ML20128M555

Forwards ISI Program Second 10-yr Interval Relief Requests NDE-18 for Units 1 & 2 & SPT-14 for Unit 2 in Support of Interval Scheduled to End 981224

ML20128M555
Person / Time
Site:	North Anna
Issue date:	02/16/1993
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
93-018, 93-18, NUDOCS 9302220208
Download: ML20128M555 (28)

Text

-

VHtOIN1 A El.ECTit!C AND POWEli COMI'ANY HICIIMOND, VHtOINI A 23U01 February 16, 1993 United States Nuclear Regulatory Commission Serial No.

93 018 Attention: Document Control Desk NL&P/EJW Washington, D.C. 20555 Docket Nos.

50 338 50 339 License Nos. NPF 4 NPF 7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 INSERVICE INSPECTION PROGRAM SECOND TEN YEAR INTERVAL RELIEF REQUESTS North Anna Power Station Unit 1 is currently in the second period of its cecond ten year interval. The interval is scheduled to end December 24,1998. The Code of reference for North Anna Unit 1 is the 1983 Edition, Summer 1983 Addendum of ASME Section XI.

North Anna Power Station Unit 2 is currently in the first period of its second ten year interval. The interval is scheduled to end December 14,2000. The Code of reference for North Anna Unit 2 is the 1986 Edition of ASME Section XI.

Relief Requests NDE 18 (applicable to Unit 1 and Unit 2) and SPT-14 (applicable to Unit 2 only) are being submitted in accordance with 10 CFR 50.55a (g) (5) for your consideration (Attachment 1). Relief Request SPT-14 is needed for North Anna Unit 2 prior to the next refueling outage, which is currently scheduled to begin on September 4,1993. The request will provide guidance for pressure testing known to be required during that outage. Examinations affected by Relief Request NDE-18 are scheduled at the end of the current Unit 1 and 2 intervals.

These requests have been approved by the North Anna Station Nuclear Safety and Operating Committee.

1 190085 oAA 93o222o208 930216 l

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i If you have any questions concerning these requests, please contact us.

1 Very truly yours,

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W. LYStewart Senior Vice President Nuclear Attachments 1

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United States Nuclear Regulatory Commission Region ll 101 Marietta Street, N.W.

Suite 2000 Atlanta, GA 30323 M.. M. S. Lesser

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NRC Senior Rosident inspector North Anna Power Station t

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l NORTH ANNA UNIT 1 RELIEF REQUEST NDE-18 1

1. Identification of Components System:

Chemical and Volume Control (CH)

Component:

Regenerative deat Exchanger (1-CH E 3) i Welds /Comoonents Q13crintion

- Code item #

Class 3

tubesheet to head B2.60 -

1 7

tubesheet to head B2.60 1

11 tubesheet to head 82.60 1

4 circumferential head B2.51.

.1 8

circumferential head B2.51 1

12 circumferential head 82.51

-1 13 nozzle-to vessel B3.150 11 14 nozzle to vessel B3.150 1

15 nozzle to vessel B3.150 1-16 nozzle to vessel B3.150 1

17 nozzle to vessel B3.150 1

18-nozzle to vessel B3.150 1

13NIR nozzle inside radius B3.160 1

14NIR nozzle inside radius

.83.160 1

15NIR nozzle inside radius B3.160 1-16NIR nozzle inside radius B3.160 1

17NIR nozzle inside radius B3.160 1

5 18NIR nozzle inside radius B3.160 1-W S-1 we!ded attachment 88.40 1

WS2 welded attachment B8.40 WS-3 welded attachment B8.40 1L

?

WS4 welded attachment B8.40 -

1 WS-5 welded attachment-88.40.

1 WS6 welded attachment B8.40 1-1 circumferential head C1.20 2

5 circumferential head C1.20 2.

9 circumferential head C1.20 -

2 2

tubesheet to shell C 1.30.

