ML20155D971

Forwards Request for Relief from Various Exam Requirements of ASME Section XI Re Hydrostatic Pressure Testing.Exams Cover Chemical & Vol Control,Feedwater Check Valves & RHR Class 1 Components.Fee Paid

ML20155D971
Person / Time
Site:	North Anna
Issue date:	10/03/1988
From:	Cartwright W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
88-467, NUDOCS 8810120054
Download: ML20155D971 (34)

Text

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VinorxrA Er.ncTurc Ann Powan Coxprxy Hicuwoxo.Vinoix:A g u g e,i  !

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October 3, 1988 k

United States Nuclear Regulatory Coomssion Serial No.18-467 Attention: Document Control Desk PES /IS!/DJF  :

Washington, D.C. 20555 Docket No. 50-339  ;

License No. NPF-7 Gentlemen: j VIRGINIA ELECTRIC AND POWER COMPANY f i'

ERTff~ ANNA POWER STATION UNIT 2 A5ME SECTION XI RELIEF REQtEST5 j in accordance with 10 CFR 50.55a paragraph (g)(5), we are requesting relief l from various examination requirements of ASME Section XI for North Anna Unit

. 2. The particular relief requests are identified in Attachments 1 and 2.  !

l Attachment 1 addresses relief requests from hydrostatic pressure testing j requirements, and Attachment 2 addresses relief requests from NDE examinatiorM.

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l The relief requests for NDE examinations are to the 1974 Edi tic,n of ASME  ;

Section XI through the Sumer 1975 Addenda. The hydrostatic pressure tests i' are performed in accordance with the 1977 Edition of the Code through the Summer 1979 Addenda as documented in the NRC letter to Virginia Electric and ,

Power Company dated June 7,1982. Relief is requested from this edition of the code for hydrostatic pressure testing, 1

i i North Anna Unit 2 will begin its first interval ten year ISI examinations during the scheduled February 1989 outage and complete these examinations during the next outage presently scheduled for August 1990. Review of these ,

' relief requests is desired in time to support the 1989 outage.  !

If you have any questions or require further information, please advise.

Enclosed is a check in the amount of $15') for the review application fee. ,

Very truly yours, i

[

j fM h~J p- y %eg all W. R. Cartwright ' '

l Vice President - Nuclear l '8l 1

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r Attachments h

, Of110120054 081003 l Pl>R ADOCK 05000339 i P PNV ,

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cc: U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, N.W.

Suite 2900 Atlanta, Georgia 30323 Mr. J. L. Caldwell NRC Senior Resident inspector North Anna Power Station

COMMONWEALTH OF VIRGlNIA )

)

COUNTY OF HENRICO )

The foregoing docuT.ent was acknowledged before me, in and for the County and Commonwealth aforesaid, today by W. R. Cartwright who is Vice President -

NJClear, of Virginia Electric and Power Company. He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this J _ day of Ofg h ; , 19 88 ity Commission expires: _ 1e httu aa 1, Z6 , 19 90 .

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Oleu Notary Pu pcaL_

(SEAL)

l Attachn.ent 1 i RELIEF RE0 VESTS FOR 10-YEAR HYDROSTATIC TESTS Relief Reauest DescriqLLgn hgg SPT-1 Chemical and Volume Control - Over 1 pressurize Reactor Coolant Pump Seal SPT-2 Chemical and Volume Control - Check 2 valves prevent lines from being pressurized SPT 3 Chemical and Volume Control - Check 3 valves prevent lines from being pressurized SPT-4 Feedwater - Check valves allow over- 4 pressurizing the Steam Generator spi-5 Feedwater, Chemical and Volume 5 Control, and Safety injection - Pump Suction Lines cannot be isolated SPT 6 Reactor Coolant - Reactor Vessel 6 Inspection SPT-7 Safety injection - Check valves allow 7 pressurizing Reactor Coolant System SPT-8 Safety injection - Overpressurizing 8,9 Reactor Ceolant System SPT-9 Safoty injectiun - Check valves allow 10 overpressurizing the primary system ,

SPT-10 Main Steam (Steam) Decay Heat 11-14 l Release, Feedwater, Chemical Feed,  !

Blowdown - Test per Steam Generator Manu'acturers recommendations for piping that cannot be isolated SPT-ll Component Cooling, Chemica' 'nd Volume 15 Control Fuel Pit Cooling, dafety '

Injection, Quench Spray, Recirculation Spray, Service Water, Sampling - Test Pressure Limit SPT-12 Residual Heat Removal - Class 1 16 Components f

REllEERE00EST SPT RE0VE!T TO TEST AT LOWER PRESSURE

1. 333113 - Chemical and Volume Centrol

2. Comoontnts - Piping on drawing 12050 FM-95C between the pumps and first flange from pumps.

PVMPS LIBE 2-RC-P 1A 2" CH-414-1502 2-RC-P-18 2"-CH 415 1502 2-RC P-lc 2"-CH 416 1502

3. Code Reauirements - Class 1 System Hydrostatic Test per IWB-5222, 1977 Edition through Sum.ner 1979 Addenda of Section XI, 1.10 times system nominal operating pressure.

4. Ergp e d Alternative Examination The normal system leakage test after each refueling is an adequate examination.

5. Reason for Relief Pressurizing the piping listed above will also pres >urize the number one seal of the reactor coolant pumps. This could potentially damage the numbe: one seal.

Relief was granted to Surry Power Station Unit 2 per Safety Evaluation  :

Report dated January 24, 1986 for the same situacion as described above.

Relief was also granted to North /,nna Power Station Unit 1 per Safety Evaluation Report dated July 13, 1987 (TAC No. 64718) for the same situation as described above.

