AUT0ftATIC ACTIOt1 N/A 5.
IMMEDIATE ACTI0'!S 5.1 Carry out G ergency Procedure 2502 (Emergency Shutdown).
5.2 Five seconds after verifying CEA's fully inserted and verifying l
a low pressuri:er pressure safety injection actuation, trip l
all four reactor coolant pumps.
(CO3).
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OP 2506 Page 6 Rev. 8 Date:
5.3 If steam generator pressure is less than 500 psig, verify auto closure, or manually close, the main steam isolation valves (MSIV's)(C05).
5.4 Check Panel C0lX to vn
- v that all equipment associated with the following signals is in the accident mode.
Safety injection actuation signal (SIAS).
Containment isolation actuation sicnal (CIAS).
Fnclosure building filtration signal (EBFAS).
I 5.5 Monitor pressurizer pressure t a determine the blowdown rate, which is indicative of the magnitude of the pipe break (C02/3).
5.6 As pressurizer pressure decreases, verify the following:
5.6.1 High pressure safety injection (HPSI) flow starts at approximately 1100 psig (CGl).
5.6.2 Low pressure safety injection (LPSI) flow starts at l
approximately 200 psig (C01).
5.6.3 Safety injection (SI) tank level and pressure start decreasing at approximately 200 psig (C01).
5.7 Check containment pressure (C01):
5.7.1 Verify containment isolation at a containment pressure of 5 psig or pressurizer pressure at 1600 psia.
(C0lX).
5.7. 2 Verify containment spray actuation at 27 psig l
(C01X).
5.8 Initiate site radiation emergency plan (Emergency Procedure l
2501).
6.
SUBSEQUENT ACTIONS 6.1 Refer to Figure 7.1 to identify the emergency.
If the emergency can not be clearly identified assume a LOCA and continue with this procedure.
If a steam leak or steam generator tube rupture is indicated refer to E0P 2509/2515.
N77 288
OP 2506 Page 7 Rev. 8 D' ate :
fiOTE:
Whenever hot leg temperature (T ) is specified the H
in-core thermocouple temperature should also be monitored. This will be particularly useful if TH is off-scale. To denand incore thermocouple reports from the computer, enter "X SP 1 SP 0 E0T".
b.2 Accomplish the following to ensure core cooling is established.
6.2.1 If pressurizer pressure and level control have not been reestablished.
6.2.1.1 Verify that steam dump and bypass valves or atmospheric dump valves are maintaining T at saturated conditions for the pressurizer H
pressure (C05/C03).
CAUTI0ft:
Ensure steam generator feed rates do not cause RCS cooldown in excess of 100 F/hr.
- 6. 2.1. 2 Continue feeding the steam generators at a minimum flow of 300 GPM per steam generator (600 GPM if only one steam generator is available) until level has returned on scale and is increasing. Then feed as necessary to return level to between 705 and 80% (C05).
6.2.1.3 Verify natural circulation has initiated f
and a AT of greater than 10 F is indicated (C03).
6.2.1.4 Ensure continued required injection by HPSI, LPSI and charging pumps (C01/3).
- 6. 2.1. 5 Monitor the voiding in the core by observing 6 and excore nuclear indication.
If natural circulation appears to be degraded or lost ( $ > 50'F) ensure the above actions maintain saturated conditions with a constant T by boiling reflux in the steam g
generators (C03).
J l'
1797 289
OP 2506 Page 8 Rev. 8 Date:
6.'2.1.6 Within one hour comt.ence a cooldown per OP 2207, Plant Cooldonn, with the following thanges.
6.2.1.6.1 Maintain steam generator level between 703 and 80:; (C05).
6.2.1.6. 2 I f break flow exceeds sa fety injection ficw maintain pressure by lowering Tg (saturated condition)
(C05).
6.2.1.6.3 If break flow is less than injection flow control pressure by throttling the injection flow i
valves (C01).
6.2.1.6.4 If LOCA conditions result in alignments which do not permit completion of certain steps, delete those steps.
fiOTE:
Use the Subcooling Meter only when I
T > 515 F.
For lower g
temperatures refer to Figure 7.2 to determine subcooling margin.
6.2.1.7 If pressure control is reestablished per i
step 6.2.1.6.3, maintain T at least 50"F H
subcooled for the remainder of the cooldown (C05/C03).
