ML19105B050

Superseded Pages Per Revision 1 to Response to NUREG 0737 Post TMI Requirements

ML19105B050
Person / Time
Site:	Surry
Issue date:	06/18/1981
From:	Virginia Electric & Power Co (VEPCO)
To:	Office of Nuclear Reactor Regulation
References
NUREG 0737
Download: ML19105B050 (74)

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PURPOSE:

The purpose of this report to the Nuclear Regulatory Commission is to provide the total response to and current status of implementing the commitments made by Virginia Electric and Power Company to NUREG 0737 for the North Anna and Surry Power Stations.

BACKGROUND:

NUREG 0578 and NUREG 0660:

By letters dated August 14, October 24, 25, November 26, December 17, 1979, January 10, 31, February 1, 8, 27, April 1, 10, 28, June 6, 9, 30, July 7, 14, and 25, 1980, Vepco has previously submitted commitments and documenta-tion of actions taken at North Anna and Surry Power Stations to implement the Lessons Learned Requirements.of NU REG 0578.

This information has been consolidated and revised as necessary for those requirements that were incor-porated in NUREG 0660 and subsequently reissued with clarifications and changes as NUREG 0737.

Responses to the remaining items of NUREG 0660 that have been reissued in NUREG 0737 provide documentation of previous actions and commitments to additional actions necessary to meet the new and/or clarified requirements.

NUREG 0696 and NUREG 0700:

To date the final requirements of NUREG 0"696 and NUREG 0700 incorporated in NUREG 0737 have not been issued.

In previous letters, Vepco has commented on NU REG CR/1580 and NU REG 0696.

Vepco commitments to NUREG 0696 and 0700 will be provided after issuance by the NRC of the final requirements.

SCOPE:

This report is to stand alone as Vepco's response to NUREG 0737.

The report is a compilation of previously submitted information referenced above for the items of NU REG 0578 and NU REG 0660.

Future correspondence submitted in response to NUREG 0696 and NUREG 0700 items will be included in this report.

In addition, this report

• presents the current status of Vepco's efforts in implementing the commitments.

Section E of this report, Response, follows the NRC format in NUREG 0737 of three chapters of requirements:

I.

Operational Safety; II.

Siting and Design;.

and III. Emergency Preparedness and Radiation Effects.

To facilitate review, the individual response sections include the NRC position and clarifications along with the detailed Vepco response.

Each Response section attempts to provide a status of the engineering and construction and a design description which addresses each clarification item and identifies anticipated deviations from the technical or schedular require-ments of NUREG 0737.

This report will be updated periodically by additional or replacement pages to maintain the required documentation and status of implementation.

Section D of this report presents a list of the requirements and the status of Vepco's responses.

This Status List will be updated with each addition to the report.

B-2

• =

CLARIFICA-TION ITEM I. D.1 I. D. 2 II.B.1 II. B. 2 SHORTENED TITLE Control-room design Plant safety parameter display console Reactor coolant system vents Plant shielding EXCEPTIONS

1) None at this time - NUREG 0700 has not been issued.

2)

Vecpo assumes the licensee submittal date should be "later" rather than 4/82.

1) None at this time - NUREG 0696 has not been issued.

1)

Vepco requires a relaxation of the final documentation submittal date to January 1, 1982 rather than July 1, 1981 f_or complete listing of qualifica-tion, final procedures and final electrical drawings for all units.

• 2)

Vepco requires a relaxation of the required installation date to July 1, 1982 or first refueling after Janu-ary 1, 1981, which ever is later, rather than installation by July 1, 1982.

• 1)

Vepco shielding review to the re-quirements of NUREG 0578 & 0660 did not consider HELB.

This addi-tional review is estimated to be complete by June, 1981 but may require additional equipment modifi-cations that may extend beyond July, 1982.

2)

Some systems included in the NRC clarifications were excluded from the review.

Justifications are provided.

3) Liners and seals in the recircula-tion spray valves at Surry Unit 2 will be replaced at next refueling currently scheduled for the first quarter, 1982.

This date could potentially delay beyond the July 1, 1982 requested date.

4)

Radiation zone maps for personnel access will be revised after all designs are installed.

B-5

CLARIFICA-TION ITEM SHORTENED TITLE EXCEPTIONS 11.B.3 Post-accident sampling NONE 11.B.4 Training for mitigating NONE core damage II.D.1 Relief & safety-valve

1) The EPRI program has not formally test requirements included the testing of block valves.

11.D.3 Valve position

1) Equipment is undergoing seismic and indication environmental testing and is scheduled to be completed Summer, 1981.

ll.E.1.1 Auxiliary Feedwater NONE system evaluation 11.E.1.2 Auxiliary feedwater system initiation Part 1

1)

The original design criteria of the plants satisfy the requirements and meet the intent of the clarifications.

Part 2

1)

The original design criteria of the plants satisfy the requirements and meet the intent of the clarifications.

11.E.3.1 Emergency power for NONE pressurizer heaters 11.E.4.1 Dedicated hydrogen

1) Installation beyond the requested penetrations date of July 1, 1981 could extend to Fall, 1981 if material is not received as scheduled.

11.E.4.2 Containment isolation

1) Justification is provided for the dependability existing containment pressure isola-tion setpoint.

• 2)

Category I containment radiation isolation signal is assumed not to be required for normally closed containment purge and vent valves.

3)

Original plant design criteria as described in the FSAR took excep-tions to General Design Criteria.

B-6

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CLARIFICA-TION ITEM SHORTENED TITLE EXCEPTIONS 11.F.1 Accident-monitoring

1.

Noble gas effluents

1) Low range sensitivity of Main Steam Monitors is restricted by very high background radiation.

2)

Some design information and proce-dures will not be available by Janu-ary 1, 1981 but will be available by the required installation date.

2.

Sampling & analysis

1) Final design details and analytical of effluents procedures will be available prior to the required implementation date of January 1, 1982 rather than the required documentation date of Janu-ary 1, 1981 based on the availability of vendor information.

3.

Containment

1) Transmitters qualified to IEEE-344 &

radiation monitors 323-1971 will be installed and up-graded per the IE Bulletin 79-0lB program

• The upgrade* may not be complete by January 1, 1982.

4.

Containment

1)

Best equipment available will be used pressure and upgraded under the I.E. Bulletin

• 79-0lB program. *Upgrade may not be complete by January 1, 1982.

5.

Containment water

1) Best equipment available will be used level and upgraded under the I.E. Bulletin 79-0lB program.

Upgrade may not be complete by January 1, 1982 *

6.. Containment

1) Hydrogen analyzers qualification hydrogen test has not been approved.

11.F.2 Instrumentation for

• 1) Vepco requires a relaxation of the detection of inadequate required installation date to the first core cooling refueling after system availability or January 1, 1982, which ever is later, rather than installation by January 1, 1982.

2)

Clarification item (7) is a new requirement.

Comparison of the pro-I posed Westinghouse design ~ith the Appendix B requirements is not complete.

B-7

CLARIFICA-TION ITEM SHORTENED TITLE EXCEPTIONS II.F. 2 Instrumentation for detection of inadequate core cooling (continued)

• 3) The ability of any system to provide

. II. G.1 Power supplies for pressurizer relief valves, block valves and level indicators II. K.1 IE Bulletins II.K.2 Orders on B&W plants an unambiguous indication has not been demonstrated.

Procurement has pro-ceeded on the information available.

The required installation date should be delayed until additional research is performed.

If the implementation date is not delayed, the installed level system should not be required to be changed if future research pro-vides a better system.

NONE NONE

. 13 Thermal Mechanical Supports

. 1 7 Voiding in RCS NONE NONE NONE II.K.3

.19 Bench Mark Analysis AFW Final recommendations, B&O task force

.1 PORV isolation system NONE

. 2 PORV Failures

• 1) Report to be submitted March 1, 1981, rather than January 1, 1981.

. 3 SV & RV Failures NONE

.5 Auto trip RCP's NONE

.9 PID Controller NONE

.10 Proposed Anticipatory Trip NONE

.11 Use of certain PORV's NONE

.12 Anticipatory Trip on NONE Turbine Trip B-8

CLARIFICA-TION ITEM II.K.3 SHORTENED TITLE Final recommendations, B&O task force (continued)

EXCEPTIONS

.17 ECC system outages

1)

None for North Anna III.A.1.1 III.A.1. 2 III.A. 2 III.D.1.1 III.D. 3. 3

• 2)

Report to be submitted by March 1, 1981 for Surry.

. 25 Power on Pump Seals

. 30 SB LOCA Methods NONE NONE NONE

. 31 Compliance with 10 CFR 50.46 Emergency prepared-ness, short-term Upgrade emergency support facilities Emergency preparedness Primary coolant outside containment Inplant radiation monitoring NONE

1)

None at this time - additional clarification has not been issued.

• 1) Implementation of improved meteoro-logical data capabilities must be conducted on the same schedule as portions of III. A. 1. 2.

NONE NONE III.D.3.4 Control-room

1) None for North Anna habitability

• 2)

Vepco requires a relaxation of the January 1, 1981 documentation date for Surry.

The on-site analysis and off-site analysis will be submitted by January 19 and June 30, 1980, re-spectively.

Appendix B Design Criteria

• 1) Equipment does not meet the seismic test results of R. G. 1.100.

2)

No additional vendor documentation was required for "extended range" qualification.

B-9

CLARIFICA-TION ITEM SHORTENED TITLE Appendix B Design Criteria (continued)

EXCEPTIONS

• 3)

Equipment does not meet the IEEE-323-1974 requirements of R.G. 1.89.

• 4) Existing plant systems meet original plant criteria for electrical separation but not R.G. 1.75.

5)

The individual requirements of the referenced Reg. Guides in Item (5) were not addressed.

The design was done in accordance with the requirements of 10CFR50, Appendix B and ANSI N45.2.

6)

Existing plant instrumentation used for normal and post-accident condi-tions will be reviewed as part of the Control Room Design Review,Section I.D.1 for proper identification.

B-10

PROPOSED MODIFICATIONS AND TENTATIVE OUTAGE SCHEDULE NORTH ANNA AND SURRY North Anna Unit 1 Outages:

2nd R - Second refueling outage currently scheduled for January 1, 1981.

SM - Spring, 1981 maintenance outage.

3rd R - Third refueling outage currently scheduled for Summer of 1982.

Unit 2 Outages:

FP - Fire protection modification outage scheduled for November 1, 1980.

SM - Spring, 1981 maintenance outage.

1st R - First refueling outage currently scheduled for the first quarter of 1982.

Surry Unit 1 Outage:

SGRP - Steam generator replacement outage (in progress).

Unit 2 Ou tag es :

SM - Spring, 1981 maintenance outage.

FM - Fall, 1981 maintenance outage.

5th R - Fifth refueling outage currently scheduled for the first quarter of 1982.

C-2

NUREG-0737 MODS RCS & PZR Vent (II.B.1)

Plant Shielding (II.B.2)

Hydrogen Recombiner Vault Shielding S.W. Rad Monitor Pump Mods.

S.W. Rad Monitor Shielding EST PROPOSED THI MODIFICATION SCHEDULE NORTH ANNA AND SURRY REQ HAT'L OUT.

NORTH ANNA //1 NORTH ANNA #2 SURRY #1 DATE UNIT DEL

~ 2nd R ~

3rd R 7/82 NAl Onsite Yes

• 1/82 NA2 Onsite Yes 7/82 Sl 1/81 Yes 7/82 S2 1/81 Yes 1/82 NAl Onsite No

• 1/81 NA2 Onsite No 1/82 NAl 12/80 Yes

• 1/81 NA2 12/80 Yes 1/82 Sl

"'3/81 No 1/82 S2

"'3/81 No 1/82 NAl 1/81 No

• 1/81 NA2 1/81 No 1/82 Sl

"'3/81 No 1/82 S2

"'3/81 No X

X FP SM 1st R SGR X

X X

C-3 SURRY //2 SM FM 5th R X

REMARKS DC-79-S69A - Unit 1 DC-79-69B - Unit 2 DC-80-S29 - Rx Vessel head removal required.

DC-80-S36 North Anna only.

DC-80-S25A (motors & seals)

DC-80-57 (motors)

DC-80-S33 DC-80-S56 Piping & electrical to be installed per DC-80-S57

EST REQ HAT'L our.

NORTH ANNA Ill NORTH ANNA 112 SURRY Ill SURRY (12 NUREG-0737 HODS DATE UNIT DEL

~ 2nd R SM 3rd R FP SM 1st R SGR SM

_!!!__ 5th R REMARKS Replacement of Mission 1/82 NAl 10/80 Yes X

North Anna Unit 1 only.

