ML20049J671

Responds to 810902 Telcon Re Alternate Shutdown Sys.Safe Shutdown Can Be Achieved by Use of Control Room Evacuation Panel for Postulated Fires in Control Room,Relay Room & Two Cable Spreading Rooms.Certificate of Svc Encl

ML20049J671
Person / Time
Site:	Summer
Issue date:	10/16/1981
From:	Nichols T SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML19291F826	List: ML20049J662 ML20049J665 ML20049J667 ML20049J669 ML20049J671
References
NUDOCS 8203190121
Download: ML20049J671 (4)

Text

.. ., V C A 'T u / atta N w h

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SOUTH CAROLINA ELECTRIC a gas COMPANY F'~e -

m, o, r.c c ... ,

CotumsA, SoutM CARQUNA 29288 g L T. C. NicH oLs, J R. O::tober 16, 1981 '

v.ese.w - w unem Mr. Harold R. Dentcn y

p ly;f e Office of Naclear Reactor Regulaticn U. S. Nuclear Regulatory Ccmnission Washington, D. C. 20555

Subject:

Virgil C. Summr Nuclear Station D:cket No. 50/395 Alternate Shutdown Systsu

Dear Mr. De:

W:an:

In respnse to a telephone conversation with NRC staff manhs on Septenber 2,1981, South Carolina Electric and Gas Ctropany p:x:vides the follcwing information.

Safe shutdown can be achieved utilizing the Cbntrol Rocrn Evacuation Panel (CREP) for pstulated fires in the control recru, relay recxn, and the two cable spraMing rocas (control c:rnplex). As documented in the SCE&G Fire Protection Evaluation, fires outside the control ccmplex do not regaire operation frczn CREP.

In the event of a fire .@ing utilization of the CREP concurrent with total loss of offsite pwer, the turbine edriven emergency feedwater pump, 'the correspnding ficw control valves and the steam code safety valves are the initial equipt required for maintainance of a Hot Standby conditicn for at least two hours. 'Ihe turbine driven emergency feedwater pump and diesels start .

autcrratically cn a loss of offsite pwer. An alternate control p wer -

supply is not needed since the steam valve for the turbine fails open l

. on loss of p wer and/or air. me code safety valves are autcmatic l

mechanical devices. Se flow control valves (IFV 3536-EF, IFV 3546-EF, and IFV 3556-EF) can he controlled frcm the CREP if control air is available; without air, one tran is recuired to control these i valves locally. Steam generator level indicators with circuits and pwer supplies independent of the control hiilding are available at the CREP (LI 477B, LI 4878, LI 4973). Radio ccrmunicaticn is available between the CREP and the valve operating station.

Calculations have shown that reactor coolant sys' ten makeup is not needed for nere than two hours after a reactor trip and loss of offsite pcwer. However, opting procedure EOP-8 is arranged to ensure the capability for makeup can be es*ablished within a short l

time after the start of the incident. Es+mh W hing the capability,., .1 __.g.

for makeup regaires the follcwing:

1

1. Iocal tranual tripping of the switchgear breaker for the offsite source to cne of the two. main Class lE huses and I 8203190121 820311 PDR ADOCK 05000395 0 PDR

.. - y

• i

. Mr. Harold R. Denton October 16, 1981 .

Page 2 ,

locel manual closure of the a_=Matai diesel generator l

breaker.

2. Starting a service water p ap using controls at the CREP

• (these controls have an alternate source of control that is irb--drisit of the control ccmplex.)

l 3. Starting the associated chilled water pump with local manual i

operation of the switchgear 1:neaker.

4. . Starting r. chiller frm local control switcf1 in the chiller control package and subsequent local manual cperation of the j

sWitChgear breaker.

5. Starting a charging pmnp by local manual option of the i

switchgear breaker.

6. Provide an initial source of horated water to the cfarging i

pump by opening valve XVI 8104-CS and starting boric acid transfer pzup XPP 13B-CS. -Both the valve and pznp can 1:e I

controlled frm the CREP (these controls have a source of control pwer independent of the control emplex). Also, boric acid tank level indication (LI-161 and LI-163 which I

have pwer supplies independent. of the control emplex) are

~

available at the CREP.

'7. Pressurizer level indication (LI-459B diich has );ower supply

, infependent of the control ccruplex) is available at the CREP.

1he varipus HVAC systems listed in Attachment I to the Fire P-t OLEction Evaluation are needed to ensure the long term reliable operation of the associated rocess I systern equipnent. Recent preliminary evaluations have shown that HVAC systems are not required

. for some t2me after the start of the incident.