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REllEF RE0 VEST SPT RE0 VEST TO J,fST AT LOWER PRESSURE 1, Eyita - Chemical and Volume Control

2. Component - Piping located on drawing 12050-FM-95C between the valves listing below:

Valves Ling HVC 2311 and 2 CH-341 2" CH-468 1502 2 CH-358 and 2-CH 340 3" CH-401-1502

3. Cnidftuttfren11 Class 1 System Hydrostatic Test per IWB 5222. Po=2500 psig, To=496*F, Test Pressure is 2550 psig per IWB-5222.

4. Pf.poosed Alternative Examination As an alternative, the Reactor Coolant System will be pressurized to a pressure as close as practical to 2335 psig but not less than 2300 psig while the reactor

• in a shutdown condition to create a pressure boundary at check valves 2-CH-341 and 2-CH 358. The compunents listed above will then be tr.stM to a pressure (2300 psig < test pressure < 2335 psig) as close as practical to the Reactor Coolant System pressure using a charging pump.

5. Reason for Relief Check valves 2 CH 341 and 2-CH 358 prevent the components listed above from being pressurized without pressurizing the Reactor Coolant System.

The code required test oressure of 2550 psig will overpressurize the Reactor Coolant System.

Also, the power operated r.$ lief valves (PCV-2456 and PCV-2455C) of the Reactor Coolant System are designed to limit the pressurizer pressure to a value below the fixed high-pressure reactor trip setpoint (2385 psig).

The relief valve setpoints are 2335 psig. It is not desirable to take the Reactor Coolant System above the power operated relief valve setpaint.

Similar relief was granted to Surry Power Station Unit 2 per Safety Evaluation Report dated January 24, 1986 (Docket 50-281) for components with the same design configuration. Relief was also granted to North Anna

! power Station Unit 1 per Safety Evaluation Report dated July 13, 1987 (TAC No. 64718) for the same situation as described above.

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RELIEF REQUEST SPT REQUEST TO 'aEST AT LOVER PRESSURE ,

1. System - Chemical and Volume Control ,

2. Components - Piping between the valves listed below located on drawings 12050-FM-95C and 12050-FH-95B.

Valves Line l

! 2-CH-358, 2-CH-HCV-2311, and 2-CH-MOV-2289A 3/4"-CH-640-1502 [

2"-CH-458-1502 3"-CH-401-1502 j 3"-CH-479-1502 1 3"-CH-819-1502  ;

! 3. Code Requirerents - Class 2 System Hydrostatic Test per IWC-5222. Since  ;

! there are no relief valves for the above components, test preasure per IWC-5222 is 3419 psig.

4. Proposed Al 'ernative Examination As an alternative, the Reactor Coolant System vill be pressurized to a pressure as close as practical to 2335 psig but not less than 2300 psig while the reactor is in a shutdown condition to create a pressure boundary [

at check valves 2-CH-341 and 2-CH-358. The components listed above will t

i then be tested to a pressure (2300 psig < test pressure < 2335 psig) as close as practical to the Reactor Coolant System pressure using a charging j i pump.

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5. , Reason for Relief  !

Check valves 2-CH-341, 2-CH-340 and 2-CH-358 prevent the components listed l'

above from being pressurized vitbout pressurizing the Reactor Coolant {

1 System. The Cods required test pressure of 3419 psiF vill overpressurize i t

j the Reactor Coolant System. Also, the power operated relief valves i (PCV-2456 and PCV-2455C) of the Reactor Coolant System are designed to

limit the pressurirer pressure to a value below the fixed high-pressurn reactor trip setpoint (2385 psig). The relief valve metpoints are 2335 t psig. It is not desirsble to take the Reactor Coolant System above the t l

power operated relief valve setpoint.

i Relief was granted to North Anna Power Station Unit 1 per Safety [

Evaluation Report dated July 13, 1987 (TAC No. 64718) for the same l situation as described above. ,

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RELIEF RF0 1 VEST SPT 4 RE0 VEST TO TEST IN ACCORDANCE WITH WESTINGHOUSE TECHNICAL MANUAL

1. Lyll.ed!) - Feedwater

2. Cs_mponents - Piping between the valves listed below located on drawing 12050-FM 74A.

Valve ConnectIna Line Valve 2 FW-64 3"-WAPD 410-601 to 2-FW 66 3" WAPD-409 601 2-FW 66 3"-WAPD 409 601 2 FW-70 2-FW 96 3"-WAPD 412-601 to 2 FW 98 3" WAPD 411-601 2-FW 98 3"-WAPD-411 601 2 FW.102 .

. 2 FW-128 3"-WAPD 414-601 to 2-FW-130 1

3"-WaPD 413-601 2 FW-130 4"-WAPD-413-601 2 FW-134 2 FW-278 4"-WAPD 439-601 to 2 FW 66 3"-WAPD 409-601

3. Code _Bequirements - Class 2 System Hydrostatic Test per IWA 5513(d) and IWC-5222. P 1400 osig, d

T <200*F, d

Test Pressure is 1540 psig per

IWC-5222.

4. P_tspgig!1 Al t e rn a t i v e _ E x a m i n a t i o n l

Since the components listed above cannot be pressurized without pressurizing the steam generator, they must be tested per the required manufacturer's hydrostatic test method for the steam generators. 1 Therefore, the proposed alternative examination is the examination '

described in the Westinghouse Technical Manual for the secondary side of the steam generator. The excmination is to pressurize the secondary side of the steam generator to 1356 psig, holi for 30 minutes, and then reduce to the design pressure (1085 psig) for 3 1/2 hours. A VT-2 examination will then he performed.

5. B. talon for Reliti Due to check valves 2 FW-134, 2-FW-102, and 2-FW-70 the piping listed above cannot be pressurized without pressurizing the steam generators. l The code required %st prassure of 1540 psig would overpressurize the i steam generator.

i Relief was Granted to North Anna Power Str. tion Unit 1 per Safety Evaluation Report dated July 11 1987 (TAC No. 64718) for the same i i

situation as described above. ,

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RELIEF .8[@E}_T_SpT RE0 VEST TO TEST AT LOWER PRESSUWE

1. Sylt.gm - Feedwater, Chemical and Volume Control and Safety Injection

2. CWlonents - Centrifugal pumps and discharge piping to first isolation -

valve on drawings 12050-FM 74A, !?050-FM 958, 12050 FM 96A.

