ML19105B047

Superseded Pages Per Revision 3 to Response to NUREG 0737 Post TMI Requirements

ML19105B047
Person / Time
Site:	Surry
Issue date:	06/28/1982
From:	Virginia Electric & Power Co (VEPCO)
To:	Office of Nuclear Reactor Regulation
References
NUREG 0737
Download: ML19105B047 (46)

Text

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RESPONSE TO NUREG 0737

\\PosT TMI - REQUIREMENTS*

\\

REVISION 0 REVISION 1 REVISION 2

/

CLARIFICA-TION ITEM SHORTENED TITLE I.A.1.1 I.A.1.2 I.A.1.3 I.A.2.1 I.A.2.3 I.A.3.1 I.C.1 I.C.2 I.C.3 I.C.4 I. C.5 I.C.6 Shift technical advisor Shift supervisor responsibilities Shift manning Immediate upgrad-ing of RO and SRO training and qualifications Administration of training programs Revise scope and criteria for licensing exams Short-term accident and procedures review Shift & relief turnover procedures Shift-supervisor responsibility Control-room access Feedback of operating experience Verify correct performance of operating activities B-4 EXCEPTIONS NONE NONE NONE NONE NONE NONE NONE NONE NONE NONE NONE NONE j

CLARIFICA-TION ITEM SHORTENED TITLE EXCEPTIONS I.D.1 Control-room design

1)

Although, the NRC issued "interim requirements" (NUREG-0700) in September 1981, Vepco cannot commit resources to the project until the 2

NRC integrates the schedule require-ments of the control room review with other related issues, (i.e., upgrading emergency procedures, emergency response facilities, etc.).

I.D. 2 Plant safety parameter display console NONE I

1 II. B.1 Reactor coolant system

1)

Vepco requires a relaxation of the vents final documentation submittal date to January 1, 1982 rather than July 1, 1981 for 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.

11.B.2 Plant shielding

1)

Some systems* included in the NRC 1

clarifications were excluded from the review.

Justifications are provided.

2)

Liners and seals in the recircula-tion spray valves at Surry Unit 2 will be replaced at next refueling 2

currently scheduled for the last quarter, 1981.

3)

Radiation zone maps for personnel access will be revised after all 1

designs are installed.

• 4)

Delivery of some equipment will require that installation be beyond 2

the required implementation date.

B-5

i I

CLARIFICA-TION ITEM II.B.3 II.B.4 II.D.1 II.D.3 II.E.1.1 II.E.1.2 II.E.3.1 SHORTENED TITLE EXCEPTIONS I

Post-accident sampling

• 1)

North Anna Unit 1 Containment Training for mitigating core damage Relief & safety-valve test requirements Valve position indication Auxiliary Feedwater system evaluation Auxiliary feedwater system initiation Emergency power for pressurizer heaters Atmospheric Sampling System uses a temporary sample return line until the permanent Hydrogen Analyzer return lines are installed by 7-1-82.

• 2)

The Containment A tmopsheric Samp-ling System permanent supply and return lines for North Anna Unit 2 and Surry Unit 1 will be installed at the respective first scheduled outages after 1-1-82.

3)

QA Category II control switches will be used until QA Category I switches are available.

• 4)

The delay on major equipment delivery will result in the system testing not being complete prior to the implemen-'

tation date of 1-1-82.

The system will be operable by 7-1-82.

NONE

1)

The EPRI program has not formally included the testing of block valves.

1) Equipment is undergoing seismic and environmental testing and is scheduled to be completed Fall, 1982.

1)

Surry AFW Pump Auto Control (Loss of all AC) Modification will be com-plete by 1-1-82 with the exception of installing qualified pressure regu-lations.

Non-qualified regulators will be used until the other are available.

• 1)

Surry AFW Flow Transmitters will be upgraded per the I.E. B. 79-0 lB schedule.

NONE B-6 2

2

CLARIFICA-TION ITEM SHORTENED TITLE EXCEPTIONS

-11.E.4.1 Dedicated hydrogen NONE 1

penetrations 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.

II.F.1 Accident-monitoring

1.

Noble gas effluents

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

• 2)

The delay on major equipment delivery will result in installation beyond the implementation date of 1-1-82.

2.

Sampling & analysis *1)

The delay on major equipment delivery of effluents will result in installation beyond the 2

implementation date of 1-1-82.

3.

Containment

' 1)

Detectors are installed.

System not radiation monitors complete until control room panels are delivered, which may be after 1-1-82.

4.

Containment

1)

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

Upgrade will 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 will not be complete by January 1, 1982.

6.

Containment

1)

Hydrogen Analyzers installed by hydrogen 1-1-82.

Analyzers functional by 2

7/82.

B-7

CLARIFICA-TION ITEM II.F.2 II.G.1 II.K.1 II.K.2 II. K.3 SHORTENED TITLE EXCEPTIONS Instrumentation for

1) Extension requested until July, 1982.

detection of inadequate core cooling

2)

Clarification item (7) is a new re-quirement.

Comparison of the proposed Westinghouse design with the Appendix B requirements is not complete.

• 3)

The ability of any system to provide 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.

