ML20062N641

Forwards Detailed Relief Request from Requirements of Section XI of ASME Code for Unit 2,per Commitment Made During 940106 Telcon

ML20062N641
Person / Time
Site:	North Anna
Issue date:	01/07/1994
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
94-008, 94-8, NUDOCS 9401180108
Download: ML20062N641 (19)

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VIRGINIA ELucTHIC AND POWER COMI%NY Ricnwoxu, Vino NIA 200 61

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January 7,1994 l

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U.S. Nuclear Regulatory Commission Serial No.94-008

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Attention: Document Control Desk NL&P/JDH R1 i

Washington, D.C. 20555 Docket Nos.

50-339 3

License Nos. NPF-7

!j Gentlemen:

i VIRGINIA ELECTRIC AND - POWER COMPANY NORTH ANNA POWER STATION UNIT 2 ASME SECTION XI RELIEF REQUEST-On' January.6,1994 at approximately 2300 brs., a conference call was held between NRC and Virginia Electric and Power Company to discuss proposed relief from the.

requiremeBis of Section XI of the ASME Code for North Anna Unit 2. The proposed relief involved alternative examination requirements from those prescribed in Section iWA-5250(a)(2) for various bolted connections. The detailed relief request is attached.

It is requested that NRC approve the requested relief no later than 1000' hrs. on-1

' January 7,1994. That time is based on a determination that a required Code examination conducted pursuant to North Anna Unit 2 Technical Specification (TS)

Surveillance Requirement 4.0.5.a had not been performed as required.

The y

determination was made at 1000 hrs. on January 6,- 1994.

North Anna. TS l'

Surveillance Requirement 4.0.3 provides for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period (i.e., until 1000 hrs. on January 7,1994) whenever it is determined that a surveillance requirement has not

-i been performed.-

During the conference call on January 6,1994, we committed to additional' compensatory measures which are documented in the attached relief request.- This l

letter also transmits two supporting documents: a Justification for Continued Operation and a Safety Evaluation. (Please note that the additional actions committed to during j

the conference call and documented in the relief request will also be incorporated in revisions of the JCO and SE).

130032 94o1180108 94o107 4*

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'DR ADOCK.0500o339 PDR

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The relief request, Justification for Continued Operation, and Safety Evaluation have.

been approved by the Station Nuclear Safety and Operating Committee.

If you have any further questions, please contact us.

j Very truly yours, LN a

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W L. Stewart jg Senior Vice President - Nuclear Attachments cc:

U.S. Nuclear Regulatory Commission l

Region ll 101 Marietta Street, N.W.

t Suite 2900 Atlanta, Georgia 30323 Mr. R. D. McWhorter NRC Senior Resident inspector North Anna Power Station

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RELIEF REQUEST I.

IDENTIFICATION OF COMPONENTS i

SYSTEMS: Reactor Coolant (RC), Chemical Volume &. Control (CH),

Safety Injection (SI), and Residual Heat Removal (RH)

Component #

1. Bolts Bolt Size Bolt Material Code Class 2-RH-25 16 7/8" A193 B16 2

2-CH-RV-2382A 8

5/8" B16 2

2-RH-1 20 1 1/8" A193 B7 2

2-RC-MOV-2535 12 3/4" A193 B7 1

2-RC-HCV-2556B 8

7/8" A193 B7 1

2-SI-TV-2884C 8

3/4" A193 B16 3

2-CH-TK-2 16 1 1/8" A193 B7 2

P II.

IMPRACTICAL CODE REQUIREMENTS The 1986 edition of ASME Section XI requires in IWA-5250(a) (2) that if leakage occurs at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100.

t III. BASIS FOR RELIEF North Anna Unit 2

is currently conducting 100%

power operations.

During the documentation review for.. the Code required NIS-1 report for the recent Unit 2 refueling outage, 4

it was discovered that several bolted connections had experienced evidence of leakage during Code pressure testing.

The Code requirement to remove the' bolting and perform the visual (VT-3) examination was not performef to the extent required by the Code at the above identified locations.

A justification. for continued operation (J.C.O.)

has been completed (attached) for the identified locations. The J. C. O.

describes the extent of examinations performed, and the operability of the component at each location.-

Performing the Code requirements at this time would be impossible ( subatmospheric containment, high pressure and.

temperature systems) to perform without unit shutdown.-

Alternatively, requiring unit shutdown to complete this requirement would be impractical considering the proposed alternative.

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ALTERNATE REQUIREMENTS The bolting identified shall be removed and visually (VT-3) examined at the next suitable outage.