-2 6

tubesheet-to shell C1.30 2

10 tubesheet to-shell C1.30 2

(Drawing attached)

II. Impractical Code Requirements -

Examination Categories B B, B D (Ir.spection Program B), B H, and C A require-that volumetric and surface examinations be performed as indicated by the Code item numbers above.

Page 1 of 3

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4 UNIT 1 REllEF REQUEST NDE-18 Continued:

111. Basis For Relief The regenerative heat exchanger (1 CH E 3) provides preheat for the normal charging water going into the reactor coolant system (RCS). The preheat is derived from normal letdown water coming from the RCS. Charging and letdown constitute the normal chemical and volume control within the RCS. The heat exchanger itself is actually three heat exchangers in series interconnected with piping. This fact was previously utilized in limiting examinations to one of the heat exchangers as allowed by the Code. The heat exchanger has an outside shell diameter of 9.55 inches. The shells were manufactured with ASTM A351 CFS type material. The heads were manufactured with ASTM A240 TP304 material. The 3 inch nozzle necks were manufactured with ASTM A182 F304 material. Until very recently the regenerative heat exchanger was entirely classified ASME Class 2 for inservice inspection activities. However, a reanalysis changed the classification of the letdown side of the heat exchanger to ASME Class 1. This action significantly increases the examination requirements associated with this heat exchanger.

Nozzles, which were previously exempt under Class 2 requirements are now required to be examined. Additionally, all Class 1 nozzios are required to be examined, and the examinations are not limited to one heat exchanger.

The nozzle to vessel welds and nozzle inside radius sections for this vessel were not designed for ultrasonic examination from the outside diameter of the vessel.

The small diameter of the vessel and nozzles along with_ the cast stainless steel vessel shell prevents a meaningful ultrasonic examination of these components.

The Code required volumetric examination on the vessel head circumferential welds is limited due to the weld crown, radius of the closure caps, and the nozzles.

The Code required volumetric examination of the tubesheet welds is limited by the weld crown and is obstructed by a support clamp. This clamp must be mechanically removed prior to the wolds' examination. Additionally weld 11 is partially obscured by the six integral attachments, which are additionally themselves butted up against a clamp. It is estimated that between 21 and 42 percent of the circumferential welds could be examined, and 42 percent of the tubosheet welds could be examined, if the clamps are removed. Weld 11 would be significantly less due to the integral attachment location. Previous partial examinations completed on these welds have identified no problems.

An ALARA evaluation has been conducted on each activity assuciated with these examinations. A table is provided documenting these results, it is estimated that more than 26 man-rem will be required to complete these examinations over the interval. This estimate assumes optimum inspection and preparation times. If difficulties are encountered, a corresponding increase in dose would be expected.

Shielding is not considered practical since the source of radiation is the component receiving the examinations. Considering the examination limitations previously discussed, expending this much dose is deemed impractical.

Page 2 of 3

UNIT 1 REllEF RF._ QUEST NDE-18 Continued:

IV. Alternate Requirements Technical Specifications requires that the Reactor Coolant System Leak Rate be limited to 1 gallon per minuto unidentified leakage. This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Additionally, the containment atmosphere particulate radioactivity is monitored every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. As a result, new leakage is rapidly identified and located during operation. Leakage identified on this component can be easily isolated by two upstream valves with manual operation from within the control room. Additionally, the valves receive an automatic control signal to close on inventory loss based on pressurizer level. However, these valves could not be used as the Class 1 boundary valves due to their nonsafety-related actuation.

Correspondingly, as a result of the reclassification to Class 1, this component will receive a system leakage test prior to start up after each refueling outage. During this system lea'Kage test the component will receive a visual (VT 2) examination.

The support structure will receive a visual (VT-3) examination to the extent required by the Code without insulation removal.

