Text

Rev 2 to ISI Program Second 10-Year Interval Vogtle Electric Generating Plant Units 1 & 2
	ML20209H395
Person / Time
Site:	Vogtle
Issue date:	02/15/1999
From:	; SOUTHERN NUCLEAR OPERATING CO.
To:	;
Shared Package
ML20209H392	List: ML20209H388; ML20209H395; ML20209H409;
References
	NUDOCS 9907200212
	Download: ML20209H395 (29)
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SOUTHERN NUCLEAR OPERATING COMPANY INSERVICE INSPECTION PROGRAM SECOND 10-YEARINTERVAL VOGTLE ELECTRIC GENERATING PLANT UNITS 1 AND 2 SNC-GENERAL OFFICE SNC - VOGTLE PLANT PREP'D REV'D

APPV, APPV.

APPV.

BY BY BY VOGTLE MGR.

PLANT REV DATE DESCRIPTION (ITS)

(ITS)

PROJECT ENGRG GEN.

(NMS)

SUPP.

MGR.

0 5/9/97 BASIC ISSUE MB JAE/HPW N

JBB AGM 8'Ippr$0'd" 1

2/15/99 Withdrew RR-3, RR-8, RR-13, l

RR-15. Revised RR-4, RR-7, RR-LCV-1016-J 10, RR-11, RR-12, RR-14, RR-16, RR-22. See Summary of Changes.

d$

RR-26, Added RR-31 AMM akPk rg% Approved 2

2/15/99 Revised RR-6 (editorial only),

V-1016-M m

1 i

O Multiple preparers and reviewers. Preparation / review sheets on file.

9907200212 990714 DR ADOCK 05 4y4

E 1

l VEGP-1 AND 2 INSERVICE INSPECTION PROGRAM i

Summary of Changes Revision 2 l

l

1. Revised Table 1 to add "RR-31" for Code Item Bl.30.

2. Revised the summary of" Requests for Relief" to the show the current revision and NRC status for each requesti

3. Added Code Category reference in summary of" Request for Relief" for RR-22 to reflect

~ NRC clarification of Safety Evaluation dated 03/24/98 and added 1" Interval relief request reference.

4. Revised Relief Request RR-6 to include an editorial change on the NRC caveat to perform a VT-1 examination of the Steam Generator nozzle inner radius.

5. Revised Relief Request RR-26 for Class 2 and Class 3 components to use the guidance of the proposed Code Case N-533 revision and commit to the 40-month frequency restriction (with a noted exception) on insulation removal.

6. Added Relief Request RR-3 ? to provide an alternative schedule for examining the Reactor Pressure Vessel shell-to-flange weld.

vprgsoc. doc Page1 ofI

i VEGP-1 AND VEGP-2 INSERVICEINSPECTION PROGRAM 1

LIST OF EFFECTIVE PAGES Distribution List Revision 1 Class 3:

Page 4-1 Revision 0 Table of Contents Revision 0 Page 4-2 Revision 0 Page 4-3 Revision 0 Introduction -

Page 4-4 Revision 0 Page 1-1 Revision 0 Page 1-2 Revision 0 Component Supports:

Page 1-3.

Revision 0 Page 5-1 Revision 0 Page 1-4 Revision 0 Page 5-2 Revision 0 Page 1-5 Revision 0 Page 1-6 Revision 0 Classes MC and CC:

Page 1-7 Revision 0 -

Page 6-1 Revision 0 Class 1:

Requests for Relief:

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VEGP-1 AND VEGP-2 INSERVICE INSPECTION PROGRAM LIST OF EFFECTIVE PAGES (Continued)

Page 7-24 Revision 2 Page 7-61 Revision 1 Page 7-25 Revision 1 Page 7 62 Revision 1 Page 7-26 Revision 1 Page 7-63 Revision 1 Page 7-27 Revision 1 Page 7-64 Revision 1 Page 7-28 Revision 1 Page 7-65 Revision 1 Page 7-29 Revision 1 Page 7-66 Revision 1 Page 7-30 Revision 1 Page 7-67 Revision 1

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~ Page 7-36 Revision 1 Page 7-73 Revision 0 Page 7 Revision 1 Page 7-74 Revision 0 Pr.ge 7-38 Revision 1 Page 7-75 Revision 0 Page 7-39 Revision 1 Page 7-76 Revision 0 Page 7 Revision 1 Page 7-77 Revision 0 Page 7-41 Revision 1 Page 7-78 Revision 0 Page 7-42' Revision 1 Page 7-79 Revision 0 Page 7-42a Revision 1 Page 7-80 Revision 0 Page 7 Revision 1 Page 7-81 Revision 0 Page 7-44 Revision 1 Page 7-82 Revision 0 Page 7-45 Revision 1 Page 7-83 Revision 0 Page 7-46 Revision 1 Page 7-84 Revision 0 Page 7-47 Revision 1 -

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VEGP-1 AND VEGP-2 INSERVICE INSPECTION PROGRAM T.IST OF EFFECTIVE PAGES (Continued)

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e 7.0 Requests for Relief

~ Requests for relief have been prepared from information determined in support of Preservice Inspection activities and Inservice Inspection activities conducted during the first ten-year inspection interval at VEGP-1 and 2. The subject requests for relief apply to both VEGP units unless specifically denoted otherwise. Please refer to the following for a detailed listing of requests for relief. The actual requests for relief follow the listing. Please note that in cases where the use of an ASME Section XI code case is requested, a copy of the code case is provided solely for the convenience of the NRC reviewer.