6.2.2 If pressurizer pressure and level control have been reestablisned.
6.2.2.1 Verify that steam dump and bypass valves or atmospheric dump valves are controlling Tavg (C05).
1797 290
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OP 2506 Page 9 Rev. 8 Date:
CAUTI0ti:
Ensure steam generator feed flow rates do not adversely affect pressurizer pressure and level control, nor result in a RCS cooldown in excess of 100 F/hr.
6.2.2.2 Continue feeding the steam generators at a minimum of 300 GPM per steam generator (600 GPM if only one steam generator is available until level is on-scale and increasing. Then feed as necessary to return level to between 70i' and 80% (C05).
6.2.2.3 Verify natural circulation by observing a AT of between i0 F and 50 F with TH stable and T constant or decreasing C
(C03).
fl0TE:
Use the Subcooling Meter only when TH > 515 F.
For lower temperatures refer to Figure 7.2 to determine subcooling margin.
6.2.2.4 Verify a lack of voiding by maintaining T at g
6.2.2.5 Ensure continued required injection by the HPSI, LPSI and charging pumps (C01).
6.2.2.6 Within one hour commence a cooldown per OP 2207, Plant Cooldown, with the following changes.
6.2.2.6.1 Maintain steam generator level bstween 70% and 80% (C05).
6.2.2.6.2 liaintain pressurizer pressure by securing HPSI pumps and/or throttling injection valves while maintaining T at least g
50 F subcooled (C01).
- ~ [' [)
8 L //8
OP 2506 Paae 10 Rev. 8 Date:
6.2.2.6.3 If LOCA conditions result in alignments which do not permit completion of certain steps delete those steps.
6.3 If core uncovery is indicated by noting superheated core outlet temperatures (T or incores), verify proper operation H
of the safety injection system and the steam generators (C01/C02/C05).
6.4 If neither steam generator is operable per step 6.3 continue maximwn safety injection available and open the PORV's (C01).
6.5 Rev'eu the ia iowing potential leak paths and isolate as required.
6.5.1 Monitor quench tank level, pressure ard temperature, PORV acoustic position indication and PORV discharge pipe temperature.
If a stuck PURV is indicatcd (RCS pressure less than 2300 psia) close the blocking valve of the stuck open/ leaking PORV.
(C03) The other isolation valve should be open to provide PORY relief protection.
6.5.2 Verify the letdown line is isolated (CO2).
Verify the RCS sample lines are isolated (locall.
6.6 If RCS integrity is reestablished, attempt to return normal pressurizer level and pressure control by using the charging pt. 7s and high pressure safety injection pumps to restore pressurizer level to between 40% and 50% and the ormsurizer heaters to elevate pressure.
Increase pressurizer temperature to at least 50 F above reactor coolant system hot leg temperature (C01/2/3).
NOTE:
The following stec is accomplished.to prevent contain-ment equipment damage due to unnecessary containment spray.
6.7 Reset the containment spray actuation modules and secure the containment spray pumps if the following conditions exist.
(C01/ local)
- 7 7
292
OP 2506 Page 11 Rev. 8 Date:
6.7.1 Containment spray pumps started due to an appa,ent pressure spike.
6.7.2 Containment pressure is less than 10 psig.
(C01 )
6.7.3 The pump is not operating on containment sump recircu-lation.
(Col) 6.8 Secur e all unneeded secondary ecuipment such as main feed pumps, condensate pumps, heater drain pumps, etc.
6.9 Re-affirm that all necessary emergency equipment is functioning by 'checiiag C0lX against other available indicators.
6.10 Check rac'ation monitoring panel and secure control room access doors (RC-14).
6.11 Monitor boric acid tank levels and, when depleted, secure both boric acid pumps and all three charging pumps (C02/3).
6.12 Monitor refueling water storage tank (RWST) level and, when level decreased to 9.5 percent, manually initiate sump recircu-lation actuation signal ( SRAS) by depressing both push-buttons on C01, and obtain the keys from the operational key locker for 2-SI-659 and 2-SI-660; place in the key lock switches on C01 and select the "0PER" position.
NOTE:
SRAS will automatically occur at 9.5 percent level in RWST if the manual initiation signal is not given.