DC-80-S23 Check Valves Valve Bodies Replacement of Hisaion 1/82 NAl Onsite Yes X

DC-80-S24A - Unit 1 Replace 13 Check Valve Bearings DC-80-S24B - Unit 2 replace 13

• 1/81 NA2 Onsite Yes X

X DC-80-S24C - Requires 2 unit outage for 2 valves or entry into Tech Specs.

action statement.

1/82 Sl 11/80 Yes X

DC-80-53 1/82 S2 11/80 Yes X

Replace Charging Pump 1/82 Sl Onsite No siirry only.

DC-80-S58 Bearing Gaskets 1/82 S2 Onsite No Replace Mech Seal -

1/82 Sl Onsite No Surry only.

DC-80-SSS Charging Pump Cooling Water Pump 1/82 S2 Onsite No Replace LHC Ball Valves 1/82 SI

"'2/81 Yes X

Surry only.

DC-80-S67 (Primary Drain Tank) 1/82 S2

"'2/81 Yes X

Replace Charging Pump 1/82 SI Onsite No Surry only.

DC-80-S66 Seal Water Cooler Rings 1/82 S2 Onsite No Charging Pump Lube 1/82 Sl 2/81 Yes X

Surry only.

DC-80-54 Oil Cooler TCV Mods 1/82 S2 2/81 Yes X

C-4

EST REQ MAT'L OUT.

NORTH ANNA ffl NORTH ANNA 112 SURRY ffl SURRY f/2 NUREG-0737 MODS DATE UNIT DEL

~ 2nd R SM 3rd R FP SM 1st R SGR SM FM, '5th R REMARKS Replace Liners & Seals 1/82 SI 12/80 Yes X

Surry only.

DC-80-S69 in Recirc. Spray Plug (valve vendor Rep. required)

Valves 1/82 S2 12/80 Yes X

INSTALLATION AFTER 1/82 FOR S2.

Safeguards Area Vent 1/82 NAl 3/81 No North Anna only.

Fan Motors Replacement DC-80-S03 1/82 NA2 Onsite No H2 Recombiner Gas 1/82 NAl 4/81 Yes X

DC-80-21A Reach Rods Cooler Outlet Mod DC-80-21B Shield Wall 1/82 NA2 4/81 DC-80-21C Control Panel/Trickle Htr.

(Trickle heater not required by 1/82)

Post-Accident Sampling (II.B.3)

RCS & Containment 1/82 NAl 3/81 Yes X

DC-80-S48 - Outage required for some Sampling Facility containment tie-ins only.

• 4/81 NA2 3/81 Yes X

1/82 Sl 3/81 Yes X

DC-80 Outage required for some containment tie-ins only.

1/82 S2 3/81 Yes X

Sample System Isolation 1/82 NAl 1/81 Yes X

DC-80-S32A - Unit 1 Valves 1/82 NA2 1/81 Yes X

DC-80-S32B - Unit 2

• 4/81 SI 1/81 Yes X

DC-80-60A - Unit 1 1/82 S2 1/81 Yes X

DC-80-60B - Unit 2 C-5

EST REQ HAT'L OUT.

NORTH ANNA #I NORTH ANNA /12 SURRY Ill SURRY /12 NUREG-0737 HODS DATE UNIT DEL

~ 2nd R SH 3rd R FP SH 1st R SGR SH FH 5th R REHARICS Saaple Containment 1/82 NAl Onsite Yes X

DC-80-60 Return Line 1/82 NA2 Onsite Yes X

1/82 Sl 1/81 Yes X

DC-80-78 1/82 S2 1/81 Yes X

AUJ: Feedwater Mods (II.E.1.1)

Puap Auto Control 1/82 Sl 10/81 Yes X

Work includes replacing PCV & HOV in steam line to turbine driven pump.

1/82 S2 10/81 Yes X

Misc. AUJ:iliary Auxiliary Feedvater Hods (II.E.1.1)

AFW Flow Restricting R

Sl 5/81 Yes X

Orifices R

S2 5/81 Yes X

ECST Level Alarm 7/81 NAl 11/80 CR X

DC-80-S02 Unit 1 DC-80-S42 Unit 2 7/81 NA2 11/80 CR X

7/81 Sl 11/80 CR X

DC-80-S37 7/81 S2 11/80 CR X

C-6

L lfUREG-0737 MODS Dedicate H2 Penetration (II.E.4.1)

Replacement of H2 Analyzer & Hydrogen REQ DATE UNIT 7/81 NAl Recombiner Valves

• 1/81 NA2 Containment Isolation (II.E.4.2)

Diverse Signal to Condenser Air Ejector T.D. AFW SI Reset CRDH Fans SI Reset Cond Air Ejector SI Reset 7/81 7/81 1/81 1/81 1/81 1/81 1/81 1/81 1/81 1/81 St S2 SI S2 NAl NA2 NAl NA2 HAI HA2 EST KAT'L OUT.

NORTH ANNA 111 DEL

~ 2nd R ~

3rd R 1/81 Yes X

1/81 Yea 1/81 Yea 1/81 Yea Onaite Yea Onaite Yea Onaite No Onaite No Onsite No Onaite No Onaite No Onaite Ho HORTH ANNA ll2 SURRY #1 SURRY 112 FP SM 1st R SGR SH FM 5th R X

X X

X X

C-7 REMARKS DC-80-S31A&B Entry into Tech Spec action statement required.

PAM&C panel must be installed.

DC-79-62B Entry into Tech Spec action statement required.

PAM&C panel must be installed.

Surry only.

DC-80-S90 S2 - KAY REQUIRE AN OUTAGE PRIOR TO 1-1-81.

DC-79-S75 DC-79-S76 DC-80-Sll

EST REQ HAT'L OUT.

NORTH ANNA fll NORTH ANNA f/2 SURRY /fl SURRY /12 NUREG-0737 MODS DATE UNIT DEL

!@ 2nd R SM 3rd R FP SM 1st R SGR SM FM 5th R REMARKS Increased Range of Rad Monitors (II.F.1)

Containment High Range 1/82 NAl 11/80 Yes X

X DC-80-S35A Unit Ill Containment.

DC-80-S35B Unit f/2 Containment.

• 1/81 NA2 11/80 Yes X

X DC-80-S35C Outside Containment.

Non-outage - C.R. Equipment ~s/81.

1/82 Sl 12/80 Yes X

Electrical termination problems.

1/82 S2 12/80 Yes X

DC-80-S52 - Penetration problem.

Non-outage - C.R. Equipment ~s/81.

Electrical termination problems.

Main Steam Effluent 1/82 NAl

~7/81 No DC-80-S37B - Electrical upgrade &

recorders.

• 7/81 NA2

~7/81 No 1/82 Sl

~7/81 No DC-80-64B - Electrical upgrade &

recorders.

1/82 S2

~7/81 No Steam Driven Aux Feed 1/82 NAl 3/81 No DC-80-S37A

• 7/81 NA2 3/81 No 1/82 Sl 3/81 No DC-80-64C 1/82 S2 3/81 No Process & Vent Rad 1/82 NAl

~7/81 No DC-80-S37C

• 7/81 NA2

~7/81 No 1/82 Sl

~1111 No DC-80-64A 1/82 S2

~7/81 No C-8

EST REQ MAT'L OUT.

NORTH ANNA {11 NORTH ANNA /12 SURRY /11 SURRY /12 HUREG-0737 HODS DATE UNIT DEL

~ 2nd R ~

3rd R FP SH 1st R SGR SH FM 5th R REMARICS Containment Accident Monitoring (II.F.1)

Containment Pressure 1/82 NAl 1/81 Yea X

DC-79-S67 - Recorders on PAH&C panel.

Outage required for installation of

• 1/81 NA2 1/81 Yes X

panel.

Non~qualified indicators will be used until qualified indicators are available.

1/82 51 12/80 Yea X

DC-79 Recorders on PAH&C panel.

Outage required for installation of 1/82 S2 12/80 Yes X

panel.

Non-qualified indicators will be used until qualified indicators are available.

Contai1111ent Water 1/82 51 11/80 Yes X

Surry only.

Gems transmitters under-Level going qualification.

Hay install 1/82 S2 11/80 Yes X

without qualification.

Containment Hydrogen 1/82 NAl 1/81 Yes X

DC-79-S68 Monitor

• 1/81 NA2 1/81 Yes X

1/82 Sl 1/81 Yes X

DC-79-62A 1/82 52 1/81 Yes X

PAH&C Panel 7/81 NAl 12/80 Yes X

DC-80-529 - Installation required for completion of containment hydrogen,

• 1/81 NA2 12/80 Yes X

pressure & RCS Vent.

7/81 51 1/81 Yes X

DC-80-SS - Installation required for completion of containment hydrogen, 7/81 52 1/81 Yes X

water level, pressure & RCS Vent.

C-9

EST REQ KAT'L OUT.

NORTH ANNA lll NORTH ANNA 112 SURRY 111 SURRY f/2 NUREG-0737 MODS DATE UNIT DEL

~ 2nd R SH 3rd R FP SH 1st R SGR SH FM 5th R REMARKS Inadeguate Core Cooling (11.F.2)

Redundant RCS Wide

• 1/81 SI

"'4/81 Yee X

Surry only. DC-80-44 Input to eub-cooling meter.

WORK ON ELECTRONICS

• 1/81 S2

"'4/81 Yee X

WILL EXTEND BEYOND 1-1-81 DUE TO MATERIAL DELIVERY.

Reactor Veaeel Level 1/82 HAI

"'9/81 Yee X

Dc-xx-xxx (Weetinghouee level syatem)

• 1/81 NA2

"'9/81 Yee X

1/82 Sl

"'9/81 Yea X

DC-XX-XX Sixth refueling outage if not available for work during 1/82 Sl

"'9/81 Yea X

SGR.

(Weetinghouee level system)

• Denotea a requeet for reviaed implementation date.

Reference Vepco lettere dated November 20, 1980 (Serial No. 924), and November 18, 1980 (Serial No. 922).

C-10

STATUS LIST OF RESPONSES TO NUREG 0737 POST-TM! REQUIREMENTS FOR OPERATION REACTORS NRC NRC Post-Imple-Tech Clarifi-Implemen-Pre-Imp le-mentation Spec.

cation tation mentation Review Revision Submittal Vepco Item Shortened Title Descri~tion Schedule Aeeroval Reguired Reguired Reg. BI Remarks I.A.1.1 Shift technical advisor

1.

On duty 1-1-80 No Yes Yes 1-1-80 Complete

2.

Tech specs 12-15-80 Yes No Yes 9-1-80 Submitted

3.

Trained per LL Cat B 1-1-81 No Yes Yes 1-1-81 Submitted

4.

Describe long-term 1-1-81 No No No 1-1-81 program I.A.1.2 Shift supervisor Delegate non-safety 1-1-80 No Yes No 1-1-80 Complete teaponsibilitiea duties I.A.1.3 Shift manning

1.

Limit overtime 11-1-80 No Yes No 11-1-80 Complete

2.

Min shift crew 7-1-82 No Yes Yes 11-1-80 Complete I.A.2.1

~mmediate upgrading

1.

SRO exper 5-1-80 No Yes No None Complete of RO and SRO

2.

SRO'a be RO's 12-1-80 No Yes No None Complete training and

3.

Three mo trng 8-1-80 No Yea No None Complete qualifies tions on shift

4.

Modify training 8-1-80 No Yes No 8-1-80 Complete

5.

Facility 5-1-80 No Yes No None Complete certification I.A.2.3 Administration of Instructors complete 8-1-80 No Yes No None Complete training programs SRO exam D-2

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

cation tation mentation Review Revision Submittal Vepco Item Shortened Title Descrietion Schedule Aeeroval Re9uired Re9uired Re9. B:r:

Remarks I.A. 3.1 Revise scope and

1.

Increase scope 5-1-80 No No No None Complete criteria for

2.

Increase passing 5-1-80 No No No None Complete licensing exams grade

3.

Simulator exams 6-1-80 No No No None Complete 10-1-81 No No No None Complete I.C. l Short-term accident and

1.

SB LOCA 6-1-80 No Yes No None Complete procedures review

2.

Inadequate core cooling

a.

Reanalyze and 1-1-81 Yes No No 1-1-81 propose guidelines

b.

Revise procedures lat refueling Yes No No TBD outage after 1-1-82

3.

Transients & accidents

a.

Reanalyze and 1-1-81 Yes No No TBD propose guidelines

b.

Revise procedures 1st refueling outage after 1-1-82 I.C,2 Shift & relief turnover Implement shift turnover 1-1-81 No Yes No 1-1-80 Complete procedures checklist I.C.3 Shift-supervisor Clearly define superv &

1-1-80 No Yes No 1-1-80 Complete responsibility oper responsibilities I.C.4 Control room access Establish authority 1-1-80 No Yes No 1-1-80 Complete limit access D-3

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

cation tation mentation Review Revision Submittal Vepco Item Shortened Title DescriJ!tion Schedule AJ!l!roval Reguired Reguired Reg. Bl'.