Requests for Relief Summary RR Code Subject 1st Interval NRC Status No.

Category (s)

RR No.

(See Note 1)

RR-1 N/A VEGt'-210 Year Update N/A Approved Rev.0 12/31/98 RR-2 B-D, B-F RPV Nozzle Schedule N/A Approved Rev.0 (Code Case N-521) 12/31/98 RR-3 B-A RPV Mechanized Limitations RR-2, RR-3, Withdrawn Rev.1 RR-5 RR-4 B-A, B-H RPV Closure Head Limitations RR-7, RR-52 Approved Rev.1 (manual) 12/31/98 RR-5.

B-G-1 RPV Closure Head Nuts N/A Approved Rev. 0 03/24/98 RR-6 B-B Steam Generator (Class 1) Welds RR-19, RR-42 Approved Rev.2 and Inner Radii Limitations 12/31/98 RR-7 B-B, B-D, B-F Pressurizer Weld Limitations RR-12, RR-14, Approved Rev.1 RR-15 12/31/98 RR-8 B-H Pressurizer Support Limitations RR-10, RR-16 Withdrawn q

Rev.1 RR-9 B-J Primary Loop Piping Calibration RR-22, RR-23 Approved Rev.O Block and Scan Limitations 12/31/98 RR-10 B-J Primary Loop Piping Weld RR-17, RR-24 Approved Rev.1 Limitations 12/31/98 RR-11 B-J Primary Loop Piping Branch RR-21 Approved Rev.1 Connection Limitations 12/31/98 RR-12 B-J 10" Safety Injection Weld RR-26 Approved Rev.1 C-F-1 Limitations 12/31/98 RR-13 B-J Piping (Class 1) Weld Limitations N/A Withdrawn Rev.1 RR-14 C-A, C-B, C-G Vessel (Class 2) Weld RR-28, RR-29, Approved Rev.1 Limitations RR-30, RR-32 12/31/98 RR-15 C-F-1 Piping (Class 2) Weld Limitations RR-35, RR-36, Withdrawn Rev.1 RR-37 l

1 rrsum. doe 7-1 Rev.2 l

1 l

Requests for Relief Summary (continued)

RR Code Subject ist Int. RR No.

NRC Status No.

Category (s)

(See note 1)

RR-16 C-F-2 Piping (Class 2) Weld Limitations RR-34 Approved Rev.1 12/31/98 RR-17 C-F-1 Piping (Class 2) Weld Selection N/A Approved Rev.O Criteria 10/24/97 IG-18.

C-F-1 NSCW Piping Classification N/A Approved Rev.0 12/31/98 RR-19 C-F-1, C-F-2 Piping Longitudinal Welds N/A Approved Rev.0 (Code Case N-524) 03/24/98 RR-20 B-II, C-C, D-A, Integral Attachment Examinations RR-61 Approved Rev.O D-B, D-C (Code Case N-509) 03/24/98 RR-21 C-A, C-C Vessel (Class 2) Exemptions RR-62 Approved Rev.0 (Code Case N-408-2) 12/31/98 RR-22 ASME 1" and Under Piping Exemption RR-64 Approved Rev.1 (Non B-Q)

(Code Case N-544) 03/24/98 RR-23 B-E, B-P, C-H, Hydrostatic Testing RR-60 Approved Rev.O D-A, D-B, D-C (Code Case N-498-1) 03/24/98 RR-24 ASME Hydrostatic Testing RR-59 Approved Rev.0 (Code Case N-416-1) 03/24/98 RR-25 ASME Bolted Connections N/A Approved Rev.0 (Code Case N-566) 10/24/97 RR-26 ASME Insulation at Bolted Connections N/A Approved Rev.2 (Code Case N-533) 12/31/98 RR-27 ASME VT-2 Personnel Requirements N/A Approved Rev.0 (Code Case N-546) 12/31/98 RR-28 ASME NIS-2 Requirements for Snubbers N/A Approved Rev.0 (Code Case N-508-1) 12/09/97 RR-29 ASME Inservice Testing and Examination RR-43 Approved Rev.O Requirements for Snubbers 12/31/98 RR-30 ASME NIS-1 and NIS-2 Reporting N/A Approved Rev.0 Requirements 12/31/98 (Code Case N-532)

RR-31 B-A RPV Shell to Flange Weld N/A Approved Rev.2 12/31/98 Notes:

(1) Code Case " Request (s) for Relief" are authorized for the current interval or until such a time as the Code Cases are approved for general use by reference in NRC Regulatory Guide 1.147. After that time, SNC will continue to use the Code Cases with the lirnitations, if any, listed in NRC Regulatory Guide 1.147.

rrsum. doe 7-2 Rev. 2 l

SOUTHERN NUCLEAR OPERATING COMPANY

' VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-6

'L1 ' System /Connonent(s) for Which Reliefis Reau== tad:

Volumetric examination of steam generator pressure-retaining welds and nozzle inner radius sections (Class 1) as identified in Attachment 1 to this request for relief.