. The design nethod for depressurization in the. event that shutdown .

utilizing the CREP is required is the use of the Iressurizer power operated relief valves (PORV's). Valves PCV 445 A and PCV 444 B can both be opted frcIn the CREP, have electrical Jower sources independent of the control emplex and have nitrogen m0ators i

which makes their operation independent of the instrument air system.

The staff questioned whether we rely on making any repairrf8f equipnent as a result of fire rather than provide separat. ten er an alternate system. A fire in the area of the A train RER pump cooling unit could pssibly damage cables for toth trains of the cooling' unit. A temprary cable is provided therefore to repair the damaged -

, q.

B train cables. Shutdown of the plant under this scenario is frcm# -

the control rom.

The staff -requested that this repair be identified on the cold -

shutdown equipnent table (Refer to May 14, 1981 letter to NRC).

Since the purpose of this table, was to identify equipnent and

w ,

.h r Mr. Harold R. Denton October 16, 1981 Page 3 circuits necessary for shutdown given a fire anywhere in the plant, we Fefer not to cmplicate the table further with notes such as this. 'Ihis particular repair is identified in the Fire Frvdan Evaluation.

'Ihe staff requested that documentation of a geviously performed test (on the implerrentation of EOP-8) be provided. 'Ihe test that was *d$

perfonned was a walk through test and was not properly documented.

It was emf 4rmed that the actions could be acocuplished in a gtk reasonable time. A more detailed test will be conducted in the upaming hot functional test.

Testing will be conducted to denonstrate that the plant can be maintained in Hot Standby, and cooled down ap;roximately 50 F frm Hot Standby. 'lhis testing will document the time to energize the Control Rocm Evacuation Panel and to implement plant operating sucedure EOP-8, " Control Roca Evacuation". 'Ihis testing will also denonstrate the capability to initiate Residual Heat Removal frca outside the control Roca by implementing plant operating s c-w%ve GOP-8, " Plant Shutdchn frcxn Hot Standby to cold Shutdown with (batrol Room Inaccessible". Please find. attached t.1ese two operating gucedures (EOP-8 and GDP-8) that will be implemented to'am14=h this testing. EOP-8 is in draft form and will be approved in approximately one week. 'Jhere will he a separate Startup g#re to document the results.

Reganiing the staff position that an independent source range monitor be provided at the CEEP, it is the position of SCE&G that this is unnecessary W we of the increase in shutdown margin before ecoldown. An inadvertant baron dilution, considering a loss of offsite power, is an unlikely accident because the available scrarces of water are borated. 'Ibe reactor make-up pumps would not supply non-borated water since there is a fail close valve in the flow path.

'Ibe only water available for make-up to the Reactor Coolant System (RCS) is borated either from the Ecric Acid Tanks, as de.,cribed in the gvcedure or frca the Refueling Water Storage Tank (RNST). To accumcdate for the system leakage by mainta.ining pressurizer level, make-up as described results in a significant increase in shutdown margin. Additional make-up recpired to -

acwmcdate RCS volume contraction during cooldown using the boric -

acid storage will insure that the reactor is maintained in a a suberitical condition.

We trust that the information provided above will resolve y T remaining concerns on our ali.ernate shutd:mn capability and with the information provided in our letters dated July 16,'1981, and

. g 4-

>* o Mr. Fe.rold R. Den *m October 16, 1981

• Page 4 ,

August 21, 1981, should allow you to close SER open itens 1.6.10. If you have any questions, please let us know.

Very truly yours, /

/

. . .A &

. T. C. Nichols, Jr.

NEC:'IQhlkb Attachment cc: V. C. Sumar G. H. Fischer -

T. C. Nichols, Jr.

E. N. Cyrus J. C. Ruoff D. A. Nauman W. A. Williams, Jr.

R. B. Clary

• O. S. Bradham A. R. Koon '

M. N. Browne B. A. Bursey J. L. Skolds J. B. Knotts, Jr.

H. E. Yocan

! J. B. Cookinham NPCF Pile N

1 i

6 Attachm:nt 5 sY .

SOUTH CAROLINA ELECTRIC & gas COMPANY h post orrsct som to.

CoLumalA, Soutn CARoLthA 29218 T. C. Nicuots. Jn.

v.c e. w. , use." November 14, 1980 e,u. o . , .

Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Com=ission Washington, D.C. 20555

Subject:

Virgil C. Su=mer Nuclear Station, Unit #1 Emergency Operating Procedures .

