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Forwards Omitted Pages & Corrections to Re Response to NUREG-0737, Post-TMI Requirements
	ML18139B945
Person / Time
Site:	Surry, North Anna, 05000000
Issue date:	07/13/1982
From:	Leasburg R; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton, Eisenhut D; Office of Nuclear Reactor Regulation
References
	RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.1, TASK-2.F.2, TASK-2.G.1, TASK-2.K.1, TASK-2.K.2, TASK-2.K.3.25, TASK-3.A.1.1, TASK-3.D.1.1, TASK-3.D.3.3, TASK-3.D.3.4, TASK-TM 367A, NUDOCS 8207190160
	Download: ML18139B945 (8)
	v • d • e

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND., VIRGINIA 23261 R.H.LEASBUBO VxcB PRESJDBNT NUCI.,BAR 0PERATXONS July 13, 1982 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation Attn:

Mr. Darrell G. Eisenhut, Director Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Gentlemen:

Serial No. 367A NO/DWL:acm Docket Nos. 50-280 50-281 50-338 50-339 License Nos. DPR-32 DPR-37 NPF-4 NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY NUREG-0737 RESPONSE-REVISION 3 CORRECTIONS In our letter dated June 28, 1982 (Serial No. 367) Vepco supplied Revision 3 to our document entitled "Response to NUREG-0737 Post-TMI Requirements."

Several pages of this revision were inadvertently deleted in some copies of the June 28, 1982 submittal due to a duplication oversight.

These pages (D-7 through D-10) are attached.

Additionally, pages I.C.5-3 and II.K.3.25-3 are also attached.

Page I.C.5-3 corrects a typing error and page II.K.3.25-3 provides a clarification of a previously ambiguous statement.

Please incorporate these corrections into your Revision 3 change pages which were previously provided.

8207190160- 820713 PDR ADOCK 05000280 P

PDR

CORRECTIONS TO REVISION 3 "RESPONSE TO NUREG - 0737"

NRC Clarifi-Implemen-Pre-Imple-cation tation mentation Itea Shortened Title Description Schedule Approval 11.F.1 Accident-monitoring

2.

Iodine/particulate 1-1-82 E No (continued) sampling

3.

Containment high-1-1-82 E No range monitor

4.

Containment pressure 1-1-82 No

5.

Containment water 1-1-82 No level

6.

Containment hydrogen 1-1-82 E No II.F.2 Instrumentation for

1.

Subcool meter 1-1-80 No detection of inadequate

2.

Tech spec (LL Cat A) 12-15-80 Yes core cooling

3. Install level 1-1-82 E No instruments (LL Cat B) 11.G.1 Power supplies for

1.

Upgrade to emerg 1-1-80 No pressurizer relief sources valves, block valves

2.

Tech specs 12-15-80 Yes and level indicators II.K.1

' IE Bulletins 79-05, 06, 08 Bulletin No specific D-7 NRC Post-Imple-Tech mentation Spec.

Review Revision-Reguired Reguired Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes No Yes Yes Yes Yes Yes No Yes Yes No Submittal Reg. Bf 1-1-81 submittal if devia-tion from position 7-1-81 submittal if devia-tion from position 1-1-82 1-1-82 1-1-82 1-1-80 9-1-80 1-1-81 Submittal if devia-tion from position 1-1-80 9-1-80 Bulletin specific Vepco Remarks Submitted-A Submitted-A Subaitted-A Submitted-A Submitted-A Closed-A(l,2,8)

Closed-A(4,5,6)

Submitted-A Closed-A(l,2,8)

Closed-A(4,5,6)

NRR has evaluated Vepco responses-A

.J' 3

l

'NRC

'NRC Post-Imple-Tech Clarifi*

Implemen*

Pre-Imple* mentation Spec.

cation tation mentation Review Revision Subllittal Vepco Itea Shortened Title Descri2tion Schedule A22roval Reguired Reguired Reg. By Remarks 11.1.2 Orders on B&W plants

13. Thermal mechanical 1-1-82 No Yea As required 1-1-82 Subaitted-A report

17. Voiding in RCS

b.

1-1-82 No Yea No 1-1-82 Submitted-A 11.J:.3 Final recoaaendationa,

1. Auto PORV* isolation Not required by II.K.3.2-A B&O task force

2.

Report on PORV 1-1-81 No.

Yea No 1-1-81 Submitted-A failures

3. Reporting SY & RV 1-1-81 No Yea Yea 1-1-81 Cloaed*A(15,16) failures.& challenges

5. Auto trip of RCPa

a.

Propose 7-1-81 E No Yea No 2*15-81 Subaitted*A modifications

b. Modify 3-1-82 Yea No Yea 7-1-81 E If required-A

9.

PID controller 1-1-81 No Yea No 12-1-80 Closed-A( 13, 17)

10.

Proposed anticipatory Plant Yea No Yea Plant Hot Applicable-A trip modifications apecifc specific

11..Justify use of Plant Ho Yea Ho Plant Hot Applicable-A certain PORV specific specific

12. Anticipatory trip 3

on turbine trip

a.

Confirmation or 1-1-81 Ho Yea Ho 1-1-81 Cloaed-A(13,17)

b.

propose mo.dificati~na Modify 1st refuel Yea Ho Yes lat refuel Hot Applicable-A 6 mo after tech spec amend staff approval request

17.