.IL Code Reanirement:

ASME Section XI Category B-B and ASME Section XI Category B-D, Table IWB-2500-1 require volumetric examinations ofpressure-retaining welds in the steam generator and nozzle inner radii. Applicable examination volume is shown in ASME Section XI Figures IWB-2500-6 and IWB-2500-7.

IIL Code Reauirement from Which Reliefis Reauested:

i Reliefis requested from performing the Code-required volumetric examinations of the components identified in Attachment I to this request for relief.

IV.? Basis for Relief:

(a) Physical limitation of the steam generator tube sheet obstructs and/or prohibits transducer movement along the required scan region of the channel head to tube sheet weld. Full Code

. coverage is not possible in the vicinity of the obstructions.

(b) ' The steam generator primary side nozzles are integrally cast as part of the channel head.

- The steam generator nozzle radius section cannot be volumetrically examined from outside of the nozzle or channel head because the rough, as-cast contact surface is not suitable for ultrasonic coupling, and the geometric configuration requires' an excessively long test metal distance resulting in high ultrasonic attenuation. The inside of the nozzle and channel head areas are covered with cladding in the "as-welded" condition; therefore, meaningful volumetric examination cannot be performed from the "as-welded" surface. Even with proper preparation of the inside surface for volumetric examination, an adequate examination of the area ofinterest (base metal just below the cladding) could not be achieved due to the resulting ultrasonic response at the clad-to-base metal interface. Refer to Attachment 2 for a depiction of the nozzles at VEGP.

1 V.

Alternate Examination:

1 (a) No alternate examination is proposed. The volumetric examination of the referenced steam j

generator tube sheet welds are being conducted to the fullest extent practical.

'.(b) No alternate examination is proposed for steam generator primary side nozzles inner radii.

1 1

7-20 Rev. 2

l SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-6 (continued)

V.

Alternate Examination (continued):

(NOTE: The Safety Evaluation Report provided in the NRC letter dated December 31, 1998, approves this request for reliefprovided that the nonle IR sections receive a VT-1 visual examination as performed during the first 10-year interval.).

VI. Justification for Granting Relief:

The examinations of the steam generator tubesheet welds are being conducted to the extent practical.

Compliance with the requirements of ASME Section XI for the specific nozzle inside radius section is impractical. Due to the high radiation field present (generally greater than 10 REM / hour), visual examinations are not practical and dose acquired is contrary to the principles of"As Low As Reasonably Achievable"(ALARA). During the First Ten-Year Interval, visual examinations and the Code-required pressure tests were performed and no evidence of degradation was observed. (NOTE: Subsequent to submitting RR-6 to the NRC for review and approval, the NRC advised SNC in their Safety Evaluation Report referenced herein that RR-6 was approved provided that VT-1 examination of the steam generator nozzle IR section is performed.)

Relief was initially granted by the NRC during the First Ten-Year Interval for those

- welds / components in Attachment I having requests for relief submitted for them. These l

included First Ten-Year Interval Requests for Relief # RR-19 and RR-42. NRC approval was documented in correspondence dated November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively.

Although there are physical obstructions which limit and/or prohibit the amount of examination coverage for those components identified in Attachment 1, reasonable assurance still exists that an acceptable level of quality and safety will be maintained since there have been no catastrophic failures of steam generator primary side nozzle inner radii nor steam generator tubesheet welds.

As a result, SNC requests that relief be authorized pursuant to 10 CFR 50.55a(g)(6)(i) since it is impractical to perform the examinations as required by the Code.

i VII. Implementation Schedule:

The subject examinations will be performed to the fullest extent practical during the Second Ten-Year Interval which commenced May 31,1997.

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T-SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE El ECTRIC GENERATING PLANT UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-6 (continued)

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F i

SOUTIf ERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-26 L

System / Component (s) for Which Reliefis Reauested:

ASME Class 1,2, and 3 bolted connections in borated systems. This includes the following systems which are borated for the purposes of reactivity control:

Reactor Coolant System, Chemical and Volume Control System, e

Residual Heat Removal System, e

Safety Injection System, and e

Nuclear Sampling System - Liquid.

IL Code Reautrement:

Subparagraph IWA-5242(a) of the 1989 Edition of ASME Section XI states in part "For systems borated for the purposes of controlling reactivity, insulation shall be removed from pressure retaining bolted connections for visual examination VT-2."

IIL Code Reauirement from Which Reliefis Reauested:

Reliefis requested from removing insulation from pressure-retaining bolted connections and

. performing VT-2 visual examinations for the purpose ofdetecting boric acid residue when systems are at the pressure and temperature requirements ofIWA-5000, IWB-5000, IWC-5000, and IWD-5000.

IV. Basis for Relief:

i Subparagraph IWA-5242(a) specifies that insulation must be removed from pressure-retaining bolted connections for VT-2 visual examination during the performance of system pressure testing. This is applicable to the following systems:

Reactor Coolant System (System consists of Class 1 and 2 components),

Chemical and Volume Control System (System consists of Class 1,2, and 3 components),

e Residual Heat Removal System (System consists of Class 1 and 2 components),

e Safety Injection System (System consists of Class 1 and 2 components), and e

Nuclear Sampling System - Liquid (System consists of Class 2 components).