6.13 When SRAS is initiated, verify the following:
6.13.1 Both LPSI pumps stop (C01).
6.13.2 Both SI/CS pumps minimum recirculation valves close (2-SI-659, 2-SI-660)(C01 ).
6.13.3 Both containment sump outlet header isolation valves open (2CS-16.1A, 2CS-16.lB) (C01).
6.13.4 Both shutdown cooling heat exchanger RBCCW outlet valves open (C06).
793
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OP 2506 Page 12 Rev. 8 Date:
6.13.5 If any HPSI pump flow is less than 30 GPM, turn off all charging pumps and all but one HPSI pump to ensure pump minimum flow.
If 50 F subcooling can not be maintained restart HPSI pumps as necessary.
6.14 When safety injection tanks are empty, install the breaker closing coils and close the four tank isolation valves (C01).
6.15 Within twc hours of LOCA initiation, start boron precipitation control per OP 2214.
6.16 After 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when pressurizer pressure and level control is re-established and pressurizer pressure exceeds 300 psig, perform the fo'llowing:
6.16.1 Continue cooldown of the RCS and obtain a RCS sample.
If RCS activity is satisfactory initiate shutdown cooling per OP 2310.
If necessary reduce pressure by throttling the HPSI injection valves.
Ensure the 50 F subcooling is maintained.
6.16.2 If 6.16.1 cannot be completed, continue decay heat removal using auxiliary feed and steam dump.
6.16.3 If 6.16.1 and 6.16.2 cannot be completed, open the pressurizer PORV's and secure post-incident boron precipitation control.
6.17 If pressurizer and level control cannot be re-established and pressurizer pressure remains below 300 psig, continue safetl<
injection and post-incident boron precipitation control.
6.18 After carefully evaluating the indicated paraneters, secure the specified equipment only if other abnormal conditions require.
NOTE:
When securing any equipment, if possible, first reset the actuation module on the safeguard panel.
If after securing the equipment, coi Jitions worsen to the point where the prerequisites are no longer met, restart the secured equipment.
1-n, 294
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OP 2506 Page 13 Rev. 8 Date:
6.18.1 Secure HPSI pumps only when all the following condi-tions exist (C01).
6.18.1.1 The HPSI pumps suction is still aligned to the R!iST (no SRAS) (C01).
6.18.1.2 All RCS hot leg and cold leg temperatures are at least 50*F subcooled (C03).
6.18.1.3 Pressurizer level has been restored and responds to charging pump operation.
(C03).
6.18.1.4 The ~ reactor is shutdown as indicated by all rods inserted and cold shutdown baron concentration (by sample). (C04).
6.18.1.5 Core cooling is being provided by the steam generators or shutdown cooling.
For steam generator operation natural circulation (Th-Tc
-<50 F) is required and the steam generators must have steam and feed flow and.evel between 70% and 80%.
(C01/5).
6.18.2 Secure HPSI pumps if minimum pressure temperature limits as specified in technical specifications are approached (Tc< 275 F) (C03/1).
(This step's require-ments are to be followed regardless of the status of the conditions specified in 6.16.1.)
6.18.3 Secure the charging pumps only if T is greater than h
50 F below saturation temperature, pressurizer level is restore <J and responds to charging pump operation, and pressure is greater than 1600 psia and-increasing.
6.18.4 Secure other safeguard equipment only after determining other equipment is providing the desired function or the function is no longer required.
6.19 Initiate post-incident hydrogen control system.
Refer to OP 2313C (Containr.ent Post-Incident Hydrogen. Control).
G;7 295
OP i.506 Page 14 Re'i. 8 Date:
6.20 Continue to' operate the control room ventilation in the recirculation mode in accordance with OP 2315A.
7.
FIGURES 7.1 Break Identification Chart 7.2 Saturation Curve JK:amc 1797 296
OP 2506 Page 15 mm o
e LD D'
D 97 P.ev. 8 o
Date:
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_ _. 7-1
_77._
Pl en t Supe. *'i nic odeo t EHCLOSURE 3 STAT I0'l PE:0CFO 'P.E COVER StiEET A.
IDEtiT_IFICAT10:1 i' umber OP 2502 Rev._
9 Title EP.ERGEllCY SHUTDOWtl (REACTOR TRIP)
Prepared Bv J. J. Kelley, Jr.
B.