Remarks I.C.5 Feedback of operating License to experience procedures implement 1-1-81 No Yes No None

)

I.C.6 Verify correct Revise performance 1-1-81 E No Yes No None performance of procedures operating activities I.D. l Control room design Preliminary assessment TBD 4/82 E Response after issue of NUREG 0700 I.D.2 Plant safety parameter

1. Description TBD Later Response after display console

2.

Installed TBD issue of NUREG

3.

Fully implemented TBD 0696 II.B.l Reactor coolant system

1.

Design vents 7-1-81 No Yes No 7-1-81 Submitted vents

2. Install vents 7-1-82 E Yes No Yes 7-1-81 (LL Cat B)

3.

Procedures 1-1-82 Yes No Yes 1-1-81 E II.B.2 Plant shielding

1.

Review designs 1-1-80 No Yes No 1-1-81 Submitted

2.

Plant modifications 1-1-81 No Yes No 1-1-81 Submitted (LL Cat B) submittal if deviation from position

3.

Equipment 6-30-82 E No Yes No 1-1-82 qualification D-4

NRC NRC Poet-Imple-Tech Clarifi-Implemen-Pre-Imple-mentation

Spec, cation tation men'tation Review Revision Submittal Vepco Item Shortened Title Deecri(!tion Schedule Ael!roval Re9uired Reguired Reg. BI Remarke II.B,3 Poet-accident sampling

1.

Interim system 1-1-80 No Yee No 1-1-80 Complete

2.

Plant modifications 1-1-82 No Yes Yes 1-1-81 Submitted submittal if deviation from position II.B.4 Training for mitigating

1.

Develop training 1-1-81 No Yes No 1-1-81 Complete core damage program

2.

Implement program

a.

Initial 4-1-81 No Yes No None

b.

Complete 10-1-81 No Yes No None II.D.l Relief & safety-valve

1.

Submit program 1-1-80 No Yee No 1-1-80 Complete test requirements

2.

RV & SV testing (LL Cat B)

a.

Complete testing 7-1-81 No No No 7-1-81

b.

Plant-specific 10-1-81 Yee Yes TDD 1-1-82 report

3.

Block-valve testing 7-1-82 Yee Yee TDD 7-1-82 II.D.3 Valve position

l. Install direct 1-1-80 No Yes Yee 1-1-80 Complete (quali-indication indications of fying equipment) valve position

2.

Tech Specs 12-15-80 Yee No Yes 9-1-80 Submitted II.E.l.l Auxiliary Feedwater

l.

Short-term 7-1-81 Yes Yes Item Plant Submitted system evaluation Specific Specific II.E.l.2 Auxiliary feedwater

l. Initiation system initiation

a.

Control grade 6-1-80 Yes No Yes 1-1-8.0 Complete

b.

Safety grade 7-1-81 No Yes Yes 1-1-81 Complete D-5

NRC NRC Post-Imple-Tech Clarif.i-Imp,Iemen-Pre-Imple-mentation Spec.

cation tat ion mentation Review Revision Submittal Vepco Item Shortened Title Descri2tion Schedule A22roval Reguired Reguired Reg. B:r; Remarks II.E.1.2 Auxiliary feedwater

2.

Flow indication system initiation

a.

Control grade 1-1-80 No Yes Yee 1-1-80 Complete (continued)

b.

LL A tech specs 12-15-80 Yes No Yes 9-1-80 Submitted

c.

Safety grade 7-1-80 No Yes Yee 1-1-81 Complete II.E.3.1 Emergency power for

1.

Upgrade power supply 1-1-80 No Yes Yes 1-1-81 Complete pressurizer heaters

2.

Tech specs 12-15-80 Yes No Yes 9-1-80 Submitted II,E,4,1 Dedicated hydrogen

1.

Design 1-1-80 Yes No No 1-1-80 Complete penetrations

2.

Install 7-1-81 No Yes No 7-1-81 Submitted II.E.4,2 Containment isolation 1-4 Imp. diverse 1-1-80 No Yes Yes 1-1-80 Complete dependability isolation

5.

Cntmt pressure eetpoint

a.

Specify pressure 1-1-81 No Yes No 1-1-81 Complete

b.

Modifications 7-1-81 Yes No Yes 1-1-81 None identified

6.

Cntmt purge valves l".'"1-81 No Yes Yes 1-1-81 Submitted

7.

Radiation signal 7-1-81 No Yes Yes 7-1-81 Assume not on purge valves applicable

8.

Tech specs 12-15-80 Yes No Yes 9-1-80 Submitted II.F.1 Accident-monitoring

1.

Noble gas monitor 1-1-82 No Yes Yes 1-1-81 Submitted Submittal if devia-tion from position

2.

Iodine/particulate 1-1-82 No Yes Yes 1-1-81 Submitted sampling submittal if devia-tion from position D-6

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

cation tat ion mentation Review Revision Submittal Vepco Item Shortened Title Descrietion Schedule Aeeroval Reguired Reguired Reg. Bl Remarks 11.F.l Accident-monitoring

3.

Containment high-1-1-82 No Yes Yes 7-1-81 Submitted (continued) range monitor submittal if devia-tion from position

4.

Containment pressure 1-1-82 No Yes Yea 1-1-82 Submitted 5,

Containment water 1-1-82 No Yee Yes 1-1-82 Submitted level

6.

Containment hydrogen 1-1-82 No Yes Yes 1-1-82 Submitted 11.F.2 Instrumentation for

1.

Subcool meter 1-1-80 No Yes Yes 1-1-80 Complete detection of inadequate

2.

Tech spec (LL Cat A) 12-15-80 Yes No Yes 9-1-80 Submitted core cooling

3. Install level 1-1-82 E No Yes Yes 1-1-81 Submitted instruments Submittal (LL Cat B) if devia-tion from position 11.G.l Power supplies for

l. Upgrade to emerg 1-1-80 No Yes Yes 1-1-80 Complete pr~ssurizer relief sources valves, block valves

2.

Tech specs 12-15-80 Yes No Yes 9-1-80 Submitted and level indicators 11.K.l IE Bulletins 79-05, 06, 08 Bulletin No Yes No Bulletin NRR has evaluated specific specific Vepco responses 11.K.2 Orders on B&W plants

13.

Thermal mechanical 1-1-82 No Yes As required 1-1-82 report

17.

Voiding in RCS

b.

1-1-82 No Yes No 1-1-82

19.

Benchmark analysis

b.

1-1-82 No Yes No 1-1-82 of seq. AFW flow D-7

NRC NRC Post-Imp le-Tech Clarifi-Implemen-Pre-Imp le-mentation Spec.

cation tat ion mentation Review Revision Submittal Vepco Item Shortened Title DescriJ:!tion Schedule AJ:!J:!roval Reguired Reguired Reg. Bl'.

Remarks 11.K.3 Final recommendations,,

1. Auto PORV isolation B&O task force

a.

Design 7-1-81 Yes No Yee 7-1-81 If required by 11.K.3.2

b. Teet/install 1st refuel Yes No Yes 7-1-81 6 mo after staff approval 2~

Report on PORV 1-1-81 E No Yes No 1-1-81 E Submittal by failures 3-1-81

3.

Reporting SV & RV 1-1-81 No Yes Yes 1-1-81 Initiate data failures & challenges beginning 4-1-80

5.

Auto trip of RCPs

a.

Propose 7-1-81 No Yes No 2-15-81 modifications

b.

Modify 3-1-81 Yes No Yes 7-1-81 If required

9.

PID controller 1-1-81 No Yes No 12-1-80 Complete

10.

Proposed anticipatory Plant Yes Nd Yes Plant N.A.

trip modifications specifc specific

11.

Justify use of Plant No Yes No Plant N.A.

certain PORV specific specific

12. Anticipatory trip on turbine trip

a..confirmation or 1-1-81 No Yes No 1-1-81 Complete propose modifications

b.

Modify 1st refuel Yes No Yes 1st refuel N.A 6 mo after tech spec amend staff approval request

17.

ECC system outages 1-1-81 E No Yes As required 1-1-81 E Submittal by

25.

Power on pump seals 3-1-81

a.

Propose mods 1-1-82 No Yes Np 1-1-82

b.

Modifications 7-1-82 Yes No No 7-1-82 D-8

NRC NRC Poet-Imp le-Tech Clarifi-Implemen-Pre-Imp le-mentation Spec.

cation tation mentation Review Revision Submittal Vepco Item Shortened Title Deecrietion Schedule Aeeroval Reguired Reguired Reg. Bl Remarks 11.K.3 Final recommends tions,..

30.

SB LOCA methods B&O task force

a.

Schedule outline 11-15-80 No Yes No 11-15-80 Complete (continued)

b.

Model 1-1-82 Yee No No 1-1-82

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. l Emergency preparedness, Short-term improvements Complete No Yes No Complete Complete short-term 111.A.l.2 Upgrade emergency

1.

Interim TSC OSC & EOF Complete No Yes No Complete Complete support facilities

2.

Design TBD TBD TBD TBD TBD

3.

Modifications TBD TBD TBD TBD TBD III.A.2 Emergency preparedness

1.

Upgrade emergency 4-1-81 No Yes Yes 1-2-81 Procedures to be plans to App. E, submitted 3-1-81 10 CFR 50

2.

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

1.

Leak reduction Complete No Yes Yes Complete Complete containment

2.

Tech specs 12-15-80 Yes No Yes 9-1-80 Submitted 111.D.3.3 Inplant radiation

1.

Provide means to Complete No Yes No Complete Complete monitoring determine presence of radioiodine n-Q

I

• NRC NRC Post-Imp le-Tech Clarifi-Implemen-Pre-Imp le-mentation Spec.

cation tation mentation Review Revision Submittal Vepco Item Shortened Title DescriJ:!tion Schedule AJ:!J:!roval Re9uired Reguired Reg. Bl Remarks III.D.3.3 Inplant radiation

2.

Modifications to 1-1-81 No Yes Yes 1-1-81 Complete monitoring (continued) accurately measure I2 III.D.3.4 Control-room

1.

Review 1-1-81 E No Yes No 1-1-81 E habitability

2.

Modification 1-1-83 No Yes Yes 1-1-81 i:-

Note E - Indicates those implementation dates to which Vepco has taken an exception.

D-10

I. A.1.1 SHIFT TECHNICAL ADVISOR

1.

Shift Technical Advisors have been on duty at North Anna Units 1 and 2 and Surry Units 1 and 2 since January 1, 1980.

Details of the qualifica-tions of currently assigned ST A's are given on pages I. A. 1.1-5 and I. A.1.1-6.

Current Technical Specifications for both North Anna Units 1 and 2 require the presence of an ST A on shift whenever either unit is in

. operation.

Similar provisions have been proposed for the Surry 1 and 2 Technical Specifications.

2.

STA training per the requirements of Lessons Learned Category B will be complete by December 17, 1980 for both stations.

This will successfully rrieet the January 1, 1981 requirements.

Details of the training conducted at each station are given on pages I. A.1.1-5 and I. A.1.1-6.

3..

The long-term ST A training program has been developed and is currently under internal review.

The program will be submitted for NRC Staff review prior to January 1, 1981

• I. A.1.1-4

I.C. l GUIDANCE FOR THE EVALUATION AND DEVELOPMENT OF PROCEDURES FOR TRANS.IENTS AND ACCIDENTS By January 1, 1981, the Westinghouse Owners Group will submit a detailed description of a program to comply with the requirements of I. C.1 for both Inadequate Core Cooling and Transients and Accidents.

The submittal will identify previous Owners Group submittals to the NRC, which are believed to comprise the bulk of the required information.

Additional effort required to obtain full compliance with Item I. C. l (with a proposed schedule for completion) will also be identified.

This approach was discussed during a November 12, 1980 meeting between Westinghouse Owners Group representatives at the NRC Staff, and is consistent with the alternate requirements on page I.C.1-4.

I. C.1-6

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

Procedures for the operation of the system organization and the North Anna SES have already been put into effect.

Procedures for the Surry SES are under final development and will be in effect prior to January 1,.1981

• I. C.5-3

1.

The majority of electrical equipment in these systems are not quali-fied to meet the integrated radiation dose to which they would be exposed in processing and concentrating the highly radioactive water or gas.

2.

There is extensive piping for the recovery systems throughout the auxiliary building.

The resulting dose rate from all these systems operating simultaneously would severely limit access for the required operation.