. Docket No. 50-395

Dear Mr. Denton:

Enclosed are the following V.C. Su=mer Nuclear Station Emergency Operating Procedures: E0P-1, " Safety Injection Actuation"; E0P-5, " Reactor Trip"; E0P-13,

" Natural Circulation"; and E0P-14, " Inadequate Core Cooling". These procedures incorporate comments generated by the Nuclear Regulatory Co= mission and their Consultant, Pacific Northwest Laboratories, and have been demonstrated to be functional on the Westinghouse Simulator at Zion, Illinois.

Two (2) remaining areas of concern expressed by the NRC representative are:

a) Justification for de-energizing Balance of Plant 480V buses XSWlD1 and XSWlC1 in response to ATUS situations. <

b) Relationship of the Shift Technical Advisor in i=plementing proce-dures.

Enclosure I shows the electrical loads on 480V buses XSWlB1 and ICl. As l shown, none of the loads are required for the safe shutdown of the unit. Further, due to the distribution of Balance of Plant electrical loads, no secondary plant l systems will be made completely inoperable except for the Rod Drive MG Sets which will trip the Reactor and the Main Turbine EHC Fluid Pumps which will trip

, the Main Turbine. The Reactor Trip and Turbine Trip are the points of de-energi-l zing buses XSWlB1 and XSWlC1. Please note that these buses are completely inde-I pendent of the ESF Electrical Distribution System. Also, during plant startup operations, buses have been tripped with no effect shown on the remainder of the l BOP Distribution System. Therefore, South Carolina Electric and Gas Company l feels that this action for an ATWS situation is the fastest, easiest method of initiating a Reactor / Turbine Trip when automatic or manual trips do not function properly. It does not rely on communicating with an outside operator and does not rely on that outside operator going from wherever he may be to the appropri-l ate panel to open a breaker.

The Shift Technical Advisor (STA) will be an integral non-operating part of each shift. He has received the training to analyze unusual plant situations, I

and since he has no operating duties, can monitor plant conditions for indica-tions that the operators either overlook or may be outside the scope of the pro-l cedures they may be using. His main function is to perform this monitoring, (1)

(> n i , , ,

i l " hJ 'l ([3h

4 V.C. Summ2r Nuclear Station Emergency Operating Procedures Mr. Harold R. Denton - NRC November 14, 1980 analyze the conditions, and advise the Shift Supervisor of developing problems.

He can further reco= mend corrective actions to mitigate the problem. However, it would be inappropriate to insert specific signoff points in the procedures for the STA for three main reasons. First, it would imply that his signoff would be required before proceeding further, in effect, giving the STA command authority that he does not have. Secondly, it would imply that the only place he would have any say in implementing the procedure or offering advice would be at that specific point and this is also not true. Finally, it would single out the STA for a unique case, which he may be; however, a case could then be made that each me=ber of the team is unique with specific duties and therefore should have signoff points. It is much cleaner to treat the signoffs as being directed to the " primary user" while other persons report conditions and actions to him.

In reference to the E0P-14 " Inadequate Core Cooling", South Carolina Electric and Gas Company is pursuing with Westinghouse Owners Group the basis for their 1200 F minimum limit on the Incore Thermocouples as a symptom of inadequate core cooling.

Very truly yours, Tc . nN A -

T.C. Nichols, Jr. g.4 e

ENCLOSURES KWW/ ARK /bsg cc: Messrs. B.A. Bursey V.C. Su=mer G.H. Fischer W.A. Williams T.C. Nichols E.H. Crews D.A. Nauman O.S. Bradham O.W. Dixon J.B. Knotts R.B. Clary J.L. Skolds A.R. Koon NPCF/Whitaker file (2)

A '/

0 gp/ , Enclocuro I, XSW1B1 and XSW1C1 Electrical Loads XSW1B1

1. Turbine Closed Cycle Cooling Pump B. Pump A is still availIable from XSW1A1.

2. Alternate Feed to Substation. Normal feed is from XSW1A1.

3. Main Condenser Vacuum Pump B. Pump C will be de-energized on IC1, but Pump A is still available. Turbine is tripped and decreasing vacuum is not a major concern at this time.

4. Generator Stator Cooling Water Pump B. Pu=p A is energized from XSWlAl.

5. Rod Drive MG Set B. - Desired to be de-energized.

6. EHC Fluid Pump B. - Desired to be de-enerE ized.

7. Auxiliary Condenser Vacuum Pump B. Pump A will be de-energized on XSWIC1, however, auxiliary condensers are open to the main condensers.

8. Main Transformer Coolers. With turbine trip heat load on, main transformer will be greatly reduced so the cooling fans are not essential.

