ML18151A974

Technical Evaluation Rept on Third 10-Yr Interval ISI Program Plan:Virginia Electric & Power Co,Surry Power Station,Unit 1.

ML18151A974
Person / Time
Site:	Surry
Issue date:	02/13/1995
From:	Beth Brown IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML18151A975	List: ML18151A974
References
CON-FIN-L-2556 INEL-94-0164, INEL-94-164, NUDOCS 9502160299
Download: ML18151A974 (53)

Text

e INEL-94/0164 TECHNICAL EVALUATION REPORT ON THE THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:

VIRGINIA ELECTRIC AND POWER COMPANY, SURRY POWER STATION, UNIT 1, DOCKET NUMBER 50-280 B. W. Brown E. J. Feige K. W. Hall A. M. Porter February 1995 Idaho National Engineering Laboratory Lockheed- Idaho Technologies Company Idaho Falls, Idaho 83415 Prepared for the Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Idaho Operations Office Contract DE-AC07-941D13223 FIN No. L2556 (Task Order 29c )

ABSTRACT This report presents the results of the evaluation of the Surry Power Station, Unit 1, Third JO-Year Interval Inservice Inspection Program Plan, Revision 0, submitted July 16, 19Q3, including the requests for ~elief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, requirements that the licensee has determined to be impractical. The Surry Power Station, Unit 1, Third 10-Year Interval Inservice Inspection Program Plan, Revision O is evaluated in Section 2 of this report. The Inservice Inspection (ISi) Program Plan is evaluated for (a) compliance with the appropriate edition/addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISi-related commitments identified during previous Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this report.

This work was funded under:

U.S. Nuclear Regulatory Commission FIN No. L2556, (Task Order 29c)

Technical Assistance in Support of the NRC Inservice Inspection Program ii

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SUMMARY

The licensee, Virginia Electric and Power Company, has prepared the Surry Power Station, Unit 1, Third 10-Year Interval Inservice Inspection Program Plan, Revision 0, to meet the requirements of the 1989 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI. The third 10-year interval began October 14, 1993 and ends October 13, 2003.

The information in the Surry Power Station, Unit 1, Third 10-Year Interval Inservice Inspection Program Plan, Revision 0, submitted July 16, 1993, was reviewed. Included in the review were the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical. As a result of this review, a request for additional information (RAI) was prepared describing the information and/or clarification required from the licensee in order to complete the review. The licensee provided the requested information in the submittal dated September 12, 1994.

Based on the review of the Surry Power Station, Unit 1, Third 10-Year Interval Inservice Inspection Program Plan, Revision 0, the licensee's response to the Nuclear Regulatory Commission's RAI, and the recommendations for granting relief from the ISI examinations that cannot be performed to the extent required by Section XI of the ASME Code, no deviations from regulatory requirements or commitments were identified in the Surry Power Station, Unit 1, Third 10-Year Interval Inservice Inspection Program Plan, Revision 0,

• except those noted in the evaluations of Requests for Relief 1 and 11.

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.e CONTENTS ABSTRACT ii

SUMMARY

. iii

1. INTRODUCTION 1

2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN 4 2.1 Documents Evaluated 4 2.2 Compliance with Code Requirements 4 2.2.1 Compliance with Applicable Code Editions 4 2.2.2 Acceptability of the Examination Sample 4 2.2.3 Exemption Criteria . 5 2.2.4 ~ugmented Examination Commitments 5 2.3 Conclusions 6

3. EVALUATION OF RELIEF REQUESTS 7 3*.1 Class 1 Components . 7 3.1.1 Reactor Pressure Vessel 7 3.1.1.1 Request for Relief SR-001, Examination Category 8-F, Item 85.10, Reactor Vessel Nozzle-to-Safe End Butt Welds . . . . . . . . . . . . . . . . . . . . . 7 3.1.1.2 Request for Relief SR-007, Weld Reference System for The Reactor Vessel and Vessel Nozzle Area 9 3.1.2 Pressurizer 10 3.1.2.1 Request for Relief SR-003, Examination Category B-D, Item 83.120, Pressurizer Surge Nozzle Inside Radius Section . . . . . . . . . . . . 10 3.1.3 Heat Exchangers and Steam Generators . . . . . . 12 3.1.3.1

\

Request for Relief SR-002, Examination Category B-D, Item B3.140, Steam Generator (Primary Side) Nozzle Inside Radius Section 12 3 .1. 4 Piping Pressure Boundary . . . . 15 3.1.4.1 Request for Relief SR-008, Examination Category 8-J, Selection Criteria for Class 1 Piping Welds for Examination ............ . 15 iv

e 3.1.5 Pump Pressure Boundary (No requests for relief) 3.1.6 Valve Pressure Boundary (No requests for relief) 3.1.7 General (No requests for relief) 3.2 Class 2 Components . 17 3.2.1 Pressure Vessels (No requests for relief) 3.2.2 Piping (No requests. for relief) 3.2.3 Pumps 18 3.2.3.1 Request for Relief SR-004, Examination Category C-G, Item C6.10, Pump Casing Welds . . . . . . . . . . . 18 3.2.4 Valves (No requests for relief) 3.2.5 General (No requests for relief) 3.3 Class 3 Components (No tequests for relief) 3.4 Pressure Tests 3.4.1 Class 1 System Pressure Tests 20 3.4.1.1 Request for Relief l, Examination Category 8-P,.Item 815.51, System Hydrostatic Test, as Modified by Code Case N-498, for Class 1 Safety Injection (SI) Piping 20 3.4.1.2 Request for Relief 2, Examination Category B-P, Item B15.51, System Hydrostatic Test, as Modified by Code Case N-498, for Cl ass 1 Residual Heat Removal (RHR)

Piping. . . . . . . . . . . . . . . . . . . 22 3.4.2 Class 2 System Pressure Tests (No requests for relief) 3.4.3 Class 3 System Pressure Tests 23 3.4.3.l Request for Relief 3, IWD-5223, System Hydrostatic Test of Class 3 Circulating and Service Water System Piping Upstream of the First Isolation Valve . . . 23 3.4.3.2 Request for Relief 4, IWD-5223, System Hydrostatic Testing of Class 3 Component Cooling Water System Piping . . . . . . . . . . . . . . . . . . . . . . 25 3.4.3.3 Request for Relief 5, IWD-5223, System Hydrostatic Test for Portions of the Class 3 Servi.ce Water System 25 3.4.3.4 Request for Relief 6, IWD-5223, Class 3 System Hydrostatic Test of the Auxiliary Feedwater System . . 27 V

3.4.3.5 Request for Relief 7, Examination Category 0-A, Item 01.10, Hydrostatic Testing of Class 3 Circulating and Service Water Systems . . . . . . . . . . . . . . . 29 3.4.3.6 Request for Relief 8, Examination Category 0-A, Item 01.10, Hydrostatic Testing of Class 3 Pressure Retaining Components in the Circulating and Service Water System 32 3.4.4 General 34 3.4.4.1 Request for Relief 9, IWA-5214, Hydrostatic Testing of Repairs and Replacements . . . . . . . . . . . . . . . 34 3.4.4.2 Request for Relief 10, IWA-5250(a)(2), System Pressure Test Corrective Measures for Leakage at Bolted Connections 34 3.4.4.3 Request for Relief 11, IWA-5242, System Pressure Tests bf Insulated Bolted Connections 36 3.5 General 38 3.5.1 Ultrasonic Examination Techniques 38 3.5.1.1 Request for Relief SR-005, ASME Section V, Article IV, Figure T-441.1 and Section XI, Appendix III, Figure III-3230-2, Requirements for Ultrasonic Calibration Blocks ........... . *38 3.5.2 Exempted Components (N6 requests for relief) 3.5.3 Other 40 3.5.3.1 Request for Relief SR-006, IWA-2610 Weld Reference System for Class 1 and Class 2 Piping, Vessels, and Components . . . . . . . . . . . . . . . . . . . . . 40 3.5.3.2 Request for Relief SH-1, Examination Categories F-A, F-B, and F-C, Items Fl.IO through F3.50, Component Supports 42

4. CONCLUSION 44

5. REFERENCES 46 vi

TECHNICAL EVALUATION REPORT ON THE THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:

VIRGINIA ELECTRIC AND POWER COMPANY, SURRY POWER STATION, UNIT 1, DOCKET NUMBER 50-280

1. INTRODUCTION Throughout the service life of a.water-cooled nuclear power facility, 10 CFR 50.55a(g)(4) (Reference 1) requires that components (including supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class l, Class 2, and Class 3 meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components (Reference 2), to the extent practical within the limitations of design, geometry, and materials of construction of the components. This section of the regulations also requires that inservice examinations of components and system pressure tests conducted during successive 120-month inspection intervals comply with the requirements in the latest edition and addenda of the Code incorporated by reference in

. 10 CFR 50.SSa(b) on the date 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed therein. The components (including supports) may meet requirements set forth in subsequent editions and addenda of this Code that are incorporated by reference in 10 CFR 50.SSa(b) subject to the limitations and modifications listed therein, and subject to Nuclear Regulatory Commission (NRC) approval.

The licensee, Virginia Electric and Power Company, has prepared the Surry Power Station, Unit 1, Third 10-Year Interval Inservice Inspection Program Plan, Revision 0, (Reference 3), to meet the requirements of the 1989 Edition of the ASME Code Section XI. The third 10-year interval began October lf 1993 and ends October 13, 2003.

As required by 10 CFR 50.55a(g)(5), if the licensee determines that certain Code examination requirements are impractical and requests relief from them, the licensee shall submit information and justification to the NRC to support that determination.

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Pursuant to 10 CFR 50.55a(g)(6), the NRC will evaluate the licensee's determination that Code requirements are impractical to implement. The NRC may grant relief and may impose alternative requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

Alternatively, pursuant to 10 CFR 50.55a(a)(3), the NRC will evaluate the licensee's determination that either (i) the proposed alternatives provide an acceptable level of quality and safety, or (ii) Code compliance would result in hardship or unusual difficulty without a compensating increase in safety.