Shielding for the recovery system piping and components would be very difficult and in some cases may be impossible to install due to the arrangement of the piping and equipment.

B.

Modifications - Shielding or Equipment Changes for Reduction of Personnel Exposure As a result of the plant radiation and shielding review, we have identi-fied additional shielding and plant modifications to reduce the personnel exposure and equipment irradiation qualification required by NUREG 0578 which is more conservative than subsequent clarifications (NUREG 0737 allows the shielding to be based on a 30-day average base).

North Anna Only

1.

The p_ost-accident hydrogen recombiner vault requires shielding modifications to limit radiation exposure to the operators at the vault while realigning and operating the recombiner, and to reduce the levels in the continuous occupancy areas.

2.

Manual valves, located in high radiation zones, which ~must be operated to line up and op,erate the post-accident hydrogen recom-

biner, will be replaced with environmentally qualified remotely operated valves, such as direct-acting solenoid valve or air-operated valves.

Surry Only

1.

In order to automatically adjust cooling water temperature to charg-ing pumps, automatic temperature control valves are being added to the service water lines in the charging pump cooling water subsystem.

North Anna and Surry

1.

Shielding of portions of the lines added as part of the new post-accident sampling system may be required.

2.

Shielding for the Post-Accident Sampling Facility is required.

3.

Although no access is required in the lower level auxiliary building or safeguards building to mitigate an accident, the drain system for the auxiliary building sump and the safeguards bu_ilding sump will be modified such that these sumps can be pumped to the affected unit's containment instead of to the high or low level waste tanks.

This would eliminate a significant potential source of activity in the basement of the auxiliary building.

II. B.2-11

4.

Sampling procedures have

• been modified and temporary shielding employed to limit dose rates at the present sample facility.

5.

Additional shielding, area relocation, or procedural modifications are being evaluated to limit radiation dose rates in the technical support center, the operational support center, the counting lab, and the security control center.

6.

System modifications to permit i11terfacing with external process systems designed and shielded after the accident are being made.

The external process system design would be based on the extent of the accident and would utilize the most current technology avail-able at the time of the accident.

II. B. 2-12

C.

Modifications - Equipment Qualification The evaluation of radiation environmental qualification of equipment is proceeding slowly because of the difficulty in obtaining vendor data on older plants.

The mechanical equipment review for LOCA is complete for North Anna and Surry.

Vepco has reported the results of the electrical equipment review in conjunction with the responses to I.E. Bulletin 79-0lB.

Any necessary modifications will be made as material becomes available.

Replacement of, or shielding for, material with insufficient radiation resistance in the following equipment has been identified to date and is in progress as noted.

See Section C for proposed installation scheduled.

These materials meet the requirements of the FSA R but not the extended requirements of NUREG 0578, NUREG 0660, NUREG 0737 and IE Bulletin 79-0lB.

1.

Replacement Safeguards area ventilation fan motors have been re-ceived.

( North Anna)

2.

Stainless steel bearings for component cooling water and service water insert check valves to replace teflon lug and plate bearings have been ordered.

(North Anna and Surry)

3.

Replacement service water radiation monitor pump motors have been delivered to North Anna.

New pumps and motors are being pur-chased for Surry.

4.

Replacement mechanical seal bellows for the service water radiation monitor pumps have been ordered. * (North Anna)

5.

Additional shielding is being designed for the service water radi-ation monitors at North Anna.

Relocation of the service water radiation monitors to a lower background radiation area is required at Surry.

6.

7.

8.

9.

10.

11.

Replacement 0-rings in the high head safety injection pump seal cooler are being ordered.

(Surry)

Valve seat replacements are on order for component cooling water valves to the reactor coolant pumps.

(North Anna Unit 1)

Charging pump gaskets and mechanical seals on cooling water pump on charging pump are on order.

(Surry)

Containment Isolation Valve Buna-N diaphragms will be replaced with qualified. m_aterial during normal maintenance.

(Surry)

Outside recirculation sprary pump plug valve seats (teflon) will be replaced at Surry.

Electrical equipment as identified in response to I.E. Bulletin 79-0lB.

II. B. 2-13

D.

Additio11,al Information The basis for systems excluded :

1.

The design of the Reactor Coolant System (RCS) and supporting systems is such that the letdown portion of the Chemical and Volume Control System (eVCS) is not required to take the plant to a safe shutdown (hot standby) condition or mitigate the effects of a LOCA.

The use of the eves letdown could create significant radiological problems.

The letdown portion of the CSVS was not considered in this review for the following detailed reasons.

At TMI-2, high airborne activity levels and high radiation levels outside the containment resulted from using systems which carry highly radioactive fluids from inside the containment to other build-ings.

One of the lessons learned from TMI-2 is to isolate from the containment all nonessential systems.

This is also a requirement of NUREG 0737,Section II.E.4.2.

The letdown and normal charging portions of the eves is automatically isolated by the phase A con-tainment isolation signal.

Use of the letdown portion of the eves presents the potential for increasing activity levels outside the containment.

This portion of the eves is not required to mitigate an accident and this is kept isolated from the containment.

The eves letdown provides various functions during normal opera-tion.

First, it provides a method for the reduction of RCS water inventory.

If *the accident resulted from a ruptured pipe in the RCS, this function is not desired immediately.

If, ~s at TMI-2, the integrity of the primary system is reestablished, other methods of reducing RCS inventory are available.

The pressurizer PORV's and safety valves are designed to reduce pressure resulting from in-creased system water level.

The letdown system provides for control of chemistry and radio-activity (i.e. RCS hydrogen inventory, boric acid dilution and filtration, and ion exchange).

Boric acid addition to ensure proper concentration in the RCS is provided by' the Safety Injection System.

Hydrogen addition is not required to maintain plant safety after an accident, as it is added only for long term corrosion control.

Degassing of the RCS is normally provided through the letdown portion of the eves.

The installation of the RCS head and pres-surizer venting system, required by NUREG 0578 and NUREG 0737, provides this function during accident conditions.

The eves letdown also provides some ability to clean up the RCS by means of filtration and demineralizaion.

This function is not re-quired nor desired during the mitigation phase of an accident.

A large increase in the radiation levels in the eves would lead to high radioactivity levels outside the containment. on resin beds and filters for which the system is not designed

• II. B. 2-14

There are additional problems associated with using the CV CS letdown during an accident.

Overpressure in the volume control tank, which could result from RCS degassing, is relieved auto-matically, first to the Boron Recovery System via the Vent and Drain System, then at a higher pressure to the liquid waste system.

In addition, there is an automatic deverting valve, which sends RCS water to the Boron Recovery System if the level in the volume control tank gets too high.

These parts of relief and automatic diversion would result in spreading radioactive fluids and gases in a large amount of piping and equipment throughout the Auxiliary building with the resulting increase in area dose rates, while*

deriving no substantial benefit.

Taking all these facts into consideration, we believe that the let-down portion of the eves should not be used during or after an accident and thus we have not considered that source in our shield-ing review.

2.

The Residual Heat Removal (RHR) system was not considered in our shieldmg review.* This system was not considered because all piping in this system is located inside the containment.

Therefore, the source activity for the containment also includes any source derived from the RHR system inventory.

3.

The Boron Recovery, Liquid Waste, Solid Waste, and Gaseous Waste systems were not considered in the shielding review for the following reasons:

a.

These systems were not designed or arranged to accommodate the activity levels that could be present after an accident, but rather to operate at the design conditions of one percent failed fuel as discussed in FSAR Section 11.

The calculated activity concentration based on Regulatory Guide 1.4 and TID-14844 of the influent to the Liquid Waste. or Boron Recovery system is approximately 2000 uci/cc, even after six months of radioactive decay.

Thus, concentrated effluent in the evaporators of these systems would be so highly radioactive that shielding, processing and handling of the waste by conventional methods would not be possible.

The area radiation dose rate from the concentrated waste and storage tanks would severely limit access to parts of the Auxiliary Building.

It is proposed to keep this waste inside containment until the recovery phase when cleanup operations begin.

b.-

Modifications are being designed to add connections to these and other systems to permit interfacing with an external waste processing system specifically selected for post-accident clean-up, during the recovery phase, of radioactive fluids and gases resulting from the accident.

II.B.2-15

Inclusion of all essential sources in the review:

All essential system piping and equipment required to mitigate the effects of a LOCA which contain or could contain highly radioactive fluids were considered as sources in our shielding review.

These systems include; the High Head Safety Injection (HHSI) portions of the eves and SI Systems, the Low. Head Safety Injection (LHSI) System, Recirculation Spray System, Sample System, and Containment Atmosphere Cleanup (hydrogen recombiner) System.

In addition, other systems which are not required to mitigate a LOCA and are not required by NUREG 0737, but which could contain significant radioactivity, were considered such as drain lines and standing water in sumps and waste tanks.

All branch connections to and from these systems were considered as sources to the first isolation val-ve.

Other sources such as the shine from the contain-ment dome, shine through containment penetrations., and shine through the personnel hatch were considered.

The location of field run pipe, which is part of the systems listed above, was considered in our analysis.

As noted in the response to the North Anna FSAR Comment 12.3, the routing and location of radioactive piping is such that the piping is in shielded areas.

The exact routing of our field run pipe is not critical in the production of the radiation zone maps.

The highest activity level in each zone is calculated and that level is considered for the entire zone.

For instance, the highest activity may be 12 inches from a pipe, regardless of its exact location within the zone.

Indirect radiation was not considered as a source.

Buildup factors in shield walls are considered but scatter over walls or through labyrinth*

doorways was not considered.

Airborne activity was not considered as a source in our shielding review.

Consideration of all vital areas:

Vital areas for personnel exposure are defined as those areas which re-quire continuous or frequent occupancy in order to control, monitor, and evaluate the accident.

These areas include the Control Room, Technical Support Center, the Counting Lab/Health Physics area, the Operational Support Center, and Security Control Center.

In addition, any area which requires access to perform manual operation

• of equipment in systems which are used to mitigate the accident were considered.

Vital areas for equipment qualification inlcude all areas in which mitigating equipment is located.

Zone maps have been developed for many non-vital areas as. well.

These areas include the entire Auxiliary Building, Main Steam Valve House, Quench Spray Pump House, Safeguards Building, Service Building, and selected areas in the yard.

a.

Post-Accident Sampling Modification The interim modifications made to the sampling system and the pro-posed long term modifications are designed to minimize the exposure to personnel during sampling using time, distance and shielding.

Shielding for portions of the interim sampling system lines has been installed.

Shielding for portions of the long term post-accident sampling system lines and the sampling facility is required.

II.B.2-16

b.

Technical Support Center

c.

Sufficient shielding will be designed into the technical support center to limit personnel exposure to acceptable levels.

Operations* Support Center (North Anna)

The Operations Support Center at North Anna will be located in an area with acceptable radiation dose rates.

d.

• Counting Laboratory

e.

f.

Shielding for some instruments in the counting laboratory or reloca-tion of the counting equipment is necessary due to background activities.

Hydrogen Recombiner Modifications (North Anna Only)

The hydrogen recombiner system at North Anna is external to the containment.

Because of the potential for a large dose contribution after post-accident operation modifications are being made.

The vault will be modified to provide a monitored release path for leakage from the hydrogen recombiner system to the Auxiliary ventilation system.

The post-accident hydrogen recombiner vault requires shielding modifications to limit radiation exposure to the operators at the vault while realigning and operating the recombiner and to reduce the levels in some of the vital areas.

Relocation of the recombiner control panel is also required.

Containment Atmosphere Cleanup System Manual valves, located in high radiation zones, which must be operated to line up and operate the post-accident recombiner at North Anna, and hydrogen analyzer, will. be replaced with environ-mentally qualified, remote operated valves.

Section II.E.4.1 shows the proposed modifications in the valve arrangement for the containment atmosphere cleanup system.

These modifications provide a double valve barrier between the accident unit's containment atmosphere which is being processed by either hydrogen recombiner, all other systems, and the unaffected contain-ment.

Remote operators will be provided for those valves where personnel access is restricted by post accident radiation levels.

II. B. 2-17

g.

Security Control Center The emergency procedures will be revised to incorporate instruction on relocation of the security boundary if radiation doses in the yard are not acceptable.

In addition, Vepco has committed to implementation of the following modifications at North Anna Units 1 & 2 and Surry Units 1 & 2.

These modifications are not required to validate the results of the shielding review or satisfy the requirements of NUREG 0578 but would reduce personnel exposure during the Recovery Phase.

Waste cleanup system tie-ins

( Additional design information in Attac}J.ment E to July 7, 1980 letter)

Auxiliary building and safeguards building sump drain modifications (Design information provided in April 1, 1980 letter)

II. B. 2-18

11.B.2 DESIGN REVIEW OF PLANT SHIELDING AND EQUIPMENT QUALIFICA-TION - EXCEPTIONS The shielding review is in compliance with or exceeds all the requirements of this Section except as noted below:

1.