Page 3 of 3

1 -C H-E-3 EXAMINATIONS MAMREM ESTIMATE Work Task Man Hours (hrs)

Dose Rate (R/hri Manrem insulation removal / install 5.3 0.500 2.650 Scaffolding install / removal 1.0 0.350 0.350 Clamp removal / install 2.0 1.000 2.000 Weld Prep 3.5 1.000 3.500 HP coverage 31.0 0.010 0.310 Nozzle-to-vessel inspection (UT) 6.0-1.000 6.000 1

Nozzle inside radius inspection (UT) 4.5 1.000 4.500 Circumferential/

Tubesheet inspection (UT) 4.0 1.000 4.000 Welded Attachment inspection (PT) 3.0 1.000 3.000 Total Estimate - 26.310

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I JOB MAN-REM PROJECTION DATE:12/04/92 ALARA EVALUATION NUMBER:

UNIT #

1 JOB DESCRIPTION: ISI/NDE REGEN HEAT EXCHANGER INSPECTIONS 1

WORK MAN DOSE MAN-REM /

TASK TASK GROUP HOURS RATE WORKGROUP TOTAL TOP Hx:

ISI/NDE

- UT (2) NOZ TO VESSEL WELDS 2.0 1.000 2.000

- UT (2) INNER RADIUS WELDS 1.5 1.000 1.500 MID He

- UT (2) NOZ TO VESSEL WELDS 2.0 1.000 2.000

- UT (2) INNER RADIUS WELDS 1.5 1.000 1.500 BOTTOM Hx:

- UT (2) NOZ TO VESSEL WELDS 2.0 1.000 2.000

- UT (2) INNER RADIUS WELDS 1.5 1.000.

1.500-

- UT (4) CIRCUMFERENTIAL WELDS 4.0 1.000

'4.000

- PT (6) WELDED ATTACHMENT LUGS 3.0 1.000 3.000 17.500

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SUPPORT ACTIVITIES:

- lNSUL. RMVlJINSTALL INSUL.

5.3 0.500 2.650

- SCAFFOLDING INSTALtJRMVL CARP.

1.0 0.350 0.350

- WELD PREP PF 3.5 1.000 3.500

- HP COVERAGE HP 31.0 0.010 0.310

- (3) CLAMP RMVLJINSTALL PF 2.0 1.000 2.000 8.810-NOTE: UT INSPECTION FOR CIRCUMFERENTIAL WELDS INCLUDE (3) PASSES: 0,45, AND 60 DEGREES MAN-REM TOTAL 26.310 REM MAN-HOUR TOTAL 60' HOURS

[/

" IAl/ ARA SIGNATURE

4 NORTH ANNA UNIT 2 RELIEF REQUF.ST NDE 18 l.

Identification of Components System:

Chemical and Volume Control (CH)

Component: Regenerative Heat Exchanger (2 CH E 3)

Welds /Comnonents DescrIntion Code item # -

Class

.i 8

tubesheet to head B2.60 1

10 tubesheet to head 82.60 1

12 tubesheet to head B2.60 1

2 circumferential head B2.51 1

4 circumferential head B2.51 1

6 circumferential head B2.51 1

13 nozzle to vessel B3.150 1

14 nozzle to vessel B3.150 1

15 nozzle to vessel B3.150 1-16 nozzle to vessel B3.150 1

17 nozzle to vessel B3.150 1

18 nozzle to-vessel B3.150 1

13NIR nozzle inside radius 83.160 1

14NIR nozzle inside radius B3.160 1

15NIR nozzle inside radius B3.160 1

16NIR nozzle inside radius 83.160 1

17NIR nozzle inside radius B3.160 1.

18NIR nozzle inside radius 83.160 1.

WS1 welded attachment B8.40 1

WS2 welded attachment B8.40 1

WS3 welded attachment B8.40 1

WS4 welded attachment 88.40 1

WS5 welded attachment 88.40 1

WS6 welded attachment 88.40 1

1 circumferential head C1.20 2

3 circumferential head C1.20 2

5 circumferential head C1.20 2

7 tubesheet to shell C1.30 2

9 tubesheet-to shell C1.30

'2 11 tubesheet to-shell C1.30 2

(Drawing attached)

II. Impractical Code Requirements Examination Categories B B, B-D (Inspection Program B), B H, and C-A require that volumetric and surface examinations be performed as indicated by the Cade item numbers above.