?_Ey]sq I have retic..ed the ebove procedure and have found it to be satisfactory.
T ITL E_
SIGNATURE DATE LEP6.Ri!1ENT HEAD C.
U:;REVIE'.lED SAFETY QU.ISTIO:1 EVALUATIO:: DOCU:'.EliTATION REOUIRED:
(Significant change in procedure r.ethod or scope as described in FSAR)
YES [ ]
I'0 '[ ]
(if yes, document in PORC/ SOP.C meeting minutes)
E!; VIRO:D;E;!TAL li' PACT (Adverse environmental impact)
YES [ ]
!!O [ ]
(If yes, document in PORC/SORC meeting miautes)
D.
PORC/SCRC APPROVAL PORC/50RC l'eeting t; umber E.
A_PPROVAL AtiD ii'.PLEi'.E:!TATIO!!
The. attached procedure is hereby approved, and effective on the dates belcw:
plant Superintendent / Unit Superintendent Approved Date Ef fecti ce Date SF-301 Rev. 2 17^7 299
OP 2502 Page 1 Rev. 9 Date:
EMERGENCY SHUTDOW1 (REACTOR TRIP)
PAGE NO.
EFFECTIVE REVISI0rl DATE 1
9 2
9 3
9 4
9 5
9 6
9 7
9 8
9 9
9 10 9
11 9
12 9
1777 300
OP 2502 Page 2 Rev. 9 Date:
1.
OBJECTIVE 1.1 To provide a procedure to bring the plant to a hot shutdown condition after an emergency shutdown has been initiated automatically or manually by operator action.
2.
DISCUSSION 2.1 Any shutdown of the' plant accomplished by rapid insertion of the control elements (also called " Reactor Trip" or Scram") is considered an emergency shutdown.
Emergency shutdown will be automatically initiated by the Reactor Protection System (RPS) whenever certain continuously monitored parameters surpass a
. predetermined setpoint.
Emergency shutdown can also be initiated manually by operator action if unstable plant conditions warrant such action.
The operator is alerted to unstable plant conditions by several "Pretrip" alarms. Additionally, due to the fail safe nature of the RPS, a malfunction of the system can also cause an emergency shutdown.
A reactor trip occurring early in core life or after a prolonged plant shutdown leaves little decay heat to be removed.
This creates the possibility of cooling the reactor coolant by the rapid insertion of feedwater (or addition of too much feedwater) to the steam generators following a reactor trip.
This cooling could result in a reduction of shutdown margin.
To help avoid this rapid over feed of water, the feed water regulating valves to the steam generators will-automatically ramp down to a 5% open position and will then automatically go to manual control.
After an emergency shutdown the reactor may not be restarted until a Plant Incident Report (Station Form SF-1001) has been completed and all requirements of administrative control procedure ACP-QA-1.07 have been met.
,-n 29 JU 1
OP 2502 Page 3 Rev. 9 Date:
3.
SYMPTCMS OF EMERGEllCY SHUTDOWl1 3.1 Major Symptoms 3.1.1 All control element assembly (CEA) electrical limit lights on as indicated on core matrix.
(C04) 3.1.2 CEA position iniication shows all full length CEA's on the bottom as indicated by position step indicators and netrascope.
(C04).
3.1.3 Reactor power indicators and recorders show rapid decrease in reactor power.
(C04).
3.1.4
" Reactor Trip" annunciator alarm on panel C04 (red alarmlight).
3.1.5 Annunciator drops verifying reactor trip circuit breakers (TCB's) (8) have tripped (C04).
3.1.6 Green indicating lights illuminated for TCB's (8)(C03).
3.2 Other Symotoms 3.2.1 Any one or more of the following annunciator alarms (red alarm lights) on panel C04.
Approximate Trip Set Point 1.
Steam Generator Low Level 36.0%
2.
Steam Generator Low Pressure 500 PSIA 3.
Reactor Coolant low Flow 4 pump Operation 91.7%
4.
Pressurizer High Pressure 2400 PSIA 5.
Thermal Margin / Low Pressure Variable 6.
fluclear Instrumentation High Power Power Q 7.
Turbine Trip, Low Hydraulic Fluid Pressure 500 PSIG 8.
Local Power Density Refer to Limit Lines of Tech.
Spec. Figure 2.2-1 and 2.2-2.