A TV camera will be installed to monitor the VCT (2-CH-TK-2) manway for leakage.

Additionally, a calculation will be performed to determine the.

minimum bolting required to maintain the structural integrity of the VCT manway.

The reactor coolant leakage will be monitored daily, which is accelerated from that required by the Technical Specifications (once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />).

If the reactor coolant leak. rate exceeds 0.4

gpm, then an investigation to ' determine the source of leakage will be initiated.

Any leakage found during the' interim at the identified locations will be evaluated considering the potential for bolting degradation, and appropriate actions will be taken.

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Justification For Continued Operation "N

POW 06 PAP-3001

1. JCO Number

2. Revision

3. Appleable Station

4. Appleable Unit

@ North Anna Power Station Unit 1

@ Unit 2 94-01 0

OSurry Power Station Common

5. Original Expiration Date

6. Revised Expiraten Date

.j July 6,1994 l.

7. JCO Title l

DEVIATING CONDITION FOR BOLTING EXAMINATIONS l

8. Descripton of Abnormal Condition System walk downs are performed in accordance with Periodic Test 2-PT-48, Visual Inspection of ASME XI Class 1 Pressure Boundary Components and 2-PT-48.1, Visual Inspection of ASME XI Class 2 and 3 Pressure Boundary Components in order to satisfy the requirements of ASME XI,IWA 5241 and 5242. During these walkdowns,the presence of boric acid at tx)lted connections indicates the necessity,in accordance with ASME XI,IWA-5250 to: 1) replace the bolting or 2) remove the bolting, perform a VT-3 inspection of all bolting in the connection and evaluate in accordance with IWA-3100. In several cases, there i, insufficient documentation to confirm either bolting replacement or VT-3 examinations of bolting left in place.

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9. Required Compensatory Actons Perform a daily RCS leakrate periodic test.

If RCS unidentified leakage exceeds 0.4 gpm, then an investigation to determine the source of the leakage will be initiated.

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10. Action Plan i

1. Submit and obtain NRC approval of a relief request for not performing bolting examinations for the deviated bolting conditions until the1995 Unit 2 refueling outage.

2. Maintenance is to perform VT-3 examinations of the deviated bolting conditions during the 1995 Unit 2 refueling outage.

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11. Name of Responsible Superintendent Assigned

12. Preparers Name (Pnnt)

13. Title Jay Leberstien StaffEngineer

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15. Date

14. fepare

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16. Fjugwers Name (Pnnt)

17. Title l

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Snow Supervisor ISI/NDE/ Engr. Programs l

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18. Reviewer Sgnature

19. Date 20.SN C~ Chairman's Sgnatur

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21. Datp&j#4 I

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i Form No. 73093Dpune 90) 1

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Technical Evaluation - Justification for i

Gm ContinuedO eration P

I POW 06 4 PAP-3001 3

of 2

Page 1, JCO Number

2. Revision

3. JCO Title I

94-01 0

DEVIATING CONDITION FOR BOLTING EXAMINATIONS

4. Provide a discussion of the background and related expenences, a discussion of signifcant safety and regulatory issues, and an explanation how the mitigating actions change the safety evaluation / review of the abnormal situation from an unacceptable to an acx:eptable situation.

DISCUSSION System walk downs are performed in accordance with Periodic Test 2-PT-48, Visual Inspection of ASME XI Class 1 Pressure Boundary Components and 2-PT-18.1, Visual Inspection of ASME XI Class 2 and 3 Pressure Boundary Components in order to satisfy the requirements of ASME XI,IWA 5241 and 5242. During tl ese walkdowns, the presence of boric acid at bolted connections indicates the necessity,in accordance with ASME XI,IWA-5250 to: 1) replace the bolting or 2) remove the bolting, perform a VT-3 inspection of all bolting in the connection and evaluate in accordance with IWA-3100. In several cases, there is insufficient documentation to confirm either bolting replacement or VT-3 examinations of bolting left in place. These cases can be grouped in the following categories:

Group 1: 2-RH-25,2-CH-RV-2382A and 2-RH-1 Hese valves had their studs removed and inspected by mechanics who were not currently qualified VT-3 examiners. Two of these valves were inspected by mechanics that previously N1d VT-3 qualifications. The third valve had its studs inspected by a qualified mechanic (not VT-3 qualified). VT-2 examinations were performed on the respective systems afterwards.

Group 2: 2-RC-MOV-2535,2-RC-IICV-2556B and 2-SI-TV-2884C Dese valves had their studs inspected in place by a VT-3 qualified examiner. VT-2 exarninations were performed on the respective systems afterwards.