Power supplies for NONE pressurizer relief valves, block valves and level indicators IE Bulletins NONE Orders on B&W plants

.13 Thermal Mechanical Supports NONE NONE I

.17 Voiding in RCS Final recommendations, B&O task force

.1 PORV isolation system

• 2 PORV Failures

.3 SV & RV Failures B-8 NONE NONE NONE 2

2 1

CLARIFICA-TION ITEM SHORTENED TITLE EXCEPTIONS II. K. 3 Final recommendations, (Cont'd)

B&O task force

.5 Auto trip RCP's Vepco requires an extension of the July 1, 1981 date for submittal of proposed design information to three 1

(3) months from NRC approval of the analysis model used.

This re-quest is consistent with the timetable provided in Item II. K. 3. 5.

.9 PI D Controller NONE

.10 Proposed Anticipatory Trip NONE

.11 Use of certain PORV's NONE

.12 Anticipatory Trip on NONE Turbine Trip

.17 ECC system outages NONE 1

.25 Power on Pump Seals NONE

.30 SB LOCA Methods NONE

.31 Compliance with 10 CFR 50.46 NONE

/

III. A.1.1 Emergency prepared-NONE ness, short-term III. A.1. 2 Upgrade emergency NONE 1

support facilities III.A.2 Emergency

• 1)

Implementation of improved meteoro-

/

preparedness logical data capabilities must be conducted on the same schedule as portions of III.A.1.2.

III. D.1.1 Primary coolant outside NONE containment III.D.3.3 Inplant radiation NONE monitoring B-9

CLARIFICA-TION ITEM III. D. 3. 4 SHORTENED TITLE Control-room habitability Appendix B Design Criteria Appendix B Design Criteria

( continued)

EXCEPTIONS Supplemental review in progress for Unit 3 construction material affect on Unit 1 & 2 Control Room habitability.

• 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.

• 3)

Equipment does not meet the IEEE-323-197 4 requirements of R. G. 1. 89 -.

• 4)

Existing plant systems meet original plant criteria for ele_ctrical separation but not R. G. 1. 7 5.

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 2

STATUS LIST OF RESPONSES TO NUREG 0737 POST-'lMI REQUIREMENTS FOR OPERATION REACTORS NRC NRC Post-Imple-Tech Clsrifi-Implemen-Pre-Imple-mentation Spec.

cation tation mentation Review Revision Submittal Vepco Item Shortened Title Descrietion Schedule A22roval Re9.uired Re9.uired Re9.. 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 Complete

3.

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

4.

Describe long-term 1-1-81 No No No 1-1-81 Submitted 1

program I.A.1.2 Shift supervisor Delegate non-safety 1-1-80 No Yes No 1-1-80 Complete responsibilities 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 Immediate upgrading

l.

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

2.

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

3.

Three mo trng 8-1-80 No Yes No None Complete qualifications 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 I~

I

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

cation tation mentation Review Revision Submittal Vepco Item Shortened Title Descri2tion Schedule

~l!roval Reguired Reguired Reg. BI 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 1.c.,/

10-1-81 No No No None Complete 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 Submitted I 1 propose guidelines

b.

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

3.

Transients & accidents

a.

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

b.

Revise procedures lat refueling Yes No No TBD 1

outage a~ter 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 aper responsibilities I.C.4 Control roOID access Establish authority 1-1-80 No Yes No 1-1-80 Complete limit access D-3

_l

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

Spec, cation tation mentation Review Revision Submittal Vepco Item Shortened Title Descrij!tion Schedule A22roval Reguired Reguired Reg. BI Remarks I.C.5 Feedback of operating License to implement 1-1-81 No Yes No None Complete 1-1 experience procedures I.C,6 Verify correct Revise performance 1-1-81 i:

No Yes No None Complete I

2 performance of procedures operating activities I.D. l Control room design Preliminary assessment TDD TDD E Response after I 1 j

• {

issue of NUREG 0700 I.D.2 Plant safety parameter

1.

Description 6-1-81 No Yes No 6-1-81 Submitted I 2 display console 2,

Installed 10-1-82:

No Yes No 10-1-82

3.

Fully implemented 10-1-82 [

Yes No Yes 10-1-82 I 1 11.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 E (LL Cat B) 1

3.

Procedures 7-1-82 Yes No Yes 1-1-81 E 11.B.2 Plant shielding

1.

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

2.

Plant modifications 1-1-82 E No Yes No 1-1-81 Submitted (LL Cat B) submittal if deviation from 1

position

3.

Equipment 6-30-82 No Yes No 1-1-82 qualification I~

Clarifi-cation Item Shortened Title Descril!tion 11,B,3 Post-accident sampling

1.

Interim _system

2.

Plant modifications 11,B.4 Training for mitigating

1.

Develop training core damage program

2.

Implement program

a. Initial

b.

Complete 11,D,l Relief & safety-valve

1.

Submit program test requirements

2.

RV & SV testing (LL Cat B)

a.

Complete testing

b.

Plant-specific report

3.

Block-valve testing 11.D.3 Valve position

1.

Install direct indication indications of valve position

2.