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2-FW P 2, 2-FW-P 3A, 2 FW-P-3B 2-CH-P-1A, 2 CH-P 18, 2-CH P-1C i 2 SI P 1A, 2-SI-P-1B

3. Code Reouirericuli - Class 2 System Hydrostatic Tests per IWC-5222 and t IWD-5223.

4. Proposed Alternative Examination As an alternative, the test pressure for the pump discharge and associated piping extending to the first shutoff valve on the discharge side of the pump shall be the same as that required for the piping and components on the suction side of the pump. In support of this alternative, the 1980 Edition, Winter of 1981 Addenda of ASMC Section XI l (approved by the NRC) paragraph IWA 5224(d) allows. the system test >

boundary interface to be the first shutoff valve on the discharge side of thr; centrifugal pump when the primary system praute ratings on the suction and discharge sides differ.

5. Reason for Relief 1 Centrifugal pumps and the portions of the pump discaarge lines up to the  ;

first isolation valve cannot be isolated from the pump suction. If the ;

, discharge piping were pressurized to the required test pressure the

suction piping would be subjetted to a pressure far in excess if its t design with the potential for permanent damage to piping and components. l il '

Pelief was granted to North Anna Power Station Unit 1 per Sa raty Evaluation Report dated July 13, 1987 (TAC No. 64718) for the saw ,

situation as described above.

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RELIFF RE0 VEST SPT RE0 VEST NOT TO PERFORM VISUAL EXAMINATION

1. System - Reactor Coolant

2. Component - Bottom of reactor vessel.

3. Csielguirements - Visual Examination per IWA-5240 and IWB 5222 during System Hydrostatic Test.

4. Procosed Alternative Examination The proposed alternative examination is to examine the bottom of the rear. tor vessel for esidence of leakage during the 10 year vessel inspection, and to perform a visual inspection of the lower head.

5. Reason for Relief The system hydrostatic test for the Reactor Coolant System is performed at Hot Standby, MODE 3. The bottom of the reactor vessel is inaccessible due to temperature and radiological concerns.

Relief was granted to North Anna Power Station Un.t 1 per Safety Evaluation Report dated July 13, 1987 (TAC No. 64718) for the same situation as described above.

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. RELIEF RF0 VEST SPT 7 - RE0 VEST TO TEST AT LOWER PRESSURE j ,

1. System - Safety Injection l

. 2. In s nents - Piping between the following sets of valves located on station print 12050-FM 968  ;

i Valves Lines j 2-S1 92, 2-SI-90, and 2-S1 91 6" S1-531-1502 f 2 51-100, 2 SI 98, and 2 51-99 6" SI-533 1502 -

! 1 2-S1-106, 2-SI 104, and 2-SI-105 6" S1 532-1502 i

! 2-SI-l?S, 2 51 123, and 2-51-112 6"-S1-416 1502 .

i 2" SI-463 1502 t 2-51 113, 2-S1-111, and 2 S1 117 6" S1-419 1502 [

1 2" SI 459-1502  !

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2-51-113, 2 51-116, and 2-SI 124 6"-SI 421-1502  ;

2 2"-SI-461-1502 1

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3. Code Reauirement.1 - Class 1 System Hydrostatic Test per IWB 5222.  ;

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Po=2235 psig, To=160*F, Test Pressure per IWB 5222 is 2432 psig, j

4. Proposed Alternatiye Examinatiort f

I As an alternative, the Reactor Coolant System will be pressurized to a  !

l pressure as close as practical to 2335 psig but not less than 2300 psig (

j while the reactor is in a shutdown condition to create a pressure

boundary at the first valve of each set listed above. These components

! will then be tested to a pressure (2300 psig < test pressure < 2335 psig)

V as close as practical to the Reactor Coolant System pressuro using a charging pump. The Raactor Coolant System will be borated ; qual to or greater than cold shutdown boron concentration.

I j 5. Reason for Relief J

The first valve listed in each set prevents the components listed above j from being pressurized without pressurizing the Reactor Coolant System.

l The power operated relief valves - (P:V-2456 and PCV 2455C of the Reactor tooknt System) are designed to limit the pressurizer pressure to a value below the fixed high-pressure reactor trip setpoint (2385 psig), lhe 1 relief valve setpoints are 2335 psig whit.h is below the test pressure of 2432 psig. It is not desirable to take the Reactor Coolant System above

the power operated relief valve setpoint.

similar relief was granted to Surry Power Station Unit 2 per Safety j Evaluation Report dated January 24, 1986 (Docket 50 281) for components

with the same design configuration. Relief was granted to North Anna

Power Station Unit 1 per' Safety Evalustion Report dated July 13, 1987 (TAC 3,

No. 64718) for the same situation as described above.

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l RELIEF R,EgUEST SPT REQUEST TO TEST AT LOWER PRESSURE j i 1. System - Safety Injection J

2. Components - Piping and valves listed below and located on drawings i 1 12050-lh-96A and 12050-FM-96B.  ;

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I Valve Connecting Line Valve _

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HOV-2890C and MOV-2890D 10"sI-418-1502/10"-SI-624-1502 2-SI-91  ;

- to 6"-SI-531-1502 2-SI-105  !

to 6"-SI-532-1502 2-SI-99 I

, to 6"-SI-533-1502  !

MOV-2890A 10"-SI-415-1502 2-SI-112 to 6"-SI-416-1502 2-SI-117 l to 6"-SI-419-1502 to 6"-SI-530-1502 2-SI-124  :

I to 6"-SI-421-1502 i

! HOV-2890B 10"-SI-540-1502 to 6"-SI-421-1502 2-SI-124 i

i 2-S1-89 2"-SI-4S1-1502 2-SI-90 t

! 2-SI-97 2"-SI-453-1502 2-SI+98 f i 2-SI-103 2"-SI-455-1502 2-SI-104  ;

i l 3. Code RequitemenIs - Class 2 System Hydrostatic Test per IWC-5222. P" d I

2485 psig, Design Temperature is it'w than 200*t, Test pressure is 2733.5 [

! psig. .