Proposed alternatives may be used when authorized by the NRC.

The information in the Surry Power Station, Unit 1, Third JO-Year Interval Inservice Inspection Program Plan, Revision 0, submitted July 16, 1993, was reviewed, including the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical. The review of the Inservice Inspection (ISI) Program Plan was performed using the Standard Review Plans of NUREG-0800 (Reference 4), Section 5.2.4, "Reactor Coolant Boundary Inservice Inspections and Testing," and Section 6.6, "Inservice Inspection of Class 2 and 3 Components."

In a letter dated June 8, 1994 (Reference 5), the NRC requested additional information that was required to complete the review of the ISI Program Plan.

The requested information was provided by the licensee in the Virginia Electric and Power Company, Surry Power Station Unit 1, Third Interval Inservice Inspection Program Additional Information Request, dated September 12, 1994 (Reference 6). In this response, the licensee withdrew Request for Relief 4. In a letter dated June 22, 1994 (Reference 7), ** .... ...

licensee withdrew Request for Relief 9, opting to follow Code Case N-416-1.

During a conference call with the licensee on October 19, 1994, the licensee verified compliance with NRC Regulatory Guide 1.14 (Reference 8).

The Surry Power Station, Unit l, Third JO-Year Interval Inservice Inspection Program Plan, Revision 0, is evaluated in Section 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the appropriate 2

edition/addenda of Section XI, (b) acceptability of examination_sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with !SI-related commitments identified during the NRC's previous reviews.

The requests for relief are evaluated in Section 3 of this report. Unless otherwise stated, references to the Code rP.fer to the ASME Code,Section XI, 1989 Edition. Specific inservice test programs for pumps and valves are being evaluated in other reports.

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2.

EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN This evaluation consisted of a review of the applicable program documents to determine whether or not they are in compliance with the Code requirements and any previous license conditions pertinent to !SI activities. This section describes the submittals reviewed and the results of the review.

2.1 Documents Evaluated Review has been completed on the following information from the licensee:

(a) Surry Power Station, Unit 1, Third JO-Year Interval Inservice Inspection Program Plan, Revision 0, (Reference 3); and (b) The September 12, 1994 response to the NRC's request for additional information (Reference 6).

2.2 Compliance with Code Requirements 2.2.1 Compliance with Applicable Code Editions The Inservice Inspection Program Plan shall be based on the Code editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.SSa(b). Based I' I

on the starting date of October 14, 1993, the Code applicable to the . I I

third interval ISi program is the 1989 Edition. As stated in Section 1 of this report, the licensee has prepared the Surry Power Station, Unit 1, Third JO-Year Interval Inservice Inspection Program Plan, Revision 0, to meet the requirements of 1989 Edition of the Code.

2.2.2 Acceptability of the Examination Sample lnservice volumetric, surface, and visual examinations shall be performed on ASME Code Class l, 2, and 3 components and their supports using sampling schedules described in Section XI of the ASME*

Code and 10 CFR 50.55a(b}. Sample size and weld selection have been implemented in accordance with the Code and 10 CFR 50.55a(b) and appear to be correct.

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2.2:3 Exemption Criteria The criteria used to exempt components from examination shall be consistent with Paragraphs IWB-1220, IWC-1220, IWC-1230, and IWD-1220, and 10 CFR 50.SSa(b). The exemption criteria have been applied by the licensee in accordance with the Code, as discussed in the ISI Program Plan, and appear to be correct.

2.2.4 Augmented Examination Commitments In addition to the requirements in Section XI of the ASME Code, the licensee has committed to perform the following augmented examinations:

(a) Reactor vessel examinations in accordance with the requirements of NRC Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations, Revision l, (Reference 9);

(b) Volumetric examination of the reactor coolant pump flywheel high stress areas every 3 years, as well as volumetric and surface examinations with the flywheel removed at IO-year intervals. The volumetric examinatio~ of the high stress areas, required every 3-years, is scheduled in the augmented section of the program plan only during the first period. During a conference call with the licensee on October 19, 1994, it was confirmed that Technical Specification 4.2~1 requires the inspection every 3 years, satisfying NRC Regulatory Guide 1.14 (Reference 8).

(c) Examination of the portions of high energy lines specified in Technical Specification 4.15. For Surry, Unit 1, this specification applies to welds in the-Main Steam and Main Feedwater lines in the Main Steam Valve House; (d) Examinations of the portions of sensitized stainless steel specified in Section B of Technical Specification Table 4.2-1; (e) Surface examination of the loop stop valv~ ~~sc pressurization lines; (f) Volumetric and surface examination of all low-pressure turbine blades and a volumetric ~xamination of the low-pressure turbine disc bore and keyway every five years; (g) Volumetric examination of the hot leg loop stop valve stems every refueling outage and of the cold leg loop stop valve stems every other refueling outage; and 5

(h) Eddy current examination (100%), each refueling outage, of all reactor vessel in-core detector thimble tubes that are in service.

2.3 Conclusions Based on the review of the documents listed above, no deviations from regulatory requirements or commitments were identified in the Surry Power Station, Unit 1, Third JO-Year Interval Inservice Inspection Program Plan, Revision 0.

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3.

EVALUATION OF RELIEF REQUESTS The requests for relief from the ASME Code requirements that the li.censee has determined to be impractical for the third 10-year inspection interval are evaluated in the following sections.

3.1 Class 1 Components 3.1.1 Reactor Pressure Vessel 3.1.1.1 Request for Relief SR-001, Examination Category B-F, Item BS.IO, Reactor Vessel Nozzle-to-Safe End Butt Welds Code Requirement: Examination Category B-F, Item BS.IO requires a 100% volumetric and surface examination of the nozz~e-to-safe end butt welds 4-inch or larger, as defined by Figure IWB-2500-8.

Licensee's Code Relief Request: The licensee requested relief from performing the Code-required surface examinations on the following nozzle-to-safe end butt welds:

Weld# Drawing#

1-0IDM 11448-WMKS-OIOOAZ-l l-17DM 11448-WMKS-OlOOAZ-l 1-0lDM 11448-WMKS-OlOlAZ-l l-17DM 11448-WMKS-OIOIAZ-l 1-0lDM 11448-WMKS-0102AZ-l' l-17DM 11448-WMKS-0102AZ-l Licensee's Basis for Requesting Relief (as stated):

"The outside diameter volumetric examination would be extremely difficult to perform. Access to the area is *;*-,r:tricted by ...

permanent neutron shielding and support structures. Any planned removal to provide access is made difficult by the relatively small sandplug access on the floor of the refueling cavity. The difficult access restrictions are also complicated by the anticipated high dose level. The general area estimate for one nozzle ranges from 150 to 800 MR/HR. The contact estimate for one nozzle ranges from 1300 to 1500 MR/HR (top) and 4500 to 6000 MR/HR (bottom)."

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Licensee's Proposed Alternative Examination (as stated):

"Alternately it is requested that an automated examination from the inside diameter, in conjunction with the vessel examination, be accepted in lieu of the surface examination from the O.D.

This relief request was granted to permit this alternative for the second interval based upon examinati~ns from the I.D. of a Surry calibration block which demonstrated sensitivity adequate to resolve a 5% notch on the O.D. In addi~ion, this ultrasonic technique demonstrated detection of a flaw in a mock-up which was estimated to be eighty percent of the critical flaw as described in 1980 Edition, Winter 1980 Addenda of ASME Section XI, IWB-3000 Acceptance Standards for Flaw.Indications.

"The demonstration of the above ultrasonic technique was witnessed by.Authorized Nuclear Inservice Inspector at the Westinghouse Waltz Mill Calibration Facility and found to be acceptable. This was also demonstrated to the NRC staff on June 24, 1986 (reference NRC letter serial #86-759, dated Nov. 12, 1986)."

Evaluation: The Code requires 100% volumetric and surface examinations for the reactor vessel nozzle-to-safe end butt welds. The licensee proposed volumetric examination of the outside surface in lieu of the Code-required surface examination.

The licensee has demonstrated detection of flaws estimated to be 80% of critical flaw size, on the OD surface using the licensee's ultrasonic examination procedures from the ID. For all future examinations, the licensee will fully document all ultrasonic indications due to geometric reflectors as to position and signal amplitude. Hard copies of indications evaluated will be maintained by the licensee.

To consider the licensee's proposed alternative to the Code-required surface examination acceptable, the ultrasonic examination should include a full-volume examination. This volumetric examination technique will cov~1 the entire volume of the weld and include 1/2-inch of the adjacent base metal.

The licensee's proposed alternative of performing an automated examination from the inside diameter, including the full-volume, should provide an acceptable level of quality and safety.

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Conclusion:

The licensee proposed volumetric examination in lieu of the *surface examination should provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), it is recommended that this alternative examination be authorized as requested, provided a full volume examination is performed 3.1.1.2 Request for Relief SR-007, Weld Reference System for The Reactor Vessel and Vessel Nozzle Area Code Requirement: Section XI, Paragraph IWA-2610, *"Weld Reference ~ystem - General, requires that a reference system be 11 established for all welds and areas subject to surface or volumetric examinations. Each such weld and area shall be located and identified by a system of reference points. The system shall permit identification of each weld, location of each weld center line, and designation of regular intervals* along the length of the weld.

Licensee's Code Relief Request: The licensee requested relief from establishing a weld reference system for the reactor vessel including the reactor vessel nozzle area.

Licensee's Basis for Requesting Relief (as stated):

"The automated tool establishes its reference point using an existing zero reference in the reactor vessel. This point allows the device to repeat examination locations without the necessity of any other reference systems. It accomplishes this by use of an electronic encoder system which provides for sufficient repeatability."

Licensee's Proposed Alternative Examination (as stated):

"The automated vessel tool examinations will continue to establish it's reference system based. upon the existing zero reference. No other system is planned or deemed necessary."