Tech Specs 11.E.l.l Auxiliary Feedwater

1.

Short-term system evaluation

2.

Long-term

~

I i I I

I l

lmplemen-tation Schedule 1-1-80:

1-1-82 E 1-1-81 '

4-1-81 10-1-81 1-1-80 4-1-82 7-1-82 7-1-82 1-1-80 12-15-80 7-1-81 1-1,82 D-5 NRC Pre-lmple-mentation Al!l!roval No No No No No No No Yes Yes No Yee Yes Yes NRC Post-lmple-Tech mentation

Spec, Review Revision Submittal Vepco Reguired Reguired Reg, BI Remarks Yes No 1-1-80 Complete I

Yes Yes 1-1-81 Submitted 2

submittal if deviation from position Yes No 1-1-81 Complete I

Yes No None Complete 1

Yee No None Yee No 1-1-80 Complete No No 4-1-82 Yee TBD 7-1-82 Yee TBD 7-1-82 Yes Yes 1-1-80 Complete (quali-fying equipment)

No Yes 9-1-80 Submitted Yes Item Plant Submitted Specific Specific I

Yes Item Plant Submitted 1

Specific Specific

I NRC NRC Poet-Imple-Tech Clarifi-Implemeri-Pre-Imple-mentation Spec.

cation tation mentation Review Revision Submittal Vepco Item Shortened Title DeecriJ!tion Schedul~

• A22roval Reguired Reguired Reg. Bz Remarks ILE, 1. 2 Auxiliary feedwater

1.

Initiation system initiation

a.

Control grade 6-1-80 Yee No Yee 1-1-80 Complete

b.

Safety grade 7-1-81 No Yee Yee 1-1-81 Complete

2.

Flow indication

a.

Control grade 1-1-80 I

No Yes Yes 1-1-80 Complete

b.

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

c. Safety grade 7-1-80 No Yes Yee 1-1-81 Complete 11.E.3.l Emergency power for

1.

Upgrade power supply.

1-1-80 No Yee Yee 1-1-81 Complete pressurizer heaters

2.

Tech specs 12-15-80 Yee No Yee 9-1-80 Submitted 11.E.4.l Dedicated hydrogen

1.

Design -

1-1-80 Yee No No 1-1-80 Complete penetrations

2.

Install 7-1-81 No Yee No 7-1-81 Submitted 11.E.4.2 Containment isolation 1-4 Imp. diverse 1-1-130 No Yee Yes 1-1-80 Complete dependability isolation

5.

Cntmt pressure setpoint

a.

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

b.

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

6.

Cntmt purge valves 1-1-81 No Yes Yee 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 Yee No Yes 9-1-80 Submitted II.F.l Accident-monitoring

1.

Noble gas monitor 1-1-82 E.

No Yee Yee 1-1-81 Submitted I 2 Submittal if devia-tion from position D-6

. i

,I NRC I

NRC Post-lmple-Tech I

Clarifi-lmplemen...

Pre-lmple-mentation Spec.

.i

.l cation tation. :

mentation Review Revision Submittal Vepco Item Shortened Title Description Schedule' Approval Reguired Reguired Reg. B):

Remarks Iodine/particulate i

I 2 11.F.l Ac~ident-monitoring

2.

1-1-82 g No Yes Yes 1-1-81 Submitted I

(continued) sampling submittal I"

if devia-I tion from position I

3.

Containment high-1-1-82 E No Yes Yes 7-1-81 Submitted 2

range monitor submittal I

if devia-tion from position "i

4.

Containment pressure*

1-1-82 No Yes Yes 1-1-82 Submitted J

5.

Containment water 1-1-82 No Yes Yes 1-1-82 Submitted i

level I

j

6.

Containment hydrogen 1-1-82 E No Yes Yes 1-1-82 Submitted 2

I I 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

1.

Upgrade to emerg 1-1-80 No Yes Yes 1-1-80 Complete pressurizer 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 I~

• I

• I NRC NRC Post-Imple-Tech Clarifi-Implemeri-Pre-Imple-mentation Spec.

cation tation mentation Review Revision Submittal Vepco Item Shortened Title Descri2tion Schedule Ap2roval Reguired Reguired Reg. Bz Remarks II.K.2 Orders on B&W plants

13.

Thermal mechanical 1-1-82 '

No Yea As required 1-1-82 report 17, Voiding in RCS

b.

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

2 II.K.3 Final recommendations,

1.

Auto PORV isolation Not required by I II.K.3,2 1

B&O task force

2.

Report on PORV 1-1-81 No Yes No 1-1-81 Submitted t

failures J.

Reporting SV & RV 1-1-81 No Yes Yes 1-1-81 Initiate data failures & challenges beginning 4-1-80 5,

Auto trip of RCPa a,

Propose 7-1-81 E No Yes No 2-15-81 Submitted 1

modifications

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

9.

PID controller i

1-1-81 No Yea No 12-1-80 Complete

10.

Proposed anticipatory Plant

• Yes No 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 lat refuel Yes No Yes lat refuel N.A 6 mo after tech spec amend staff approval request L

17.

ECC system outages 1-1-81 No Yes As required 1-1-81 Submitted

25.