4, Proposed Alternative Examination [

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! As an alternative, the Reactor Coolant System will be pressurized to a I

pressure as close as practical to 2335 psig but not less than 2300 psig I while the reactor is in a shutdown condition ts create a pressure boundary at check valves 2-SI-92, 2.SI-100 2-SI-106, 2-SI-125, 2-SI-113 and 2-SI-118. These components will then be tested to a pressure (2300 psig <

test pressure s' 2335 psig) as close as practical to the Reactor Coctant f System presture us!ng a tert pump.

5. _Reaspn for l'.slie f i

Check valves 2-SI-92, 2-SI-100, 2-SI-105. 2-SI-125, 2-SI-113 and 1-SI-118

! prevent the components listed above from being pressurized without

! pressurizing the Reactor Coolant System. The Code required test pressure of 2733.5 psig will overpressur2:e the Reactor Coolant System.

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The power operated relief valves (PCV-2456 and PCV-2455C) of the Reactor Coolant System are designed to limit the pressurizer pressure to a value below the fixed high-pressure reactor trip setpoint (2385 psig). The relief valve setpoints are 2335 psig which is below the test pressure of 2733.5 psig. It is not desirable to take the Reactor Coolant System

, above the power operated relief valve setpoint.

Similar relief was granted to Surry Power Station Unit 2 per Safety

, Evaluation Report dated January 24, 1986- (Docket 50-281) for components with the same design configuration. Relief was also granted to North Anna Power Station Unit 1 per Safety Evaluattor, Report dated July 13, 1987 (TAC No. 64718) for the same situation as described above.

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REllEF RE0 VEST SPT RE0 VEST TO TEST AT LOWER PRESSVRE

1. System - Safety injection  ;

2. Comoonent - Piping between the sets of valves listed below located on drawing 12050 FM 96B.

Valves Ling ,

MOV-2865A and 2-51-151 12"-S1-523-1502 2-S1-151 and 2-SI-149 3/4"-S1-478-1502 MOV 28658 and 2 S1-168 12" SI-524-1502 2-S1-168 and 2-S1-166 3/4" S1-484 1502 MOV 2865C and 2-51 185 12" S1-525-1502 2-51-185 and 2-51-183 3/4"-51 480-1502

3. Code Reoutrements - Class 2 System Hydrostatic Test per IWC-5222. P a-2485 psig I <200*F, Test pressure per the code is 2733.5 psig since thsre is d

no over-pressure protection for the above components.

4. Proposed Alternative Examination ,

As an alternative it it requested that the Class 2 components listed above be tested per IWB 322". The nominal operating pressure is 660 psig

. and temperature is 120*F. Thus, testing per IWB-5222 would require a i test pressure of 724 psig. This should be adequate considering the j nominal operating conditions.  ;

5. Reason for Relief theck valves 2 51-151, 2-51-168, and 2-51-185 at the class 1 and 2 system '

boundaries prevent the pressurization of the above components without pressurizing the primary system. The required test pressure is 2733.5 psig as stated above, which would over-pressurize the primary system.

Relief was granted to North Anna Power Station Unit I per Safety Evaluation Report dated July 13, 1987 (TAC No. 64718) for the same situation as described above.

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F RELIEF REQUEST SPT REQUEST TO TEST IN ACC0& DANCE WITH WESTINGHOUSE TECHNICAL MANUAL

1. System - Main Steam (Steam) Decay Heat Release feedwater Chemical Feed Blowdown

2. Components - Steam generator and piping located on drawings 12050-FM-708, 12050-F?%74A, 12050-FM-898, 12050-FM-98/., and 12050-FM-102A, t

Component Connected Piping Component 2-RC-E-1A 32"-SHP-401-601 SV-MS-201A to 32"-SilP-422 601 SV-MS-202A SV-MS-203A SV-MS-204A SV-MS-205A i

to 6"-SHP-437-601/1"-SHP-484-601 PCV-MS-201A to 3"-SHP-464-601/1"SHP-478-601 2-MS-335 2-MS-18 to 1 1/2"-SHPD-406-601 2-MS-22 to 1/2"-SHPD-471-601 2-MS-26 2-RC-E-1A 32"-SHP-401-601 't-MS-35 2-NRV-MS-201A j 3"-SHP-460-601 2-MS-344 1

2-RC-E-1A 32"-SdP-401-601 to 32"-SHP-422-601

! to 3"-5HP-445-601 2-MS-344 to 3"-SHP-562-601 NRV-MS-203A to 1"-SHP-571-601/1"-SHP-555-601 2-MS-346 2-MS-348 2-RC-E-1A 2"-55-620-601/1"-55-753-601 2-55-83 2-RC-E-1A 32"-SHP-401-601 to 32"-SHP-422-601 to 3"-SDHV-401-601 i to 4"-SDHV-404-601 2-MS-20 i

2-RC-E-1A 16"-WFPD-424-601 2- FW-62 l 2-FW-70 to 3/4"-CFPD-401-902 2-W1-42 i

l 2-RC-E-1A 2"-WGCB-404-601 2-B0-4 2"-WGCB-405-601 2-80-1 7

1"-WGCB-406-601 2-B0-2 i

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COMPONENTS CONNECTED PIPING COMPONENTS, 2-RC-E-1B 32"-SHP-402-601 SV-MS-201B to 32"-SHP-423-601 S V-MS-202B SV-MS-203B SV-MS-204B SV-MS-205B to 6"-SHP-438-601/1"-SHP-485-601 PCV-MS-201B to 3"-SHP-465-601 2-MS-333 2-MS-57 to 1 1/2"-SHPD-408-601 2-MS-60 to 1/2"-SHPD-473-601 2-MS-64 2-RC-E-1B 32"-SHP-402-601 2-MS-73 NRV-MS-201B 3"-SHP-461-601 2-MS-353 2-RC-E-1B 32"-SHP-402-601 to 32"-SHP-423-601 to 3"-SHP-446-601 NRV-MS-203B to 3"SHP-461-601 2-MS-353 to 3"-SHP-563-601 2-MS-356 to 1"-SHP-503-601 2-MS-357 2-RC-E-1B 2"-S5-625-601/1"-55-754-601 2-SS-84 I