Evaluation: The Code requires a reference system that provides for identification of eacn weld, location of each weld center 9

line, and designation of regular intervals along the length of the weld. For the reactor pressure vessel examination, an automated vessel tool is used, that establishes its own reference system based on the existing references in/on the vessel. This is a valid reference system, that allows for repeatability of inspections. This alternative meets the.intent of the Code and will provide an acceptable level of quality and safety.

Conclusion:

Use of the automated vessel tool provides an adequate and repeatable reference system. Therefore, pursuant to 10 CFR 50.55a(g)(3)(i), it is recommended that this proposed alternative be authorized.

3.1.2 Pressurizer 3.1.2.1 Request for Relief SR-003, Examination Category 8-D, Item 83.120, Pressurizer Surge Nozzle Inside Radius Section Code Requirement: Examination Category 8-D, Item 83.120 requires a 100% volumetric examination of the pressurizer surge nozzle inside radius section as defined in Figure IW8-2500-7.

Licensee's Code Relief Request: The licensee requested relief from performing the Code-required volumetric examination of the pressurizer surge nozzle inside radius section.

Licensee's Basis for Requesting Relief (as stated):

"The Surry Unit 1 pressurizer surge line nozzle is integrally cast into the bottom pressurizer head. The nozzle is located under the pressurizer skirt and is surrounderl hy 78 heater penetrations. Interference from the heater penetrations and heater cables, as well as the location of the nozzle under the pressurizer skirt restricts the access to the nozzle. This limits the examiners ability to manipulate the search unit to examine the nozzle inner radius.

"The only viable ultrasonic technique currently available to examine nozzle inner radii involves the fabrication pf calibration blocks that closely simulate the 0. D. and I. D.

nozzle geometry. This is necessary so that search units can b~

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e produced that will interrogate the inner radius section at precise angles. Also, in order to obtain meaningful results, the nozzle material grain structure must be such that an adequate signal-to-noise ratio can be obtained over a long metal path distan~.

"Integrally cast nozzles contain limitations such as an irregular O. D. profile, a rough surface condition, and an attenuating grain structure. The irre~ular surface condition causes the beam angle to vary from point to point around the nozzle. The attenuating grain structure results in a low signal-to-noise ratio at the nozzle inner radius. Limited access to the nozzle as well as the limitations imposed by the material conditions, area dose rates and the complicated nature of the examination technique would make evaluation of any indications very difficult.

"Any examination on this nozzle could only be described as "best effort", and not commensurate with the anticipated exposure estimate of 9 man-rem to perform this examination."

In the September 12, 1994, response to the NRC RAI, the licensee submitted the following additional information:

"An I.D. visual {VT-1) examination is considered impractical, since the area in question is covered by a welded retaining basket. This basket has only 3/8 inch (nominal) holes, making any penetration into the area of interest extremely difficult.

Additionally, the area is partially covered by a thermal sleeve."

Licensee's Proposed Alternative Examination (as stated):

"A visual (VT-2) examination of the pressurizer surge line nozzle area will be performed during the normally scheduled pressure test (Class 1) each refueling."

Evaluation: The Code-required volumetric examination of the inner radius section of the pressurizer surge nozzle is impractical since such integrally cast nozzles contain limitations such as an ir. ~~ular OD profile, a rough surface condition, and an attenuating grain structure. Limited access to the nozzle, limitations imposed by the material conditions, area dose rates, and the complicated nature of the examination technique would make evaluation of any indications very difficult. Therefore, the Code-required examinatton is impractical for the surge nozzle. To perform the Code-required 11

examination, design modifications and/or replacement of the pressurizer would be required. Imposition of the Code requirement would cause a burden on the licensee. In addition, a visual examination from the I6 is impractical due to interference from the welded retaining basket. However, continued volumetric examination of other Surry, Unit 1 pressurizer nozzle inner radius sections should provide reasonable assurance that unacceptable generic degradation does not exist.

Conclusion:

Based on the above evaluation, the Code-required volumetric examination of the pressurizer surge nozzle inside radius section is impractical to perform at Surry, Unit 1.

Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), it is recommended that relief be granted as requested.

3.1.3 Heat Exchangers and Steam Generators 3.1.3.1 Request for Relief SR-002, Examination Category 8-0, Item 83.140, Steam Generator (Primary Side) Nozzle Inside Radius Section Code Requirement: Examination Category 8-0, Item 83.140 requires a 100% volumetric examination of the steam generator primary side nozzle inside radius section as defined by Figure IWB-2500-7.

Licensee's Code Relief Request: The licensee requested relief from performin~ the Code-required volumetric examination on the following steam generator primary side nozzle inner radius sections:

Mark# Componer~ M Drawing#

1-0lANIR 1-RC-E-lA 11448-WMKS-RC-E-lA.2 1-0lBNIR 1-RC-E-lA 11448-WMKS-RC-E-lA.2 l-02ANIR l-RC-E-18 11448-WMKS-RC-E-18.2 l-02BNIR l-RC-E-18 11448-WMKS-RC-E-18.2 l-03ANIR 1-RC-E-IC 11448-WMKS-RC-E-lC.2 l-03BN1R 1-RC-E-lC 11448-WMKS-RC-E-lC.2 12

e e Licensee's Basis for Requesting Relief (as stated):

"The only viable ultrasonic technique currently available to examine nozzle inner radii involves the fabrication of calibration blocks that closely simulate the O. D. and I. D.

nozzle geometry. This is necessary so that search units can be produced that will interrogate the inner radius section at preci~e angles. When the beam varies it is not possible to locate indications or discriminate between flaws and geometry.

Also, in order to obtain meaningful results, the nozzle material grain structure must be such that a relativity high signal-to-noise ratio can be obtained over the required metal path distance. Additionally, for nozzles with a complex 0. D.

profile, examination personnel need training on the proper placement and manipulation of the search unit.

"Virginia Power has previously assessed the feasibility of performing examinations on the North Anna Unit 2 cast primary nozzle inner radii. Virginia Power performed examinations on a Westinghouse Model 44 channel head which was used to train welders for the North Anna Unit 1 steam generator replacement.

This channel head is made from the same cast material (ASTM 216-WGG) as Model 51 generators which are currently installed in Surry Unit 1. We believe that the general surface profile and acoustic properties are representative of the Surry Unit 1 steam generators ..

"Examination Results

1. "Comparison of Material Noise "Figures A and B on Attachment I' depict the respective responses from notches 4 and 6 from calibration block VPSGINRl. This calibration block was manufactured to examine the North Anna Unit 1 replacement steam generator forged carbon steel primary nozzle inner radii. Both notch responses exhibit a high signal to noise ratio with little evidence of material noise. Figure A on Attachment 2' depicts the material noise from the Model 44 Primary nozzle at the same sensitivity level. At this sensitivity level, there is no evidence of clad roll. The first indication of sound penetration (Attachment 1, Figure B) appeared at 12db above the reference level when evidence of clad roll was

... .:! tected.

2. "The 0.0. surface of the nozzle had an irregular contour which is typical of large cast products. Due to the surface contour, it was necessary to apply a lot of couplant to maintain contact with the examination surface. Therefore, "Attachments were provided in the licensee's submittal, but not included in this report.

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e we could not determine where the sound beam was directed with respect to the inner radius.

"Conclusion "The Surry Unit 1 steam generator primary nozzle inner radii were not designed for ultrasonic examination from the 0. D. The nozzles are integrally cast into the channel head. Therefore, the rozzles contain examination limitations such as an irregular

0. D. profile, a rough surface condition, and an attenuating grain structure. The irregular surface causes the beam angle to change from point to point around the nozzle. The varying beam angle combined with a relatively low signal-to-noise ratio makes evaluation of the results extremely difficult. Furthermore, it would be unduly difficult to design a practical search unit for this configuration. As a result of the above limitations, it is our opinion that a full scale cast mock-up of the nozzle would be necessary to develop an inner radius examination technique, and to correspondingly provide appropriate training for the examination personnel. As such, the Code prescribed examination is deemed impractical."

.Licensee's Proposed Alternative Examination (as stated):

"As an alternative, the areas will be visually (VT-1) examined from the nozzle I.D. using direct or remote methods per the schedule shown in Table IWB-2412-1."

Evaluation: The steam generator primary nozzle sections at Surry, Unit l, are integrally cast with the channel head. This design results in a complex component geometry and rough surface conditions due to the as-cast surface that, along with the excessively long metal path, restrict volumetric examination of the nozzle inside radius sections from the external surface. The steam generator nozzle design, therefore, makes the Code-required examination impractical to perform. To examine the nozzle inside radius sections in accordance with the Code requirements, design modifications would be required.

The licensee has committed to perform a VT-1 visual examination of the nozzle inside radius sections from the inside surface using direct or remote techniques. Any system used for this examination should have a color capability to improve the ability to detect rust on the surface of the stainless steel cladding.

This examination will provide adequate assurance that signHicant 14

e patterns of degradation will be detected, provided a color system is used, and removed or repaired prior to the return of the steam generators to service.

Conclusion:

Based on the above evaluation, the Code-required volumetric examination of the steam generator* nozzle inside radius sections is impractical to perform at Surry, Unit I.

Considering the examinations that are being performed and the impracticality of meeting the Code requirements, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i), provided that any system used for the VT-I visual examination has color capability.

3.1.4 Piping Pressure Boundary 3.1.4.1 Request for Relief SR-008, Examination Category B-J, Selection Criteria for Class 1 Piping Welds for Examination Code Requirement: Examination Category 8-J, Table IWB-2500-1 gives the following nates as requirements in selecting Class I piping welds for examination:

"(1) Examinations shall include the following:

"(b) All terminal ends and joints in each pipe or branch run connected to other components where the stress levels exceed either of the following limits under loads associated with specific seismic events and operational conditions:

"(l) primary plus secondary stress .intensity range of 2.4Sm for ferritic steel and austenitic steel

"(2) cumulative usage factor U of 0.4

"(2) The initially selected welds shall be reexamined during each inspection interval."