Not all systems listed in the clarification were assumed to contain the Accident Level Radiation Source Term as described in the "Basis for Systems Excluded" and "Inclusion of all Essential Sources in the Review".

2.

Equipment qualification review for mechanical equipment required to miti-gate a High Energy Line Break (HELB) for effects of increased radiation is considered to be a new requirement.

This review is estimated to be completed by June 1, 1981 and may require additional equipment modifica-tion that could extend beyond the equipment qualification date of June 30, 1982.

The LOCA review is expected to be more conservative; therefore, few additional modifications are expected.

3.

The replacement of recirculation spray valve liners and seals at Surry Unit 2 is scheduled for the next refueling outage (currently scheduled for the first quarter of 1982).

This replacement requires extended access to the containment and vendor representatives to make the changes.

This modification should be complete by the required date of July 1, 1982 unless the refueling is extended by unscheduled outages during which the vendor representatives would not be available.

4.

The radiation zone maps for personnel access will be revised after all the post-TM! modifications are complete on January 1, 1982 to ensure the

• impact of all changes are incorporated.

This revision will take a few months and will replace the radiation zone maps prepared by January 1, 1980 that identified the required modifications for personnel access.

II.B.2-19

II. B.4 TRAINING FOR MITIGATING CORE DAMAGE

1.

A training program to teach the use of installed equipment and systems to control or mitigate accidents in which the core is severely damaged has been developed for use at both North Anna and Surry.

The program is currently being revised to incorporate newly acquired data.

2.

The program has already been implemented and taught at both stations.

Attendance was required for all licensed operator, licensed senior opera-tors and non-licensed operators.

Personnel identified by the Emergency Plan qualified to becorne Emergency Directors are required to attend, as are all Shift Technical Advisors and Nuclear Training Coordinators.

The training program has been incorporated into the operator training and retraining programs

• II. B.4-3

I I

The automatic initiation signals and circuits shall be designed so that their failure will not result in the loss of manual capa-bility to initiate the AFW system from the control room.

Response

Vepco has verified that the automatic start AFW signals and asso-ciated circuitry are safety grade.

The AFW system is initiated automatically by a safety injection signal, a loss of off site power, or a low-low steam generator level.

These actuation signals are testable and these signals are the system actuations on which the FSAR Chapter 15 accident analysis is based.

The AFW system is

• also automatically initiated on loss of the main feedwater pumps in anticipation of low steam generator level.

This anticipatory actua-tion is not testable during normal operation.

All initiation signals and circuits are designed to prevent a single failure from causing a loss of the AFW system.

ADDITIONAL SHORT-TERM RECOMMENDATIONS The following additional short-term recommendations resulted from the staff's Lessons Learned Task Force review and the Bulletins and Orders Task Force review of AFW systems.

1.

Recommendation The licensee should provide redundant level indications and low level alarms in the control room for the AFW system primary water supply to allow the operator to anticipate the need to make up water or transfer to an alternate water supply and prevent a low pump suction pressure condition from occurring.

The low level alarm setpoint should allow at least 20 minutes for operator action, assuming that the largest capacity AFW pump is operating.

Response

An additional level transmitter has been installed on the Emergency Condensate Storage Tank (ECST) for North Anna Unit 2 and is expected to be installed for Unit 1 during the upcoming refueling outage.

This instrument will provide the operator with both ECST level indication and a low level alarm in the main control room.

The alarm will be set to alert the operator* of ECST low level at least twenty (20) minutes before the ECST could be emptied by the largest auxiliary feedwater (AFW) pump.

Until the additional level transmitter is installed, operating per-sonnel have been instructed on how to correlate pump suction pressure with ECST level.

The operation of the AFW pump distorts suction pressure such that accurate level indicatipn is difficult; however, the suction pressure indicator is still useful to allow the operator to anticipate the need to make up water or transfer to an alternate water supply and prevent a low pump suction pressure condition from occurring.

ll.E.1.1-5

2.

Recommendation The licensee should perform a 72-hour test on all AFW system pumps, ff such a test or continuous period of operation has not been accomplished to date.

Following the 72-hour pump run, the pumps should be shut down and cooled down and then restarted and run for one hour.

Test acceptance criteria should include demonstrating that the pumps remain within design limits with respect to bearing/bearing oil temperatures and vibration and that pump room ambient conditions (temperature, humidity) do not exceed environmental qualification limits for safety-related equipment in the room.

Vepco indicated to the staff th~t a potential conflict existed between the requirement for a 72-hour endurance test on the turbine driven AFW pump and the LCO in Technical Specification 3.7.1.2.

The Additional Short-Term Recommendations No. 2 was then revised.

The revision was issued as follows:

Revision to Recommendation No. 2 of "Additional Short-Term Recom-mendations" Regarding Auxiliary Feedwater Pump Endurance Test*

The licensee should perform an endurance test on all AFW system pumps.

The test should continue for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after achiev-ing the following test conditions:

Pump/ driver operating at rated speed, and Pump developing* rate discharge pressure and flow or some higher pressure at a reduced flow but not exceeding the pump vendor's maximum permitted discharge pressure value for a 48-hour test.

For turbine drivers, steam temperature should be as close to normal operating steam temperature as practicable but in no case should the temperature be less than 400°F.

Following the 48-hour pump run, the pumps should be shut down and allowed to cool down until pump temperatures reduce to within 20°F of their values at the start of the 48-hour test and at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> have elapsed.

Following the cool down, the pumps should be restarted and run for one hour.

Test acceptance criteria should include demonstrating that the pumps remain within design limits with respect to bearing/bearing oil temperatures and vibration and that ambient pump room conditions (temperature, humidity) do not e~ceed environmental qualification limits for safety-related equipment in the room.

The licensee should provide a summary of the conditions and results of the tests.

The summary should include the following:

1) A brief description of the test method (including flow schematic diagram) and how the test was instrumented (i.e., where and how bearing temperatures were measured), 2) A discussion of how the test II.E.1.1-6

conditions (pump flow, head, speed and steam temperature) compare to design operating conditions,

3) Plots of bearing/bearing oil temperature vs. time for each bearing of each AFW pump/ driver demonstrating that temperature design limits were not exceeded,

4) A plot of pump room ambient temperature and humidity vs. time demonstrating that the pump room ambient conditions do not exceed environmental qualification limits for safety-related equipment in the room, 5) A statement confirming that the pump vibration did not exceed allowable limits during tests.

Response

The motor driven and turbine driven pumps have been endurance tested for North Anna Unit 2 and the motor driven pumps have been tested for Unit 1.

These results have previously been transmitted to the NRC.

The turbine driven pump for Unit 1 is expected to be tested prior to January 1, 1981 and results will then be submitted.

3

• Recommendation The licensee should implement the following requirements as specified by Item 2.1.7.b on page A-32 of NUREG-0578:

"Safety grade indication of auxiliary {eedwater flow to each steam generator shall be provided in the control room

• The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements for the auxiliary feedwater system set forth in Auxiliary Systems Branch Tech-nical Position 10-1 of the Standard Review Plan, Section 10. 49."

Response

Modifications are complete for upgrading the safety-grade indication of AFW flow from semi-vital bus power to vital bus power.

4.

Recommendation Licensees with plants which require local manual realignment of valves to conduct periodic tests on one AFW system train and which have only one remaining AFW train available for operatio~ should propose Technical Specifications to provide that a dedicated indi-vidual who is in communication with the control room be stationed at the manual valves.

Upon instruction from the control room, this -

operator would re-align the valves in the AFW system train from the test mode to its operational alignment.

Response

Periodic testing does not require local manual realignment of valves.

Also, there are three AFW trains available.

Therefore, no further action is required.

11.E.1.1-7

LONG-TERM Long-term recommendations for improving the system are as follows:

1.

Recommendation GL The licensee should upgrade the AFW system automatic initia-tion signals and circuits to meet safety grade requirements.

Response

The AFW system automatic initiation signals and circuits are pre-sently designed to meet safety grade requirements.

Item 3 of the Position corresponds to Enclosure 2 of the NR C Letter to Vepco dated September 28, 1979.

These requirements can be found along with the corresponding response in Attachment A.

II.E.1.1-8

II.E.1.1 AUXILIARY FEEDWATER SYSTEM EVALUATION SURRY UNITS 1 & 2 As stated in the clarification, the Staff reviewed Items 1 and 2, a,nd issued letters to those plants that required the implementation of certain short and long-term AFW system upgrade requirements.

This was the September 25, 1979 letter from the NRC to Vepco for Surry Units 1 and 2. of this letter listed those short and long-term requirements, while Enclosure 2 requested additional information concerning reevaluation of the AFW system flow rate design bases and criteria (Item 3 of Position).

The short and long-term requirements are listed below with a description of the modifications implemented, as required.

SHORT-TERM

1.

Recommendation G S-1

2.

The licensee should propose modifications to the Technical. Specifica-tions to limit the time that one AFW system pump and its associated flow train and essential instrumentation can be inoperable.

The outage time limit and subsequent action time should be as required in current Standard Technical Specifications; i.e., 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, respectively.

Response

A modification to the Technical Specifications to provide limited conditions of operation of the Auxiliary Feed water System have been submitted.

This modification will limit the time that one AFW system pump and its associated flow train and essential instrumentation cah be inoperable.

The limits of the Standard Technical 'Specifications were utilized

• Recommendation G S-4 Emergency procedures for transferring to alternate sources of AFW supply should be available to the plant operators.

These proce-dures should include criteria to inform the operator when, and in what order, the transfer to alternate water sources should take place.

The following cases should be covered by the procedures:

The case in which the primary water supply is not initially available.

The procedures for this case should include any operator actions required to protect the AFW system pumps against self-damage before water flow is initiated; and, The case in which the primary water supply is being depleted.

The procedure for this case should provide for transfer to the alternate water sources prior to draining of the primary water supply *

Response

Procedure modifications have been made to provide operators with guidance to diagnose availability of the primary water supply and protect the AFW system pumps against self-damage before water flow II. E.1.l-9

3.

is initiated.

The procedure also provides a prioritized list of alternate water sources and defines when and how to shift to the alternate sources as the primary source is depleted.

Recommendation GS-5 The as-built plant should be capable of providing the required AFW flow for at least two hours from one AFW pump train independent of any alternating current power source.

If manual AFW system initiation or flow control is required following a complete loss of alternating current power, emergency procedures should be estab-lished for manually initiating and controlling the system under these conditions.

Since the water for cooling of the lube oil for the turbine-driven pump bearings may be dependent on alternating current power, design or procedural changes shall be made to eliminate this dependency as soon as practicable.

Until this is done, the emergency procedures should provide for an individual to be stationed at the turbine-driven pump in the event of the loss of all alternating current power to monitor pump bearing and/or lube oil temperatures.

If necessary, this operator would operate the turbine-driven pump in an on-off mode until alternating current power is restored.

Adequate lighting powered by direct current power sources and communications at local stations should also be provided if manual initiation and control of the AFW system is needed.

(See Recommendation GL-3 for the longer-term resolution of this concern. )

Response

Our primary water source provides sufficient water volume to insure we can supply 700 gpm for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; independent of any alternating current power source.

No change is therefore re-quired.

Procedure modifications have been made to give the operator guidance for controlling feed to the steam generators manually and how the steam driven auxiliary feedwater pump can be manually started and controlled if necessary.

Each of our Auxiliary Feedwater pumps is cooled by a flow path from its own discharge back to its suction, independent of any alternating current power.

Therefore, no change is necessary.

The requirement to station an operator or rµodify procedures in the event of the loss of all AC power to insure bearing cooling is also unnecessary, as the cooling water is provided by the individual pump.

The concern for the operator to operate the turbine-driven pump in an on-off mode until alternating power is restored has been incorporated in the manual procedure as described in Item b above.

. II.E.1.1-10

/

/.

Emergency (D. C.) lighting has been provided to give sufficient lighting to manually control the turbine-driven AFW pump and manually control pump discharge valves.

Sound power phone communication is already available in the area of the* AFW pumps and discharge valves.

This circuit has been checked to be operable and head/hand sets have been provided to insure we have ready communication with the Control Room should it be necessary.

We feel one general comment is necessary.

The probability of a total loss of all normal and emergency alternating current sources would seem very small.

Only a series of failures without operator action would allow a complete loss of AC power.

With the present capability to supply the affected. unit's S/G's with auxiliary feed-water from the unaffected unit, it is extremely unlikely that we could deteriorate to having only steam driven auxiliary feedwater pumps available to both or either unit.

4.