Group 3: 2-CH-TK-2 Manway A VT-2 examination was performed and there was no evidence of active leakage. Leakage from this system is monitored.

SIGNIFICANT SAFETY AND REGULATORY ISSUES There are no significant safety issues involved with this deviating condition. Each component that had identified leakage was repaired. Also, each component identified above was VT-2 examined to ensure there was no continued leakage. In addition, all Classi systems, were tested during the RCS hydro that was performed following the Unit 2 RTD manifold work. Also, the at pressure walkdown at the end of the outage did not indicate any leakage concerns in the repaired areas, ne total RCS leakrate has been very low.

He deviating condition requires the submittal of a relief request from the requirements of ASME XI. IWA-5250. De relief request is needed for the remainder of the Unit 2 operating cycle until the bolting examinations can be performed.

S. Preparer's Name (Pnnt)

6. Signature

7. Date Jay Lelersben Attach this Evaluation to the Justification For Continued Operation (Form Number 730930)

Form No.730933(June 90)

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Technical Evaluation - Justification for 4

Continued O eration P

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va, POW 06 VPAP 3001 Page of 2

2

1. JCO Number

2. Revision

3. JCO Title 94-01 0

DEVIATING CONDITION FOR BOLTING EXAMINATIONS

4. Provde a discussion of the background and related expenences, a discussion of sgnrfcant safety and regulatory issues, and an explanation how the mitigating actions r.fiange the safety evaluation / review of the abnormal situation from an unacceptable to an acx:eptable siNation.

REQUIRED COMPENSATORY ACTIONS Perform a daily RCS leakrate periodic test.

If RCS unidentified leakage exceeds 0.4 gpm, then an investigadon to determine the source of the leakage will be initiated.

ACTION PLAN

1. Submit and obtain NRC approval of a relief request for not performing bolting examinadons for the deviated bolting conditions until the 1995 Unit 2 refueling outage.

2. Maintenance is to perform VT-3 examinations of the deviated bolting conditions during the 1995 Unit 2 refueling outage.

DISCUSSION OF MITIGATING ACTIONS By implementing the required compensatory actions, increased leakage will be identified by the licensed operators and quickly dispositioned. The heightened awareness of the operators regarding leakage will ensure compliance with the Technical Specification allowable leakage limits. Therefore the potential effects will be bounded by the associated accident analysis described in the UFS AR.

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5. Preparer's Name (Print)

6. Signatur 1

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7. Date

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34) Iller 11Jen

/g f Attach this Evaluation to the Justification $r htinued Operation (Form Number 730930)

Form No.730923(June 90) i

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Safety Evaluation Page 1 of 12 Vinowinmm coy og VPAP-3001

1. Safety Evaluaton Number

2. Appbcable Station

3. Applicable unrt j

@ N rth Anna Power Station O Unit 1 Unit 2 1

94-SE-JCO-Surry Power Station O unit i uni: 2 Part A-Resolution Summary Report l

4. List the governing documents for whch this safety evaluaton was performed.

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DR N-94-16 S. Summanze the change, test, or expenment evaluated.

System walkdowns are performed in accordance with PT-48 and FT-48.1, which are assigned to Maintenance Engineering. During these walkdowns, the presence of boric acid at bolted connections indicates the necessity to either replace the bolting, or remove the i

bolting components and perform a VT-3 inspection of all bolting components in the connection, and evaluate for acceptability.

6. State the purpose for this change, test, or experiment.

In several cases for the Unit 2 outage, there is insufficient documentation to confirm either bolting replacement or VT-3 examinations of bolting material left in place.

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7. List the hmiting cond:tions and >pecial requirements identified or assumed by this safety analysis.

j Perform a daily RCS Leakrate F"T.

• If RCS unidentified leakage exceeds 0.4 gpm, then an investigation to determine the source of the leakage will be initiated.

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8. Will the proposed activity /condaion result in or constitute an unreviewed safety question or Yes @ No require a hcensing amendment?