15

e Licensee's Code Relief Request: The licensee requested relief from using the Code weld selection requirements l(b). and 2 of Examination Category B-J, Table IWB-2500-1.

Licensee's Basis for Requesting Relief (as stated):

"The second interval selection was based upon the 1974 Edition with Summer 1975 Addenda (74/S75) of ASME Section XI. As a result notes l(b) and 2 cannot be applied without some programmatic additions and modifications. In addition, although stress and utilization calculations exist for Surry Unit 1, no correlation exists with actual weld locations. Total reuse of the second interval plan is not desirable, since even though the 74/S75 requirements were met, distribution of welds selected did not equitably cover certain line functions and designs" Licensee's Proposed Alternative Examination (as stated):

"ISi Class l piping welds will be selected for examination such that 25% of the total number of welds are examined during the interval. The 25% sampling will include terminal ends as they appear on our plant isometrics as no corresponding stress calculations exist (ref. NUREG-0800, BTP 3.6.2-13). Terminal ends will include extremities of piping runs connected to vessels and pumps. Pipe integral attachments that act as rigid constraints to piping motion and thermal expansion shall have the first weld upstream and downstream of the integral attachment selected. In piping runs which are maintained pressurized during normal plant conditions for only a portion of the run, the first normally closed valve is a terminal end. The weld on the high pressure side of the valve will be included in the interval selection. Additionally all branch connections will be selected.

Terminal end welds are identified in the third interval plan by the designation "Terminal End" in the comments column. The welds selected will be evenly distributed based upon line size, line function, and line design to the extent practicable. These selected welds will be examined in future successive inspection intervals to the extent allowed by code editions approved at that time."

Evaluation: The licensee has committed to select Class 1 piping welds such that 25% of the total number of welds will be examined. Because stress data are not available for individual Class 1 piping welds, terminal ends and branch connections (considered to typically be high stress) will be selected for inspection. This commitment eliminates the need to perform the stress calculations required in the selection criteria for 16

e l Class 1 piping welds found in Note l(b). Note 2 requires reexamination of welds inspected during the first interval.

Since the first interval sample was selected based.on the 74 Edition, Summer 75 Addenda, the selection criteria did not include all of the terminal ends and branch connections. The licensee's approach to selection of Class 1 piping welds would provide a sample of welds considered susceptible to inservice degradation and therefore, should provide an acceptable level of quality and safety.

Conclusion:

Stress data for Class 1 piping welds required by the 1989 Edition is not available for weld selection. The licensee's approach to the selection of Class 1 piping welds (terminal ends), has a sound engineering basis and provides an acceptable level of quality and safety. Therefore, pursuant to 10 CFR S0.55a(a)(3)(i), it is recommended that the proposed alternative be authorized.

3.1.5 Pump Pressure Boundary (No requests for relief) 3.1.6 Valve Pressure Boundary (No requests for relief) 3.1.7 General (No requests*for relief) 3.2 Class 2 Components 3.2.1 Pressure Vessels (No requests for relief) 3.2.2 Piping (No requests for relief) 17

3.2.3 Pumps 3.2.3.l Request for Relief SR-004, Examination Category C-G. Item C6.10, Pump Casing Welds Code Requirement: Examination Category C-G, *item C6.10 requires a 100% ID or OD surface examination of the Class 2 pump casing welds as defined in Figure IWC-2500-8.

Licensee's Code Relief Request: The licensee requested relief from performing a Code-required surface examination on the following pump casing welds in the outside recirculation spray

{RS} and safety injection {SI} pumps:

ComRonent Weld Drawing#

l-RS-P-2A 2-01 11448-WMKS-RS-P-2A l-RS-P-2A 2-02 11448-WMKS-RS-P-2A l-RS-P-2A 2-03 11448-WMKS-RS-P-2A l-RS-P-2A 2-04 11448-WMKS-RS-P-2A l-RS-P-28 2-01 11448-WMKS-RS-P-28 l-RS-P-28 2-02 11448-WMKS-RS-P-28 l-RS-P-28 2-03 11448-WMKS-RS-P-28 l-RS-P-28 2-04 11448-WMKS-RS-P-28 1-SI-P-lA 2-01 11448-WMKS-SI-P-lA 1-SI-P-lA 2-02 11448-WMKS-SI-P-lA 1-SI-P-lA 2-03 11448-WMKS-SI-P-lA 1-SI-P-lA 2-04 11448-WMKS-SI-P-lA l-SI-P-18 2-01 11448-WMKS-SI-P-lB l-SI-P-18 2-02 11448-WMKS-SI-P-18 l-SI-P-18 2-03 11448-WMKS-SI-P-18 1-SI-P-lB 2-04 11448-WMKS-SI-P-18 Licensee's Basis for Requesting Relief {as stated}:

"These pumps are vertical, two-stage, centrifugal pumps, with an extended shaft and casing to allow suction from the containment sump. The motor and mechanical seals of the pumps are located at approximately the 12 foot elevation and the bottom of the casing is located at approximately the -30 foot elevation. The welds .

identified are at the bottom of the pump casing, and are embedded within the concrete building structure. This makes the welds inaccessible from the outside. The small diameter of the casing

{24 inch. 0. D.), and the pump shaft prevent examination from the inside diameter:"

18

e Licensee's Proposed Alternative Examination (as stated):

"A visual examination (VT-1) of the accessible portions of the I.D. of the pump casing welds will be performed only if the pump is disassembl~d and the pump shaft removed for maintenance."

Evaluation: The above listed pump casing welds are all encased in concrete prevent a surface examination from the OD. 1,e approximately 42 feet of 2-inch diameter casing between mechanical seal and the pump casing weld prevent surface examination from the ID. The inaccessibility of the welds, therefore, makes the Code-required surface examination impractical. Extensive modifications would be required to provide acce~s for examination. Imposition of the requirement on Virginia Electric and Power Company would cause a considerable burden on the licensee.

The licensee has proposed to perform a VT-1 visual examination of the accessible portions of the inside surfaces of the pump casing welds when a pump is disassembled and the pump shaft removed for maintenance. The pump casing welds would possibly deteriorate due to wear, corrosion, erosion, ardcracking. This VT-1 examination will enable significant degradation to be detected and, therefore, provide reasonable assurance of operation~l readiness.

Conclusion:

Based on the above evaluation, the surface examination of the subject pump casing welds is impractical to perform at Surry, Unit 1. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), it is recommended that relief be granted as requested.

3.2.4 Valves (No requests for relief) 3.2.5 General (No requests for relief) 3.3 Class 3 Components (No requests for relief) 19

e 3.4 Pressure Tests 3.4.1 Class 1 System Pressure Tests 3.4.1.1 Request for Relief 1, Examination Category B-P, Item 815.51, System Hydrostatic Test, as Modified by Code Case N-498, for Class 1 Safety Injection (SI) Piping Code Requirement: Table IWB-2500-1, Examination Category B-P, requires a system hydrostatic test in accordance with IWB-5222, System Hydrostatic Test, for Class I pressure-retaining piping.

Code Case N-498 requires that:

(1) A system leakage test (IWB-5221) shall be conducted at or near the end of each inspection interval, prior to start-up.

(2) The boundary subject to test pressurization during the syste~ leakage test shall extend to all Class 1 pressure retaining components within the system.

(3) Prior to performing a VT-2 visual-examination, the system

.shall be pressurized to nominal operating pressure for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated systems and 10 minutes for noninsulated systems. The system shall be maintained at nominal operating pressure during the performance of the VT-2 visual examination.

Licensee's Code Relief Request: The licensee requested relief from performing the hydrostatic test at the pressure required by Table IWB-5222-1, as modified by Code Case N-498, for the Class 1 piping in the following six areas between check val* - in the safety injection system:

1) l-SI-79 AND l-SI-235, l-SI-241

2) l-SI-82 AND l-SI-236, l-SI-242

3) l-Sl-85 AND l-SI-237, l-SI-243

4) l-Sl-88 AND 1-SI-238

5) l-SI-91 AND l-SI-239

6) l-SI-94 AND l-SI-240 20

Licensee's Basis for Requesting Relief (as stated):.

"The double check valve combination prevents pressurization of the area in between the check valves when conducting pressure tests on the primary system."

Licensee's Proposed Alternative Examination (as stated):

"During the performance of the Class l, Code Case N-498 pressure test, the area described above will be tested at a pressure 100 psig less than the RCS normal operating pressure. 1he establishment of a 100 psig differential pressure across the affected valves would prevent any uncontrollable dilution of the primary during testing."

Evaluation: The Code requires a system hydrostatic test of Class 1 piping at a test pressure above normal operating pressure. Code Case N-498, Alternative Rules for JO-Year Hydrostatic Pressure Testing for Class 1 and 2 Systems,Section XI, Division 1, provides a system leakage test (IWB-5221) as an alternative to the system hydrostatic test.

Code Case N-498 has been approved by the NRC by incorporation into Regulatory Guide 1.147, Revision 9, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1.

In a SER dated January 24, 1984 (Unit 1), relief was granted based on a proposed alternative test pressure of 2335 psig, which appears to be above normal operating pressure. This previously proposed test pressure indicates that pressure testing of the subject portions of the Class 1 SI system is practical, and should still be considered acceptable.

The licensee is using Code Case N-~a~ as an alternative to the

• 10-year system hydrostatic test, which is acceptable if all the requirements of the Code Case are met. However, the licensee has proposed an a test pressure of 100 pounds less than operating pressure, which does not meet the alternative requirements of the Code Case. This should not be considered acceptable. If the requirements of the Code Case cannot be met, the Code-required system hydrostatic test should be performed.

21

e e

Conclusion:

Based on the above evaluation, the proposed alternative does not meet the requirements of IWA-5222 nor the requirements of the approved alternative, Code Case N-498.

Therefcire, it ~s recommended that relief be denied.