Page 1 of 3 z

MIT 2 RELIEF REQUEST NDE-18 Continued:

Ill. Basis For Rollef The regenerative heat exchanger (2 CH E 3) provides prohoat for the normal charging water going into the Reactor Coolant System (RCS). The preheat is derived from normal letdown water coming from the RCS, Charging and letdown constitute the normal chemical and volume control within the RCS. The heat exchanger itself is actually three heat exchangers in series interconnected with piping. This fact was previously utilized in limiting examinations to one of the heat exchangers as allowed by the Code. The heat exchanger has an outside shell diameter of 9.55 inches. The shells were manufactured with ASTM A351 CF8 type material. The heads were manufactured with ASTM A240 TP304 material. The 3 inch nozzle necks were manufactured with ASTM A182 F304 material. Until very recently the regenerative heat exchanger was entirely classified ASME Class 2 for inservice inspection activities. However, a reanalysis changed the classification of the letdown side of the heat exchanger to ASME Class 1. This action significantly increases the examination requirements associated with this heat exchanger.

Nozzles which were previously exempt under Class 2 requirements are now required to be examined. Additionally all Class 1 nozzles are required to be examined, and the examinations are not limited to one heat exchanger.

The nozzle to-vessel welds and nozzle inside radius sections for this vessel were not designed for ultrasonic examination from the outside diameter of the vessel.

The small diameter of the vessel and nozzles along with the cast stainisss steel vessel shell prevents a meaningful ultrasonic examination of these components.

The volumetric examination on the vessel head circumferential welds as required by the Code is limited due to the weld crown, radius of the closure caps, and the nozzles. The Code required volumetric examination of the tubesheet welds is limited by the weld crown and is obstructed by a support clamp. This clamp must be mechanically removed prior tc the welds' examination. Additionally weld 12 is partially obscured by the six integral attachments which are themselves butted up aga!nst a clamp.

It is estimated that between 21 and 42 percent of the circumferential welds could be excmined, and 42 percent of the tubesheet welds could be examined, if the clamps are removed. Weld 12 would be significantly l

less due to the integral attachment bcation.

Previous partial examinations completed on these welds have identified no problems.

An ALARA evaluation has been conducted on each activity associated with these l

examinations. A table is provided documenting these results. It is estimated that more than 32 man rom will be required to complete these examinations over the interval. This estimate assumes optimum hospection and preparation times. If difficulties are encountered, a corresponding increase in dose would be expected.

Shielding is not considered practical since the source of radiation is the I

component receiving the examinations. Considering the examination limitations previously discussed, expending this much dose b deemed impractical.

Page 2 of 3 I

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UNIT 2 RELIEF REQUEST NDE 18 Continued:

IV. Alternete Requirements Technical Specifications requires that the Reactor Coolant System Leak Rate be limited to 1 gallon por minuto unidentified leakage. This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Additionally, the containment atmosphere particulate radioactivity is monitored overy 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. As a result, new leakage is rapidly identified and located during operation. Leakage identified on this component can be easily isolated by two upstream valves with manual operation from within the control room. Additionally the valves receive an automatic control signal to close on inventory loss based on pressurizer level. However, these valves could not be used as the Class 1 boundary valves due to their nonsafety related actuation.

Correspondingly, as a result of the reclassification to Class 1, this component will receive a system leakage test prior to start up after each refueling outa00. During this system leakage test the component will receive a visual (VT 2) examination.

The support structure will receive a visual (VT 3) examination to the extent required by the Code without insulation removal.