7-105 Rev. 2

SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO, RR-26 (continued)

IV. Basis for Relief (continued):

Class 1 components:

Table IWB-2500-1, Examination Category B-P requires a system leakage test (IWB-5221) and corresponding VT-2 visual examination on Class I components each refueling outage prior to plant startup. This system leakage test is performed in Mode 3 when the Reactor Coolant System is at Nominal Operating Pressure (= 2235 psig) and Nominal Operating Temperature

(= 550'F to 650*F). The majority of the Class I components are in the Reactor Coolant System; however, some portions of the Class 1 boundary extend to include portions of Safety Injection, Chemical and Volume Control, and Residual Heat Removal systems. All Class I components are in containment. The removal and installation ofinsulation during the performance of system pressure testing inside containment presents the following hazards:

Increased potential for debris to be in containment which could migrate to the Containment Emergency Sumps and restrict the suction of the Emergency Core Cooling System during accident conditions, such as a Loss of Coolant Accident (LOCA). All debris is required to l

be removed from containment prior to entering Mode 4, i

Increased potential for personnel heat stress since the containment ambient temperature e

may be a high as 100*F, 3

Increased potential for personnel burn injuries due to installation ofinsulation in proximity ofextremely hot components, Increased personnel safety hazard since ladders would have to be used to inspect many of e

the bolted connections and replace the insulation. Temporary work platforms / scaffolding inside containment are removed prior to entering Mode 4, Increased radiation exposure to personnel since temporary shielding is removed prior to e

entering Mode 4, and Increased potential for impacting outage duration due to the amount of manpower required to support insulation removal and examinations during Mode 3 following refueling outage activities.

7-106 Rev. 2

SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-26 (continued)

IV. Basis for Relief (continuedh Class 2 components:

Table IWC-2500-1, Examination Category C-H requires a system pressure test (IWC-5221) during system functional and system inservice tests and corresponding VT-2 visual examination on Class 2 components once each inspection period. The following discusses the applicable l

systems and the basis for relief for each Class 2 system:

1.

Basis for Relief for the Reactor Coolant System (RCS):

The Class 2 portions ofRCS are located adjacent to the Class 1 boundary and are classified as Class 2 based on line size and isolation valve criteria. The system inservice tests for these Class 2 portions ofRCS are VT-2 examined in Modes 1,2, and 3 and therefore the same basis for relief as provided for Class 1 applies. These Class 2 pressure boundaries are located in containment.

2.

Basis for Relief for the Chemical and Volume Control System (CVCS):

For those portions of CVCS which are located inside containment (e.g., Charging, l

Letdown, Excess Letdown, Alternate Pressurizer Spray, Reactor Coolant Pump Seal Leakoff) the same basis for relief as provided for Class 1 above applies except that the l

system operating temperatures are less resulting in less potential for burn-related injuries.

The VT-2 examinations are performed in Modes 1,2, and 3.

For those portions of CVCS which are located outside containment, radiation levels, high component temperatures (e.g., approx. 290 F for Letdown), and availability of personnel l

during non-outage times may preclude removing insulation while the pressure-retaining bolted connections are pressurized. CVCS is inservice during power operation and, as such, many, if not all, components will be uninsulated, VT-2 examir.ed, and reinsulated when the pressure-retaining bolted connections are pressurized. However, as previously addressed, conditions may be present which may not allow insulation removal except during refueling outages.

3.

Basis for Relief for the Residual Heat Removal System (RHR):

The RHR System is placed inservice during a shutdown prior to refueling activities in Mode 4 when RCS is = 350 *F and = 350 psig. RHR remains inservice in Modes 5 and 6 7-107 Rev. 2

SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-26 (continued)

IV pasis for Relief (continuedh Class 2 components (continuedh.

3.'

Basis for Relief for the Residual Heat Removal Systern (RHR) (continued):

during the refueling outage and remains inservice in Mode 4 during startup following the refueling outage. The VT-2 examinations are performed in either Mode 4 or Mode 5 when the RCS is = 350 psig.

l For those portions of RHR which are located inside containment, the same Nis for relief l

as provided for Class 1 above applies except that the system operating ten e ratures are less, resulting in less potential for burn-related injuries. It is impractical to attempt to limit VT-2 examinations to Mode 5 in order to avoid the complications ofperforming RHR pressure tests in Mode 4. The VT-2 examinations are performed in Modes 4 or 5.

l For those portions ofRHR which are located outside containment, radiation levels, high l

component temperatures, availability of personnel, and increased thermal loads on chilled water room cooling systems may preclude removing insulation while the pressure-retaining bolted connections are pressurized. It is significantly more prudent to uninsulate, VT-2 examine, and reinsulate the pressure-retaining bolted connections in RHR when the system is not pressurized during non-outage times or during refueling outages when the system is not at the required pressure.

4.