2-RC-E-1B 32"-SHP-402-601 to 32"-SHP-423-f.01

to 3"-SDHV-402-601 to 4"-SDHV-404-601 2-MS-20 2-RC-E-1B 16"-WFPD-423-601 2- FW-102 i to 3/4"-CFPD-402-902 2-FW-94 2-WT-54 2-RC-E-1P 2"-WGCB-407-C01 2-80-13 2"-WGCB-408-601 2-80-10 1"-WGCB-409-601 2-BD-11 l

l 2-RC-E-1C 32"-SHP-403-601 to 32"-SHP-424-601 SV-MS-201C SV-MS-202C SV-MS-203C l^

SV-MS-204C SV-MS-205C to 6"-SHP-439-601/1"SHP-486-601 PCV-MS-201C to 3"-SHP-466-601 2-MS-95 2-MS-331 to 1 1/2"-SHPD-407-601 2-MS-98 to 1/2" - SHPD-475-601 2-MS-102 2-RC-E-1C 32"-SHP-403-601 2 ?S-111 NRV-MS-201C

COMPONENTS CONNECTED PIPING COMPONENTS 2-RC-E-1C 3"-SHP-462-601 2-MS-362 32"-SHP-403-601 to 32"-SHP-424-601 '

to 3"-SHP-447-601 2-MS-362 to 3"-SHP-462-601 to 3"-ShP-564-601 4

to 1"-SHP-504-601/1*-SHP-557-601 NRV-MS-203C 2-MS-365 2-MS-366 2-RC-E-1C 2"-SS-627 601/1"-SS-755-601 2-55-85 2-RC-E-1C 32"-SHP-403-601

, to 32"-SHP-424-601 to 3"-SDHV-403-601 to 4"-SDHV-404-601 2-MS-20 l i 2-RC-E-1C 16"-WFPD-422-601 2-FW-126 to 3/4"-CFPD-403-902 2- FW-134 1 2-WT-70 4

2-RC-E-1C 2"-WGCB-410-601 2-BD-22 2"-WGCB-411-601 2-80-19 1 1"-WGCB-412-601 2-80-20 }

3. Code Requirements - Cl&ss 2 System Hydrostatic Test per IWA-5213(d) and IWC-5222. For Feedwate* componentsaP =1100 psig, T -200-F, test pressure i per IWC-5222 would be 1375 psig. FortheChemical/NedComponentsP=1775 g i psig, T g-200-F, test pressure per IWC-5222 would be 1952.5 psig. The

remaining components have P =1085 psig, T d-200-F, test pressure per IWC-5222 would be 1356 psig. d l

4. Proposed Alternative Examination The Westinghouse Technical Manual for the Stean Generator requires the secondary side to be pressurized to 1356 psig, held for 30 minutes and then reduced to the design pressure (1085 psig) for a sufficient time to permit proper examination of welds, closures and surfaces for leakage or weeping.

The secondary side will be held at 1356 psig for 30 minutes and then at 1085 psig for a minimum of 3 1/2 hours in accordance with the Code. A VT-2 examination will then be perforced.

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5. Reason far Relief Westinghouse, the manufacturer of the steam generators, gives specific testing requirements for the steet generator which must also be applied to the components listed above due to the fact that the components cannot be isolated from the steam generators.

Relief was granted to North Anna Power Station Unit 1 per Safety Evaluation Report dated July 13, 1987 (TAC No. 64718) for the same situation as described above.

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l REllEF RE0 VEST SPT-ll - RE0 VEST TO TEST AT LOWER PRES 3VRE

1. Sylt_qm - Componeat Cooling, Chemical and Volume Control, Fuel Pit Cooling, Safety Injection, Quench Spray, Recirc Spray, Service Water, and Sampling.

2. Components -

Piping and components included in the system hydrostatic test boundary.

3. Code Reauirement -

Per IWA-5265(b)..."the imposed pressure on any component, including static head, will not exceed 106% of the specified test pressure for the system."

4. ProDosed Alternative Examination Hydrostatic testing of systems inat cannot be isolated to meet the system test pressure at the test boundary high point and the 106% system test pressure maximum at the test boundary low point shall be conducted by pressurizing to the system test pressure at the low point in the test boundary.

5. Reason for Relief Unisolable portlans of the various systems within the system hydrostatic test boundary e located throughout the plant such that there are variations in elevation within the boundaries that would result in imposed pressure in excess of six percent of the specified test pressure. It is Virginia Electric and Power Company's desire to limit the test pressure imposed on system components to 106% of the specified test pressure (as required by paragraph IWA-5265(b)). Thus, due to the effects of static head, portions of the piping at higher elevations will be subjected to a test pressure lower than that specified. There is no practical method for isolating the piping segments to achieve the required test pressure at all elevations.

In a Safety Evaluation Report on Duane Arnold Energy Center (Docket No.

50-331) dated March 31, 1986, relief was granted from IWA-5265(b) for situstions as described above. Relief was also granted to North Anna Power Station Unit 1 per Safety Evaluation Report dated July 13, 1987 (TAC No. 64718) for the same situation as described above.