)l) b0-
t OP 2502 Page 4 Rev. 9 date:
9.
Containment High Pressure 4.75 PSIG 10.
Reactor Coolant Pump Underspeed 829 RPM 3.2.2 The reactor has been manually tripped because TAVG had decreased to less than 515 F for greater than 15 minutes.
3.2.3 The reactor has been manually tripped because of loss of one or more reactor coolant pumps.
3.2.4 The reactor has been manually tripped because number one and/or number two steam generator level has increased to 90%.
3.2.5 The reactor has been manually tripped because of a sustained start-up rate of 2.5 decades per minute.
3.2.6 The reactor has been manually tripped because number one and/or number two steam generator level has decreased to 45% with reactor power greater than 5%.
3.2.7 Manual trip (after manual turbine trip) upon a dropped CEA and pressurizer level is less than 20%
(0.P. 2391, Dropped CEA).
4.
IMMEDIATE ACTIONS 4.1 Ensure the reactor has tripped by depressing the four (4) reactor trip pushbuttons (C04).
4.2 Verify that reactor power is decreasing.
(C04).
4.3 Verify that all full length CEA groups are fully inserted (C04).
4.4 Ensure the turbine has tripped by depressing the turbine trip button (1) on the EHC insert panel (C07).
Verify that all steam admission valves indicate closed (C07) and that generator megawatts indicate zero or negative (C07).
4.5 Ensure the generator ACB's (15G-8T-2 and 15G-9T-2) are open.
(C08)
If not open, trip by using the emeroency trip pushbuttons (C07).
(Push both buttons simultaneously).
4.6 Verify the transfer of 6.9KV and 4.16KV buses to reserve station service transformer (RSST) (C08).
4797 303
OP 2502 Page 5 Rev. 9 Date:
4.7 Verify feed water flow decreasing as the feedwater regulating valves are ramping down to the 55 open position (C05).
4.3 If the feed pumps are in manual speed control, ramp feed pump speed to minimum speed, and if two pumps are running, trip one pump.
(C05) 4.9 If the main feed pumps have tripped and/or are not available start the electric and/or steam driven auxiliary feed pumps (C05).
4.10 Adj0st feedwater flow rate to at least 300 GPM per steam generator (600 GPM if only one steam generator is available) to return steam generator water level to 70-80%.
If level has decreased below 45% limit feedwater flow to less than 600 gpm per steam generator.
If feedwater flow has been lost for greater than 15 minutes and level is less than 45% limit feedwater flow to less than 168 gpm per steam generator.
(C05)
CAUTION:
ENSURE FEEDWATER FLOW RATE IS CONTROLLED TO ALLOW PRESSURIZER LEVEL AND PRESSURIZER PRESSURE CONTROLS TO FUNCTION PROPERLY AND MAINTAIN THOSE PARAMETERS WITHIN NORMAL LIMITS, AND TO NOT ALLOW A RCS C00LDOWN IN EXCESS OF 100 F/HR.
4.11 Trip both heater drain pumps (if running).
(C05) 4.12 Trip two condensate pumps if three were running.
Trip one condensate pump if two were running.
(C05) 5.
SUBSEQUENT ACTIONS 5.1 Monitor Primary System Temperature.
5.1.1 Verify that steam dump and bypass valves are functioning to reach and maintain a no-load TAVG of 532 F (C04-C05).
5.1.2 If the main steam isolation valves have closed or the main condenser is not available, use the atmospheric dump valves to maintain TAVG at 532 F (Corresponds to 900 PSIA secondary pressure).
(C05).
^7 304
OP 2502 Page 6 Rev. 9 Date:
5.2 Monitor steam generator water level and adjust flow to maintain level between 70% and 80%.
(C05) 5.2.1 If steam generator level continues below 455 limit feedwater flow to less than 600 GPM per steam generator; if flow was interrupted for 15 minutes or longer and level is less than 45% limit flow to less than 168 GPM per steam generator.
(C05) 5.2.2 Monitor condensate storage tank level.
If low level is indicated respond per the applicable sections of OP 2319B, Condensate Storage and Transfer.
(C05).
5.2.3 If all feedwater is lost to both steam generators or if for any reason both steam generators become inoperable re'er to E0P 2521, Loss of Feedwater/ Steam Generators.
