ML18151A702

Forwards Response to Request for Addl Info Re Third Interval ISI Program

ML18151A702
Person / Time
Site:	Surry
Issue date:	11/28/1994
From:	Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
94-634, NUDOCS 9412010189
Download: ML18151A702 (28)

Text

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PRIORITY 1

.ACCELERATED RIDS PROCESSI~G-REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9412010189 DOC.DATE: 94/11/28 NOTARIZED: NO FACIL:50-281 Surry Power Station, Unit 2, Virginia Electric & Powe AUTH.NAME AUTHOR AFFILIATION O'HANLON,J.P.

Virginia Power (Virginia Electric & Power Co.)

RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

DOCKET#

05000281

SUBJECT:

Forwards response to request for addl info re third interval ISI program.

DISTRIBUTION CODE: A047D COPIES RECEIVED:LTR t ENCL SIZE:.;i_3>

TITLE: OR Submittal: Inservice/Testing/Relief from ASM~Code -

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NOTES:

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COPIES RECIPIENT LTTR ENCL ID CODE/NAME 1

0 PD2-2 PD 2

2 6

6 AEOD/SPD/RAB 1

1 NRR/DE/EMCB 1

1 NUDOCS-ABSTRACT 1

0 RES/DSIR/EIB 1

1 LITCO RANSOME,C 1

1 NRC PDR 1

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e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 November 28, 1994 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT 2 Serial No.94-634 SPS/ETS Docket No.

50-281 License No. DPR-37 THIRD INTERVAL INSERVICE INSPECTION PROGRAM ADDITIONAL INFORMATION REQUEST Your letter dated October 20, 1994 requested additional information concerning the Surry Unit 2 third interval ISi program. This additional information is provided in the Enclosure to this letter. Also, as requested in your letter, this information has been transmitted directly to Mr. Boyd W. Brown of INEL Research Center.

Should you have any questions or require additional information, please contact us.

Very truly yours, f?d-d~ 6o-r-James P. O'Hanlon Senior Vice President - Nuclear Enclosure cc:

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

Atlanta, Georgia 30323 Mr. Morris Branch

• NRC Senior Resident Inspector Surry Power Station 9412010189 941128 PDR ADOCK 05000281 Q

PDR 3*00068

ENCLOSURE 1 ADDITIONAL INFORMATION REQUESTED ON SURRY UNIT 2 THIRD INTERVAL ISi PROGRAM NRC Question 2A Address the degree of compliance with augmented examinations that have been established by the NRC when added assurance of structural reliability.is deemed necessary.

Examples of documents that address augmented examinations that may be applicable based on licensee commitments are listed below:

(1)

Branch Technical Position MEB 3-1, High Energy Fluid Systems, Protection Against Postulated Piping Failures in Fluid Systems Outside Containment; (2)

Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and lnservice Examinations;

Response

Section 2.2.6 of our program submittal states that the requirements of Regulatory Guide 1.150 will be followed.

Specific other augmented examinations are scheduled within the program plan in compliance with the station's Technical Specifications or other commitments. A summary of these requirements follow.

1) Reactor Coolant Pump Flywheel inspection (ref. T.S. Table 4.2-1, Item 1.3)

2) Low Pressure Turbine Rotor inspection (ref. T.S. Table 4.2-1, Item 1.4)

3) Sensitized Stainless Steel inspection (ref. T.S. Table 4.2-1, Item 2.1 & 2.2)

4) High Energy Lines inspection (ref. T.S. 4.15)

5) Reactor Vessel In core Detector Thimble Tubes inspection (ref. NRCB 88-09)

6) Steam Generator Feedwater Nozzles inspection (ref. IEB 79-13)

7) Special Lifting Devices inspection (ref. NUREG-0612)

8) Low Pressure Turbine Disc inspection (ref. Letter Serial# 528, dated September 22, 1981 to NRC from R.H. Leasburg) page 1 of 8

NRC Question 28 The Code of Federal Regulations, Part 10, 50.55a(g)(6)(ii)(A), requires that all licensees must augment their reactor vessel examinations by implementing once, during the inservice inspection interval in effect on September 8, 1992, the examination requirements for reactor vessel shell welds specified in Item 81.1 O of Examination Category 8-A, of the 1989 Code.