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l REllEF RE0 VEST SPT RE0 VEST TO TEST AT LOWER PRESSURE

1. System - Residual Heat Removal

2. Component - Piping located on drawing 12050-FM 94A between valves listed l below.  ;

Valves Ling MOV 2701 and MOV-2700 14" RH-401-1502

3. Code Reouirements Class 1 System Hydrostatic Test IWB-5222. Po=2235 psig, 10 650 degrees F. Test pressure per IWR 5222 is 2280 psig.

4. Proposed Alternative Examination As an alternative, the components listed above will be tested in accordance with IWC-5222. The test pressure will be 584 psig as determined by the setpoints of relief valves RV-2721A and RV-27218 (467 psig). This alternative is considered sufficient since the relief valves are set at 467 psig. As a result, lin? 14"-RH 401-1502 should not see a pressure significantly higher than 467 psig. In addition, MOV-2700 and MOV-2701 will not open if the Reactor Coolant pressure is >660 psig.

5. Reason for Relief During the system hydrostatic test of the primary system, MOV-2700 is closed in addition to MOV-2701 in order to prevent possible over-pressurization of the Residual Heat Removal System. Thus, the portion of the RHR system identified above cannot be pressurized with the primary system and due to design, it cannot be pressurized without opening one of the M0V's.

Relief was granted to North Anna Power Station Unit 1 per Safety Evaluation Report dated July 13, 1987 (TAC No. 64718) for the same situation as described above.

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l ATTActiMENT If Attached are the requests for relief from impractical Code requirements as stated in Articles IWB and IWC of the 1974 Edition of ASME Section XI with addenda through the Summer 1975 Addendum. Ex'ent of coverage obtainable is not presently available for these components. This information will not be known until the components are exainined in the future.

A process of identifying partial examination coverage will be instituted at North Anna Power Station. T h i *., process will involve alternate angle and alternate method evaluation. Once an examination of these components has been performed, the process will identify the total amount of coverage on all components in n:hich 1007. of the code required volume is not examined.

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Attachment !!

Page 1 of 13 RELIEF REQUESTS F0R >

NDE EXAMINATIONS ,

1

Re1ief Reauesi Descrintion Page I NDE-1 Bl.18, B 0 - Control rod drive housings 2 1

NDE-2 Bl.3, B-C - Head to flange weld 3 i

s NDE-3 B2.1, B B Shell to head weld ~4 l

NDE 4 B2.4, B F - Nozzel-to-safe end welds 5

' NDE 5 B2.9, B I 2, and B3.8, B-I Cladding 6 i  !

NDE-6 B5.6, B L-1, and B5.7, B L Pump Casing 7 i welds and pump casing

) NDE-7 B6.7, B-M 2 - Valve bodies 9 ;

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NDE-8 Cl.1, C-A - Head to-flange weld 10 i NDE-9 C1.1, C-A - Shell-to head weld and Shell-to- 11 l flange weld.  :

NDE ID C1.1, C-A - Shell-to-flange weld

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l NDE-Il C3.1, C G - Pump casing welds 13 I

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Attachment II Page 2 of 13 REllEF RE0 VEST NOE-1

1. IDENTIFICATION OF COMPONENTS Control Rod Drive Housing Welds

11. IMPRACTICAL CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code 1974 Edition, through the Summer 1975 Addendum, Table IWB-2500 states that tiie examination areas shall include essentially

• 4. of the weld metal and base metal for one wall thickness beyond the edge of the weld in the installed peripheral control rod drive housings only.

!!!. BASIS FOR RELIEF Several of the peripheral housings are not accessible for ultrasonic examinations per the requirements of IWB-2500 due to the closeness of the non removable insulation around and under the reactor vessel and instrumentation penetrations.

IV. ALTERNATE EXAMINATION Housings on tem inner portion of the head which are accessible will be substituted for the peripheral housings which are not accessible.

A volumetric examinatior, will be performed to the substituted housings on the inner portion of the head.

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9 Attachment 11 oa ge 3 of 13 RELIEF RE0 VEST NDE-2

1. IDENTIFICATION OF COMPONENTS Reactor Vessel (2 RC-R-1): Head To-Flange Weld II. IMPRACTICAL CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition through the Summer 1975 Addendum, requires the Category B C reactor vessel head-to-flange weld have a volumetric examination each inspection interval in accordance with subsection IWB-2500. A full volumetric examination is not practicable.

111. BASIS FOR RELIEF The geometric configuration of the reactor vessel head-to-flange weld limits the extent to which ultrasonic examinations can be performed from the flange side of the weld.

IV. ALTERNATE EXAMINATION A volumetric examination will be performed to the extent practicable.

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Attachment 11 Page 4 of 13 RELIEF RE0 VEST NDE-3

1. IDENTIFICATION OF COMPONENTS Pressurizer (2-RC-E-2): Shell-To Head Wold

11. IMPRACTICAL CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition through the Summer 1975 Addendum, requires the pressurizer shell-to head weld of Category B B have a volumetric examination each inspection interval in accordance wil.h Subsection IWB-2500. A full volumetric examination is not practicaule.

111. BASIS FOR RELIEF Examination of the vessel shell to head weld is limited by the vessel configuration. The vessel configuration restricts the scan and prevents complete examination of the volume as required by Table IWB 2500.

IV. ALTERNATE EXAMINATION A volumetric examination will be performed to the extent practicable.

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Attachment 11 i Page 5 of 13 1

RELIEF REQUEST NDE-4

I. IDENTIFICATION OF COMPONENTS I

Pressurizer (2 RC E-2)

Nozzle-To Safe End Welds I

!!. IMPRACTICAL CODE REQUIREMENTS

l Section ;*1 of the ASME Boiler and Pressure Vessel Code 1974 Edition

) through th6 Summer 1975 Addendum, requires the Category B F pressurizer nozzle to-safe end welds have a surface and volumetric examination each 3

inspection interval in accordance with subsection IWB 2500. A full

, volumetric examination is not practicable.