5.3 If using the auxiliary feed pumps to supply the steam generators, secure the normal hydrazine feed pumps, open valves 2MS-15A and 2MS-15B and then start the auxiliary chemical feed pumps.
CAUTION:
IF ALL CONDENSATE PUMPS MUST BE SECURED, REMOVE STEAM FROM THE TURBINE BUILDING PRIOR TO SECURING THE PUMPS BY CLOSING BOTH MAIN STEAM ISOLATION VALVES.
ENSURE STEAM DUMP TO CONDENSER IS SECURED AND ATMOSPHERIC DUMP VALVES ARE BEING USED TO CONTROL TAVG AND THEN BREAK CONDENSER VACUUM.
5.4 Monitor pressurizer level decreasing to reach the no load programmed level of 40%.
Switch to manual control if r.ecessary, and maintain level between 40% and 50%.
(CO2/3) 5.5 Use manual control of pressurizer sprays and heaters if necessary to return primary system pressure to normal (2250 PSIA) (C02/3).
NOTE:
Use the Subcooling Meter only when Tg1515 F.
Use Figure 6.2 when TH 1515 F to determine subcooling margin.
e, l///
303
OP 2502 Page 7 Rev. 9 Date:
5.5.1 If pressurizer prc:: cure centro! is lctt: commence a cooldown per OP 2207, Plant Cooldown and attempt to maintain T 50 F subcooled.
g 5.5.2 If saturation conditions are approached, (less than 20 F subcooled) observe reactor coolant pump parameters for indication of void formation. Void formation would be indicated by the following.
5.5.2.1 Pump current:
less than 400 amps 5.5.2.2 Steam Generator A P: < 10 psid (2 pumps)
< 20 psid (3 pumps)
< 30 psid (4 pumps) 5.5.2.3 RCS 6: greater than 5 F 5.5.2.4 RCP Vibration alarms present.
(C03).
5.5.3 If voiding is indicated and a low pressurizer pressure l
SIAS is not present, do not secure RCP's unless at least one of the following conditions exist:
5.5.3.1 One RCP per loop will continue to operate (C03) or, 5.5.3.2 The RCP's are not providing forced flow as indicated by low steam generator t. P (< 3 psid), low RCP current (<100 amps) and high reactor vessel ai (T ~ C > 10 F)
H (C03) or, 5.5.3.3 RBCCW to the RCP's has been secured and i
cannot be restored and pump temperatures exceed alarm limits.
(C03) 5.6 If the pressurizer power operated relief valves have lifted and primary pressure has decreased below 2350 psia, attempt to reset the acoustic PORV valve position moniter and monitor quench tank parameters and discharge pipe temperatures to verify the relief valves have reseated (C03).
1797 306
OP 2502 Page 8 Rev. 9 Date:
5.7 If the pressurizer power operated relief valves appear to be stuck open, close the affected PORV's isclation valve (2-RC-403 or 2-RC-405). Maintain the stuck open PORV's isolation valve closed (2-RC-403 or 2-RC-405) (C03).
5.8 If steam generator safety valves had lifted, verify that they have reseated. (C03) 5.9 Unless a low pressurizer pressure safety injection initiation has occurred, ensure at least one reactor coolant pump (RCP) per loop is running.
(l A or 1B should be one of the running pumps for pressurizer spray purposes).
(C03) 5.10 Record all relay actions on rear of C08, open 15G-2X1-4, (visually verify stabs open) reset the relays, and then reclose 15G-GT-2 and 15G-9T-2.
(C08) 5.11 Secure secondary plant equipment not necessary in maintaining a hot shutdown condition.
5.12 Verify main turbine motor suction pump running and manually start lift pumps and turning gear motor to ensure safe coast-down of main turbine generator (C07).
NOTE:
Lift pumps and turning gear will auto start if Step 5.12 is omitted.
5.13 Verify steam generator feed pumo turbine auxiliary oil pump running.
Feed pumps turbine turning gear will automatically engage at zero speed.
(C05) 5.14 Determine cause of reactor trip and correct.
5.14.1 If cause of trip was low condenser vacuum, refer to Emergency Procedure 2507 "Lcss of Condenser Vacuum".
5.14.2 If cause of trip was high containment pressure, refer to Emergency Procedure 2506 " Loss of Coolant Incident" or 2509 " Steam Line Rupture".