In addition, all previously granted relief for Item 81.10, Examination Category 8-A, for the interval in effect on September 8, 1992, is revoked by the new regulation. For Licensees with fewer than 40 months remaining in the interval on the effective date, deferral of the augmented examination is permissible with the conditions stated in the regulations.

Based on the effective date of the subject regulation and the May 1 O, 1994, starting date of the third 10-year interval of the Surry Power Station, Unit 2, please provide the staff with the projected schedule for this augmented examination and a technical discussion describing how it will be implemented at Surry power Station, Unit 2, during the third interval. Describe the intended approach and any specialized techniques or equipment that will be used to complete the required augmented examinations. Include an estimate on the percentage of volumetric coverage that can and will be achieved.

Response

Virginia Electric and Power Company will complete the additional examinations required by 1 O CFR 50, 50.55a(g)(6) (ii)(A) during the refueling outage currently scheduled to start on February 4th, 1995. The ultrasonic techniques used for the augmented examinations will comply with the 1989 Edition of ASME Section XI and Regulatory Guide 1.150. Essentially 100% of the required volume is expected to be examined. An automated inspection tool will be used to perform the examinations from the inside surface.

NRC Question 2C Portions of Code Class 2 piping welds in the Residual Heat Removal, Emergency Core Cooling, and Containment Heat Removal Systems are critical to the safe shutdown of the plant. It has been recognized that current Code examination requirements exclude selection of thin-wall piping welds (<3/8 inch) in the subject systems. As a result, flaws in thin-wall piping welds would not be detected until through-wall leakage occurs.

In the review of the Licensee's program, it has been noted that Class 2 piping welds <3/8 inch are included in the total Class 2 piping weld population but are excluded from examinations.

Considering the safety significance of the subject systems, describe your plans for performing volumetric examinations on a sample of thin-wall piping welds to assure the continued integrity of the subject systems.

Response

The Containment Spray (CS) System distribution of selections are located in page 2 of 8

the suction piping as a result of using the Code required.375 wall-thickness limitation. However, weld 1-22 on line 8"-CS-133-153 is scheduled for a PT/UT (surface/volumetric) examination for Surry's Technical Specification augmented inspection requirements (sensitized stainless steel). This weld is located on the discharge side of containment spray pump, 2-CS-P-1A. Additionally one of the welds selected on the suction side of the pumps will be replaced with one selected on the discharge side of 2-CS-P-1 B.

This weld will receive a volumetric and surface examination. Code requirements would be maintained.

The Safety Injection (SI) System includes both suction and discharge selections, however any thin-walled piping excluded because of the Code required.375 wall-thickness limitation in item C5.10.

Two additional welds will be selected on certain thin-walled (<.375 inch) SI piping. These welds will receive a volumetric and surface examination. Code requirements would be maintained.

NRC Question 2D In Request for Relief SR-002, the licensee requested relief from performing a volumetric examination of the steam generator, primary side nozzle inner radius sections. The licensee has deemed the volumetric examination impractical due to factors that include component geometry, long metal paths, and material attenuation. However, in the conclusion the licensee states that to perform an examination, a mockup would be required and training provided to examination personnel.

Many ISi examinations require mockups as well as additional training for examination personnel to meet Code requirements and to ensure a meaningful examination. The Licensee's position that the subject examinations are impractical is not supported by the basis for relief provided. Provide further discussion to support a conclusion that the Code-required examinations are impractical.

Response

It is specified in Request for Relief SR-002 that a full scale mock-up would be necessary to design an inner radius technique and to correspondingly provide appropriate training for examination personnel. We did not intend to imply that a full scale mock-up would ensure a meaningful examination of the Surry steam generator primary nozzle inner radius to be performed. To the contrary, it is believed that the estimated $70,000 required for a mock-up and the exposure anticipated in performing the examination is not justified given the likely indeterminate results that can be expected from the examination.