3.4.1.2 Request for Relief 2, Examinati0n Category B-P, Item BlS.51, System Hydrostatic Test, as Modified by Code Case N-498, for Class 1 Residual Heat Removal (RHR) Piping.

Code Requirement: Table IWB~2500-l, Examination Category B-P, requires a system hydrostatic test in accordance with IWB-5222, System Hydrostatic Test, for Class 1 pressure-retaining piping.

Code Case N-498 requires* that:

(1) A system leakage test (IWB-5221) shall be conducted at or near the end of each inspection interval, prior to start-up.

(2) The boundary subject to test pressurization during the system leakage test shall extend to all Class 1 pressure retaining components within the system.

(3) Prior to performing VT-2 visual examination, the system shall be pressurized to nominal operating pressure for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated systems and 10 minutes for noninsulated systems. The system shall be maintained at nominal operating pressure during the performance of the VT-2 visual examination.

Licensee's Code Relief Request: The licensee requested relief from performing the hydrostatic test at the pressure required by Table IWB-5222-1 for the Class* nyR piping between MOV-1700 and ~

MOV-1701.

Licensee's Basis for Requesting Relief (as stated):

"During a normal hydrostatic test of the primary, MOV-1700 is closed in addition to MOV-1701. This prevents pressurization of MOV-1701 and the piping between the two MOVs. Both valv~s are 22

e cl~sed to prevent possible ove~ pressurization of the Residual Heat Removal System."

Licensee's Proposed Alternative Examination (as stated):

"This area will be tested in conjunction with the normal Class 2, N-498 pressure test at the pressure required by the adjoining Cl ass 2 syste,11."

Evaluation: As shown in drawi.ng 11448-CBM-87A-3; it is impractical to pressurize the piping between valves MOV-1700 and MOV-1701. The subject valves are pressure interlocked for automatic closure to prevent accidental over pressurization of the attached Class 2 piping in the RHR system. This safety feature would have to be bypassed to allow MOV-1700 to remain open during RCS pressurization, defeating the designed safeguard.

The licensee has proposed testing this section of piping to the requirements of Code Case N-498-1 pressure test for Class 2 piping. The proposed alternative test will detect dny through-wall cracks and provide adequate assurance of operational readiness.

Conclusion:

Based on the above evaluation, the hydrostatic test as required by Section XI of the ASME Code fo, the Class 1 piping between the subject MOVs is impractical to perform at Surry, Unit 1. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), it is recommended that relief be granted as requested.

3.4.2 Class 2 System Pressure Tests (No requests for relief)

3. 4. 3 Cl ass 3 System P* * * '51.ire Tests 3.4.3.l Request for Relief 3. IWD-5223. System Hydrostatic Test of Class 3 Circulating and Service Water System Piping Upstream of the First Isolation Valve Code Requirement: IWD-5210(a) requires that the pressure-retaining components within the boundary of each system specified 23

e in the Examination Categories of Table IWD-2500-1 be pressure tested and examined in accordance with Table IWD-2500-1 during the system hydrostatic test [IWA-52ll(d)]. The system hydrostatic test shall be conducted in accordance with IWA-5000, as applicable.

Licensee's Code Relief Request: The licensee requested relief ffom performirig the hydrostatic test for the Class 3 Circulating and Service Water System piping upstream of the first isolation valve.

Licensee's Basis for Requesting Relief (as stated):

"The Code addresses the problem of performing hydrostatic test on open ended portions of discharge lines beyond the last shut-off valve in non-closed systems in IWD-5223(d). A similar problem exist for the intake piping at Surry Unit 1 as it is non-isolatable for the increased pressure requirements of a hydrostatic test. 11 Licensee's Proposed Alternative Examination (as stated):

"As an alternative, the requirements applied to open ended portions of discharge lines (IWD-5223(d)) will be applied. In this case confirmation of adequate flow during system operation shall be acceptable in lieu of system hydrostatic test. 11 Evaluation: IWD-5210(a) requires that the pressure-retaining components within the boundary of each system be pressure tested.

For open-ended systems, the piping beyond the last discharge valve is exempted from the hydrostatic test requirements. The problems in testing are the same for suction piping prior to the first isolation valve as for open-ended systems beyond the last discharge *_*'*,e. The only practical test to verify the operability of that portion of such a system is a flow test.

==

Conclusion:==

The licensee's proposed confirmation of flow in the Circulating and Service Water systems for the open-ended intake piping before the first shut-off valve should provide an acceptable level of quality and safety. Therefore, ~uriuant to 24

• . e 10 CFR SO~SSa(a)(3)(i), it is recommended that the proposed alternative be authorized.

3.4.3.2 Request for Relief 4, IWD-5223, System Hydrostatic Testing of Class 3 Component Cooling Water System Piping NOTE: In the September 12, 1994, response to the NRC's request for additional information, *the licensee withdrew Request for Relief 4, based on their reassessment of ASME Code developments.

3.4.3.3 Request for Relief 5, IWD-5223, System Hydrostatic Test for Portions of the Class 3 Service Water System Code Requirement: IWD-5210(a) requires that the pressure-retaining components within the boundary of each system specified in the Examination Categories of Table IWD-2500-1 be pressure tested and examined in accordance with Table IWD-2500-1 during the system hydrostatic test [IWA-521I(d)]. The system hydrostatic test shall be conducted in accordance with IWA~SOOO, as applicable.

IWD-5223(a) requires that the system hydrostatic test pressure be at least 1.10 times the system pressure for systems with design temperature of 200°F or less, and at least 1.25 times the system pressure for systems with design temperature above 200°F. The system pressure shall be the lowest pressure setting among the safety or relief valves provided for over-pressure protection within the boundary of the system to be tested. For systems (or portions of systems) not provided with safety or relief valves, the system design pressure shall be substituted for the system pressure (PsJ.

Licensee's Code Relief Request: The licensee requested relief fro~ meeting the system test pressure requirements of IWD-5223(a) for the Class 3 Service Water System piping used for the cooling 25

of component cooling water for the charging pumps and lube oil for the charging pumps.

Licensee's Basis for Requesting Relief (as stated):

"The service water system on the above mentioned drawing is used for cooling component cooling water for the charging pumps and lube oil for the charging pumps. The system*was designed without the use of a safety or relief valve due to the low pressure output of the charging pump service water pumps which is 60 psig.

(max). The only other possible pressure source for the system would occur if an extensive heat exchanger leak occurred at either 1-SW-E-lA or 18 between component cooling and service water. This pressure could be no more than 57 psig the maximum discharge pressure of l-CC-P-2A or 28. Using a design pressure (PD) of 100 psig would be excessive for this system as the maximum pressure potential is 60 psig. Additionally this piping connects to fiberglass reinforced plastic piping, which has a design pressure of 60 psig. In some instances this piping could be included in the hydrostatic test boundary.

Licensee's Proposed Alternative Examination (as stated):

"As an alternative, it is requested that 60 psig be used as this systems PD value."

Evaluation: IWD-5210(a) requires that the pressure-retaining components within the boundary of each system be pressure tested.

The licensee has proposed to use 60 psig as the design pressure for the Class 3 Service Water System piping and components located on prints 11448-CBM-0718-3 and 11448-CBM-O?lD-3. The licensee has pointed out that using a design pressure (PD) of 100 psig would be excessive for the piping and components being pressure tested in the Service Water System. The licensee proposed alternative pressure of 60 psig for the system PD value is based on maximum design pressures associated with non-isolable

,interconnected piping and components. Based on the review of the licensee's request for relief, it appears that the licensee's proposed design pressure of 60 psig should be sufficient pressure to test the subject system.

Conclusion:

As a result of the ~bove evaluation, It has been determined that the proposed test pressure will provide an 26

acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i) it is recommended that the proposed alternative be authorized.

3.4.3.4 Request for Relief 6, IWD-5223, Class 3 System Hydrostatic Test of the Auxiliary Feedwater System Code Requirement: IWD-5210(a) requires that the pressure-retaining components within the boundary of each system specified in the Examination Categories of Table IWD-2500-1 be pressure tested and examined in accordance with Table IWD-2500-1 during the system hydrostatic test [IWA-52ll(d)]. The system hydrostatic test shall be conducted in accordance with IWA-5000, as applicable.

IWD-5223(a) requires that the system hydrostatic test pressure be at least 1.10 times the system pressure for systems with design temperature of 200°F or less, and at least 1.25 times the system pressure for systems with design temperature above 200°F. The system pressure shall be the lowest pressure setting among the safety or relief valves provided for over-pressure protection within the boundary of the system to be tested. For systems (or portions of systems) not provided with safety or relief valves, the system design pressure shall be substituted for the system pressure ( psv) .

Licensee's Code Relief Request: The licensee has requested relief from performing the Code-required system hydrostatic test for the following three portions of the Class 3 Auxiliary Feedwater (AFW) system; ~-

27

Line No. between Valve

• and Valves

1) l-WAPD-15-601 l-FW-145 l-FW-149 l-FW-146 l-FW-609

2) l-WAPD-17-601 l-FW-160 l-FW-163 l-FW-161 l-FW-608

3) l-WAP0-20-601 l-FW-175 l-FW-179 l-FW-176 l-FW-607 Licensee's Basis for Requesting Relief (as stated):

"Three pressure reducing orifices 1-FW-RO-lOOA, 1-FW-RO-lOOB, and 1-FW-RO-lOOC provide during normal operation a pressure drop from approximately 1200 psig to 110 psig. The system design takes advantage of this pressure drop by incorporating lower pressure rated piping downstream of the orifices. However the higher pressure rated piping continues beyond the orifices for some distance. The actual pipe design pressure class change occurs at downstream Check Valves l-FW-148, l-FW-162, and l-FW-178 and manual Valves, l-FW-146, l-FW-161, and l-FW-176. The check valve operation, however, does not allow separation of the lower pressure class system, when testing the higher class piping in accordance with IWD-5223. IWD-5223 requires a test pressure of 1576 psig on the higher pressure class components in question.