9. Preparer Name (Print) 10.

reparer ature _ /

11.Date / /

Jay 1xbersuert

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12. Cognizant Supervisor Name (Pnnt) 13.

oce:zro - ennsor Signature

14. Date f'

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c T. Snow

15. Dispositon

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% Approved O Disapproved C Approved as Modified Requires Further Evaluaton

16. SNSOC Chairman Signaturh>

17.kate/94 l

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Comments 1

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j Note Attach a Copy o art A. Resoluton su Send A Copy of rt A to tecensing for Send a copy of t

,l cornpleted S#oty F Send the compk d Safety Evawon Use "Spiety Eva,,: ton. Supplemental hi i

Key: MSRC-Management Safety Rev6ew Committee Form ho. 7309'6<w= 93) l

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Safety Evaluation Page 2 of 12 GOV 02 VPAP-3061 Part A - Resolution Summary Report

18. Summarize from Part D, Unreviewed Safety Queston Determination, the major issues considered; state the reason the change, test, or experiment should be ahowed; and state why an unreviewed safe.y question daes or does not exist (a simple conclusion statement is insuffeient).

System walkdowns are performed in accordance with I'T-48 and PT-48.1, which are assigned to Maintenance Engineering. During these walkdowns, the presence of boric acid at bolted connections indicates the necessity to either replace the bolting, or remove the toldng components and perform a VT-3 inspection of all bolting components in the connection, and evaluate for acceptability. In several cases for the Unit 2 outage, there is insufficient documentation to confirm either bolting replacement or VT-3 examinations of bolting material left in place.

Failure of the fasteners due to boric acid degradation would result in increased leakage, but woulo respond as a leak-before-bmak -

mechanism in the same fashion as other piping joints in the plant. No catastrophic failure will occur. Failure would occur slowly and would be indicated by increased operational leakage which would be observed by an increase in the sump pumping frequency, an increase in the vent stack radiation levels, and/or an increase in containment radiation Icvels. Any VCT leakage would be identified during performance of the RCS Leakrate PT. Although the level of radioactivity in the Containment would increase, the increase would be small and would be contained. Operator identification of an increase in the leakrate would occur quickly and would prevent Unit etwration outside of the limits allowed in Tech Specs. For components located in the Auxiliary Building, leakage would result in increased Aux Building doses which would be picked up by installed radiadon detectors. This would prevent operation outside of analyzed limits for dose and exposure.

Studs were inspected in place as a minimum except on the VCT manway where a VT-2 inspection indicated no evidence of active leakage. No defects were identified. Class I systems were tested during the RCS hydro following the RTD manifold work. The at-pressure walkdown at the end of the outage did not indicate any leakage concems in the repaired areas. Operational leakage is currently lo'. Any fastener failure is expected to develop as a leak-before-break failure and will be detected prior to exceeding Tech Sper imits for operational leakage. Operational leakage will therefore not exceed limits used as a precursor event for accident analyses. For these reasons, the probability of occurence of accidents or rnalfunctions of equipment previously analyzed will not increase.

Failure of the fasteners will not cause a catastrophic failure of any equipment required for the mitigation of analyzed accidents.

Increased leakage at bolted connections may occur, but will not create a situation which is not tuunded by existing accident analyses. For these reasons, the consequences of accidents or malfunctions of equipment previously identified will not increase.

No fastener failures are expected since the inst'ections which were performed were conducted by knowledgeable craft and no defects were found. Leakage which may occur is bounded by existing analyses. Therefore, the margin of safety as mpresented in the Tech Specs is not reduced.

For these reasons, an Unreviewed Safety Question does not exist.

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f orm No. 730916Nar 93)

Safety Evaluation Page 3 of 12 GOV 02 VPAP-3QO1 Part B Applicable References.

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19. Identify aopicable Updated Final Safety Anaysis Report (UFSAR) sections.

5.2.5 Reactor Coolant System - Inservice Inspection.

20. Identify appbcable Technical Specihcatons sections.

3.4.10.1 4.0.3 l

4.0.5 r

21. Identify any other references used in this review.

2-PT-48 2-PT-48.1 DR N-94-16 Part C-items Considered By This Safety Evaluation ?

1 m.M happformnt 0 P.gs d Adenr' o No m r T4.oded f or F aptm ovv i k. ~%i'+w I. A e m L;g +wd f ' O.

orm No 730928 Doate Aste a : tw Ar sworod ~Yee Reqa* D.oy Ano 'y App ac e pi. 4 7 ' F qw 12 of '?:

22. Will the operation of any system or component as described in the Safety Anatysis Report be aftered?

Yes @ No This includes abandonment of equ pment or extended periods of equipment out of service.

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Explain.

No changes to the operation of systems or components ir expecte<*. No excessive leakage is currently identified. Increased leakage.

due to fastcaer failure is not expected to occur catastrophicany, but rather in the same fashion as any other compnent which is failing '

I due to reasons other than boric acid degradation.