In order to obtain meaningful results from this examination, one must be able to determine precisely where the ultrasonic beam is directed with respect to the nozzle inner radius. Surface contour variations (i.e., lumpy surface) of the cast steam generator head cause the refracted angle of the ultrasonic beam to vary from point to point around the nozzle. Additionally, the examination must be conducted over a metal path of 12" to 17" which magnifies the beam angle error page 3 of 8

at the inner radius. Given the irregular surface condition and long metal paths, we cannot predict the location of the ultrasonic beam at the inner radius. The uncertainty of the beam angle and low signal-to-noise ratio expected in this material will make evaluation of indications impractical. We expect that non-relevant indications do exist on the inside of the nozzle due to clad geometry and imperfections at the clad interface or within the clad. Because we can not accurately plot the position of the ultrasonic beam, we may not be able to properly determine these indications to be non-relevant. This could result in a required subsequent analysis, or other more extreme actions to resolve indications which are non-relevant. We know of no other nondestructive means that may be employed evaluate the expected non-relevant indications.

Virginia Power performs inner radius examinations where meaningful results are expected or can be obtained. As an example, Surry plans to examine the steam generator feedwater and main steam nozzle inner radius sections, and also the 5 nozzle inner radius sections on top of the pressurizer. Likewise, the North Anna Unit 1 replacement steam generator primary nozzles being made of forged material instead of the cast material used at Surry, are examined on the nozzle inside radius sections, since meaningful results can be obtained.

It is our opinion that the visual (VT-1) specified in Request for Relief SR-002 provides an alternative that will be more effective than the Code required volumetric examination as a means to monitor the integrity of the steam generator primary nozzle.

NRC Question 2E In Request for Relief SR-003, the licensee described the limiting factors for the examination of the pressurizer surge nozzle inside radius section and concluded that any examination of this nozzle would be a "best effort." A "best effort" volumetric examination would provide a level of assurance that flaws are not initiating in the inside radius section. Discuss the extent of Code coverage that can and will be achieved by performing a "best effort" examination.

Response

Please find attached (attachment 1) Westinghouse design drawings as wall as copies of photographs taken of the bottom head of the pressurizer at Surry Unit

2. These drawing and photographs clearly show the limitations to access the surge line nozzle at Surry due to the heater cables and vessel skirt. Due to ALARA concerns, the "best effort" examination referred to in Request for Relief SR-003, was an attempt to perform the examination with the heater cables connected. Upon further consideration, we have determined that it would not be practical to attempt this examination without disconnecting the heater cables.

With the heater cables disconnected, we expect that essentially 100% of the examination volume could be examined.

An estimated total dose of approximately 9 man-rem would be required to remove and reinstall the heater cables and insulation, and to prepare and examine the surge line nozzle inner radius section. We do not believe that the gain in system integrity attained by page 4 of 8

e performing the code required volumetric examination is commensurate with the anticipated exposure required to perform the examination.

The visual (VT-2) examination specified by Request for Relief SR-003 is the only practical examination that may be performed on the pressurizer inner radius. It is not possible to place a radiographic source in this configuration and expect to perform acceptable radiography. An O.D. surface examination would require even more surface preparation with consequential a higher doses to accomplish. A visual (VT-1) examination from the I.D. is considered impractical, since the area in question is covered by a welded retaining basket and the inner radius is partially covered by a thermal sleeve.

It is our opinion that sufficient alternatives to the Code requirements exist to monitor the integrity of the pressurizer surge line nozzle. In addition to the visual (VT-2) specified by Request for Relief SR-003, the integrity of the nozzle is monitored in accordance with Technical Specifications which requires leak rate monitoring of the Reactor Coolant System through calculation and radionuclide measurement. In addition, the 5 nozzle inner radii on the top of the pressurizer will receive a Code examination which provides assurance of integrity from a sampling perspective.

NRC Question 2F In Request for Relief SR-005, the licensee proposed the use of the existing ultrasonic calibration blocks without modifying them to satisfy current Code requirements. It is expected that licensees will meet the requirements for the applicable interval by upgrading calibration block designs. Provide a list of calibration blocks and describe where the existing designs differ from the applicable Code requirements for the third interval. Describe how the existing calibration blocks will provide the same level of examination quality as those designs currently required by the Code in effect.

Response

Please find attached (attachment 2) a list of calibration blocks which are used to perform ultrasonic system calibration for third interval ultrasonic examinations at Surry Units 1 and 2. As indicated by Request for Relief SR-005, some of the calibration blocks differ slightly from the design recommended by the 1989 Edition of ASME Section XI, Article Ill, Figure 111-3230-2 for piping blocks and the design shown by ASME Section V, Article 4, Figure T 441.1 for vessel blocks.