5.14.3 If cause of trip cannot be corrected without plant cooldovin, initiate plant operating procedure 2205
" Plant Shutdown".
1797 307
OP 2502 Page 9 Rev. 9 Date:
5.15 Fill out Station Form SF-1001 (Plant Incident Report) and notify higher supervision in accordance with ACP 1.07.
5.16 As soon as conditions permit reset the turbine as follows:
5.16.1 Reset and close the reactor protective system switch-gear breakers.
(C04) 5.16.2 Reset the turbine and verify the intermediate stop valves (4) stroke to the full open position.
(C07) 5.16.3 If the intermediate stop valves do not open, EHC pressure can be increased to provide additional force per the following:
5.16.3.1 Loosen the lock nut on the pressure compensator at the pump.
5.16.3.2 Note original EHC pressure.
NOTE:
The pump discharge relief valves are set at approximately 2000 PSIG.
5.16.3.3 Turn the compensator clockwise until pressure equals 1900 PSIG.
5.16.3.4 Af ter the valves open or if the extra pressure does not open the valves return the pressure to its original value.
5.16.4 After the intermediate stop valves open the turbine may be tripped.
(C07) 5.17 If condenser vacuum is being maintained, 'onitor turbine steam seal header pressure and adjust as necessary to maintain 4 PSIG on the header.
(C07) 5.18 Notify Chemistry to perform an isotopic analysis for Iodine within a 2 to 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> period from the time of the trip if reactor power was greater than 15t; at the time of the trip.
5.19 Monitor wide range power instrumentation.
If the count rate for any channel decreases below.5 CPS and the channel is required to meet Technical Specification minimum channel requirements, reconnect the B-10 detector by depressing the B-10 off pushbutton.
17?7 308
OP 2502 Page 10 Rev. 9 Date:
5.20 Monitor system voltage and if necessary request CONVEX to adjust system voltage to ensure adequate in house voltage.
(C03) 6.
FIGURES 6.1 Break Identification Chart 6.2 Saturation Curve JK:amc 17?7 309
OP 2502 Page 11 Rev. 9 Date:
BREAK IDEl:TIFICATICli CilART Per Level Changing and Pr.r. Press. Rapidly Decreasing s
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ENCLOSURE 4 ATTENDEES AT THE EXIT If4TERVIEW DURING NRR'S B&OTF AUDIT OF StiALL-SREAK LOCA PROCEDURES AT MILLST0i;E 2
_0N DECE!'BER 19, 1979 NNECo.
E. C. Farrell (Plant Superintendent )
J. J. Kelley W. McRoy D. Wright R. Gouzza S. Myers liRC R. Zimmerman (IE Resident Inspector)
E. L. Connor (NRR ORPM)
P. Boehnert (ACRS Staff)
G. B. Kelly (B&OTF)
K. R. Mahan (B&OTF)
I. Villaiva (B&OTF) 1797 312
ENCLOSURE 5 M
$5.7/,'.'[.I.E.d[
27 December 1979
'.5'?5:,b:.?'b::'
?
L L I 13 M. Connor ln b YL.
N
.C.
Farrell N
Instrument Power Supplies Attached are the requested instrurwint power supplies for your information RCT/JJK/per L
l l
a.
Reactor Cooling Systent Pressure P 100K VA 10 and VR 11 Vital (Computer Indicator On.!
Nonvital (No Control Room Indicator / Control Functions)
P 100Y VA 20 and VA 21 Vital (Computer Indicator Oni Nonvital (No Control Toom Indicator / Control Functions)
P 102A VA 10 Vital P 1028 VA 20 Vital P 102C VA 30 Vital P 102D VA 40 Vital P 103 VA 20 Vital P 103-1 VR 11 Nonvital PR 10 2B-1 VR 21 NortVital PR 100 VR 11 Nonvital b.
Wide Ran~e RTDs TR 115 VR 11 Nonvital TR 125 VR 21 Nonvital T
101 VR 21 Nonvital T
109 VR 21 Nonvital T
121X VR 21 Nonvital I
c.
Core Thermalcouples self powered d.
Steam Generator Level LI 1111A VA 10 Vital LI 1113a VA 20 Vital LI lil3C VA 30 Vital LI lil3D TA 40 Vital LT ll23A VA 10 Vital LT 1123B VA 20 Vital LT 5272 VR 11 Nonvital LT 5274 VR 21 Nonvital i
e.