Recommendation G S-6 The licensee should confirm flow path availability of an AFW system flow train that has been out of service to perform periodic testing or maintenance as follows:

Procedures should be implemented to require an operator to determine that the AFW system valves are properly aligned and a second operator to independently verify that the valves are properly aligned.

The licensee should propose Technical Specifications to assure that prior to plant startup following an extended cold shutdown, a flow test would be performed to verify the normal flow path from the primary AFW system water source to the steam gene-rators.

The flow test should be conducted with AFW system valves in their normal alignment.

Response

-The periodic test (PT-15) for testing the operability of the Auxiliary Feedwater pumps have been modified to provide for a second operator to verify that the valves manipulated as part of the test are in proper alignment following the completion of the test.

Our present start-up procedure, OP-1.4, provides for an actual flow verification of the Auxiliary Feedwater systems prior to taking the reactor critical.

A change to our Technical Specifications has been proposed to require the flow test, to verify normal flow path for the primary AFW system water source to the steam generators prior to plant start-up following an extended cold shutdown.

II.E.1.1-11

5.

Recommendation GS-7 The licensee should verify that the automatic start AFW signals and associated circuitry are safety grade.

If this cannot be verified, the AFW system automatic initiation system should be modified in the short-term to meet the functional requirements listed below.

For the longer term, the automatic initiation signals and circuits should be upgraded to meet safety grade requirements as indicated in Recommendation GL-5.

The design should provide for the automatic initiation of the auxiliary feed water system flow.

The automatic initiation signals and circuits should be designed so that a single failure will not result in the loss of auxiliary feedwater system function.

Testability of the initiation signals and circuits shall be a feature of the design.

The initiation signals and circuits should be powered from the emergency buses.

Manual capability to initiate the auxiliary feedwater system from the control room should be retained and should be imple-mented so that a single failure in the manual circuits will not result in the loss of system function.

The alternating current motor-driven pumps and valves in the auxiliary feedwater system should be included in the automatic actuation (simultaneous and/or sequential) of the loads to the emergency buses.

The* automatic initiation signals and circuits shall be designed so that their failure will not result in the loss of manual capa-bility to initiate the AFW system from the control room.

Response

The current design of the Auxiliary Feedwater System provides for automatic initiation.

All initiation signals and circuits are designed to prevent a single failure from causing a loss of the Auxiliary Feedwater System.

The Auxiliary Feedwater System is initiated automatically by a safety injection signal, loss of offsite power, and on low-low steam generator level of any one steam generator.

These actuation signals are testable and these signals are the system actuations on which the FSAR Chapter 14 safety analysis is based.

The Auxiliary Feedwater System is also automatically initiated on loss of the main feedwater pumps in anticipation of low steam* generator level.

This anticipatory actuation is not testable during normal operation.

II.E.1.1-12

6.

7.

All initiating circuits which automatically start the Auxiliary Feed water System, are powered from vital buses and are backed-up by the emergency power system.

The capability presently exists to manually initiate the A uxil-iary

• Feedwater System from the control room.

A single failure in the manual circutis will not result in the loss of system function.

The AC motor feed pumps in the Auxiliary Feedwater System are automatically,initiated.

The motor operated valves required to establish generators are left in the open position and also receive automatic signals.

These valves are under strict administrative control and can be operated from the control room.

The motor operated valves are powered from the emer-gency bus.

The capability of cross-connecting auxiliary ' feedwater and supplying Auxiliary Feed from the unaffected unit has been installed at Surry.

The valves receive automatic signal~ to open during certain steam rupture conditions.

They are powered from the vital bus and are controlled manually from the control room.

The same flow indications, individual steam

• generator isolation valve, and flow paths are utilized.

The automatic signals are designed in such a manner that their failure will not result in the loss of manual capability to start

• the Auxiliary Feedwater System.

The automatic initiation circuits are presently safety-grade equipment and meet the long-term requirements.

Recommendation Procedures should be established to lock open and periodically verify open position of all manual AFWS valves inside containment.

Response

Our startup procedure (OP-lB) presently verifies the open position of all manual AFWS valves inside containment.

In addition, our checklist (CL-53) which is done periodically, checks the Auxiliary Feedwater Pump Manual Discharge valves in the open position.

Procedures have been modified to lock open the manual valves inside the containment.

Recomendation The licensee should require staggering of the periodic pump train tests (e.g., one train at North Anna is tested every 10 days rather than all three trains tested at once on a monthly basis).

This reduces the potential for inadvertently leaving closed the discharge valves of all trains after test.

TL F.. 1. 1-U

Response

Our periodic testing has been staggered to test the motor driven and steam driven auxiliary feedwater pumps at different times to reduce the. potential for inadvertently leaving closed the discharge valves of all trains after a test.

8.

Recommendation Emergency procedures should be available to the operators for operating the AFWS of one unit such that it is supplying water to the steam generator(s) of the opposite unit in the event that such an. operating mode should be necessary.

Response

Emergency procedures have been modified to give operational guid-ance to utilize the other unit's AFWS should this condition become necessary.

ADDITION AL SHORT-TERM RECOMMENDATIONS The following additional short-term recommendations resulted from the staff's Lessons Learned Task Force review and the Bulletins and Orders Task Force review of AFW systems *

1.

2.

Recommendation The licensee should provide redundant level indications and low level alarms in the control room for the AFW system primary water supply to allow the operator to anticipate the need to make up water or transfer to an alternate water supply and prevent a low pump suction pressure condition from occurring.

The low level alarm setpoint should allow at least 20 minutes for operator action, assuming that the largest capacity AFW pump is operating.

Response

An additional safety-grade level transmitter will be installed on the Emergency Condensate Storage Tank (ECST).

This instrument" will provide the operator with both ECST level indication and a low level alarm in the main control room.

The alarm will be set to alert the operator of ECST low level at least twenty (20) minutes before the ECST could be emptied by the largest auxiliary feed water ( AFW) pump.

This modification should be fully implemented on Unit 1 during the steam generator replacement outage and on Unit 2 by th.e required implementation date.

The existing level indication loop will be upgraded at the same time.

Recommendation The licensee should perform a 72-hour test on all AFW system pumps, if such a test or continuous period of operation has not been accomplished to date.

Following the 72-hour pump run, the II.E.1.1-14

• - - ****------ ~*- **--*----=-~----

-~~ ~---~

pumps should be shut down and cooled down and then restarted and run for one hour.

Test acceptance criteria should include

• demonstrating that the pumps remain within design limits with respect to bearing/ -bearing oil temperatures and vibration and that pump room ambient conditions (temperature, humidity) do not exceed environmental qualification limits for safety-related equipment in the room.

The NRC, in a February 8, 1980 letter to Vepco, revised the pump endurance test requirements reducing the test from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

The revision was issued as follows:

Revision to Recommendation No. 2 of "Additional Short-Term Recom-mendations" Regarding Auxiliary Feed water Pump Endurance Test The licensee should perform an endurance test on all AFW system pumps.

The test should continue for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after achiev-ing the following test conditions:

Pump/ driver operating at rated speed_, and Pump developing rate discharge pressure and flow or some higher pressure at a reduced flow but not exceeding the pump vendor's maximum permitte~ discharge pressure value for a 48-hour test.

For turbine drivers, steam temperature should be as close to normal operating steam temperature as practicable but in no case should the temperature be less than 400°F.

Following the 48-hour pump run, the pumps should be shut down and allowed to cool down until pump temperatures reduce to within 20°F of their values at the start of the 48-hour test and at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> have elapsed.

Following the cool down, the pumps should be restarted and run for one hour.

Test acceptance criteria should include demonstrating that the pumps remain within design limits with respect to bearing/bearing oil temperatures and vibration and that ambient pump room conditions (temperature, humidity) do not exceed environmental qualification limits for safety-related equipment in the room.

The licensee should provide a summary of the conditions and results of the tests.

The summary should include the following:

1) A brief description of the test method (including flow schematic diagram) and how the test was instrumted (i.e., where and how bearing temperatures were measured), 2) A discussion of how the test conditions (pump flow, head, speed and steam temperature) compare to design operating conditions,

3) Plots of bearing/bearing oil temperature vs. time for each bearing of each AFW pump/ driver demonstrating that tempera~ure design limits were not exceeded, II.E.1.1-15

4) A plot of pump room ambient conditions do not exceed environ-mental qualification limits for safety-related equipment in the room,

5) A statement confirming that the pump vibration did not exceed allowable limits during tests.

Response

The motor driven and turbine driven pumps have been endurance tested for both Surry Units 1 and 2. These results have previously been previously transmitted to the NRC.

3.

Recommendation The licensee should implement the following requirements as specified by Item 2.1.7.b on page A-32 of NUREG-0578:

"Safety grade indication of auxiliary feedwater flow to each steam generator shall be provided in the control room.

The auxiliary feedwater flow instrument channels shall be powered from

• the emergency buses consistent with satisfying the emergency power diversity requirements for the auxiliary feedwater system set forth in Auxiliary Systems Branch Tech-nical Position 10-1 of the Standard Review Plan, Section 10.49."

Response

To meet the diversity requirements of ASTB-10-1, the Auxiliary Feedwater Flow Indication power supplies have been moved to an existing cabinet which meets the diversity requirements.

4.

Recommendation Licensees with plants which require local manual realignment of valves to conduct periodic tests on one AFW system train and which have only one remaining AFW train available for operation:-should propose Technical Specifications to provide that a dedicated indi-vidual who is in communication with the control room be stationed at the manual valves.

Upon instruction from the control room, this operator would re-align the valves in the AFW 'system train from the test mode to its operational alignment.

Response

Our plant does not require local manual realignment of valves to conduct periodic tests on the AFW system.

In the event that a periodic test is performed with only one AFW system train available, a dedicated individual who is in communication with the Control Room will be stationed at the manual valves and upon instruction from the Control Room, would re-align the AFW *valves to their operational alignment.

To insure this, a precaution has been added to the Periodic Test Procedure.

II.E.1.1-16

LONG-TERM Long-term recommendations for improving the system are as follows:

1

• Recommendation 2

• GL At least one AFW system pump and its associated flow path and essential instrumentation should automatically initiate AFW system flow and be capable of being operated independently of any alternating current power source for at least two hours.

Conversion of direct power to alternating current is acceptable.

Response

Our present design has at least one AFW system pump and flow path which is automatically initiated upon a loss of all AC power.

How-ever, in order to control the amount of flow delivered, the pump must be manually operated in an on-off mode.

In order to provide automatic, or in this case, remote on-off operation from the control room, a design modification must be made.

Hence, an analysis is in progress to determine the best method for providing this capability.

The analysis and subsequent modifications should be completed and/or implemented by January 1, 1982.

Recommendation GL The license should upgrade the AFW system automatic initiation signals and circuits to meet safety-grade requirements.

Response

The AFW system automat.ic initiation signals and circuits are pre-sently designed to meet safety grade requirements.

3.,

Recommendation The AFWS flow control valves for both the motor and turbine pump trains are AC powered, normally open, fail as-is motor operated valves which are located inside containment.

Also, manual, normally open valves are located inside containment.

The AFW design should be reevaluated, including the possibility of relocating the valves outside containment, assuming an accident inside containment which necessitates AFWS operation and which creates a containment environment (humidity, radiation) that precludes access to the valves.

The reevaluation should consider the following:

a.

A possible common mode failure (environmentally induced) casuing spurious closure or failure of the MOV's in a throttled posi-tion.

b.

An AFWS line break downstream of the MO V's and failure of the MOV's to operate.

II.E.1.1-17

Response

As a result of Vepco's I.E. Bulletin 79-0lB Review for Surry Units 1 and 2, it has been determined that the AFW flow control valves have environmental qualification data requiring further corroborating information from the manufacturers before

• final judgement can be made.

These valves are of the same general type and were purchased during the same time period as those which have been determined to be qualified.

We do not foresee any problem in having these com-.

ponents qualified for nuclear use.

Futhermore, these components will have performed their accident safety function prior to receiving a radiation dose above the threshold of 2500 rads.

Since we did not receive vendor qualification documentation in time for the December 1, 1980 update to I.E. Bulletin 79-0lB, we have placed a purchase order for

• replacement operators. If in the interim, prior to installation, we receive substantiation of qualification for the presently installed operators, the existing operators will not be replaced.

Therefore, only those installed operators that are proven not to be qualified will be replaced during the first outage of suffi-cient duration upon receipt of material.

Item 3 of the Position corresponds *to Enclosure 2 of the NRC letter to Vepco dated September 25, 1979.

These requirements along with the corresponding response can be found in Attachment B.