Please note that current recommended block design for piping was first incorporated into the 1986 Edition of ASME Section XI. The following states the specific differences:

1.)

The notches in piping and vessel blocks which are less than 1" thick are not staggered.

2.)

The notches in piping blocks are located at one times the block thickness from the end of the block instead of at 1 1/2" minimum from the end.

page 5 of 8

e Therefore, blocks less than 1 1 /2" thick will be closer than recommended.

3.)

The vessel calibration blocks for the reactor vessel head-to-flange weld, steam generator primary side tubesheet-to-head weld, and pressurizer welds are partially clad instead of fully clad as depicted by Figure T-441.1 of ASME Section V, Article 4.

The differences in piping blocks described by 1. and 2. above will in no way adversely affect the use of the existing blocks. The Code required calibrations can be accomplished with the existing blocks without any effect on sensitivity or resolution. The design of the existing blocks can not be altered to meet the present Section XI requirements without degrading the effective use of the blocks. In addition, the existing piping blocks still meet the current requirements of ASME Section V, Article V for design of piping blocks.

The difference in the vessel blocks described by 3. above will in no way adversely affect the use of the existing blocks. The portion of the block which lies under the hole and notch reflectors is clad and is therefore representative of the component to be examined. Also, the existing vessel calibration blocks have an advantage over the block depicted by Figure T 441.1 in that the 5/4 T calibration can be accomplished directly from the unclad portion of the clad side of the block instead of through the clad as described by Nonmandatory Appendix B, B-22 of Section V, Article 4.

In order to fully comply with the recommendations of Figure 111-3230-2 and Figure T 441.1, it would be necessary to replace 14 piping calibration blocks.

Because replacing the existing blocks will in no way improve the level of examination quality, it is our position that replacement of the existing blocks is unnecessary and therefore impractical.

NRC Question 2G:

Request for Relief 1 addresses the pressure test of the piping between valves MOV-2700 and MOV-2701. The licensee stated that valves MOV-2700 and MOV-2701 are closed for the pressure test of the Class 1 side to avoid overpressurization of the Class 2 side.

Explain why MOV-2701 is not an acceptable pressure boundary for the pressurization of the segment between the subject valves, which is required to be tested at the higher of the operating pressures when implementing Code Case N-498. Describe how pressurization of the Class 1 segment of line between the subject valves at the lower pressure provides the same level of assurance of structural integrity as the Code-required test pressure.

Response

The area in question can only be tested by defeating the pressure interlock, which prevents the opening of MOV-2700 at the pressure required to perform the Class 1 N-498 test (approx. 2235 psig.). This interlock is a design safety function to prevent the overpressurization of the residual heat removal system page 6 of 8

e (600 psig. system). However, since valve MOV-2701 is designed to be a RCS pressure boundary valve. Our position is not that the valve will not perform it's intended pressure boundary function, but that challenging the boundary is unnecessary and potentially puts the residual heat removal system at risk, if internal valve leakage occurs.

It is our position that the Class 2 N-498 test will identify any through-wall pressure boundary leakage and would not result in reduction of assurance of structural integrity.

NRC Question 2H:

In Request for Relief 4, is the licensee proposing to use the lower pressure associated with the auxiliary feedwater pump on the high pressure side and take credit for the pressure test on both sides? If so, how does this lower pressure provide reasonable assurance of component integrity for the high pressure side?

Response

The test pressure used corresponds with the pressure associated with the discharge of the auxiliary feedwater pump. This should be approximately 1200 psig on the upstream side of the pressure reducing orifice. The downstream side of the pressure reducing orifice should have a pressure of approximately 110 psig. A VT-2, visual examination conducted at these pressures should provide reasonable assurance of component integrity and the identification of any through-wall leakage. Complying with the Code prescribed test pressure on this piping would necessitate a design change. This relief request was addressed recently and approved for Surry Unit 2 in the second interval program (reference Relief Request 23, NRC Serial# 94-272, dated April 14, 1994).