Basis for Relief for the Safety Injection System (SI):

The system pressure tests performed on SI are either system functional tests or system inservice tests as follows:

- Some of the system functional tests are pedormed during Modes 1,2, and 3 when RCS l

pressure is greater than SI pump discharge pressure. VT-2 examinations are performed on portionn f SI during various activities and tests which require a SI pump to be in operation. The scope of these VT-2 examinations includes components located both inside l and outside containment. The performance of the VT-2 examinations during these activities is generally performed in less than one hour to minimize run time on the SI pumps.

The remainder of the system functional tests are performed during Mode 6 and defueled conditions with the reactor vessel head removed. VT-2 examinations are performed on portions of SI which are pressurized during check valve flow testing activities which j

7-108 Rev.2

)

l 1

I l

SOUTHERN NUCLEAR OPERATING COMPANY l

VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-26 l

(continued)

IV. Basis forRelief(continued):

Class 2 components (continued):

4.

Basis for Relief for the Safety Injection System (SI) (continued):

involve injection ofwater into the reactor pressure vessel. The scope of these VT-2 examinations includes components located both inside and outside containment. The l

performance of the VT-2 examinations during these activities is generally performed in less than one hour to minimize run time on the applicable pumps and to minimize the impact on critical path testing during refueling outages.

Some of the system inservice tests are performed during Modes 1, 2, or 3 on portions of SI which are pressurized by the SI accumulator tanks. The SI accumulator tanks are generally depressurized during refueling outages. The scope of these VT-2 examinations includes components which are located only inside containment.

l The remainder of the system inservice tests are performed on portions of SI which are pressurized by the static head of the refueling water storage tank. The VT-2 examinations on these portions of SI are generally performed during power operation (Mode 1) but may be performed in other Modes if tank levels are adequate. The scope of these VT-2 examinations includes components which are located only outside containment.

l 1

For those portions of SI which are located outside containment, radiation levels and availability of personnel during non-outage times or during system functional testing may preclude removing insulation while the pressure-retaining bolted connections are pressurized.

For those portions of SI which are located inside containment and VT-2 examined during Modes 1,2, and 3, the same basis for relief as provided for Class 1 above applies except that the system operating temperatures are less, rem! ting in less potential for burn-related l

injuries.

For those portions of SI which are located inside containment and VT-2 examined during Mode 6 and defueled conditions with the reactor vessel head removed, containment i

radiation levels and availability of persormel during system functional testing may preclude

{

removing insulation while the pressure-retaining bolted connections are pressurized.

1 7-109 Rev. 2 l

SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE EI.ECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-26 l

(continued)

IV. Basis for Relief (continued):

Class 2 components (continued):

5.

Basis for Relief for the Nuclear Sampling System - Liquid:

The liquid portions of the Nuclear Sampling System are used for providing samples for analysis purposes of the RCS, CVCS, and RHR. This system is located both inside and l

outside containment and is subject to the same system pressure tests as the systems for which it is used to provide samples. Therefore, the same basis for relief as discussed above for RCS, CVCS, and RHR is applicable to the liquid portions of the Nuclear Sampling System.

Class 3 components:

Table IWD-2500-1, Examination Category D-A requires a system inservice test (IWD-5221) and corresponding VT-2 visual examination of Class 3 components once each inspectior period.

l Subparagraph IWA-5242(a) is applicable to the boric acid storage tank and boric acid transfer portions of CVCS. System inservice tests are performed as follows:

l Some of the system inservice tests are performed on portions of CVCS which are pressurized by the static head of the boric acid storage tank. The scope of these VT-2 examinations includes components which are located only outside containment. The VT-2 examinations on these l

portions of CVCS are generally performed during power operation (Mode 1) but may be performed in other Modes if tank levels are adequate. These system inservice tests are generally performed during power operation and, as such, many, if not all, components will be uninsulated, VT-2 examined, and reinsulated when the pressure-retaining bolted connections are pressurized.

However, availability of personnel during non-outage times may preclude removing insulation while the pressure-retaining bolted connections are pressurized.

i The remainder of the system inservice tests are performed on portions of CVCS which are pressurized when a boric acid transfer pump is operating. The scope of these VT-2 examinations includes components which are located only outside containment. The VT-2 examinations on l

these portions of CVCS are generally performed during power operation (Mode 1) with a boric acid transfer pump running with system valves aligned in a recirculation flowpath which precludes injecting high concentrations of borated water into CVCS and ultimately into the RCS.

l The boric acid transfer pumps are operated as necessary to perform system functions and j

necessary testing and, as such, are not continuously in operation. Since these pumps are not 7-110 Rev.2

SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-26 (contir.ned)

IV. Basis for Relief (continuedh Cass 3 comnonents (contin _p.gD:

continuously in operation, availability of personnel during non-outage times may preclude removing insulation while the pressure-retaining bolted connections are pressurized.

The 1983 Edition through Summer 1983 Addenda of ASME Section XI was applicable for the First Ten-Year Interval at VEGP. IWA-5242 of the 1983 Edition through Summer 1983 Addenda of ASME Section XI did not require insulation removal; therefore, this request for reliefwas not needed at VEGP during the First Ten-Year Interval.

V.

Alternate Examination:

Class 1 components:

Insulated Class 1 pressure-retaining bolted connections will be uninsulated and VT-2 examined once each refueling outage while the connections are at atmospheric or static pressures. The bolted connections will be examined again with the insulation installed during the regularly scheduled system pressure test at nominal system operating temperature and pressure as required by Table IWB-2500-1, Examination Category B-P.