Power on pump seals

a.

Propose mods 1-1-82 No Yes No 1-1-82 I

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

I I i

i j

I l

I I l I

i I

I NRC NRC Post-Imple-Tech Clarifi-Implemen-Pre-Imple-mentation

Spec, cation tation menta:tion Review Revision Submittal Vepco Item Shortened Title Descri(!tion Schedule'.

A(!Eroval Reguired Reguired Reg

• B:z:

Remarks II.K.3 Final recommendations,

30.

SB LOCA methods I

I B&O task force a.

Schedule outline 11-15-80:

No Yee No 11-15-80 Complete

. i (continued)

b.

Model 1-1-82 Yes No No 1-1-82 C,

New analyses 1-1-83 or Yes No No 1-1-83 or 1 yr alter 1 yr afte*r 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. l. l F.mergency preparedness, Short-term improvements Complete No Yes No Complete Complete short-term IIl,A, 1,2 Upgrade emergency 1,

Interim TSC OSC & EOF Complete\\

No Yes No Complete Complete I 2 support facilities

2.

Design 6-1-81 Yes No No 6-1-81 Submitted 3,

Modifica tioils 10-1-82 No Yee Yes 10-1-82 1

111.A.2 Emergency preparedness

1.

Upgrade emergency 4-1-81 No Yes Yes 1-2-81 Submitted plane to App. E, 10 CFR 50

2.

Meteorological data 6-1-83 No Yee Yes 1-2-81 Staged implements-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 Yee No Yes 9-1-80 Submitted III.D.3.3 I1wlant radiation

1.

Provide means to Complete; No Yes No Co.iiplete Complete monitoring determine presence of radioiodine D-9

~'

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

cation tation mentation Review Revision Submittal Vepco Item Shortened Title Descril!tion Schedule AJ!l!roval Reguired Reguired Reg. BI Remarks III.D.3.3 Inplant radiation

2.

Modifications to 1-1-81 i No Yes Yes 1-1-81 Complete monitoring (continued) accurately measure 12 III.D.3.4 Control-room

l.

Review 1-1-81 No Yes No 1-1-81 Submitted I 2 habitability

2.

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

i E - Indicates those implementation and/or submittal dates to which Vepco;has taken an exception.

D-10

I.A.1.3 SHIFT MANNING Technical Specifications, Section 6.2.2, for both North Anna Units 1 and 2 require a minimum shift crew and overtime restrictions in accordance with the requirements of I.A.1.3.

Administrative procedures have been revised at Surry Units 1 and 2 to limit overtime in accordance with these requirements.

In addition, a request to change Technical Specifications was submitted on November 14, 1980, which would specify a minimum shift crew consistent with the requirements on page I. A.1.3-4.

I.A.1.3-11

I. A. 2.1 IMMEDIATE UPGRAD_ING OF REACTOR OPERATOR AND SENIOR REACTOR OPERATOR TRANING AND QUALIFICATIONS The upgraded requirements for senior reactor operator license applicants have been implemented effective December 1, 1980.

The experience of ea.ch appli-cant is checked by the station Supervisor - Nuclear Training against the criteria in this section.

The previous requirement that applicants for SRO licenses have 4 years of responsible power plant experience, of which at least 2 years shall be nuclear power plant experience (including 6 months at the specific plant) and no more than 2 years shall be academic or related technical training, has also been implemented.

Current applicants for licenses are certified to be competent by the Executive Vice President - Power.

The training _program for SRO/RO candidates includes training in heat transfer, fluid flow, thermodynamics, and plant transients.

I.A.2.1-21

I.C.l GUIDANCE FOR THE EVALUATION AND DEVELOPMENT OF PROCEDURES FOR TRANSIENTS AND ACCIDENTS On December 15, 1980, the Westinghouse Owners Group submitted a detailed description of a program to comply with the requirements of I. C.1 for both Inadequate Core Cooling and Transients and Accidents (Ref:

Letter No.

OG-47).

The submittal id~ntified 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.1 ( with a proposed schedule for completion) was also identified.

Additionally, on March 18, 1981, an update of the status of the Owners Group I. C.1 activities was submitted to the NRC (Ref:

Letter No. OG-54).

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.

The NRC review of the Westinghouse Owners Group proposed program ( OG-54) identified several staff concerns.

These concerns were outlined in letters from Mr. Robert A-. Clark (dated June 9, 1981 for North Anna) and Mr. Steven A. Varga (dated June 10, 1981 for Surry) to Mr. J. H. Ferguson. of Vepco.

In response, the Westinghouse Owners Group evaluated and revised its proce-dures development program and forwarded to the NRC a revised program on July 7, 1981 by letter from Mr. R.

• W. Jurgenson ( AEP, Owners Group Chair-man) to Mr. S. H. Hanauer of the NRC (Ref:

Letter No. OG-61).

This revised prgoram was received and considered acceptable by the NRC.

NRC acceptance of the Owners Group revised program is documented in a letter from Mr.. D, G. Eisenhut to Mr. R. W. Jurgenson dated September 18, 19Sl.