111. BASIS FOR RELIEF j Examination of the pressurizer nozzle-to safe end welds is limited by

the geometry and surface condition of the nozzle. The physical geometry l of the nozzle-to-safe end weld permits ultrasonic examination of only

the weld and the pipe side of the weld.

IV. ALTERNATE EXAMINATijk l

l The surface examination will be performed in accordance with the Code t

requirements and a volumetric examination will be performed on the weld i and the base metal on the pipe side of the weld to the extent l

practicable.

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Attachment !!

Page 6 of 13 RELIEF RE0 VEST NDE-5

1. IDENTIFICATION OF COMPONENTS Pressurizer (2 RC-E-2) and Steam Generators (2 RC-E-1A, 2-RC-E-18, and 2 RC-E-lC): Interior Clad Surfaces II. IMPRACTICAL CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition through the Summer 1975 Addendum, requires the steam generator and pressurizer interior clad surfaces of Category B 1-2 have a visual examination of 100% of the clad patch areas each inspection interval in accordance with subsection IWC 2500. A visual examination is not practicable.

111. BASIS FOR RELIEF Subsequent edition and addenda to the ASME Code, which have been approved by the NRC for incorporation into 10 CFR 50.55a, have deleted the cladding examination.

Recognizing this deletion and the intent of the ASME Section XI examination to provide monitoriig of component degradation over the plant's service interval, it is our position that the radiation exposure and cost associated with the cladding examinations are not commensurate with the increase in safety realized. The clad examination results obtained during the first inspection interval will not be directly comparable to examination results in later intervals.

Based on the preceding factors, we request relief from the remaining Pressurizer and Steam Generator cladding examinations at North Anna Power Station Unit 2 during the first ten year inservice inspection interval.

IV. ALTERNATE EXAMINATION i

No additional alternate examinations: later editions and addendum of ASME Section XI, approved by the NRC and incorporated into 10 CFR 50.55a, no longer require cladding examinations.

Relief was granted tc North Anna Power Station Unit 1 per Safety Evaluation Report dated July 13, 1987 (TAC No. 64718) for the same situation as described above.

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Attachment 11 Page 7 of 13 i

i RELIFF REQUEST NDE 6 l

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!. IDENTIFICATION OF COMPONENTS Reactor Coolant Pump (2 RC P 1A, 2 RC P 1B and 2 RC-P lC)

Pump casings  !

I and pressure retaining welds in pump casings j i

II. IMPRACTICAL CODE REQUIREMENTS )

Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition i

. through the Summer 1975 Addendum, requires a volumetric examination be t

> performed during each inspection interval on 100% of the pressure  :

] retaining welds of Category B L 1 in at least one pump in each group of  :

1 pumps performing similar functions in the system (e.g., reactor coolant l l pumss) and a visual examination of one pump casing of Category B L-2 in t

eac1 group uf pumps performing similar functions in the system be j performed.  !

1 III. BASIS FOR RELIEF l l

The North Anna Power Station Unit 2 reactor coolant pumps are Westinghouse Model ^33 controlled leakage pumps. The Model 93 pump's ,

j casing is fabricated by welding two stainless steel castings together. ,

a Thus, there is one circumferential pressure boundary weld in the pumps  !

that is to be examined in accordance with Category B-L-1. -

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s i Since the installation of these pumps, it has been recognized that a j

! volumetric examination of the casing welds is not practical with today's t ultrasonic techniques.  !

) The physical properties of the stainless steel casting and weld j l

material preclude a meaningful ultrasonic examination. The capability l 1 to examine these pump casing welds in the field did not exist until l recently. In the spring of l'a!, an examination was performed on one of l the reactor coolant pumps at 4.he R.E. Ginna plant using the miniature I l

linear accelerator (MINAC), which was built under an EPRI sponsored '

program. This equipment has been made available to other utilities, and currently constitutes the only viable examination method for the  !

volumetric examination of reactor coolant pump welds. l The volumetric examination method is radiographic and is performed by i placing the MINAC inside the pump casing and placing film on the outs'.rie i of the pump. To perform the examinattoa, the pump must be completely disassembled, including removal of the diffuser adapter. This amount of j

! disassembly is far beyond the amount of disassembly performed for normal  ;

i maintenance. Insulation must also be removed from the exterior of the '

} pump casing.

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t Attachment II Page 8 of 13 The examination has been performed at four different sites, all of which I

, have the Westinghouse Model 93 pump. The MINAC examination was performed 4

at Ginna in the spring of 1981, at Point Beach Unit 1 in the fall of  ;

1981, at Turkey Point Unit 3 early in 1982 and at H. B. Robinson Unit 2 .

later in 1982. No problems with the welds were found at any of the  !

sites. A review of the original radiographs of the Point Beach Unit 1 ,

pump was performed prior to the MINAC examination, and all the lanomarks '

found were identified during field examination with no apparent change.

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.ne successful performance of this volumetric examination using the MINAC

! at four different sites demonstrates that the method is capable of satisfying ASME Section XI examination requirements. However, the per-formance of the exar.ination has shown there is a relatively high radia- ,

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tion exposure associated with it. The total exposure associated with 4

insulation removal, disassembly, examination and reassembly of the pump i has averaged about 40 man-rem per pump.  ;

, i There have been no defects identified by the four examinations performed

on these pumps to date. A volumetric examination has attempted at North j Anna in 1982. A radioactive source was placed within the pump casing and j film around the outside. The developed film did not meet the density 1

requirements for an acceptable examination. This examination was i attempted twice at Surry. Both examinations yielded similar results.

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I The pump casing examinations are also not justified from a cost / benefit l perspective. The pump disass?mbly, examination and reassembly is esti-  !

mated to cost $750,000.  ;

IV. ALTERNATE EXAMINATION i

A visual examination of the external surfaces of one pump's casing weld and a surface examination to the extent practicable of the external  !

casing weld of one pump will be performed to the extent and frequency of l Category B-L-1, I Relief was granted to North Anna Power Station Unit 1 per Safety ,

i Evaluation Report dated July 13, 1987 (TAC No. 64718) for the same i situation as described above.  ;

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J Attachment !!