Cass 2 and 3 components:

Insulated Class 2 and 3 pressure-retaining bolted connections will be uninsulated and VT-2 examined once per inspection period, when the connection is not at pressure. In addition, where this request for reliefis applied, the frequency between the insulation removals (including the VT-2 examination) will not exceed forty (40) months except as follows.- If a reactor refueling outage is in progress or is scheduled to start within six months when the 40 months expires,

- insulation removal and the VT-2 examination would be allowed to be deferred provided that they are performed prior to plant startup following the reactor refueling outage. These examinations may be performed when the connections are not at the pressures required by IWA-5000, IWC-5000, and IWD-5000 The bolted connections will be examined again with the insulation installed during the regularly

- scheduled (once per inspection period) system pressure test as required by Table IWC-2500-1, Examination Category C-H, and Table IWD-2500-1, Examination Category D-A.

7-111 Rev. 2

SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-26 (continued)

VL Justification for Grantine Relief:

Class I components:

The following items arejustification for granting relief for Class 1 Components:

1.

Evidence ofleakage through pressure-retaining bolted connections which are in systems which are borated for the purpose of controlling reactivity is readily detectable by visual observation when systems are not at operating tem,erature and pressure. The boric acid concentrations are sufficiently high such that bor acid residues will be present ifleakage has occurred at the pressure-retaining bolted cosaiection.

2.

In addition, the ASME Section XI Code Committee has issued Code Case N-533 (copy l

provided as Attachment I to this request for relief) which allows as an alternative for Class 1 pressure-retaining bolted connections that insulation may be removed and VT-2 examined when the connection is not pressurized. The Code Case also requires that any evidence ofleakage be evaluated in accordance with IWA-5250. Refer to Request for Relief RR-25 for details concerning relief from IWA-5250(a)(2).

3.

Compliance with the Code presents the hardships previously discussed which are:

Increased potential for debris to be in containtnent which could migrate to the Containment Emergency Sumps and restrict the suction of the Emergency Core Cooling System during accident (LOCA) conditions. All debris is required to be removed from containment prior to entering Mode 4, Increased potential for personnel heat stress since the containment ambient e

temperature may be as high as 100 F, Increased potential for personnel burn injuries due to installation ofinsulation in e

proximity of components greater than 500 F, l

Increased personnel safety hazard since ladders would have to be used to inspect e

many of the bolted connections and replace the insulation. Temporary work platforms / scaffolding inside containment are removed prior to entering Mode 4, Increased radiation exposure to personnel since temporary shielding is removed prior e

to entering Mode 4, and 7-111a Rev.2 l

SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-26 (continued)

VL Justification foi Grantine Relief (continued):

Increased potential for impacting outage duration due to the amount of manpower requir4 to support insulation removal and examinations during Mode 3 following rebeling o 2tage activities.

4.

For the reasons discussed above, SNC has determined that implementation of the proposed alternatives to the Code requirements provides an acceptable level of quality and safety and therefore requests that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).

Class 2 and 3 components:

The proposed alternative provides the following two-phase methodology for ensuring the integrity of the pressure-retaining bolted connections.

1.

Removing insulation and performing a visual examination (at the modified time period specified in the Alternate Examination), when the pressure-retaining bolted connection is not at pressure, will allow for detection ofpreviously occurring leakage through the presence of boric acid cryr tals. Boric acid concentrations are sufficiently high such that boric acid residues will be present and can be visually observed ifleakage has occurred at the pressure-retaining bolted connection.

2.

Performing a system pressure test on the bolted connection with the insulation in place, utilizing the Code specified holding time to allow time for any leakage to penetrate the insulation, will provide a means of detecting any significant leakage.

The proposed alternative provides reasonable assurance that the structural integrity of the pressure-retaining bolted connections will be maintained, thereby, continuing to provide an acceptable level oflevel of quality and safety. Therefore, approval of this proposed alternative should be authorized pursuant to 10 CFR 50.5Sa(a)(3)(i).

7-111b Rev. 2

l l

l i

SOUTlIERN NUCLEAR OPERATING COMPANY VOGTIE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL JtEOUEST FOR RELIEF NO. RR-26 i

(continued)

VIL Imolementation Schedule:

This request for reliefis applicable to the Second Ten-Year Interval which commenced May 31, i

1997.

VIIL Status:

The Class 1 portion of this request for relief was previously approved by the NRC on October 24,1997 and remains virtually unchanged (except for minor editorial changes) due to this revision to RR-26.

J Reliefis requested by this revision of RR-26 for Class 2 and 3 as discussed herein.

l i

7-111c Rev. 2

SOUTHERN NUCl. EAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-26 (continued)

ATTACHMENT 1 CASE N-533 CASES OF ASME 80tLEn A.9 PRESSURE VESSEt. Coot Approval Date: Maren 14.19sg See Nurnensed aiene Apr orpeuses and any reeninneaan desea, Case N.533 Alternative Requirements for VT 2 Visual Examination of Class 1 lasulated Fiessoas.