23. Will the activity after the performance characteristics of any safety related system or component?

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i Action statements, jumpers, and temporary modifications should be reviewed.

Explain.

I No changes in the perfonuance of any of the components in the DR have been identified. No changes are expected due to leakage.

Should failure of the fasteners begin to occur,it would be identified as increased operational leakage and would be subject to the limits prescribed in Tech Specs. No catastrophic failures are expected.

24. Will the abitty of operators to control or monitor the plant be reduced in any way? Explain.

Yes @ No l

The discrepancies involve bohing material caly. No control or monitoring equipment is involved.

25. is a temporary modif caton involved? [ Commitment 3.2.15]

yes no Are testing requirements as stated for the temporary modifcation adequate to ensure operability after

] yes O No installation, as well as after removal of the temporary modification? Explain.

@ N/A No temporary modifications are involved.

F orrn luo. 730916(Mar 93)

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Safety Evaluation Page 4 of 12 GOV 02 VPAP-3001

' Part C -items Considered By This Safety Evaluation (Continued) '

26. Could the propnsed actrvrty affect reactivity? Reactivity is affected by such items as: RCS temperature, Yes @ No dilution, or flow; boric acid concentrations or volumes; RWST or accumulator boron concentration; main steam flow or instruments that measure main steam flow; main steam pressure; nuclear instrumentation; calorimetric power monitoring; rod control system: fuel, and fuel components. Explain. The Reactor Engineer must approve the explanation for *Yes" answers. (Commitments 3.2.9 & 3.2.14]

No changes are made to any equipment or systems which can affect reactivity. No catastrophic failures are expected which would insert positive or negative reactivity. Fastener failure would be similar to increased operational !eakage and would be observed and controlled by Tech Spec surveillance.

26A. Reactor Engineer Signature 268. Datn 2L Wdi the activity significantly increase the potential for personnel injury or equipment damage?

O Yes @ No No operation is expected which would cause failures other than a leak-before-break incident. Each of these connections is a bolted connection and is subject to this kind of failure regardbss of the quality of the fastener.

28. Will the activity create or increase the levals of radiation or airborne radioactivity?

Yes @ No Will that change result in a significant unreviewed environmentalimpact, a significant increase in C Yes No occupational exposure, or significant change to the dose to operators performing tasks outside the fittered air boundary during a design basis accident (GDC-19)? Explain. The superintendent-Radiological Protection must approve the explanation for "Yes" answers.

No fastener degradaticm is cunently indacated since the operationalleakage is very low (for components subject to RCS leakage). FaJure of the fasteners would occur slowly and would be indacated by increased operationalleakage which would be observed by an increne in the sump pumpmg frequency, increased Auxihary Building exhaust radiation monitor activity and/or an increase in contamment radiation levels. Any VCTleakage would be identtfied during performance of the RCS trakrate PT. Although the level of radmactivity in the Containment would increase.the increase would be small and would be contained. Operator idenufication of an increase in the leakrate would occur quickly and would prevent Unit operation outside of the hmits allowed in Tech Specs. trakage of any component in the Auxihary Building would result in increased Aux Building doses which would be picked up by insta!1ed radiation detectors. This would prevent operation outside of ana! ped hnuts for dose and exposure.

28A Supenntendent Radiological Protection 288. Date

29. Could the activity change or decrease the ettoctiveness of the emergency plan?

Yes @ No Expla;n. The coordinator-Emergency Preparedness must approve the explanation for "Yes" answen No changes are proposed to systems required for communications, notification, or dose assessment i

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29A. Emergency Preparedness Coordinator Signature 29B. Date

30. Will the consequences of failure for this activr y affect tne abihty of systems or components to Yes @ No perform safety functions? Describe the modes and consequences of failure considered during this evaluatiort Failure of the fasteners due to boric acid degradation would result in increased leakage, but would respond as a leak-before break mechanism in the same fashion as other piping joints in the plant. No catastrophic failure is expected to occur. Increased i

operational leakage would alert the operator to fastener failure and Unit operation would be bounded by existing Tech Specs.

Key: RCS-Reactor Coolant System; RWST-Refueling Water Storage Tank F a" N' 73o6 M*' 83) c mr-

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Safety Evaluation Page 5 of 12 GOV 02 VPAP-3001 Part C 4 Items Considered By This Safety Evaluation (Continued):

31 A. Will tne activny cause equipment to be exposed (or potentialty exposed) to adverse condaions, O..Yes No including those created by temperature, pressure, humidity, radiation or meteorological conditions?