NRC Question 21:

Recent incidence of degraded bolting have reinforced the requirement to remove a bolt for a VT-3 visual examination as part of the evaluation process.

(

Reference:

NRC Event 26899, dated 03/08/94, and 26992, dated 03/25/94). Because degradation rates cannot be reliably predicted and bolting material records may not be accurate, direct visual examination and immediate corrective action for leakage at bolted connections is warranted.

For Request for Relief 5, it is not apparent that the licensee intends to remove at least one bolt nearest the source of leakage for a VT-3 visual examination as part of each evaluation. Verify that at least one bolt, closest to the source of leakage, will be removed for a VT-3 visual examination for each leakage occurrence as part of the evaluation.

page 7 of 8

Response

It is our intent to remove at least one bolt as the Code describes.

NRC Question 2J:

For Code Class 1 integral attachment welds to piping, pumps, and valves, the Code does not require examinations for the third and fourth interval when implementing Inspection Program B. Examination of integral attachments in Code Class 2 and 3 systems is required in the third and fourth interval. ASME Code Case N-509 (approved November 25, 1992 by ASME) provides for continued examination of Class 1 integral attachments for the life of the plant as well as readjustments in the sample inspection requirements for Code Class 2 and 3.

Describe your plans with respect to examination of Code Class 1 integral attachment welds during the third inspection interval or implementation of this Code Case.

Response

As indicated in Relief Request SR-009, it is our intent to implement Code Case N-509 pending your approval, which addresses the continued examination of Class 1 integral attachments.

NRC Question 2K:

Verify that no additional requests for relief are required at this time. If additional requests for relief are required, the licensee should submit them for staff review.

Response

Our letter dated October 11, 1994 (Serial No.94-552) added Relief Request SH-1 for Surry Unit 2. This relief request addressed snubber visual inspection frequency.

page 8 of 8

IDENTITY VIR-1A VIR-2 VIR-3 VIR-4 VIR-4A VIR-5 VIR--6 VIR-7 VIR-9 VIR-10 VIR-11 VIR-14 VIR-15 NOMINAL SIZE/SCHEDULE

2. 758'T X 12"L X 3"W 30" 1.1'T 14" SCH 140 1.25'T 14" SCH 80. 750'T 14" SCH 80. 750'T 14" SCH 40.438'T 12" SCH 140 1.125'T 12" SCH 40S.375'T 1 O" SCH 140 1.00'T 1 O" SCH 120.844'T 10" SCH 40S.365'T 6

11 SCH 160.719'T 6

11 SCH 120.562'T SURRY NUCLEAR POWER PLANT UNIT #1 AND #2 CALIBRATION BLOCKS IDENTITY OR HEAT NUMBER PIPING BLOCKS 5160C-1 3G5682 26374-2 L45865 48082 71771 F0959 805222-1 061232 6448 1971-12-1-2 M2060 M9948 MATERIAL SA351 GR CF8A SA515 GR 70 CS SA376 TP 316 SS SA106 GR B CS SA355 GR P22 SA358 TP 316 SS SA312 TP 304 SS SA312 TP 304 SS SA312 TP 304 SS SA312 TP 304 SS SA312 TP 316 SS SA376 TP 304 SS SA376 TP 316 SS COMPONENT /SYSTEM Reactor Coolant Pipe (from ELL Side) 30" Mainsteam Piping 14" Feedwater Piping 14" Feedwater Piping 14" Feedwater (80-Repl)

Piping 14" SCH 40 Piping 12" SCH 140 Piping 12" SCH 40S Piping 1 O" SCH 140 Piping Seal Water Injection Filter 1 O" SCH 40S Piping 6" SCH 160 Piping 6" SCH 120 Piping

J SURRY NUCLEAR POWER PLANT UNIT #1 AND #2 CALIBRATION BLOCKS NOMINAL IDENTITY OR IDENTITY SIZE/SCHEDULE HEAT NUMBER MATERIAL COMPONENT /SYSTEM VIR-16 6