9. Open Cycle CW Booster Pump B. Main cooling loads have been removed on plant shutdown. ,

10. Feed to XMC-1B1X. This motor control center has an ATU (Automatic Transfer Unit) which will receive power from an alternate source without loss of function.

XSW1C1

1. QA Office Building. - Not required for a plant operation.

2. Exhaust Hood Spray Pump B. - Not required during shutdown, Pump A is avail-able from 1A1.

3. Rod Drive MG Set A. - Desired to be de-energized.

4. EHC Fluid Pump A. - Desired to be de-energized.

5. Main condenser Cleaning Pit Sump Pump. - Not required for shutdown.

6. Motor Control Center XMCIC1X. This motor control center has an ATU (Automatic Transfer Unit) which will receive power from an alternate source without loss

~

of function.

7. Main Condenser Vacuum Pump C - Normally not running. Vacuum Pump A still available from XSW1A1.

(1 of 2)

. /

Enclo ura I XSW1Cl (contd.) ,

8. Turbine Closed Cycle Cooling Pump A. - Pump B is still available.

9. Auxiliary Condenser Vacuum Pump A. - Pump B is also de-energized on XSW1B1, however, auxiliary condensers are open to the main condenser to help maintain vacuum during feed pump coastdown.

10. Open Cycle CW Booster Pu=p A. Main cooling loads have been removed on plant shutdown. System is e!;uipped with a bypass areund the pumps to maintain cooling water flow /from the circulating water system.

c G

l (2 of 2)

~

.-s' ,

g_ j Attachm::nt 6 south CAROLINA ELECTRIC a gas COMPANY Cot.umseA, south CAnouMA 29218 T. C. NacMo ts, J R.

v,u no. . ... .. o- February 11, 1982 wu. onum.,

Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Washington, D. C. 20555

Subject:

Virgil C. Summer Nuclear Station Docket No' 50/395 Schedules

Dear Mr. Denton:

South Carolina Electric & Gas Company, acting for itself and as agent for S6uth Carolina Public Service Authority, advises that the expected fuel load readiness date for the Virgil C. Sumer Nuclear Station is being extended to March 31, 1982. This delay is attributable to a final review of operating procedures, Code "N" Stamping of systems and receipt of vital materials impacting on completion of items affecting fuel load.

Very truly yours, T. C. Nichols, Jr.

MBW:TCN:rh cc: V. C. Summer T. C. Nichols, Jr.

G. H. Fischer H. N. Cyrus H. T. Babb D. A. Nauman M. B. Whitaker, Jr.

W. A. Williams, Jr.

O. S. Biadham R. B. Clary M. N. Browne A. R. Koon G. J. Braddick J. L. Skolds B. A. Bursey-J. C. Ruoff J. P. O' Reilly NPCF

' **N File

d o UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

SOUTH CAROLINA ) Docket No. 50-395 OL ELECTRIC & GAS COMPANY )

(Virgil.C. Summer )

Nuclear Station, Unit 1) )

CERTIFICATE OF SERVICE I hereby certify that copies of "' Applicants' Response in Opposition to Intervenor's Motion to Reopen for Consideration of New Contention' with Af fidavits and Attachments" in the above captioned matter, were served upon the following persons by deposit in the United States mail, first class postage prepaid this lith day of March 1982.

Herber; Grossman, Esq. Chairman, Atomic Safety and Chairman, Atomic Safety Licensing Board Panel and Licensing Board U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission Washington, D.C. 20555 Washington, D.C. 20555 George Fischer, Esq.

Dr. Frank F. Hooper Vice President and Group School of Natural Resources Executive - Legal Affairs University of Michigan South Carolina Electric &

Ann Arbor, Michigan 48109 Gas Company Post Office Box 764 Mr. Gustave A. Linenberger Columbia, South Carolina 29202 Member, Atomic Safety and Licensing Board Panel Steven C. Goldberg, Esq.

U.S. Nuclear Regulatory Office of the Executive Commission Legal Director Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Y W Chairman, Atomic Safety and Mr. Brett Allen Bursey Licensing Appeal Board Panel Route 1, Box 93-C U.S. Nuclear Regulatory Little Mountain, S.C. 29076 Commission Washington, D.C. 20555 John C. Ruoff Post Office Box 96 Mr. Chase R. Stephens Jenkinsville, S.C. 29065 Docketing and Service Section Office of the Secretary Robert Guild, Esq.

U.S. Nuclear Regulatory 314 Pall Mall Commission Columbia, South Carolina 29201 Washington, D.C. 20555 Richard P. Wilson, Esq.

Assistant Attorney General South Carolina Attorney General's Office P.O. Box 11549 Columbia, South Carolina 29211 Jed C. Sanford )