I.C.1-6 1

2

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

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

I. C.5-3

-- - I 1

/

I.D.1 CONTROL ROOM DESIGN REVIEWS Vepco has reviewed and provided comments on the draft document NUREG/

CR-1580, "Human Engineering Guide to Control Room Evaluation".

North Anna Units 1

& 2 control room has been reviewed and modified-- to reflect the human engineering recommendations of the NRC presented in the

_ Essex Corporation Report. Surry Units 1 & 2 control room has been reviewed to establish short-term improvements.

One of these improvement modifications will be made on Unit 2 during the Fall 1981 refueling outage and is scheduled for the Spring 1982 maintenance outage on Unit 1.

Other modifications will be evaluated using mock-ups on the simulator prior to implementation in the Control Room.

I.D.1-4 2

1.D.1 CONTROL ROOM DESIGN REVIEWS - EXCEPTIONS In September 1981, the NRC issued the "interim requirements" (NUREG 0700) for the control room design review.

Vepco is concerned that the present "interim requirements" may be superseded by later final requirements that would adversely affect Vepco resource commitments.

Until the NRC issues specific schedule requirements on the subject, Vepco will initiate only a minor effort at this time in response to NU REG 0700.

I. D.1-5 2

II. B.1 REACTOR COOLANT SYSTEM VENTS - EXCEPTIONS At the present time, priority is being given to the development of the North Anna Unit 1 final design to support the implementation of the vent systems during the* refueling outage scheduled to begin January 1, 1981.

The electri-cal schematics providing qualified valve cable entry seals are not yet complete.

After the North Anna Unit's electrical schematics have been finalized, emphasis will be put on the electrical schematics for Surry Unit 1.

As a consequence, we do not expect to have all the electrical schematics for all units by the required documentation date of July 1, 1981.

However, the schematics will be similar and will be provided for the individual units as soon as they become available.

Vepco will* not be able to support complete documentation of the test and operating procedures and supporting analysis as required by Position 2 and document item (3) by the required date of July 1, 1981.

This information will be developed as the RCS vent and reactor vessel level system designs are developed and the systems are installed.

The information should be available by January 1, 1982 to allow NRC review prior to implementation of the systems.

By. July 1, 1982, Vepco intends to provide a list of vent system components (valves, cable, connections, etc.) and a description of the test programs that have been or will be performed to demonstrate compliance with NRC's order of May 27, 1980.

The availability of_ this information cannot be assured at this time because the vent system components are under various stages of procure-ment.

Installation of the RCS vent system connection requires removal of the reactor vessel head.

Vepco proposes to install this portion of the system during refueling outages.

As shown in Section C, the next refueling outage for North Anna 2 and Surry 2 are currently scheduled for the first half of 1982.

Those dates would support inst~.llation of the system lJy July 1, 1982.

How-ever, any unscheduled outages could extend the refueling beyond the required installation date.

Therefore, Vepco requests a change to the required instal-lation date for the RCS Vent System to be July 1, 1982 or the first refueling outage after January 1, 1981, which ever is later

• II.B.1-10

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.

2.

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

Manual valves, located in high radiation zones, which must be operated to line up and operate the post-accident hydrogen recom-biner and hydrogen analyzer 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.

2.

Manual valves, located in high radiation zones, which must be operated to line up and operate the post-accident hydrogen analyzer will be replaced with environmentally qualified remotely operated valves, such as direct-acting solenoid valve or air-operated valves.

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.

II. B. 2-11

/

1

c.

3

• Al though 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 building sump is being studied for modification so 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.

4.

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

5.

6.

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.

System modifications to pe~init interfacing with external process systems to be 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 available at the time of the accident.

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 !iS material becomes available.

In addition, Vepco has concluded its review of the radiation effects on mechanical equipment required to mitigate a high energy line break

( HELB) for North Anna and Surry Power Stations.

This review indicates that no equipment modifications are required to mitigate the HELB tran-sient.

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 FSAR 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 purchased for Surry as part of IE Bulletin 79-0lB Review, (see exception).

II.B.2-12 I z 1

2

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.

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

(Surry)

7.

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

(North Anna Unit 1) 8.

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

(Surry)

9.

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

( Surry)

10.

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

11.

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

• and N UREG 0588.

12.

The low head safety injection and putside recirculation spray pumps will have Torlon to replace Teflon for support pads.

(North Anna and Surry)

D.

Additional Information The basis for systems excluded:

1.

The design of the Reactor Coolant System (RCS) and supporting systems is s1:1ch that the letdown portion of the Chemical and Volume Control System (CVCS) 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 11.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.

II. B. 2-13 1 2 I* l

3.

The Boron Recovery, Liquid Waste, Solid Waste, and Gaseous Waste systems were not considered in the shieldmg 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.

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 (HHS!) portions of the CVCS 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 valve.

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

The locapon 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.

II. B. 2-15

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.

b.

c.

d.

e.

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.

Technical Support Center 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.

Counting Laboratory 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.

II. B. 2-16

r

f.

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 op~rated valves.

Also, for more reliable system operation, dedicated penetrations will be provided for both of the H2 analyzers.

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.