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L JOB MAN-REM PROJECTION DATE: 12/04/92 ALARA EVALUATION NUMBER:

UNIT #

2 JOB DESCRIPTION: ISI/NDE REGEN HEAT EXCHANGER INSPECTIONS -

WORK MAN DOSE MAN-REM /

TASK TASK GROUP HOURS RATE WORKGROUP TOTAL TOP Hx:

ISI/NDE

- UT (2) NOZ TO VESSEL WELDS 2.7 1.000 2.700

- UT (2) INNER RADIUS WELDS 2.2 1.000 2.200 MID Hx:

- UT (2) NOZ TO VESSEL WELDS 2.7 1.000 2.700

- UT (2) INNER RADIUS WELDS 2.2 1.000 2.200-BOTTOM Hx:

- UT (2) NOZ TO VESSEL WELDS 2.7 1.000 2.700

- UT (2) INNER RADIUS WELDS 2.2 1.000 2.200

-- UT (4) CIRCIIMFERENTIAL WELDS 5.3 1.000 5.300

- PT (6) WELDED ATTACHMENT LUGS 3.0 1.000 3.000 23.000 SUPPORT ACTIVITIES:

- INSUL. RMVL/ INSTALL INSUL 5.3 0.650 3.445

- SCAFFOLDING INSTALURMVL CARP.

0.0 0.000 0.000

- WELD PREP PF 3.5 1.000 3.500

- HP COVERAGE HP 36.0 0.010 0.360

- (3) CLAMP RMVUINSTALL PF 2.0 1.000 2.000 9.305 NOTE: UT INSPECTION FOR CIRCUMFERENTIAL WELDS INCLUDE (3) PASSES: 0,45, AND -

60 DEGREES MAN-REM TOTAL -

32.305 REM MAN-HOUR TOTAL 70 HOURS q

U Al' ARA SIGNATURE

NORTH ANNA UNIT 2 RELIEF REQUEST SPT-14 0

1.

Identification of Components Class 1,2, and 3 pressure retalning botting (North Anna Unit 2).

i ll. Impractical Ccde Requirements IWA-5250 (a) (2) states, "if leakage occurs at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and 4

evaluated in accordance with IWA-3100...."

lit. Basis For Relief Typically pressure testing required by the Code will identify mechanical connection leakage from gasket or packing sources. This leakage is-evaluated routinely during or following the test, and additional torquing-or other corrective measures on the bolting is applied to eliminate the leakage.

Generally, this type leakage is not active long enough to cause component damage.

In some instances testing occurs while containment is subatmospheric and just prior to reactor criticality. Requiring that bolting be removed in these instances for minor leakage would severely interrupt the normal start up schedule. Additionally, the Code requirement does not incorporate previous NRC commitments associated with boric acid wastage on bolting. North Anna will examine safety-related components for boric acid leakage at the start of a reactor refueling. Each component affected by boric acid leakage will be evaluated for potential damage, and corrective action will be taken as appropriate. In accordance with commitments made previously to NRC Generic Letter 88 05 and in relief request SPT-12, a supplemental examination of Class 1 Category B-G-1 and B G 2 bolting is also conducted every refueling outage. This supplemental examination is in excess of the normal inservice inspection program requirements. The current Code requirement assumes no additional examination or evaluation has taken place, and the Code makes the conservative assumption that leakage found has occurred for a sufficient time period to cause damage.

This approach is impractical, when compared to the alternative arrangements proposed.

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- RELIEF - REQUEST SPT-14 Continued:

IV. Proposed' Alternate Requirements For_ bolting examined during pressure tests (Class 1 system leakage, Class 1 hydrostatic, and Class 2 pressure tests scheduled.In association with the Class 1 tests) which are conducted just prior to start up in subatmospherico conditions and in which leakage is identified near the bolted connections, bolting will be evaluated for removal need. ' The removal need will be based -.

upon the' extent of the leakage, correction requirements, and results of-previous examinations conducted during:that refueling' outage. iThisi evaluation shall be subject to the review of the Authorized Nuclear Inservice-Inspector (ANil). Bolting in any situation requiring removal and visual (VT-

3) examination may be limited to one bolt nearest the leakage source.Dif that bolt has evidence of degradation, then all other bolting:in the connection shall.be removed, visually (VT-3) examined, and evaluated to -

i the Code requirements. :The limitation of selecting only one bolt. initially is; the same as the Code requirements found in the 1992 Edition of ASME-Section XI, IWA-5250 (a) (2).

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