II.E.1.1-18

Part 2:

Auxiliary Feed water System Flowrate Indiciation Surry Units 1 & 2 The Auxiliary Feedwater System Flowrate Indication requirements are consis-tent with those set forth in General Design Criterion 13 to provide the capability in the control room to ascertain the actual performance of the AFWS when it is called to perform its intended function.

Those requirements for Westinghouse and Combustion Enginering Plants, as found in the clarification, are listed below with the corresponding response.

1.

2

• 3

• Requirement To satisfy these requirements, Westinghouse and CE plants must provide, as a minimum, one auxiliary feedwater flowrate indicator and one wide-range steam generator level indicator for each steam generator or two flowrate indicators.

Response

One auxiliary feedwater flowrate indicator and one wide-range steam generator lever indicator is provided for each steam generator on both Surry Units 1 and 2's main control boards.

Requirement The flow indication system should be environmentally qualified.

Response

Vepco is presently participating in a transmitter Qualification Program which is scheduled to be completed by the end of 1981.

Upon suc-cessful completion of the test program and receipt of materials, these transmitters will be replaced at the first outage of sufficient duration.

Transmitters which fall into this category for auxiliary feedwater flow indication are FT-FW-lOOA, B, C for Surry Unit 1 and FT-FW-200A, B, C for Surry Unit 2.

Flow indicators on the main control board (FI-FW-lOOA, B, & C and FI-FW-200A, B, C) are type Westinghouse VX-252.

In a letter to Vepco from Westinghouse dated September 2, 1980, Westinghouse stated that their VX-252 instruments are Class IE and meet the intent of the NRC programs including their Directive 79-0lB released January 14, 1980.

Since the control room :will not be subject to a change in environment during accident conditions (including accidents requiring the auxil-iary feedwater system), the flow indicators are environmentally qualified.

Requirement The flow indication system should be powered from highly reliable, battery backed non-Class IE power source.

II.E.1.2-P2-7

U.E.4.1 DEDICATED HYDROGEN PENETRATIONS.

North Anna Design The original North Anna design uses redundant external Hydrogen Recombiners shared between Units 1 and 2.

The Hydrogen Recombiner line takes suction from the same penetration used for the suction of the Containment Vacuum pumps, the Hydrogen Purge lines and the Hydrogen Analyzers.

Two contain-ment isolation valves located outside of containment will be installed on each of these lines.

Since radioactive gases could be flowing through these penetra-tions during the post-accident mode, these systems were considered to become extensions of containment and the modifications listed below were proposed.

The discharge line from the hydrogen recombiner shares the same penetration with the discharge line from the hydrogen analyzer.

Containment isolation is provided by a check valve inside containment and two remote manual valves outside containment.

The combined hydrogen recombiner suction and discharge line is sized such that the flow requirements for the use of the combustible gas control system are satisfied.

The attached basic flow diagram shows the. modified containment atmosphere clean-up system layout.

This modification will enable the control room operator to line up the system using remotely operated valves and establish flow from the containment to the hydrogen analyzer and the hydrogen recombiner without exposure to high radiation.

The modified system will retain the design basis for single failure criteria.

Valves will be added to the existing valves relocated in the suction of the containment atmosphere purge blower to provide double barrier isolation.

Also, the Post-Accident Sampling System, Containment Vacuum

& Gaseous Waste System connections will be provided with double isolation valves.

Throttle valves will be added in the suction lines from Unit 2 containment to the hydrogen recombiners to balance the system ensur-ing the required flow rate can be established.

The pipe run from Unit 2 containment to the recombiners is considerably shorter than from Unit 1 containment; therefore, throttling for Unit 2 may be required to provide the proper flow balance from both units.

The System under.normal plant operations will be lined up with all containment isolation valves closed except the containment isolation valves for the Contain-ment Vacuum system which are periodically opened during vacuum pump *opera-tion.

Under accident conditions,* the Containment Vacuum isolation valves will receive a containment isolation Phase "A" signal to close.

The operator will open from the control room either the "A" or "B" train isolation valves to place the hydrogen recombiner and/or hydrogen analyzer in operation.

The non-accident unit as well as all tie-in systems (Containment Vacuum, Post-Accident Sampling and Gaseous* Waste) will be isolated from the containment atmosphere by a double valve barrier.

The backup Hydrogen Purge system is presently isolated from the hydrogen analyzers and recombiners by an administratively locked closed valve.

This system is not operated during normal plant operations.

Its use would only be comtemplated if both hydrogen recombiners fail.

Vepco has committed to relocation of remote manual valves to areas accessible within five days per the requirements of the North Anna FSAR.

II.E.4.1-3

II.F.1 ATTACHMENT 1, NOBLE GAS EFFLUENT MONITOR A.

Process vent and ventilation 1 vent monitors

1.

Refer to item 4 below for the upper range capacity.

2.

The vent effluent monitors have a total range of concentration extending from a minimum design condition of 10-7 uci/cc to a maximum of 105 uci/cc - Xe133 (Table 11.F.1-1 requires a maximum of only 103 uci/cc).

Three detectors are required to cover this range and have an expected minimum range overlap of a factor of ten.

B.

Main steam SV.& DV and auxiliary FWPT exhaust monitors

1.

Ref er to item 4 below for upper range capacity.

2.

The Main Steam SV & DV and Auxiliary FWPT exhaust monitors have a total range of concentration extending from a minimum design condition of 18 uci/cc to a maximum of 109 uci/cc (Kr85).

Three detectors are required to cover this range and have a range overlap.

1.

The gaseous effluent monitors meet the following requirements specified in Table II. F.1-1:

a.

The following potential accident release paths are monitored:

1 Process vent stack Ventilation vent stack(s)

Main steam safety valve and atmospheric dump valve discharge Auxiliary feedwater pump turbine (FWPT) exhaust

b.

Design basis maximum range (requirements - 10+3 uci/cc) - Refer to item 4 for each effluent monitor range and radionuclide spectrum distribution.

c.

Redundancy - None

d.

The offline monitors (process vent and ventilation vents) sampling design criteria will be per ANSI N13.l-1969.

e.

The effluent monitors in item (a) will be powered by the emergency power supply system.

f.

The following calibration sources will be used:

i)

Process vent and ventilation vent monitors Beta Detector - c136 GM Detector - co60 II. F. l-Al-6

ii)

Main steam and auxiliary FWPT exhaust monitors GM Detector - cs137 Ion Chamber Detector - cs137

g.

A continuous display with recording capability is provided.

i)

Process vent and ventilation vent monitors are digital based systems which contain the following information on demand:

30 values of ten-minute averages for radiation, uci/ cc 30 values of one-hour averages for radiation, uci/cc 30 values of one-day averages for radiation, uci/cc 30 values of ten-minute averages for integrated release 30 values of one-hour averages for integrated release 30 values of one-day averages for integratec;I release Also, radiation readings will be recorded by analog strip chart recorders.

ii)

Main Steam RV & DV and Auxiliary FWPT exhaust monitors will provide continuously recorded radiation readings by analog strip chart recorders.

h.

Refer to the response to Item 4.a.i for qualification details.

i.

Refer to the response to Item 4. a for design considerations.

. 2.

The post-accident gaseous effluent monitors will be capable of functioning both during and following an accident.

The monitors are seismically installed but a seismic system design is not required.

The effluent monitors are designed to accomodate a design-basis release (10+3 uci/ cc xe133 equivalent) and respond to decreasing concentrations of noble gases down to the sensitivity (i.e., minimum detectable level, MDL) of each monitor.

Refer to Item 4.a.i for the sensitivity of each monitor type.

3.

The Main Steam RV & DV and Auxiliary FWPT exhaust monitors will be externally mounted monitors viewing the Main Steam line and Auxiliary Feed Water Pump Turbine exhaust line, respectively.

Procedures will be

_ developed to correct for the low energy gammas that the external moni-tors will not detect.

Refer to Item 4. b for a description of these procedures.

4.

The design decription of the post-accident gaseous monitors is as follows:

Process Vent and Ventilation Vent Effluent Monitors

a.

System

Description:

(i)

Kaman Sciences Corporation's model KMG-HR effluent monitoring system will be used to monitor noble gas effluents.

The system consists of noble gas monitors, particulate and Iodine collectors, isokinetic nozzle, and control room display. The noble gas mon-itors consist of three detectors, one Beta scintillation and two GM tubes, and cover the following ranges (sensitivity. to xe133):

II. F. l-Al-7

Normal Background - 1 mr/hr co60 Low Range Mid Range High Range 5.2 x 10-7 to 3.7 x io-1 uci/cc 9.6 x 10-4 to 1 x 102 uci/cc 7.1 x 10-l to 1 x 105 uci/cc Maximum Accident Background - 600 mr/hr co60 Low Range Mid Range High Range 1.2 x 10-6 to 3. 7 x 10-l uci/cc

1. 7 x 10-2 to 1 x 102 uci/cc 1.8 to 1 x 105 uci/cc NOTE:

The above sensitivity values are based on ANSI Nl3.10 -

1974 Procedures.

The energy response of the detectors are:

Low Range Mid Range High Range (To be supplied later)

Calibration - Each prototype detector/sampler design is processed through a primary radionuclide calibration.

Two or more isotopes are prepared to simulate the effluent for which the system will be used.

These isotopes are assayed by direct comparison with NBS or commerically certified isotope standards and thus become primary calibration sources.

A secondary point source is directly.referenced to the primary calibrated source and is used to calibrate each detector supplied by Kaman.

Field calibration sources and fixtures will be used to periodically recalibrate each detector and in accordance with the Technical Specification requirements.

(ii) Sampler Probe Design - The sampler probes will be designed in accordance with ANSI N13.1 and will be located down-stream of the last effluent entry point prior to discharge.

Sampler/ Detector Location -

The sampler will be located in a low background area (less than 600 mr/hr under accident conditions) such that the detector sensitivities are not adversly affected.

Refer to the list of monitor ranges for maximum background conditions provided above.

(iii) & (iv)

Location of Instrument Readouts - The readouts will be located in the control room providing the information discussed by item 1.h above.

The readings will be continuously recorded and can be obtained at least every 15 minutes during and following an accident.

The procedure for transmitting or disseminating this information will be developed and provided prior to operation.

II. F. l-Al-8

b.

(v) Source of Power -

The monitors will receive power from the station emergency power supply system.

Methods used to convert radiation readings* (uci/cc) to release rate per unit time (uci/sec):

The process vent and ventilation vent monitors are digital based systems that will receive signals from the detectors and vent flow meters, process these signals, and display results as radiation concentrations (uci/cc) or release rates (uci/sec) on demand.

The effluent concentrations and release rates (or duct flow rates) will be continuously recorded.

Periodic grab samples can be obtained for a laboratory spectral analysis to determine the radionuclide distribution as a function of time after shutdown.

Main Steam RV & DV and Auxiliary FWP.T Exhaust Monitors

a.

System Description

(i)

Externally mounted monitors will view the main steam line upstream of the safety valves/dump valves and downstream of the auxiliary feedwater pump turbine.

Nuclear Research Corporation's model TA900-TA600 area monitoring system detec-tors are located within a collimated two inch thick lead shielded enclosure.

The system consists of a three detector area monitor (two GM tubes and an Ion Chamber) and a remote readout/

control unit.

The detector range overlap and cover a dynamic range of.01 MR/hr to 10,000 R/hr, which represents the following approximate effluent concentrations based on a time zero TID-14844 nuclide source term and various background conditions:

Monitor Location Normal Background Conditions Approximate Range

Background

M. S.

1.8 x 101 to 1 x 109 uci/cc (Kr85 equiv. cone.*)

.75 MR/hr - co60 or (later)

(Xe133 equiv. cone.)

AFWPT

• • to 1 x 109 uci/cc (Kr85 equiv. Cone.*)

or (later)

(Xe133 equiv. cone.)

LOCA W/0 SI Recirc. (TMI)

Monitor Location Approximate Range

Background

2 R/hr M. S.

3.5 X 104 to 1 X or (later) 109 uc1*/cc (K 85

• -)

r133eqm~. cone.

(Xe equiv. cone.)

AFWPT

• • to 1 x 109 uci/cc (Kr85 equiv. cone.*)

or (later)

(Xe133 equiv. cone.)

11.F.l-Al-9

LOCA With SI Recirc.

Monitor Location Approximate Range *

Background

200 R/hr M.S.

3.5 x 106 to 1 x 109 uci/cc (Kr85 equiv. cone.*)

or (later)

(Xe133 equiv. cone.)

AFWPT NOTES:

• • to 1 x 109 uci/cc (Krs5*equiv. cone.*)

or (later)

(Xel33 equiv. cone.)

• The effluent concentrations are based on a time zero TID nuclide mixture converted to an equivalent Kr85 (or xe133) mixture.

• • Low range dependent upon final design Energy Dependence - +/- 20% at 80 Kev to 1.5 Mev energies.