11 SCH 80.432"T N14446 SA106 GR B CS 6

11 SCH 80 Piping VIR-17 31" 1.51' 3G8217 SA515 GR 70 SS 32" Mainsteam Piping VIR-18 4

11 SCH 160.5311' 01038 SA312 TP 304 SS CRDM VIR-19 4

11 SCH 120.4381' M6108 SA376 TP 316 SS 4

11 SCH 120 Piping VIR-20 3

11 SCH 160.4381' N7212 SA376 TP 316 SS 3

11 SCH 160 Piping VIR-21 2

11 SCH 160.3441' 01003 SA376 TP 316 SS 2" SCH 160 Piping VIR-32 6

11 SCH 40S.2801' M9959 A312 TP 304 SS 6

11 SCH 40S Piping VIR-33 8

11 SCH 40S.3221' M0937 A312 TP 316 SS 8

11 SCH 40S Piping VIR-34 16" SCH 80.8441' L21488 SA 106 GR B CS 16" Feeclwater Piping VIR-35 30 11 1.8101' 676344 SA155 KC60 CL 1 CS 30 11 Tee Mainsteam Piping VIR-44 2.3201' X 27.5"1D X 32.14"0D J6954 A376 TP 304 N RC Loop Pipe Block VIR-45F 2.6251' X 31 "ID X 36.25"0D J6959 A376 TP 304 N RC Loop Pipe Block

SURRY NUCLEAR POWER PLANT UNIT #1 AND #2 CALIBRATION BLOCKS NOMINAL IDENTITY OR IDENTITY SIZE/SCHEDULE HEAT NUMBER MATERIAL COMPONENT /SYSTEM VESSEL BLOCKS VIR-8 12" SCH 20.250T 8051726 SA312 TP 304 SS Seal Water Heat Exchanger VIR-13 8

11 SCH 120.719"T M0176 SA312 TP 304 SS Excess Letdown HT /Ex VIR-23 6.2T X 6 11W X 21.8"L B&W SA508 CL 2 RV-Closure Head-to-Flange Weld Piece 1 VIR-24 5.2T X 6"W X 18.2"L B&W SA508 CL 2 Channel Head to Tubesheet Piece 1 VIR-25 4.375T X 6 11W x 15.5"L 08366-5 SA533 GR A CL 1 Pressurizer VIR-26 3.5T X 6 11W X 18.2"L 08366-5 SA533 GR A CL 1 Steam Generator (Sec.

Side)

VIR-29A

.627T x 4'W x 9"L 42204 A240 TP 304 SS Non Regen. Heat Exchanger VIR-30

.313T X 4"W X 9"L 30106 A240 TP 304 SS Volume Control Tank VIR-12 8

11 SCH 160.906T 2626-8-1 SA312 TP 304 SS Regen Heat Exchanger VPNIR4 R4980 SA508 CL2 Steam Generator (MS and FW)

Nozzle Inner Radii

SURRY NUCLEAR POWER PLANT UNIT #1 AND #2 CALIBRATION BLOCKS NOMINAL IDENTITY OR IDENTITY SIZEfSCHEDULE HEAT NUMBER MATERIAL COMPONENT SYSTEM VPNIR5 94-892 A216 GR wee Pressurizer (Surge, Spray and and Safety and Relief)

Nozzle Inner Radii SIZING BLOCKS VP-.25-CS-03

.25" 321-0676 A36 1/4" Sizing Block CS VP-.50-CS-03

.50" 331-0889 A36 1/2" Sizing Block CS VP-. 75-CS-03

.75" 1-57528 A36 3/4" Sizing Block CS VP-1.0-CS-03 1.00" 333-0889 A36 1.0" Sizing Block CS VP-1.25-CS-03B 1.25" 85489 A36 1.25" Sizing Block CS 10%-50%

VP-1.25 CS-03A 1.25" 85489 A36 1.25" Sizing Block CS 60%-90%

VP-1.50-CS-03B 1.50" 60378 A36 1.50" Sizing Block CS 10%-50%

VP-1.50-CS-03A 1.50" 60378 A36 1.50" Sizing Block CS 60%-90%

VP-1.75 CS-03B 1.75" 72674 A36

1. 75" Sizing Block CS 10%-50%

VP-1. 75-CS-03A 1.75" 72674 A36

1. 75" Sizing Block CS 60%-90%

SURRY NUCLEAR POWER PLANT UNIT #1 AND #2 CALIBRATION BLOCKS NOMINAL IDENTITY OR IDENTITY SIZELSCHEDULE HEAT NUMBER MATERIAL COMPONENT SYSTEM VP-2.0-CS-03B 2.00" 83427 A36 2.0" Sizing Block CS 10%-50%