In the case of the containment purge blowers, an additional manual isolation valve is being installed to isolate this system from the Containment Atmosphere Cleanup System.

This valve will not be remotely operated.

Since use of the purge blower system is not considered to be requir~d in the mitigation phase of an accident, this system would be used as a back-up to the redundant H2 recombiners only if the containment atmosphere were acceptable for release.

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 Attachment E to July 7, 1980 letter)

Auxiliary building and safeguards building sump drain modifications

( Design information provided in April 1, 1980 letter)

II. B.2-17 2 -

2

II.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".

2a.

In accordance with the section titled Changes to Previous Requirements j 2 and Guidance, Vepco will install remotely operated valves for the Contain-ment Atmosphere Cleanup System for North Anna Power Station during the first sufficient outage but no later than July 1, 1982.

2b.

Qualified Radiation Monitoring Pumps & Motors are ordered.

Delivery is currently scheduled for the Summer, 1982.

The replacement of the

• equipment will be done in accordance with the guidelines and schedule set forth in I.E. Bulletin 79-01B.

The installation is scheduled for the first refueling after Summer 1982.

(Surry only) 2c.

Delivery for control valves required to automatically adjust service water to the charging pump lube oil cooler has been delayed due to a labor strike since July 19, 1981, at the vendors manufacturing plant.

The vendor has promised valve delivery in November, 1981 and should the vendor deliver the valves in November, Vepco will be able to schedule to install the valves by no later than July 1, 1982.

( Surry only) 2d.

Due to material delivery problems, the modifications :Scheduled for the hydrogen Recombiner Vault and Service Water Radiation Monitor (shield-ing) will be delayed beyond January 1, 1982.

Vepco will have these modifications completed by April 1, 1982.

(North Anna only).

2e.

Due to manufacturing problems at the vendor, the modifications to replace the teflon material on the Low Head Safety Injection & Recirculation Spray Pump Restraints will be delayed beyond January 1, 1982.

However, Vepco will make the material replacement by April 1, 1982.

(North Anna only)

3.

The radiation zone maps for personnel access will be revised after all the post-TM! modifications are complete 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. 2-18 2

II.B.3 POST-ACCIDENT SAMPLING SYSTEM - EXCEPTIONS A.

The exceptions for the containment atmospheric sample system are as follows:

B.

1.

North Anna Unit 1 shall have a temporary sample return line to the existing Hydrogen Analyzer lines.

The temporary return line will be used until the permanent Hydrogen Analyzer returns are installed by 7-1-82.

The Containment Atmospheric Sampling System supply lines for North Anna Unit 1 and Surry Unit 2 will be heat traced after 1-1-82.

2.

North Anna Unit 2 and Surry Unit 1 will utilize the Short-Term Post-Accident Sampling System connected to existing Hydrogen Analyzer tubing.

The temporarily installed Short-Term Containment Atmospheric Sampling System supply tubing will not be heat traced.

The permanent supply and return lines to the Containment Atmos-pheric Sampling Panel will be installed at the respective first scheduled unit outages after 1-1-82.

3.

QA Category I control switches for the containment isolation valves for Containment Atmospheric Sampling System are not available prior to 1-1-82.

QA Category II control switches will temporarily be installed until the QA Category I control switches are available.

The Exceptions for the Reactor Coolant Sampling System:

The Sentry Equipment Company Chemical Analysis Panels ( CAP) and Chemical Analysis Monitor Panels (CAMP) were originally scheduled for delivery in December, 1980.

Vendor delays have resulted in delivery in late October, 1981 to North Anna. and Surry Power Stations.

Because of this late delivery, the installation of mechanical and electrical tie-ins, as well as system testing, must be accomplished in an extremely short time frame.

However, Reactor Coolant Sampling Syste,m installation will be completed by 1-1-82 except system testing which will be expedited.

Vepco will have the Reactor Coolant Sampling System fully operable by 7-1-82.

During that period, the Short-Term Sampling System will be utilized.

II.B.3.l-A19 2

2

11.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.

3.

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

Personnel iden.tified by the Emergency Plan qualified to become 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.

The program for managers and technicians in the Instrumentation and Control (I&C), health physics and chemistry departments has been imple-mented in. conjunction with the Station Emergency Plan training on EPIP's specific to each technical area and will be complete prior to October 1, 1981 at both the Surry and North Anna: Power Stations.

)

II. B.4-3 1

II.D.1 PERF'.ORMANCE TESTING OF POWER RELIEF VALVE AND SAFETY VALVES As a sponsor of the EPRI PWR Safety and Relief Valve Test Program, Vepco intends to comply with the requirements of NU REG. 0578, Item 2.1. 2.

By letter dated December 15, 1980, R. C. Youngdahl of Consumers Power Company has provided the current PWR*Utilities' positions of NUREG 0737, Item II.D.1 clarifications.

Briefly, those positions are:

A.

Safe.ty and Relief Valves and Piping - the EPRI "Program Plan for Per-formance Testing of PWR Safety and Relief Valves", Revision 1, dated July 1, 1980, does provide a program that satisifes the NRC requirements.

Discussion with the NRC staff and their consultants are resolving specific detailed issues.

B.