Pressurizer Level LI 103 VR 11 Nonvital LIC 110K VR 11 Nonvital LIC 110Y VR 21 Nonvital LR 110 VR 11 Nonvital E.
Refuel trater Storage Tank Level LI 3001 VR 11 Nonvital LI 3002 VR 21 MonVital LI 3003 VR 11 Nonvital LI 3004 VR 21 MonVital q.
P ant Computer 1
B 6155 Vital
)[9[4 3 B 6156 Vital h. High Pressure Safety Injection F 311 VM 11 Nonvital F 321 VR 11 Nonvital F 331 VR 21 MonVit.a1 F 341 VR 21 Nonvital k
e 1. Aux Feedwater Flow F 5277A VR 11 IttmVital F S278A VR 21 Monvital J. PORV Acoustic Monitors VA 10 Vital k. PORVs 2-RC-402 0 12 Vital 2-RC-404 D 22 Vital L. PORV Isolatlon Valves 2-RC-403 B 5155 Vital 2-RC-405 B 6120 Vital 3. 1797 315 i
+ D"* WD CW s J 5.' k??' so, g ' c'ecr'1 censing M a s:cator i.]7 ':..'.- * ;'__,;_! a n,e'rs _2 :_a J Cc s.;mers power Com ~an.
- T~i i; - Company
- 212 'lest Michigan A.enue [' 0 uackson, Michigan 4923 .j g._j ],.11-]... Mr. Willian Cavana;c., II: Mr. Ken Morris ,. ice Chairran Executive Of rector of Generation combustion Encir.eering Owners Group and Construction Cmaha Public Pcwer District Arkansas Power & Light Company Fourth & Jones P. O. Box 551 Omaha, !!abraska 63103 Little' Rock, Arkansas 72203 Mr. John Garrity, Chairman Mr. A. E. Lundvall, Jr. Guidelines Subgroup Vice-P. resident - Supply Baltimire Gas & Electric Company Maine Yankee Atcmic Power Company Edison Crive P.. Box 1475 Augusta, Maine 04336 Baltimore, Maryland 21203 Mr. Theodore E. Short Mr. Robert T. Harris, Chairman nnalysis Subgroup Assistant General Manager il rtheast Utilities Service Co. Omaha Public Power District P. O. Box 270 ) Hartford, Connecticut 06101 0aa ie ra k 8102 Mr. David S. Van de Walle Mr. Robert H. Groce Consumers Power Ccmpany Licensing Engineer 212 West Michigan Avenue Yankee Atomic Electric Company 20 Turnpike Road Jackson, Michigan 49201 Westboro, Massachusetts 01581 Mr. William Szymczak Yankee Atcaic Power Company 25 Research Drive Mr. W. G. Counsil, Vice-P. resident Westboro, Massachusetts 01581 Nuclear Engineering & Operatior flortheast fluclear Energy' Company P. O. Box 270 Mr. J..i. tnos Hartford, Connecticut 06101 Arkansas Power & Light Company P. O. Box 551 Dr. Robert E. Uhrig, Vice-Preside Little Rock, Arkansas 72203 Advanced Systems & Technology Florida Power & Light Company P. O. Box 529100 Miami, Florida 33152
} Dist i ticn ' is t - ' E L.J.7,e: Fin s E. Case P-404 B. Campbell 357 R. Mattson F-ll;; E. Jordan, I&E D. Eisenhut 533 Local PDR D. Ross 275 J R Buchanan, NSIC D. Vassallo 278 TERA W. Kane 242 E. L. Connor T. Novak P-il32 P. Boehnert, ACRS R. Reid 330 Z. Rosztoczy P-1030 D. Ziemann 314 S. Israel P-ll32 W. Hodges P-1030 P. Matthews P-822 G. Mazetis P-1132 P. Norian P-1030 I. Villaiva 242 N. Wagner P-ll32 T. Wambach 314 R. Woodruff E/W 359 K. Mahan 357
- 8. Wilson 357 G. Kelly P-ll32 J. Lee 142 W. Gammill 266 S. Grimes 340 M
..RC PDR i. Docket (Central) Files soon%~ ^ ' * * " ' ' )797 317}}