VP-2.0-CS-03A 2.00" 83427 A36 2.0" Sizing Block CS

• 60%-90%

VP-.25-SS-03

.25" 130359 A479 TP 304 1/4" Sizing Block SS VP-.50-SS-03

.50" 34214 A479 TP 304 1/2" Sizing Block SS VP-. 75-SS-03

.75" 70777 A479 TP 304 3/4"- Sizing Block SS VP-1.0-SS-03 1.00" 36463 A479 TP 304 1.0" Sizing Block SS VP-1.25-SS-03B 1.25" 28498 A479 TP 304 1.25" Sizing Block SS 10%-50%

VP-1.25-SS-03A 1.25" 28498 A479 TP 304 1.25" Sizing Block SS 60%-90%

VP-1.50-SS-03B 1.50" A13072 A479 TP 304 1.50" Sizing Block SS

• 10%-50%

VP-1.50-SS-03A 1.50" A13072 A479 TP 304 1.50" Sizing Block SS 60%-90%

VP-1. 75-SS-03B 1.75" AH6895 A479 TP 304

1. 75" Sizing Block SS 10%-50%

VP-1. 75-SS-03A 1.75" AH6895 A479 TP 304

1. 75" Sizing Block SS 60%-90%

SURRY NUCLEAR POWER PLANT UNIT #1 AND #2 CALIBRATION BLOCKS NOMINAL IDENTITY OR IDENTITY SIZELSCHEDULE HEAT NUMBER MATERIAL COMPONENT SYSTEM VP-2.50-SS-03B 2.50" A14267 A479 TP 304 2.50" Sizing Block SS 10%-50%

VP-2.50-SS-03A 2.50" A14267 A479 TP 304 2.50" Sizing Block SS

• 60%-90%

VP-3.0-SS-03B 3.00" A14394 A479 TP 304 3.0" Sizing Block SS 10%-50%

VP-3.0-SS-03A 3.00" A14394 A479 TP 304 3.0" Sizing Block SS 60%-90%

BOLTING BLOCKS VIR-36

5. 75" Dia. x 10.5"L 112863 SA540 GR B24 Reactor Vessel Stud 80W80 VIR-37

5. 75" Dia. x 18"L 112863 SA540 GR B24 Reactor Vessel-Stud 80W8o VIR-38

5. 75" Dia. x 33"L 112863 SA540 GR 824 Reactor Vessel-Stud

• 80W8o VIR-39 4.32" Dia. x 18.25"L 112863 SA540 GR 824 Reactor Coolant Pump Stud VIR-40 4.32" Dia. x 10.63"L 112863 SA540 GR 824 Reactor Coolant Pump Stud VIR-41 4.32" Dia. x 6.81 "L 112863 SA540 GR 824 Reactor Coolant Pump Stud

SURRY NUCLEAR POWER PLANT UNIT #1 AND #2 CALIBRATION BLOCKS NOMINAL IDENTITY OR IDENTITY SIZELSCHEDULE HEAT NUMBER MATERIAL COMPONENT SYSTEM NDE-LS-88 2.62" Dia. x 22.S"L 60300 A286 Carpenter Loop Stop Valve Bolt LSVN-04 3.0" Dia. x 22.S"L 60300 A286 Carpenter Loop Stop Valve Stud RCP-03 4.5" Dia. x 30.S"L SA540 GR 823 Reactor Cool Pump

• Bolt RPV-03 6.0" Dia. x 63.25"L X7225 SA540 GR 824 Reactor Vessel Closure Head Stud

BOTTOM 'll~W*

THE HEJITERS ARE TO BE.RSSEMBLED SO THAT iHE SERTFIL,JP. ON HE/JTER.AGREES WITH THE NUMBERING, SE~VENCE SHOWN.

BOTTOM VIEW

.SP,4CE. lltE~'D. F'O" HUTER R!MOV.4L Cf;:NTER OF ~R'.11\\/IT'l'

'HEIGHT C.ONDIT/ON I.,,..

lN LSS, DR'1' l19U3~ 175" 400

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