Calibration -

Each detector is calibrated with a point ~ource (Cs137) traceable to NBS.

The frequency of the recalibrations will be in accordance with Technical Specification requirements.

Analytical methods are utilized to convert radiation readings (MR/hr) to effluent concentrations (uci/cc -

Kr85 equiv.) in lieu of a primary radionuclide calibration.

(ii) Monitor Location - The Main Steam monitors are located in the ma.:insteam valve house oriented such that one monitor "views" each relief header for a total of three monitors per unit.

The radiation background in the main steam valve house varies with the type of accident postulated.

The effect of background on monitor sensitivity for normal conditions, a LOCA with out Safety Injection recirculation, and a LOCA with Safety Injec-tion recirculation are listed in the Table above.

The Auxiliary Feedwater Pump Turbine exhaust monitors will be located in the auxiliary feedwater pump house for North Anna and on the wall outside the main steam valve house for Surry.

One monitor will "view" the turbine exhaust line for each unit.

The radiation background will be less for the Auxiliary FWPT exhaust and the effect on monitor sensitivity will be assessed.

(iii) Location of Instrument Readouts -

The readouts are located in the control room at North Anna and in the electrical switch gear room at Surry.

Both areas are accessible following an accident.

(iv) The readings will be continuously recorded in the control room and can be obtained at least every 15 minutes during and following an accident.

11.F. l-Al-10

-~-

(v)

The procedure for transmitting O! disseminating this information have been developed and issued (EPIP's) for the Main Steam effluent monitors.

Source of Power -

The monitors will receive power from the station emergency power supply system.

b.

Methods used to convert radiations readings to release rates per unit time:

The detector radiation readings (mr/hr) will be converted to effluent concentrations (uci/cc -

Kr85 equivalent) by using analytically derived conversion graphs.

The graphs display dose rate (mr/hr) versus Kr85 equivalent concentration (uci/cc) based on time zero TIO 14844 mixture.

Also correction factors as a function of time will be used to correct for changes in the radionuclide spectrum after shutdown.

Steam flow released from the safety valves or dump valves is deter-mined by the design flow through each valve times the number of valves opened.

The steam flow from the Auxiliary Feedwater pump turbine exhaust will be determined by measuring the AFW pump flow rate (GPM) and correlating the pump BHP requirement at that flow to the turbine steam requirements.

II. F.1-Al-11

11.F.1.

ATTACHMENT 1, NOBLE GAS EFFLUENT MONITOR - EXPECTIONS I.

Process Vent and Ventilation Vent Effluent Monitors Some design information will not be available by January 1, 1981 for NRC review.

In addition, procedures and operating instructions will be deve-loped as additional information is received from the vendor and will be available by the required implementation date, but not by January 1, 1981.

II.

Main Steam SV &DV and Auxiliary FWPT Exhaust Effluent Monitors Some design information will not be available for the Auxiliary FWPT exhaust effluent monitors by January 1, 1981, the conceptual design is similar to the Main Steam SV

&: DV effluent monitors but the details of shielding and mounting are still in design.

Preliminary procedures are in effect for the installed main steam SV & DV effluent monitors.

How-

ever, final procedures for both effluent monitors are still under development to adequately consider the requirements of this section.

These final procedures will be in effect by the required implementation date.

II. F.1-Al-12

II. F.1, - Sampling and Analysis of Plant Effluents

1.

The continuous collection of post accident releases of radioactive iodines and particulates meets the following requirements as specified in Table II. F.1-2:

Applicability - Process vent and ventilation vent effluent release paths.

Equipment - Kaman Sciences Model KMG-HR effluent monitoring system will continuously sample and collect particulates and iodines on shielded filters with provisions for removal and laboratory analysis.

The samples will be analyzed with site Germanium detectors for low activity samples.

The highly radioactive samples will be analyzed as discussed in the "Analysis" section below.

Design Basis Shielding Enveloped - The high range particulate and iodine collectors are mounted inside of a three inch 4II lead shielded assemblies.

A portable transfer housing containing three inches of 4II lead shielding will be used for personnel protection during the transfer of the filters.

Occupational exposure for filter removal will be negligible due to the lead shields.

Assuming a collecting time of thirty minutes at the maximum posulated accident of 102 uci/ cc (. 5 mev gamma) with a design sample flow rate of 1000 cc/min through the filters, the radiation dose rate one foot from the filter assembly will be O.1 mR/hr.

Sampling Media -

The iodine collector is a charcoal or silver zeolite cartridge with an effective absorption (for methyl iodines) of not less than 95%.

The particulate collector is a paper filter disk with an effective retention of 99% of O. 3 micron particles.

Sampling Considerations -

A.

Representative Sampling -

The sampler probes will be designed in accordance with ANSI-N13.1 and located downstream fo the last effluent point prior to discharge.

The sample piping will be routed using ANSI-N13.1, Appendix B as a guide (i.e., minimize sample line bends and lengths).

The sample flow isokinetics is addressed in item (3).

B.

Entrained Moisture - Refer to Item ( 4)

C.

Continuous Collection -

The effluent monitors will continuously operate, in accordance with Technical Specifications, as long as the exhaust flow occurs.

D.

Limiting Occupational Dose -

The particulate and iodine collector design is sufficiently shielded to minimize personnel exposure during filter transfer to the count laboratory.

Refer to* "Design Basis Shielding Envelope" response above

• II.F.l-A2-4

A detector dedicated to each collector will limit the activity on the filters to preset levels and provides indication on the rate of radio-activity buildup.

Three* foot tongs will be used to transfer the unshielded filter from the collector to the portable transfer housing thus avoiding direct contact with the filters.

Shielding will be designed to protect personnel during sample analysis.

E.

Analysis - (To be supplied later)

2.

Shielding Design Basis - Refer to Item (1)

3.

Isokinetic Sampling - The effluent monitoring sampling design is currently being formulated and will be submitted later.

4.

Entrained Moisture - Effluent streams will be analyzed to determine the potential of containing entrained moisture.

If a potential exists, then the sample lines will be heat traced to prevent condensation.

II. F.1-A2-5

II.F.1, ATTACHMENT 2, SAMPLING AND ANALYSIS OF PLANT EFFLUENTS -

EXCEPTIONS Some design information will not be available by January 1, 1981 for NRC review.

Information will be made available as it is received from the vendor.

I II. F.1-:-A2-6

11.K.3.1 INSTALLATION AND TESTING OF AUTOMATIC POWER OPERATED RELIEF VALVE ISOLATION SYSTEM Vepco will respond to the requirements of this section after the studies speci-fied in II. K. 3. 2 have been completed.

II. K. 3.1-3

11.K.3.2 REPORT ON OVERALL SAFETY EFFECT OF POWER-OPERATED RELIEF VALVE ISOLATION SYSTEM The Westinghouse Owners Group is in the process of developing a report which will address historical valve failure rate data and documentation of actions taken since the TMI-2 accident to decrease the probability of a stuck-open PORV.

This report will be used to support a decision on the necessity of incorporating. an automatic PORV isolation system, as specified in Item II. K. 3.1.

Due to the time-consuming process of data gathering, breakdown, and evalu-*

ation, it will not be possible to submit the report by January 1, 1981.

The report is scheduled for submittal on or about March 1, 1981.

II. K. 3. 2-4

11.K.3.3 REPORTING SAFETY AND RELIEF VALVE FAILURES AND CHALLENGES These requirements are being implemented at North Anna Units 1 and 2 and Surry Units 1 and 2.

North Anna Unit 2 Technical Specification 6. 9.1. 6 requires documentation of all challenges to the PORV's and Safety Valves with the routine monthly operating report.

Failures of these valves will be reported in accordance with Specification 6. 9.1. 8.

Similar provisions will be proposed for North Anna Unit 1 and Surry Techntcal Specifications on or about January 1, 1981.

II. K.3.3-2

II. K. 3.5 AUTOMATIC TRIP OF REACTOR COOLANT PUMPS DURING LOSS-OF-COOLANT ACCIDENT The Westinghouse Owners Group will submit the results of a blind post-test prediction of the LOFT L3-6 test by February 15, 1981 or two (~) months following the completion of the test, whichever is later.

This is consistent with the schedule provided on page II.K.3.5-1.

Should the NRC review of the submitted information indicate a need for auto-matic trip of the RCS pumps during a LOCA, proposed design modifications and an implementation schedule will be submitted by July 1, 1981.

II.K.3.5-3

II.K.3.17 REPORT ON OUTAGES OF EMERGENCY CORE-COOLING SYSTEMS LICENSEE REPORT AND PROPOSED TECHNICAL* SPECIFICATION CHANGES Operating records for the last five years at Surry Units 1 and 2, and since the commencement of operation at North Anna Units 1 and 2, are being re-viewed to determine the dates, lengths and causes of outages of emergency core cooling systems.

The results of this review for North Anna 1 and 2 will be compiled into a report and submitted prior to January 1, 1981.

Since the ECCS outage records for Surry cover a longer period of time and are in a form less amenable to the compilation of the required information, additional time will be required to complete the report.

Submittal of the report will be made prior to March 1, 1981.

II.K.3.17-3

• • • -- --**----------*-****-*-** **---~-........._... *------*-------*-*-*-**

III.A.1.2 UPGRADE EMERGENCY SUPPORT FACILITIES Additional clarification will be provided in the near future.

III.A,1,2-1

III.A.1.2 UPGRADE EMERGENCY SUPPORT FACILITIES Vepco will respond once the addition clarification is provided.

III. A.1. 2-2

III. A.2 IMPROVING LICENSEE EMERGENCY PREPAREDNESS - LONG-TERM

1.

The Radiological Emergency Response Plan for North Anna Units 1 and 2 was filed with the Commission on May 1, 1980, and subsequently amended in response to Staff comments.

The plan included a description of the program to provide the elements of NUREG 0654, Appendix 2.

The NRC Staff's review and approval of the North Anna plan is documented in Supplements 11 and 12 to NU REG 0053, the Safety Evaluation Report for North Anna Unit 2.

The Surry emergency plan was submitted for review on June 16, 1980.

Based on recent comments from the Staff, the majority of the first amendment is scheduled to be submitted prior to January 2, 1981.

2.

3.

The requirement to submit radiological emergency response plans has been met for both Surry and North Anna.

The Emergency Plan Implementing Procedures for both Surry and North Anna are currently under revision.

They will be submitted for review by the Staff on or prior to March 1, 1981.

Implementation of the revised radiological response plans is scheduled to be accomplished prior to* April 1, 1981.

A meteorological measurements program which incorporates the features of both element.1 and element 2 of Appendix 2 of NU REG 0654, including

~he provision of backup power* supplies tp both primary and backup meteorological instrument towers, and display of this data in the control room, will be operational by April 1, 1981.

An operable dose calculational methodolgy (DCM) shall be available for emergency use.

Implementation of element 3, Real-Time Predictions of Atmospheric Efflu-ent Transport and Diffusion, and element 4, Remote Interrogation of the Atmospheric Measurement at Prediction Systems, is related to the installa-tion of equipment to meet the requirements of NU REG 0696, "Functional Criteria for Emergency Response Facilities".

Installation of equipment required by NU REG 0696, including new computer capabilities, will not be complete prior to July 1, 1982.

Consequently, milestone 4, which requires installation of Emergency Response Facility hardware and software by March 1, 1982, must be deferred to July 1, 1982, consistent with NU REG 0696.

At that time, the Class A model should be available, in accordance with milestone 5.

Futher modifications to the Class A model, as well as implementation* of the Class B model, will take place on a schedule to be determined based on ongoing discussions with the Commission

• III. A. 2-7

1 -*

i

• III.D.3.4 CONTROL ROOM HABITABILITY REQU°IREMENTS A control room habitability review was conducted for the North Anna control room as a part of the Unit 2 full power licensing review.

Based on the reviews conducted in accordance with Standard Review Plan Sections 2. 2.1, 2.2.2, 2.2.3 and 6.4, and Regulatory Guide 1.78 and 1.95, we concluded that the control room meets the specifications and guidance in these SRP sections and Regulatory Guides; and therefore, no modifications are required.

Since the control room is common to both units, this conclusion applies to both Units 1 and 2.

Additional information required by item III.D.3.4, beyond that which was required for full power licensing is being compiled and will be submitted by January 1, 1981.

The control room habitiability review for Surry Units 1 and 2 is underway.

Based on current progress in this program, it will not be possible to complete the review by the January 1, 1981 date.

We propose to submit an analysis of habitability with respect to all on-site hazards by January 19, 1981.

The special case of chemical shipments along the James River will require additional study.

We expect to complete this portion and submit the report by June 30, 1981.

III. D. 3. 4-6