Page 9 of 13  ;

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t RELIEF REQUEST NDE-7 j

!  !. IDENTIFICATION OF COMPONENTS I

! Class ! Valve Bodies Exceeding 4 in, Nominal Pipe Size [

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11. IMPRACTICAL CODE REQUIREMENTS l

In Class I systems, valves which are greater than four inches nominal  ;

d pipe size are subject to visual examination. These valves vary in  :

size, design and manufacturer but are all manufactured from either cast  !

stainless steel or carbon steel. None of the valve bodies are welded. i f

! Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition .

I through the Summer 1975 Addendum, requires a visual examination be  !

perfnrmed on the internal pressure boundary surfaces of one valve in 5 each group of valves that are of the same constructional design, manufacturing method, and manufacturer and that perform similar j functions in the system (Category B M 2). l t

Since these examinations must be met whether or not the valves have to  !

{ be disassembled for maintenance, this requirement is considered t j

impractical. L

!!!. BASIS FOR RELIEF j The renuirement to disassemble primary system valves for the sole i purpose of performing a visual examination of the internal pressure i boundary surhees has only a very small potential of increasing plant

safety margins and a very disproportionate impact un expenditures of j plant manpower and radiation exposure, i

l The perforsance of both carbon and stainless cast valve bodies has been j excellent in PWR applications. Based on this experience and both industry and regulatory acceptance of these alloys, continued excellent service performance is anticipated.

! A more practical approach that would essentially provide t.n equivalent i sampling program and significantly reduced radiation exposure to plant j personnel is to examine the internal pressure boundary of only those valves that require disassembly for maintenance purposess This would still provide a reasonable sampling of the primary system valves and give adequate assurance that the integrity of these components is being maintained.

IV. ALTERNATE EXAMINATION The visual examination of the internal pressure boundary surfaces will be performed, to the extent practical, when a valve is disassembled for maintenance purposes.

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i Attachment !!

Page 10 of 13 REllEF RE0 VEST NDE-8

!. IDENTIFICATION OF COMPONENTS Excess Letdown fl eat Exchanger (2-CH E-4): Head-To Flange Weld

!!. IMPRACTICAL CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code,1974 Edition through the Summer 1975 Addendum, requires the excess letdown heat exchanger head to finnge weld of Category CA have a volumetric examination each inspection interval in accordance with. subsection

> IWC-2600. A full volumetric examination is not practicable.

III. BASIS FOR RElitF The configuration of the excess letdown heat exchanger head-to flange weld is such tFat a full vo'lumetric examination is impractical. The flange side slopes and the head side contains 4 nozzles such that the required volume as described in Tablo IWC-2520 is not practicable.

IV. ALTERNATE EXAMINATION A volumetric examination will be performed to the extent practicable.

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Attachment 11 Page 11 of 13 BELIEF RE0 VEST NDE-9

1. IDENTIFICATION OF COMPONENTS Non Regenerative Heat Exchanger (2 CH E 2): Shell-To-Head Weld and Shell-To F1&nge Weld

!!. IMPRACTICAL CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition through the Summer 1975 Addendum, requires the non regenerative letdown heat exchanger shell-to head wcld and shell to flange weld o' Category C-A have a volumetric examination each inspection interval in accordance with subsection IWC-2500. A full volumetric exaniination is not practicable.

III. BASIS FOR RELIEF Examination of the vessel shell to-head weld is limited by the intergral attachments and the shell to-flange weld is limited by the slope of the flange. The integral attachments and slope of the flange restrict the scan and prevent co2plete examination of the volume as required by Table IWC-2520.

IV. ALTERNATE EXAMINATION A volumetric examination will be performed to the extent practichble.

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Attachment 11 Page 12 of 13 REllEF REQUEST..NQC-1Q

1. IDENTIFICATION OF COMPONENTS Seal Water injection Filters (2 Cll FL 4A and 2-CH FL 48):

Shell-To Flange Welds.

11. IMPRACTICAL CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition through the Summer 1975 Addendum, requires the Category CA shell to-flange welds on the seal water injection filters be volumetrically examined each inspection interval in accordance with subsection IWC 2500. A full volumetric examination is not practicable.

III. BASIS FOR PELIEF The configuration of the seal water injection filter headtoflange weld is such that a full volumetric examination as required by Tab e 3 IWC 2520 is impracticable.

j IV. ALTERNATE PROVISIONS A volumetric examination will be performed to the extent practicable.

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.w Athchment 11  !

Page 13 of 13 ,

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JtELIEF REQUEST ME-11 I

I. IDENTIFICATION OF COMPONENTS l Low Head Safety injection Pumps (2 SI P 1A and 2 SI-P 18): Pump Casing

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Welds .

II. IMPRACTICAL CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code,1974 Edition I through the Summer 1975 Addendum, requires pump casing welds in Category l C G, item Number C3.1, ha e a volumetric examination each inspection interval in accordance with subsection IWC 2500. A volumetric l examination of all of the pumps casing welds is not practicable. L

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!!!. BASIS FOR REllEF -

Each of the two low head safety injection pump casings have a total of five circumferential welds and five longitudinal welds. Three of the circumferential welds and three of the longitudinal welds are completely encased in concrete and are not accessible for examination. Of the ,!

remaining two longitudinal welds, one weld is partially encased in f concrete and one weld is partially covered by a vibratien plate. {

Volumetric examinations can be perfonned on the accessible areas on both j of these longitudinal welds. The remaining two circumferential welds t are accessible for volumetric examinations, j.

IV. ALTERNATE PROVISIONS [

A volumetric examination of the accessible circumferential and longitudinal welds will be performed to the extent and frequency t described in IWC 2500. A remote visual examination of the I.D. of the i pump casing welds will be performed only if the pump is disassembled for  !

maintenance. l i

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