[ Commitment 3.2.11]

318. If "Yes", could these conditions lead to equipment failure or a dangerous atmosphere? Explain.

Yes @ No Failure to inspect or replace the fasteners identified in the DR will have no effect of the installed equipment which would be different than having random failure of good fasteners. Equipment in the vicinity of these suspect fasteners is qualified for expected environmental conditions, i

32. Could f ailure of the activity feed back into protective circuitry? Explain.

i Yes @ No Failure to perform the VT-3 inspections or fastener replacement has no interaction with protective circuitry,

33. Could f ailure of the activity feed back into control circuits important to stable plant operation Yes No (e g., feedwater control, control rods)? [ Commitment 3.2.12]

Failure to perform the VT-3 inspections or fastener replacement has no interaction with control circuitry.

34. Could the activity affect emergency diesel generator sequencing logics (including testing Yes No logics), or other logics important to safety? [ Commitment 3.2.8)

Failure to perform the VT-3 inspections or fastener replacement has no interaction with EDG sequencing logics.

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35. Could the activity cause a loss of separation of instrument channels / trains or electrical Yes No power supplies? Explain.

No physical independence issues are involved.

36. Will the activ:ty involve the addition or deletion of any loads on the Class 1E electrical Yes No distribution system? Explain.

No electrical load changes are proposed, i

f DPrn No.730916(Mar 93)

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Safety Evaluation Page 6 of 12 GOV 02 VPAP-3001 y

Part C - Items Considered By This Safety Evaluation (Continued)

37. Will the activity adversely affect the abikty of a system or component to maintain its Yes @ No integrrty or code requirements? Explain.

Failure to perform the VT-3 inspection or replacement of fasteners which had boric acid buildup will not adversely affect the integrity of the fasteners. Le maintenance engineer, during his post maintenance inspection of work performed, verifies that there are no obvious external deviations. Adequate inspections were performed during the maiatenance evolution to insure com;,anent integrity. De at. pressure walkdown at the end of the outage did not indidate any leakage in the repaired areas. The Unit is currently operating with no significant leakage.

38. Will the activity reconfigure, ehminate, or add components and/or piping to the single or two-phase Yes @ No erosion / corrosion piping inspection program? Expla,in.

No changes to the crosion/ corrosion program are proposed.

39. Will aodnional survedlance requirements, as cefined in the Technical Specif cations, be Yes No necessitated by the activity? Explain.

Tech Specs are adequate to identify potential fastener failure (increased operational leakage) and limit Unit operation to within analyzed conditions.

40. Will the apphcable Technical Specif cation basis descnpflon be a!!ered by the activity? Explain.

Yes No No changes to the bases of the Tech Specs are proposed. The fasteners were inspected in place (some were removed and inspected) and found to be in good condition. No leakage was identified during the at-pressure walkdown and operational leakage is low.

41. Will the activity result in a violation of any Limiting Condition for Operation (LCO), as defined in the Yes @ No Technical Specifications? Explain.

Unit operation will continue to be in compliance with Tech Specs. Limits on operational leakage will insure that a fastener failure cannot develop from a small leak into an unanalyzed unit transient.

42. Were any otner concerns or rtems ident;fied cunng this review? If "Yes,* explain.

Yes No F otra No, 733fa16f MW 93)

Safety Evaluation Page 7 of 12 GOV 02 I

VPAP-3001 Pan C - Items Considered By This Safety Evaluation (Continued)

!l hems 43 through 62 consider potentialimpacts. VPAP-3001 provides engineering evaluation guidelines for these items.

i If the answer to any of the questions for these items is "Yes," a detailed engineering review must be performed. The results of the i

detailed review should be documented on a supplemental page, identified by this safety evaluation number and Part C item number.

1 Wdl the activity deactivate a security-related system or breach a security barrier?

Yes @ No j

T A. Willthe activity add or eliminate a significant amount of combustible material from plant areas?

O Yes @ No l

B. Willthe activity change or affect any plant structure or barrier that acts as a fire barrier?

Yes @ No l

C. Will the activity impact the performance of an existing fire protection or detection system?

O Yes @ No D. Will the activity involve modifying any component required for Appendix R. or any Appendix R O Yes @ No i

support system such as emergency lighting or emergency power supplies?

E. Wdl the activity changs or affect system flow paths shown on Appendix R ilow diagrams Yes @ No

)

(North Anna Power Station - 11715/12050-DAR-Series and Surry Power Station -

11448/11548-DAR-Series)?