The downstream connecting piping has a design pressure of 150 psig and a corresponding test pressure of only 165 psig. As such, the connecting components would be over pressurized during the required Section XI test.

The test boundary could be backed up to manual Valves l-FW-147, l-FW-162, and l-FW-177 however hydrostatically testing this test boundary would also pressurize the auxiliary feedwater pumps and their suction connection. These pumps have welded discharge connections and cannot be isolated from'the test boundary due to the absence of a drain or vent valve in the area identified above. There is a flange connection on the suction side of each pump, however using this flange for isolation purposes is considered difficult due to the piping arrangement, and susceptibility to cold spring misalignment problems. Typically centrifugal pumps are hydrostatically tested at a pressure based on the suction side of the pump described in IWA-5224[d] of the Code, which prevents any potential over pressurization concerns, or the need to use pump flanges as isolation points.

The basis for relief then is two-fold. The first impracticality is the over pressurization of the pressure reducing orifice and the design pressure class rating change. The second impractfcality is the incorporation of the auxiliary feedwater 28

__ I

e

• pumps into the test boundary due to the lack of vent, drain, and manual isolation valves."

Licensee's Proposed Alternative Examination (as stated):

"The identified components will be tested in accordance with IWD-5222, Functional Test Requirements, in conjunction with the associated auxiliary feedwater pump at normal operating pressur~.

Evaluation: The Code requires a system hydrostatic test for the subject Class 3 lines. However, because of the difference in design pressures and the configuration of the valves, the Code-required test pressure cannot be achieved without over pressurizing the piping downstream. Therefore, the Code requirement is impractical. Imposition of the requirement would necessitate design modifications to adequately isolate the high pressure portions of the system. This would represent a considerable burden on the licensee. In lieu of the hydrostatic test, the licensee will perform a functional test at normal operating pressure.

The subject portions of the AFW system are redundant loops fabricated from I-inch diameter piping. Consequently, if a leak*

does occur, the volume will be small, because of the pipe size.

Leakage from this small piping, however, will likely be detected during the system functional test.

Conclusion:

Based on the factors presented in the above evaluation, reasonable assurance of operational readiness will be provided by the licensee's proposed examination~ Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).

3.4.3.5, Request for Relief 7. Examination Category D-A. Item DI.IO.

Hydrostatic Testing of C1ass 3 Circulating and Service Water Systems Code Requirement: Table IWD-2500-1, Examination Category D-A, Item Dl.10, requires a system hydrostatic test as specified by 29

e

• IWD-5223. IWD-5223 states that the system hydrostatic test pressure shall be at least 1.10 times the system pressure for systems with design temperatures of 200°F or l~ss, and at least 1.25 times the system pressure for systems with design temperatures greater than 200°F.

Licensee's Code Relief Request: The licensee has requested relief from performing the system hydrostatic test at the Code-required test pressure for Circulating and Service Water piping and components between the following four sets of valves:

between Valve on Line and Valve on Line

1) -l-SW-499 8"-WS-480-21X l-SW-311 8"-WS-480-21X l-SW-312 6 -WS-483.-21X 11 2-SW-331 6 -WS-484-21X 11

2) 2-SW-476 8 -WS-480-21X 11 l-SW-311 8 -WS-480-21X 11 2-SW-312 6 -WS-483-21X 11 2-SW-331 6 -WS-484-21X 11

3) l-SW-346 2" -WS-318--9107 l-SW-317 2"-WS-314-9107 l-SW-327 2"-WS-315-9107 2-SW-337 2"-WS-316-9107

4) 2-SW-344 2"-WS-319-9107 l-SW-317 2"-WS-314-9107 l-SW-327 2"-WS-315-9107 2-SW-337 2"-WS-316-9107 Licensee's Basis for Requesting Relief (as stated):

"Technical Specification 3.14C requires, "There shall be an operating service water flow path to and from one operating main control and emergency switchgear rooms air ccinditioning condenser and at least one operable service water flow path to and from at least one operable main control and emergency switchgear rooms air conditioning condenser whenever fuel is loaded in [the~

reactor core." By design there are only two supply headers and strainers associated with this common unit air conditioning

, system. This specification requires that the components identified always be operating or operable, while either of our two units have fuel in the reactor core. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> action statement for the service water headers allowing routine maintenance is provided in Technical Specification. However, challenging this action statement with a hydrostatic test (ten-year or post maintenance) is seen as excessive, potentially forcing a dual unit outage should problems occur, and impractical 30

e when compared to* the proposed alternative in this low pressure system."

Licensee's Proposed Alternative Examination (as stated):

"The 1ines and components in question are gravity fed from the in-take canal from lines providing water to both units main condensers. In lieu of the Code required nydrostatic test it is requested that a system inservice test, IWD-5221 be performed in conjunction with a visual (VT-2) examination of the accessible areas."

Evaluation: The Code requires a system hydrostatic test for Class 3 pressure-retaining components. The licensee's technical specifications require that the portions of the circulating and service water components covered in this request be operable while either of the two units have fuel in the reactor core.

Performance of the hydrostatic test would require the system to be taken out of service. Therefore, the Code-required pressure test would be impractical to perform. Imposition of the requirement could require a dual outage for the sole purpose of performing a low-pressure hydrostatic test. In lieu of this requirement, the licensee proposed to perform a system inservice test with an associated VT-2 visual examination.

The subject piping is gravity-fed lines that supply water directly to the main condensers. There are no pumps to boost the pressure above the head pressure. Since these lines are essentially not pressurized, the slight elevation in pressure will have little benefit, and any leakage that does occur should be detected by the VT-2 visual examination during the system inservice test. The licensee's proposed examination will provide reasonable assurance of the system's oper' * *1nal readiness.

==

Conclusion:==

The hydrostatic test of the above mentioned portions of.the circulating and service water components is impractical to perftirm at Surry, Unit 1. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), it is recommended that relief be granted as requested.

31

3.4.3.6 Request for Relief 8, Examination Category D-A. Item DI.IO, Hydrostatic Testing of Class 3 Pressure Retaining Components in the Circulating and Service Water System Code Requirement: Table IWD-2500-1, Examination Category 0-A, Item 01.10, requires a system hydrostatic test as specified by IWD-5223. IWD-5223 states that the system hydrostatic test pressure shall be at least 1.10 times the system pr~ssure for systems with design temperatures of 200°F or less, and at least 1.25 times the system pressure for systems with design temperatures greater than 200°F.

Licensee's Code Relief Request: The licensee requested relief from performing the Code-required system hydrostatic tests of the Class 3 Circulating and Service Water*System between the following two sets of valves:

between Valve and .Valve on

1) l-SW-MOV-102A 42"-WS-13-10 l-SW-37 30"-WS-14-10 l-SW-33 30"-WS-15-10 l-SW-29 30"-WS-16-10 l-SW-25 30"-WS-40*-10

2) l-SW-MOV-102B 42"-WS-12-10 l-SW-37 30"-WS-14-10 l-SW-33 30"-WS-15-10 l-SW-29 30"-WS-16-10 l-SW-25 30"-WS-40-10 Licensee's Basis for Requesting Relief (as stated):

"Technical Specification 3.14 requires service water flow to and from the component cooling water heat exchangers. One unit operation requires two component coolinq water heat exchangers, while two unit operation requires thre~ component cooling water ~*

heat exchangers. Additionally Technical Specifications 3.5 and 3.10 provide requirements associated with component cooling

' operability for core residual heat removal in both unit operation and refueling modes. The identified supply header to the component cooling water heat exchangers is the only source of cooling water to these components. Conducting.the Code required hydrostatic test would require isolation of all four component cooling water heat exchangers, placing the unit(s) in conflict with Technical Specifications as previously discussed."

32

Licensee's Proposed Alternative Examination (as stated):

"The lines and components identified are gravity fed from the in-take canal from lines providing water to the main condensers.

As an alternative, a system inservice test to the requirements of IWD-5221 and the corresponding visual (VT-2) examination of the accessible areas will be performed in lieu of the Code hydrostatic test." **

Evaluation: The Code requires a system hydrostatic pressure test for Class 3 pressure-retaining components. The licensee stated that the subject lines are the only source of cooling water for the Component Cooling Water Heat Exchangers (CCWHX), and that all four of the heat exchangers would have to be taken out of service to conduct the Code-required hydrostatic test. Since plant technical specifications require service water flow to and from the CCWHXs during operation and refueling modes, taking these lines out of service would be a burden causing a technical specification violation. Therefore, the Code requirements are impractical for these lines. In lieu of this requirement, the licensee proposed to perform a system inservice test, with an associated VT-2 visual examination, thus providing reasonable assurance of the system's operational readiness.

The subject piping consists of large diameter, gravity-fed lines that feed water directly to the CCWHXs. There are no pumps to boost the pressure above the head pressure. Since these lines are essentially not pressurized and contain a large volume of water, the slight elevation in pressure will have little benefit, and any leakage that does occur should be obvious and easily detected by the VT-2 visual examination during the system inservice test.

Conclusion:

Based on the above evaluation, the hydrostatic test

' of the subject portions of the circulating and service water components is impractical to perform at Surry, Unit 1:

Therefore, it is recommended that relief be granted pursuant to

• 10 CFR 50.55a(g)(6)(i). The licensee's proposed examination will 33

e provide reasonable assurance of the system's operational readiness.

3.4.4 General 3.4.4.1 Request for Relief 9, IWA-5214, Hydrostatic Testing of Repairs and Replacements NOTE: In a letter submitted June 22, 1994, the licensee withdrew Request for Relief 9.

3.4.4.2 Request for Relief 10, IWA-5250(a)(2), System Pressure Test Corrective Measures for Leakage at Bolted Connections Code Requirement: IWA-5250(a)(2), states that if leakage occurs at a bolted connection during a system pressure test, then the bolting must be removed and visually (VT-3) examined for corrosion.

Licensee's Code Relief Request: The licensee requested relief

_from performing the Code-required removal and visual (VT-3) examination of *bolting, if leakage occurs during a system pressure test of tlass 1, 2, or 3 systems.