Block Valves - The EPRI Program has not formally included the testing of block valves.

However, a small number of block valves have been tested at the Marshall Steam Station Test Facility.

By letter dated July 24, 1981, Mr. R. C. Youndahl of Consumers Power Company has provided the current PWR utilities position on Block Valve testing.

C.

A TWS Testing - PWR Utilities will not support additional efforts for A TWS valve testing until regulatory issues are resolved.

The major safety and relief valve test facility ( CE) is nearing completion and some measures were taken to provide additional test capability beyond the current program requirements.

The NRC should recognize that results from the current program are likely to provide most of the information necessary to address ATWS events (i.e.*, relief capability at high pressure).

II.D.l-4.

2

11.F.1.

ATTACHMENT 1, NOBLE GAS EFFLUENT MONITOR - EXCEPTIONS

1.

Final design details for the process vent and ventilation vent (noble gas) are available.

2.

In order to comply with 1-1-82 implementation date, an order was placed with Kaman Sciences Corporation on 11-14-80 which requires delivery date of 7-20-81.

The delivery date for the monitor units was scheduled for 8-21-81 for isokinetic nozzles and hardware with the monitor skids scheduled for the week of 9-30-81.

Rework of the microprocessor units by the vendor resulted in schedule slip to October 15, 1981.

When undergoing final system test, a circuit*

board failure resulted in further shipping delays to October 13, 1981.

The Kaman Sciences.Increased Range Radiation Monitors Skids were shipped on October 28 and 29, 1981, respectively, for North Anna and Surry Stations.

This delay on the delivery of the Kaman system will severely impact on current construction schedules and will result in installation beyond 1-1-82 implementation date.

Vepco will complete the installation of the Increased Range Radiation Monitors skids by July 1, 1982 for North Anna and April 1, 1982 for Surry.

In the interim, the Short-Term Increased Range Radiation Monitors will be utilized.

II. F.1-Al-14 2

11.F.1.

ATTACHMENT 2, SAMPLING AND ANALYSIS OF PLANT EFFLUENTS -

EXCEPTIONS

1.

Final design details for the sampling and analysis of plant effluent (radio-active iodines and particulates) are available.

2.

In order to comply with 1-1-82 implementation date, an order was placed with Kaman Sciences Corporation on 11-14-80 which requires delivery date of 7-20-81.

The delivery date for the monitor units was scheduled for 8-21-81 for isokimetic nozzles and hardware with the monitor skids scheduled for the week of 9-30-81.

The rework of the microprocessor units resulted in schedule slip to 1 10-15-81.

When undergoing final system test, a circuit board failure resulted in further shipping delays from 10-5-81 to 10-9-81 then 10-13-81.

The Kaman Sciences Increased Range Radiation Monitors Skids were shipped on October 28 and 29, 1981, respectively, for North Anna and Surry Stations.

This delay on the delivery of the Kaman system will severely impact on current construction schedules and will result in installation beyond 1-1-82 implementation date.

Vepco will complete the installation of the Increased Range Radiation Monitors skids by July_ 1, 1982 for North Anna and April 1, 1982 for Surry.

In the interim, the Short-Term Increased Range Radiation Monitors will be utilized.

II.F.1-A2-6 2

e I I

• F.1,.ATrA.Cll\\1ENI' 3, CXN.rAIN.\\'IENI' HIClI RANGE RADIATICN l\\'[NITCR EXCEPTICNS Clarification of i terns 5 and 7 of Appendix B to NUREG-0737 are presented llllder iter!lS la and lb.

The vendor (Victoreen) supplied control room panels which house the recorders, and internal wiring for the high range radiation system which failed the seismic test to the extent that the panels required corrplete redesign.

The system will be corrpleted on an expedited basis after arrival of the panels.

The panels are now scheduled to be shipped by mid-Decerri>er, 1981.

The system will be made operational after installation of the panels but no later than July 1, 1982.

All in-containment terminations will be made by January 1, 1982 using a terrporary termination procedure. Qualified terminations will be made in accordance with the schedule of IE Bulletin 79-0lB and NUREG-0588.

I I. F.1-A3-6

~... -..,._,................ ~..

~

I 2 2

II.F.l, ATTACHMENT 6., C.ONTAINMENT HYDROGEN MONITOR - EXCEPTIONS Surry Units 1 & 2 The hydrogen analyzers for Units 1 & 2 will be operational by January 1, 1982.

However, the extent of heat tracing and insulation and associated seismic design of the sample line system has not been resolved.

Resolution of this problem and the modifications required to the sample input line are being worked on an expediteq basis; however, final heat tracing of the sample input lines will not be

• completed by January 1, 1982.

The existing redundant hydrogen analyzers will remain operational in the interim.

Installation of Category I heat tracing is dependent on material delivery.

Heat tracing will be complete by 8-1-82.

North Anna 1 & 2 The 'hydrogen analyzers will be installed and operational by January 1, 1982.

However, the new sample and return lines to the containment will not be complete by January 1, 1982.

The sample and return line installation will be complete by startup from the Spring 1982 refueling outages for Units 1 and 2.

Installation of Category I heat tracing is dependent on material delivery.