F. Wdi the activity change station equipment arrangement drawings tnat show Appendix R Yes @ No ecuipment (Nortn Ar.na Power Stauon - 11715-FAR Series and Surry Power Station -

j 11448-FAR-Senes)?

4 A. Wdl the activity adversety affect any Class 1E electrical equipment located in a potentially Yes @ No harsh environment (as designated by the Environmental Zone Description)?

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B. Wdi the activity have the potential to after any of the environmental parameters identified O Yes @ No j

in the Environmental Zone Desenption?

C. Wi!l the activity have the potentialto aMect any of the Class 1E electricaldistribution systems O ves @ No i

i (e g.,4KV. 480V,120V(AC))?

i D. Wdi the activdy add, ehminate, or have the potential to affect ASME Section XI equipment adversely?

O Yes @ No E. Wdl the activity change a setpoint in the Precautions. Limitations, and Setpoints (PLS) Document 7 Yes @ No F. Wdithe acovity adversely affect equipment on the ECML or 0-List?

Yes @W i

Could the activity be adverse!y affected by a seismic event, or could the activity affect surrounding equipment during a seismic event?

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Wdi the activity change instrumentation or contro!s in the Control Room or on the auxiliary shutdown A

j panel?

[ Yes @ No i

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B. Wat the acuvdy alter the Control Room or the auxiliary shutdown panel?

Yes @ No i

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Page 8 of 12 GOV 02 VPAP 3001

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i Part C - Items Considered By This Safety Evaluation (Continued)-

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Will the activity change any of the equipment associated with the SPDS/ERF, including Yes @ No SPDS/ERF computer inputs?

1 Will the activ:ty have a significant potential to modify or add sof tware to station computers?

O Yes @ No I

A. Will the activity impact more than one-fourth of an acre of land, work in Yes @ No navigable waters, ws!!s, dams, or wetlands, and/or involve any wastes or discharges?

l B. Will the activity involve changes to site terrain, features, or structures?

Yes @ No i

l C. Will the activity have a significant potential to expose safety related equipment to flooding O Yes No via fluid system equipment / piping malfunction or f ailure?

i Wdi the act:v:ty have a sign:ficant potentialto modify equipment and/or instrumentation Yes @ No assoc:ated with Regu!atory Guide 1.97 vanables?

I l

A. Wdi the act:vity have a sign:!icant potential to increase the heating or cooling loads in plant areas O Yes @ No j

and/or to pl ant equipment?

l B. Wut the act;vity change the ex: sting ventitation system in any way?

Yes No l

C. Wdi the activity change any budding walls, ceilings, windows, doors, or floors, in a way that Yes @ No i

3 may affect existing HVAC systems?

j i

l l

l Wdi the act:vity involve heavy loads (including the transfer of heavy loads in areas housing safety Yes @ No re!ated equipment)?

6 l

Wdi detrimental materiais be introduced into the containment or other plant areas?

O Yes No Have ALARA concepts been included? (Detailed explanation not required.)

Yes O No i

j Wd! the proposed change adverse'y impact the current system / component capacities or design O Yes @ No J

pedormance?

I l

WWI the act'vity change applicable sections of the station design basis document?

O Yes No

]

l If a change to the Control Room or Safe Shutdown Panelis considered, will the enange Yes @ No

)

reed to be replicated in the simulator?

Fcrm No. 730916(Mar 93) 1 1

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Safety Evaluation Page 9 of 12 4

i GOV 02 l VPAP-3001 4

3 l Part C - Items Considered By This Safety Evaluation (Continued) l b

Wdithe activity resu't in the procurament of special nuclear materials or change the handling Yes @ No or storage of special nuclear materials?

I r8._

Will the activity affect a masonry block wall in any way, either through addition, removal, mounting of equipment, or location of safety related equipment within the vicinity of a block wall?

0 Yes No l

Will the activrty create a potential hazard / chemical re! ease?

Yes @ No f

E l

Will the activity affect station labeling?

Yes @ No l

1s Management oversight of infrequent tests or evolutions (as defined by VPAP-0108, infrequently Yes @ No

[

Conducted or Comptex Tests or Evolutions) recommended?

4 i

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)

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2 ll Form No. 730916(Mar 93)

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Safety Evaluation Page 10 of 12

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3 GOV 02 Part D Unteviewed Safety Question Determination J

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64. Which accidents prevously evaluated in the Safety Analysis Report were cons dered?

l Chapter 15 accidents.

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A. Could the activrry increase tne probability of occurrence for the accidents identified above? State i

the basis for this conclusion.