Licensee's Basis for Requesting Relief (as stated):

"Typically pressure testing required by the Code will identify mechanical connection leakage from gasket or packing sources.

Routinely during the test or following the test~ this leakage is evaluated and additional torquing or other corrective measures on the bolting is applied to eliminate the leakage. Generally though,* this type leakage is not active long enough to cause component damage. Additionally, some materials are not affected by boric acid wastage. In some instances testing occurs while containment is sub-atmospheric and just prior to reactor criticality. Requiring that bolting be removed at this time 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. Surry at the start of a reactor refueling will examine safety-related components inside containment for boric acid leakage. Each component affected by boric acid leakage will be evaluated for potential damage and corrective action at this 34

e e time. The current Code requirement assumes no examination or evaluation like this has taken place, and defaults conservatively that leakage found has been in place for a significant time period to cause damage. This approach should be considered impractical, when compared to the alternative arrangements proposed."

Licensee's Proposed Alternative Examination (as stated):

11 Bolting in situations requiring removal and visual (VT-3) examination may be limited to one bolt nearest the leakage source. If that bolt has evidence of degradation, then all other bolting in the connection shall be removed and visually (VT-3) examined and evaluated to 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). Bolting removal, however would be such that visual (VT-3) examination would identify this condition.

11 Additionally for bolting subject to boric acid wastage examined during pressure tests conducted just prior to start up (Class 1 system leakage, Class 1 hydrostatic, Class 2 pressure *test scheduled in association with the Class 1 test, and Code Case N-498 test) in sub-atmospheric conditions, leakage identified near bolted connections shall be evaluated for removal need based upon the extent of leakage, correction requirements, and other examinations conducted during that refueling outage. This evaluation shall be subject to the review of the Authorized Nuclear Inservi ce Insp_ector (ANI I)."

Evaluation: In accordance with the 1989 Edition of the Code, when leakage occurs at bolted connections, all bolting is required to be removed for VT-3 visual examination.

The licensee's proposed alternative to the removal of all bolting at bolted connections when leakage occurs, is to ~erform an evaluation of the bolted connection. The licensee's alternative to the Code requirement could potentially eliminate the removal and VT-3 visual examination of bolting at leaking bolted 1.;onnections.

Recent incidents of degraded bolting have reinforced the reasons for removing and evaluating at least one bolt at a leaking bolted connection as part of the corrective action: Because degradation rates cannot be reliably predicted and bolting materjal records may not be accurate, the removal of a bolt for evaluation and 35

immediate corrective action for leakage at any bolted connection, regardless of material, is warranted. (

Reference:

Event Report Numbers 26899 and 26992).

It is reasonable to conclude that degradation, if present, would be detected provided that the licensee removes at least one bolt nearest the source of leakage for VT-3 visual examination. The licensee's alternative, in *combination with the removal and VT-3 visual examination of at least one bolt closest to the source of leakage, should provide an acceptable level of quality and safety.

Conclusion:

Pursuant to 10 CFR S0.55a(a)(3)(i}, it is recommended that the proposed alternative be authorized provided that at least one bolt closest to the so~rce of leakage is removed and evaluated.

3.4.4.3 Request for Relief 11, IWA-5242, System Pressure Tests of Insulated Bolted Connections Code Requirement; IWA-5242(a}, states that for systems borated for the purpose of controlling reactivity, insulation shall be removed from pressure-retaining bolted connections for VT-2 visual examination.

Licensee's Code Relief Request: The licensee requested relief from the Code-required removal of insulation for the VT-2 visual examination of bolted connections in borated systems that are normally tested in sub-atmospheric conditions (Reactor Coolant, Charging, and Safety Injection Systems).

Licensee's Basis for Requesting Relief (as stated):

"In some cases the reactor coolant system and systems connected to the reactor coolant system are tested in sub-atmospheric conditions, when the reactor coolant temperature is greater than 500° Fahrenheit. Typically this testing is done just prior to startup. Removing and reinstalling insulation under these 36

conditions is difficult to perform and considered impractical, when compared to the alternate proposal."

Licensee's Proposed Alternative Examination (as stated):

"It is proposed that bolted connections on Class 1 systems containing boric acid be examined each refueling outage at zero or static pressure. The examination would b.e performed with

• insulation removed. Class 2 bolting will be examined similarly once a period. This alternative only applies to systems that are pressure tested under sub-atmospheric conditions. In addition the required testing will still be conducted with a VT-2 examination without removing the insulation."

Evaluation: Paragraph IWA-5242(a) requires the removal of insulation from pressure-retaining bolted connections in borated systems for direct VT-2 visual examination during system pressure testing. The licensee implies that removal and replacement of

• insulation for direct VT-2 visual examination of bolted cqnnections during startup is a hardship. However, the Code does not require the pressure test to be performed during startup and therefore, the licensee has not submitted adequate justification.

Recent incidents of degraded bolting have reinforced the reasons for removing insulation at bolted connections when performing the VT-2 visual examinations (

Reference:

Event Report Numbers 26899 and 26992). Because degradation rates cannot be reliably predicted and bolting records may not be accurate, the direct visual examination and immediate corrective action for leakage at bolted connections is warranted.

Conclusion:

Based on the above referenced incidence of degraded bolting and the lack of justification of impracticality, it is recommended that relief be denied.

37

.* e e 3.5 General 3.5.1 Ultrasonic Examination Techniques 3.5.1.1 Request for Relief SR-005, ASME Section V, Article IV, Figure T-441.1 and Section XI, Appendix III, Figure III-3230-2, Requirements for Ultrasonic Calibration Blocks Code Requirement: Section V, Article IV, Paragraph T-441.1.2.2 requires the calibration block to be clad to the component clad nominal thickness.

Section XI, Appendix III-3430 requires the calibration blocks to generally conform to the design layout shown in Figure III-3230-2.

Licensee's Code Relief Request: Relief is requested from Section V, Article IV, paragraph T-441.1.2.2 and Section XI, Appendix III, Figure III-3230-2 calibration block requirements.

Licensee's Basis for Requesting Relief (as stated):

"Meeting the above new ASME Section XI requirements would require the fabrication of new calibration blocks.

"Satisfactory ultrasonic system calibration can be performed with the existing calibration blocks. Use of the existing calibration blocks also allows correlation of ultrasonic data from previous interval examinations as required by IWA-1400(h). The location of the notches in the piping calibration blocks provides adequate signal separation for sweep calibration. Distance-amplitude calibration down to the clad-to-base metal interface, as delineated by Non mandatory Appendix B to Section V, Article 4, can be performed from the unclad portion of the clad side of the existing*vessel calibration blocks."

, Licensee's Proposed Alternative Examination (as stated):

"It is proposed that the existing calibration blocks be used during the third inspection interval."

38

e In the September 12, 1994, response to the NRC's RAI, the licensee submitted the following description of the existing calibration blocks' deviation from Code requirements:

"NOTE 1: Notches for piping <l" and vessels 5.l" thick are not staggered as specified by Figure III-3230-1.

Satisfactory ultrasonic system calibration can be performed with the existing calibration blocks.

"Note 2: Notches for piping calibration blocks <1% thick are located IT from the end of the block instead of a minimum of 1%" as specified by Figure III-3230-1.

Satisfactory ultrasonic system calibration can be performed with the existing calibration blocks.

"NOTE 3: Blocks VIR-23, VIR-24 and VIR-25 are partially clad instead of fully clad as shown by Figure T-441.1. The portion of the calibration blocks which contains the calibration reflectors is clad. Satisfactory ultrasonic system calibration can be performed with the existing.calibration blocks."

Evaluation: Section XI, Appendix IIl-3430 requires the calibration blocks to generally conform to* the design layout shown in Figure 111-3230-2. However, Appendix Ill-3430 also states "Alternate block design and layout may be used, provided similar beam paths are utilized." The licensee meets this requirement and therefore, relief is not required for those calibration blocks with the Code deviations of Notes 1 and 2.

Section V, Article IV, Paragraph T-441.1.2.2 requires the block to be.clad to the component clad nominal thickness. The subject calibration blocks are partially clad instead of fully clad as shown by Figure T-441.1. The portions of the calibration blocks that contain the calibration reflectors are clad. Since the calibration is performed in an area that is clad, the calibration block meets the intent of the Code requirement.

In addition, the continued use of the subject calibration blocks will provide consistent results with previous examinations. The existing blocks have been proven satisfactory for performing calibrations. Any increase in plant safety that might occur with 39

e new blocks would not compensate for the burden placed on the licensee to fabricate new calibration blocks to satisfy the current Code requirements.

Conclusion:

Based on the above evaluation, the existing calibration blocks provide for acceptable examination sensitivity. Therefore, pursuant to 10 CFR S0.55a(a)(3)(ii), it is recommended that the proposed alternative be authorized as requested.

3.5.2 Exempted Components (No requests for relief) 3.5.3 Other 3.5.3.1 Request for Relief SR-006, IWA-2610 Weld Reference System for Class 1 and Class 2 Piping, Vessels, and Components Code Requirement: Section XI, Paragraph IWA-2610, "Weld Reference Syste~ - General," requires that a reference system be established for all weJds and areas subject to surface or volumetric examination. Each weld and area shall be located and identified by a system of reference points. The system shall permit identification of each weld, location of each weld centerline, and designation of regular intervals along the length of the weld.

Licensee's Code Relief Request: The licensee requested relief from establishing a weld reference system for all welds in Class 1 and 2 piping, vessels, and components.

Licensee's Basis for Requesting Relief (as stated):

"The original construction code used at Surry Power Station, ASA 831.1-1955 with applicable nuclear code cases, did not establish a weld reference system. Immediate establishment of a weld reference system cannot be practically attained within the scope and schedule of existing outages."

40

I

• C, Licensee's Proposed Alternative Examination (as stated):

"Surry Unit 1 has recently updated it's weld isometrics, providing a detailed identification of weld location. These drawings will be used in tracking and locating welds.