ECC system outages 1-1-81 Ho Yes As required 1-1-81 Submitted-A

25.

Power on pump seals 8; Propose 11oda 1-1-82 Ho Yea Ho 1-1-82 Submitted-A

b.

Modifications 7-1-82 Yea Ho Ho 7-1-82 None Required-A D-8

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NRC NRC Post-Imple-Tech Clarifi-Implemen-Pre-lmple-mentation Spec.

cation tation mentation Review Revision Subaittal Vepco Itea Shortened Title Descrietion Schedule Ael!roval Reguired Reguired Reg. By Remarks 11.IC.3 Final reconaendations,

30.

SB LOCA methods B&O task force

a.

Schedule outline 11-15-80 No Yes No 11-15-80 Subaitted-A (continued)

b. Model 1-1-82 E Yes No No 1-1-82 E

c.

New analyses 1-1-83 or Yes No No 1-1-83 or 1 yr after 1 yr after staff approval staff approval

31. Compliance with 1-1-83 or Yes No TBD 1-1-83 CFR 50.46 1 yr after staff approval 111.A.1.1 Emergency preparedness, Short-tel'II improvements Complete No Yes No Coaplete Complete-S, Closed-NA(l short-tera 111.A.1.2 Upgrade emergency

1.

Interim Tsc*osc & EOF Complete No*

Yes No Coaplete Clo11ed-A(l,2,8) 3 support facilities

2. Design 6-1-81 Yes No No 6-1-81 Submitted-A

3. Modifications 10-1-82 E No Yes Yes 10-1-82 111.A,2 Emergency preparedness

1. Upgrade emergency 4-1-81 No Yes Yes 1-2-81 Subaitted-A plans to App. E, 10 CFR 50

2.

Meteorological data 6-1-83 No Yes Yes 1-2-81 Staged iapleaenta-tion (E)-A 111.D.1.1 Primary coolant outside

1.

Leak reduction Complete No Yes Yes Complete Closed-A(l,2,3) contai1111ent

2. Tech specs 12-15-80 Yes No Yes 9-1-80 Closed-A(4,S,6) 111.D.3.3 Inplant radiation

1. *Provide means to Complete No*

Yes No Coaplete Closed-A(!, 2, 3) monitoring determine presence of radioiodine D-9

NRC NRC Post-I11ple-Clarifi-Imple11en-Pre-Imple-mentation cation talion men ta ti on Review Item Shortened Title Descri~tion Schedule A~~roval Reg,uired 111.D.3.3 lnplant radiation

2. Modifications to 1-1-81 Ho Yea 11e>nitoring (continued) accurately measure lz 111.D.3.4 Control-rooa

1. Review 1-1-81 Ho Yea habitability

2. Modification 1-1;..33 No Yes Motes E - Indicates those illple11entation and/or submittal dates to which Vepco has taken an exception.

Closed - Indicates that actions and/or requirements were met and are documented as so.

Tech Spec.

Revision Submittal Reg,uired Reg. By Yes 1-1-81.

Ho 1-1-81 Yes 1-1-81 Submitted - Indicates that information was provided stating completion of requirements but no formal Close-out was received.

Complete - Indicates that requirements were met but no submittal was required.

S - Surry Power Station (both units)

HA -*Horth Anna Power Station (both units)

A - All units (Ref. Ho.) - Close-out doc1J11entation reference (see reference listing)

D-10 Vepco Reaarks Cloaed-A(l,2,3) r Subaitted-S, Cloaed-HA(3)

Hot Required-HA 3

0..

e I.C.5 PROCEDURES FOR FEEDBACK OF OPERATING EXPERIENCE TO PLANT STAFF The operating experience assessment function has been implemented through both a system-level Safety Evaluation and Control group and Safety Engineer-ing staffs at each station.

Procedures for the operation of the system organization, the North Anna SES and the Surry* SES were in effect prior to January 1, 1981.

Part of the operating experience assessment function is accomplished through the use of the INPO SEE-IN Program.

This program was endorsed by the NRC Staff in Generic letter 82-04.

This program has been part of the Vepco Safety Evaluation-and Control Staff procedures since Fall 1980.

I. C.5-3

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II. K.3. 25 EFFECT OF LOSS OF ALTERNATING-CURRENT POWER ON PUMP SEALS This item requires that the consequences of a loss of reactor coolant pump (RCP} seal cooling due to a loss of AC power (defined as loss of offsite power) ~or at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months be demonstrated.

During normal operation, seal injection flow from the chemical and volume control system is provided to cool the RCP seals and* the component cooling water system provides flow to the thermal barrier heat exchanger to limit the heat transfer from the reactor coolant to the RCP internals.

In the event of loss of offsite power, the RCP motor is deenergized and both of these cooling supplies are terminated; how-

' ever, the diesel generators are automatically started and seal injection flow is automatically restored within seconds.

Component cooling water to the thermal barrier heat exchanger is also automatically restored at North Anna, but must be manually reinitiated at Surry. Either of these cooling supplies is adequate to provide seal cooling and prevent seal failure due to loss of seal cooling during a loss of offsite power for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

This information was provided on December 28, 1981 (Serial No. 702).

II.K.3.25-3 3

ML18139B945

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