] yes g go Failure of the fasteners identified in DR N-94-16 is not expected. Studs were inspected in place as a minimum. No defects were

]

identified. Class I systems were tested dunng die RCS hydro following the RTD manifold work. The at pressure walkdown at the end of the outage did not indicate any leakage concerns in the repaired areas. Opea6onal leakage is currently low. Any l

fastener failure is expected to develop as a leak-before-break failure and will be detected prior to exceeding Tech Spec limits for 1

operadonal leakage. Operauonal leakage w ill therefore not exceed limits used as a precursor event for accident analyses.

I 2

j B. Could tne activity increase the consequences of the accidents ident:fied above? State the basis for this conclusion.

Yes 521 No o

i Failure of the fasteners will not cause a catastrophic failure of any equipment required for the midgation of analyzed accidents.

increawd leakage at bolted connecdons may occur, but will not create a situauon which is not bounded by existing accident i

analyses.

j i

i C Cou d tne activny create tne posstbinty for an accicent of a c:!ferent type inan was prevously Yes @ No evaluated in the Safety Analysis Report? State the basis for this conclusion.

Inetcased leakage of systems due to fastener f ailure will be captured and processed by liquid waste systems adequate to handle the fluids (radioacuve fluid, chemical content, etc.). Failure of eqwpment and components is bounded by existing analyses (VCT failure, Loss of Coolant Accidents, etc.). No new accidents are proposed.

f

65. What ma:functons of equ:pment related to safety, previousty evaivated in tne Safety Analys's Report, were considered?

Fadure of the Charging Pump suction to swap to the RWST on low VCT level as detected ' y 2-CH-LT-2112 and 2115.

r Rupture of the VCT.

i s

A. Could the activ!ty increase the procaony et occurrence 01 mattunctions ident:fied above? State the basis for this conclusion.

Yes @ No j

Should 2-CH LT-2112 develop a leak, indication would fail high and an alarm would come in to help identily this condinon, i

'Ihe probability of occurence is not mcreased since the studs were removed and inspected and no defects were found.

Manway fasteners on the VCT were VT-2 inspected and no defects were identified. No leakage has been identified.

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r wm w ras.uar ex i

e Safety Evaluation Page 11 of 12 GOV 02 VPAP-3001 Part D - Unreviewed Safety Question Determination (Continued) j B. Could the activity increase the consequences of the malfunctions icentifisc above?

State the basis for this conclusion.

g Consequences of failure of the auto swap of the Charging Pump suction will not change.

Rupture of the VCT is bounded by existing analyses.

C. Could the activity create the possibility for a mattuncton of equipment of a different type than Yes No was previously evaluated in the Safety Analysis Report? State the basis for this conclusion.

Failure of the fasteners would progress in a leak-before-break fashion and would be identified by increased sump activity or increased airborne radioactivity. Stud inspections revealed no defects, and no leakage has been identified on these repairs.

l

66. Has the margin of safety of any part of the Technical Specificatons as casenbod in the bases sectrori l

been reduced? Explain.

Yes @ No No fastener failures are expected since the inspections which were performed were Skill of the Craft inspections and no defects were found. leakage which may occur is bounded by existing analyses.

67. Does the proposed change, test. or experiment require a change to tne Techncal Specif catons?

Explain.

Yes @ No Unit operation may continue in accordance with the Tech Specs. Existing surveillances will identify fastener failure.

68. Does the proposso enange, test, or expenment involve a sgnitcant unreviewed environmental impact?

Explain.

Yes No l

j N/A

69. Does the proposed change, test, or expenment involve a sgnificant increase in occupatonal exposure?

Yes O No State the basis for this conclusion.

gjg Famusrw warn

Safety Evaluation Page 12 of 12 GOV 02 VPAP-3001 Part D - Unreviewed Safety Question Determination (Continued)

If all responses are "fJo" to Questions 64 through 69, the proposed activity may be implemented following SNSOC approval All related documentation must be retained, if a response is "Yes" to any part of Questions 64 through G7, an operating license amendment must be approved by NRC before the change. test, or experiment may be implemented.

if a response is "Yes" to Question 68 or 69, an application for an ISFS1 license amendment must be approved by NRC before the change, test, or experiment may be implemented.

70. Reviewer flame (Pnnt)

71. Reviewer Title i

d [49gog Shift Technical Advisor l

72. Reviewer Signature

73. Date i

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/ - G - 9 '/

14. Design Authority Heviewer (Jame (Pnnt)

15. Design Authorrty Heviewer htte

76. Design Autnority Reviewer Signature

77. Date i

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