Heat tracing will be complete by 8-1-82.

The existing redundant (nonqualified) hydrogen analyzers will remain operational in the interim.

II. F. 1-A 6-4 2

2

11.F.2 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING -

EXCEPTIONS NUREG 0737 requires the implementation of the reactor vessel level systems by January 1, 1982 and the documentation associated with these systems by Janu-ary 1, 1981.

Design information and supporting analyses of the reactor vessel level system has been submitted.

Vepco requests a change to the required installation date of January 1, 1982 for the Reactor Vessel Level System as follows:

Surry Unit 1 and 2 All in containment work with the exception of filling and* venting the system will be complete by January 1, 1982.

Unit 1 will be operational after the system is filled and vented during the next refueling scheduled for 11-19-82.

Unit 2 is now scheduled to be operational after the system is filled and vented *during the Spring maintenance outage.

The system will be fully operational when the reactor vessel vent system is operational July 1982.

North Anna Unit 1 and 2 System will be completed during the spring refuelings currently scheduled to be as follows :

Unit 1 21-82 Unit 2 5-82 System will be functional by July 1982 when the reactor vessel vent system is functional.

11.F.2-27 2

II. K.2.13 THERMAL MECHANICAL REPORT--EFFECT OF HIGH-PRESSURE INJECTION ON VESSEL INTEGRITY FOR SMALL-BREAK LOSS-OF COOLANT ACCIDENT WITH NO AUXILIARY FEEDWATER To completely address the NRC requirements of detailed analysis of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks, with an extended loss of all feed water, a program will be completed and documented to the NRC by January 1, 1982.

This program will consider generic Westinghouse PWR groupings.

Plant specific analyses, if required, will be provided on a schedule to be determined following additional discus-sions with the NRC.

II. K. 2.13-3

II.K.2.17 POTENTIAL FOR VOIDING IN THE REACTOR COOLANT SYSTEM DURING TRANSIENTS The Westinghouse Owners Group is currently addressing the potential for void formation in the Reactor Coolant System (RCS) during natural circulation cooldown conditions, as described in Westinghouse letter NS-THA-2298 (T. M. Anderson, W to P. S. Check, NRC).

We believe the results of this effort will fully address the NRC requirement for analysis to determine the potential for voiding in the RCS during antici-pated transients.

A report describing the results of this effort will be provided to" the NRC before January 1, 1982

• II. K.2.17-2

II.K.2.19 SEQUENTIAL AUXILIARY FEEDWATER FLOW ANALYSIS The transient analysis code, LOFTRAN, and the present small break evalua-tions analysis code, WFLASH, have both undergone benchmarking against plant information or experimental test facilities.

These codes under appropriate conditions have also been compared with each other.

The Westinghouse Owners Group will provide, on a schedule consistent with the January 1, 1982 requirement, a report adddressing the benchmarking of these codes.

II. K. 2.19-2

II.K.3.5 AUTOMATIC TRIP OF REACTOR COOLANT PUMPS DURING LOSS-OF-

. C_OOLANT ACCIDENT The Westinghouse Owners Group has submitted. LOFT test predictions in accor-dance with the schedule provided by II.K.3.5.

Currently, additional 'informa-tion, in response to NRC questions on the LOFT predictions, is being prepared and will be submitted in June, 1981.

NRC approval of Westinghouse models for the RCP trip ~valuations has not yet been received.

Therefore, Vepco's response regarding the need for automatic RCP trips during a LOCA will be provided within three (3) months of approval of the Westinghouse models.

This is consistent with the timetable of Item II.K.3.5 but not the July 1, 1981 submittal date for design information of an automatic RCP trip system.

II. K. 3. 5-3 1

II.K.3.25 EFFECT OF LOSS OF ALTERNATING-CURRENT POWER ON PUMP SEALS An evaluation of the potential for the failure of Reactor Coolant Pump seals will be conducted, and the results submitted for review by January 1, 1982.

If the evaluation identifies the need for design or procedure changes, the details of such actions will also be provided.

II.K.3.25-3

II. K. 3. 30 REVISED SMALL-BREAK LOSS-OF-COOLANT-ACCIDENT METHODS TO SHOW COMPLIANCE WITH 10CFR PART 50, APPENDIX K Westinghouse Electric Corporation has supplied a letter to the NRC dated September 26, 1980, detailing the schedule to comply with this item.

Westing-house has indicated that there should be no difficulty in submitting the Topical Report on their small break LOCA codes by the January 1, 1982 deadline

• II. K. 3.30-4

III.A.1.2 _ UPGRADE EMERGENCY SUPPORT _FACILITIES The requirements of the *Emergency Response Facilities (ERF's) have been detailed in the final version of NU REG 0696.

The schedule for ERF implemen-tation given in. Item III.A.1.2 (and NUREG 0696) will be met.

Preliminary design information (based on the draft of NUREG 0696) regarding the proposed Vepco ER F's was submitted to the NRC on December 18, 1980.

An Update of this info-rmation was provided on June 1, 1981..

Further updates will be made when infoTmation becomes available.

III. A. 1. 2-8 l 1 I 2 I 1 I 2