"In addition at the time welds are examined volumetrically for program requirements, a reference will be established for each weld, indicating a zero point and direction of ex~mination.

Welds which contain recordable indications (RI) shall be marked to ensure location of the indication, using appropriate reference marks. This reference system and marks will be permanently fixed on the weld."

Evaluation: For an operating plant, establishing a weld reference system for all welds and areas subject to surface or volumetric examination is a major effort and, in some cases, is prohibitive due to inaccessibility and/or high radiation levels.

Therefore, the Code requirement for establishing a weld reference system for all welds subject to examination in the absence of examination is impractical for an operating plant. In order to establish a weld reference system for all welds and areas subject to surface and volumetric examinations in accordance with the Code requirements, many man-hours and potential man-rem of radiation exposure would be required to perform such tasks as locating the welds, removing insulation, marking the welds, and reinstalling insulation, regardless of whether or not the weld is scheduled for examination. Imposition of this requirement on Virginia Electric and Power Company would cause a burden that would not be compensated by an increase in public health and safety.

However, as inservice examinations of Class 1 and 2 piping systems are performed, each piping weld examined ~~uld receive all of the required reference markings. Impracticality will not exist for these welds since access will have been provide to perform the examinations.

Conclusion:

Marking of all welds and areas subject to surface or volumetric examination, as required by the Code in absence of inspection is impractical at Surry; Unit l, since it is an 41

, t '-'

operating plant; However, as each Class 1 and 2 piping system is

,H! :,

\JI;* ,

examined, access for marking each weld will be provided and impracticality for that particular weld will not exist.

Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i), provided that each Class 1 and 2 piping weld examined receives all of the required refe~ence markings at the time of inservice examination to provide assurance of traceability of piping welds and repeatability of examinations.

3.5.3.2 Request for Relief SH-1, Examination Categories F-A, F-8, and F-C, Items Fl.IO through F3.50, Component Supports Code Requirement: Section XI, Table IWF-2500-1, Examination Categories F-A, F-8, and F-C, Items Fl.IO through F3.50 require VT-3 visual examination of component supports as defined by Figure IWF-1300-1.

Licensee's Code Relief Request: Relief is requested to use the alternative requirements for component supports provided under ASME Code Case N-491, "Alternative Rules for Examination of Class l, 2, 3, and MC Component Supports of Light-Water Cooled Power Plants,Section XI, Division l".

Licensee's Basis for Requesting Relief (as stated):

"Subsection IWF, IWF-1000 through IWF-3000 inclusive in the 1989 Edition of ASME Section XI, lacks a complete concise set of rules for the inservice inspection of component supports. The following areas in particular have been identified as needing clarification: *

"l) Supports exempt from examination and test; ?~ Supports ~*

selected for examination; 3) Sampling program.

"It is our opinion that the 1989 Edition of ASME Section XI, IWf-1000 through IWF-3000, is still incomplete, and as a result, extremely difficult to follow in establishing as IWF Program.

The Code has recognized this deficiency and has written Code Case N-491 into ASME Section XI, 1989 Edition, 1990 Addendum."

42

~I

. . ~

Licensee's Proposed-Alternative Examination (as stated):

I "Code Case N-491 will be used in its entirety. This Code Case provides clear concise guidance in establishing an IWF examination program. The ASME Code has accepted this guidance by incorporating it into ASME Section XI."

Evaluation: The licensee prrposes, as an alternative to the Code requirements, to apply the requirements of Code Case N-491 for the examination of component supports on Code Class 1, 2, 3, and MC piping and components. Code Case N-491 has been approved for use in Revision 10 of Regulatory Guide 1.147 and therefore relief is not required.

Conclusion:

Relief is not required.

I

.. I I

43

p

( ~'

4. CONCLUSION Pursuant to 10 CFR 50.55a(g)(6)(i), it has been determined that certain inservice examinations cannot be performed to the extent required by Section XI of the ASME Code. For Requests for Relief 2, 6, 7, 8, SR-002, SR-003, SR-004, and SR-006, the licensee has demonstrated that specific Section XI requirements are impractical and it is recommended that relief be granted. The granting of relief will not endanger life, property, or the common defense and security and is otherwise in the public interest, g1v1ng due consideration to the burden upon the licensee, that would result if the requirements were imposed on the facility.

Pursuant to 10 CFR 50.55a(a)(3)(i), it is concluded that for Requests for Reli~f 3, 5, 10, SR-001, SR-007., and SR-008, the licensee's proposed alternative provides an acceptable level of quality and safety in lieu of the Code required examination. In these cases, it is recommended that the proposed alternative be authorized.

Pursuant to 10 CFR 50.55a(a)(3)(ii), it is concluded that for Reque~t for Relief SR-005 the licensee has demonstrated that specific Section XI requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. In this case, it is recommended that the proposed alternative be authorized.

For Requests for Relief RR-1 and RR-11, it is concluded that the licensee has not provided sufficient information to support the determination that the Code requirement is impractical, and that requiring the licensee to comply with the Code requirement would result in hardship. Therefore in these cases it is recommended that relief be denied. Requests for Relief 4 and 9 were withdrawn by the licensee and deleted from the*:* Program Plan. For Request for Relief SH-1 it was concluded that relief was not required.

Based on the review of the Surry Power Station, Unit 1, Third JO-Year Interval Inservice Inspection Program Plan, Revision 0, the licensee's response to the.

NRC's request for additional information, and the recommendations for granting relief from the ISI examination that cannot be performed to the extent required by Section XI of the ASME Code, no deviations from regulatory 44

e

• requirements or commitments were identified in the Surry Power Station, Unit 1, Third JO-Year Interval Inservice Inspection Program Plan, Revision 0, with the exGeption of Requests for Relief 1 and 11, and the Plan is acceptable and in compliance with 10 CFR 50.55a(g)(4).

45

~I I

..l,'l

5. REFERENCES

1. Code of Federal Regulations, Title 10, Part 50.

2. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Division 1:

1989 Edition

3. Surry Power Station, Unit 1, Third JO-Year Interval Inservice Inspection Program Plan, Revision 0, submitted July 16, 1993.

4. NUREG-0800, Standard Review Plans, Section 5.2.4, "Reactor Coolant Boundary Inservice Inspection and Testing, and Section 6.6, "Inservice 11 Inspection of Class 2 and 3 Components," July 1981.

5. Letter, dated June 8, 1994, B. C. Buckley (NRC) to J. P. O'Hanlon (Virginia Electric and Power Company) containing request for additional information on the "Surry Power Station, Unit 1, Third JO-Year Interval Inservice Inspection Program Plan."

6. Letter, dated September 12, 1994, J. P. O'Hanlon (Virginia Electric and Power Company) to Document Control Desk (NRC) containing the response to the NRC's request for additional information.

7. Letter, dated June 22, 1994, J. P. O'Hanlon (Virginia Electric and Power Company) to Document Control Desk (NRC) withdrawing Request for Relief 9, opting to use Code Case N-416-1.

8. NRC Regulatory Guide 1.14, "Reactor Coolant Pump Flywheel Integrity,"

Revi$ion 1, August 1975.

9. NRC Regulatory Guide 1.150, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations," Revision 1, February 1983.

46

U.S. NUCI.EAR REGUI.ATORY ISSION  : 'l;?C.~7 '<<..".":::"

1....-

_ _ N,-C. Aod Vol .. S..... - *

• ti¥ ,,,Nummn_il-.J BIBLIOGRAPHIC CATA SHEET

2. TITI.E ANO SUBTITI.E INEL-94/0164 Technical Evaluation Report on the Third 10-Year Interval Inservice Inspection Program Plan: 1 OAT: RE?ORT PUBLISHEO r.t()NTH Virginia Electric and Power Company, I Surry Power Station, Unit 1

• l='i:>hr11;:irv 1 QQ4 Docket Number 50-280 4. FIN OR Gr!ANf NUMBi:rl

5. AUTH\..!1(Sl 6. TYPE OF REPORT Technical B. W. Brown, E. J. Fe,ge, 7. PERIOD COVERED 1 1 K. W. Hall, A. M. Porter

8. PER FORMING ORGANIZATION - NAME ANO AOC RESS fll NRC. ~tfltlfda DiNioll. OHO or lf.,;o,t. U..S. NUCIJMI - ~ eo,,w,,llllion. Md .lflMHlff MIIJrwa.* ii cottrrcror. /JtflYII#

LITCO P.O. Box 1625 Idaho Falls, ID 83415-2209

9. SPONSORING ORGANIZATION - NAME ANO AODR ESS '" NIIC. r,,- *-s.. - - * * : i f ~ _ . . NRC Q,.,....,, Offla i,, ~ - IJ*.S. Nw#ilr " ~ ,:-,,..;..,,.

Materials and Chemical Engineering Branch Office of Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT 12tX1_.,,._,

This report. presents the results of the evaluation of the Surry Power Station, Unit 1, Third JO-Year Interval Inservjce Inspection (JSI) Program Plan, submitted July 16, 1992, including the requests for relief from the American Society of Mechanical Engineers (ASM[J Bo,*ler and Pressure Vessel Code Section XI requirements th~t the licensee has determined to be impractical. The Surry Power Station, Unit 1, Third JO-Year Interval ISI Program Plan i~ evaluated in Section 2 of this report.

The ISi Program Plan is evaluated for (a) compliance with the appropriate edition/addenda of Section XI, (b) µcceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commission (NRC) reviews. The r~quests for relief are evaluated in Section 3 of this report.

~-

12. KEY WOROSJOESCR!PT~RS fLi1t-lll'p1t--Mlliailt~11110UDlff,,..,.,,..rt.1 13. AVAILA81UTY STATEMENT Un 1 i mited 1'. SECURITY CI..ASSIFIC:,,.TION

/Thu,,_/

Unclassified (TltioReoonJ Unclassified

15. NUMBER OF PAGES

~6. PRICE NRC: FO,-M 33!1 12-8111