Retalaing Belsed Conoscalens Secties XI, Diviales 1 lapiry What alternadve sugeressamu any be used in lieu of those of !WA 3242(a) to remove insulsdom from Class I pressure ratamag behad -a is pMem a VT 2 visual==am?

~

&p+ It is the eptaion of the Coannities that, as an alternatrve to the requirements of IWA-5242(a) to remcvs insulation froic C!sn 1 pressure-retaining bolted cona-== to performs a VT 2 visual amand.

nation, the following ;erpirements shall be met.

(s) A system pressurs test ad YT 2 visual asam.

inscion shall be performed each refueling outage without rerooral of insulation.

(b) Each refueling outage the insulation shaR be removed froen the bolted connection, and a VT 2 vis.

ual e===ination shall be performed.The commestion i

is not required to be pressunned. Any evidemos et leakage shall be svaluated in accordamos with IWA.

s25a.

7-112 Rev.2

SOUTHERN NUCLEAR OPERATING COMPAN s' VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-31 I.

System / Component (s) for Which Reliefis Requested:

Volumetric examination of Reactor Pressure Vessel (RPV) Shell-to-Flange (B-A) welds,11201-V6-001-WO3 and 21201-V6-001-WO3.

II.

Code Requirement:

Item No. Bl.30, Examination Category B-A, Table IWB-2500-1 of ASME Section XI requires a volumetric examine. tion of RPV Shell-to-Flange welds as described in Footnotes 3 and 4 of the referenced table. Footnote 3 states: "If partial examinations are conducted from the flange face, the remaining volumetric examinations required to be conducted from the vessel wall may be performed at or near the end of each inspection in;erval." Footnote 4 states: "The examination of shell-to-flange welds may be performed during the first and third inspection periods in conjunction with the nozzle examinations of Examination Category B-D (Program B). At least

{

fifty percent (50%) of the shell-to-flange weld length shall be examined by the end of the first inspection period, and the remainder by the end of the third inspection periods."

III. Code Requirement from Which Reliefis Requested:

1 As an alternative to the existing schedule, reliefis requested to allow the reschedulinE of fifty percent (50%) of the examinations of the RPV Shell-to-Flange weld (flange face) from the end of the first inspection period to at or near the end of the ten-year inspection interval. Footnote 4, as referenced above, requires at least 50% of the rhell-to-flange weld length to be examined by the end of the first inspection period. This rescheduling would allow examination of the RPV Shell-to-Flange and other RPV welds or appurtenances, i.e., nozzle-to-shell welds, nozzle inside radius, and nozzle-to-safe end welds, as addressed in Request for Relief RR-2, to be performed as part of the " Ten-Year Inservice Inspection" at which time the RPV is usually examined. Refer to the attached Tables 1 and 2 which show the current schedule and the proposed schedule, respectively.

IV. Basis for Relief:

To comply with Table IWB-2500-1 during the First Ten-Year Interval,50% of the RPV Shell-to-Flange weld length was scheduled and examined from the flange face during the first inspection period. The remaining 50% of the weld length was scheduled and examined from the flange face during the third inspection period as allowed by Footnote 4 of Table IWB-2500-1. With the partial examinations being conducted from the flange face, the remaining volumetric examinations from the vessel wall, were scheduled and examined at or near the end of the First Ten-Year Interval as allowed by Footnote 3 of Table IWB-2500-1.

7-127 Rev. 2

SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-31 (continued)

IV. Basis for Relief (continued):

At the end of the First Ten-Year Intervals during the IR6 and 2R6 maintenance / refueling outages at VEGP-1 and 2, respectively, one hundred percent (100%) of the RPV Shell-to-Flange weld length was examined from the flange face, and 100% of the same weld length was examined from the vessel wall. These examinations were part of those performed for the First Ten-Year Interval.

Southern Nuclear Operating Company has concluded that rescheduling of these examinations such that they are performed at or near the end of the inspection interval will have little, if any, effect on the quality of examinations while providing a substantial benefit to SNC. This conclusion is based on the following:

1. Examination Quality - This request for reliefinvolves rescheduling the first period examinations (50% of the RPV Shell-to-Flange weld length, from the flange face) such that the entire weld length will be examined at or near the end of the inspection interval. Fifty percent of the weld length examined from the flange face during the first inspection period of the First Ten-Year Interval was re-examined during maintenance / refueling outages IR6 and 2R6. This was done voluntarily by SNC in order to "re-zero" the examination so that no g

mere than ten years would elapse before being examined again at or near the end of the Second Ten-Year Interval contingent upon NRC approval of this request for relief. By "re-zeroing" thGPV Shell-to-Flange weld during the First Ten-Year Interval such that no more than ten yv[ Aculd ebse before being examined again at or near the end of the Second Ten-Year Interval, the proposed rescheduling of this examination should have no effect on the quality of the overall RPV Shell-to-Flange examinations.

2. Examination Continuity - The potential evaluation of flange face examinations presents difficulties due to limited transducer manipulation, long metal paths, and extensi.ve beam spread limitations (Refer to Fi ure I which depicts the RPV Shell-to-Flange weld). As a b

result, accurate sizing technology is not as practical for examinations conducted from the flange face when compared to those conducted from the vessel wall. By using proven autometed sizing technology from the vessel wall such as that used during the " Ten-Year Inservice Inspection", more accurate sizing results are achievable. With approval of this j

request for relief, SNC would only use a RPV flange face examination tool and a mechanized RPV inspection tool once at or near the end of the inspection interval. In addition, approval of this request for relief would allow SNC to re-synchronize the examination schedule, thus providing a readily available means of characterizing any indications observed from the flange face examination, should they occur.

l 7-128 Rev. 2

r.

SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-31 (continued)

In addition to the foregoing, SNC is anticipating NRC approval to re-schedule the RPV Examination Category B-D and B-F welds or appurtenances, i.e., Nozzle-to-Vessel, Inside Radius Sections, and Nozzle-to-Safe End welds, to the end of the inspection interval, as submitted to the NRC in Request for Relief RR-2. In that request for relief, SNC requested permission from the NRC to use ASME Section XI Code Case N-521," Alternative Rules for Deferral ofInspections of Nozzle-to-Vessel Welds, Inside Radius Sections, and Nozzle-to-Safe End Welds of a Pressurized Water Reactor (PWR) Vessel". Use of that Code Case would allow for the rescheduling of Examination Category B-D and B-F welds or appurtenances to be examined at or near the end of the inspection interval provided that certain conditions were met, e.g., no inservice repairs or replacements by welding, no identified fisws or relevant conditions that currently require successive inspections, and the unit is not in the first inspection interval.

Although a different Examination Category is involved, i.e., B-A, than those addressed in Code Case N-521, the VEGP-1 and 2 RPV Shell-to-Flange welds meet those same conditions in that examinations from both the flange face (100%) and the vessel wall (100%) were performed at the end of their respective first inspection intervals. In addition, no inservice repairs or replacements by welding have occurred and no flaws or relevant conditions that require successive inspections were identified for that particular weld on either VEGP unit.

Finally, the proposed rescheduling allows SNC significant opportunities for savings in contractor cost, critical path time, radiation exposure, and internal manpower requirements while still maintaining compliance with the examination requirements of the Code. Approval of this request for relief for the RPV Shell-to-Flange (B-A) welds,' in conjunction with approval of Request for Relief RR-2, would allow the B-A, B-D, and B-F welds or appurtenances to be examined at the same time. By performing the examinations at the same time, additional radiation exposure reduction can be realized, as well as reductions in mobilization and coordination efforts, set-up time, and examination time. In addition, performance of the examinations at one time at or near the end of the Second Ten-Year Interval constitutes a Cost-

- Beneficial Licensing Action (CBLA) in that savings in excess of $100,000 are expected to be realized over the remaining lives of the two VEGP units due in part to savings from not having to 1

mobilize the RPV inspection vendor in the first inspection period of the inspection intervals.

I V.

Alternate Examination:

The volumetric examinations of the RPV Shell-to-Flange welds will continue to be perfonned at VEGP-1 and 2. However, the entire length of these welds will be scheduled and examined from the flange face at or near the end of the Second Ten-Year Interval, instead of 50% of the weld length being examined from the flange face in the first inspection period and the remainder in the third inspection period. The entire length of the RPV Shell-to-Flange welds will be scheduled and examined from the vessel wall at or near the end of the Second Ten-Year Interval.

7-129 Rev. 2

SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2

. SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-31 (continued)

VI. Justification for Granting Relief:

The proposed alternative will provide an acceptable level of quality and safety. The welds on VEGP-1 and 2 will still be volumetrically examined during the Second Ten-Year Interval except that the subject welds will have their entire length examined from both the flange face and vessel wall at or near the end of the interval rather than examining just a portion of the welds from the flange face during the first inspection period. Approval of the proposed alternative will not result in more than ten years elapsing between the RPV Shell-to-Flange weld examinations conducted during the First Ten-Year Interval and those scheduled for examination during thc Second Ten-Year Interval. Therefore, it is requested that this alternative be authorized pursuant to 10 CFR 50.5Sa(aX3Xi). Denial of this request for relief would eliminate potential opportunities for savings in contractor costs, radiation exposure, and internal manpower requirements.

VII. Implementation Schedule:

This request for reliefis applicable to the Second Ten-Year Interval which commenced May 31, 1997.

l 7-130 Rev.2 l

SOUTilERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-31 TABLE 1 -(Current Schedule)

INTERVAL 1 INTERVAL 2 INTERVAL 3 INTERVAL 4 1

2 3

1 2

3 1

2 3

1 2

3 A 50% of B-50% of A-50% of B-50% of A-50% of B-50% of A-50% of B-50% of weld weld.

weld weld weld weld weld weld length.

length length length length length length length from from from from from from from from Flange Flange Flange Flange Flange

- Flange Flange Flange Face Face Face Face Face Face Face Face C-100%

C-100%

ofwcld ofweld ofweld ofweld length length length length from from from from Vessel Vessel Vessel Vessel Wall Wall Wall Wall

}

TABLE 2 -(Proposed Schedule)

INTERVAL 1 INTERVAL 2 INTERVAL 3 INTERVAL 4 I

I 2

3 1

2 3

1 2

3 1

2 3

A-50% of B-50% of B-50% of B-50% of B-50% of weld weld weld weld weld length length length length length from from from from from Flange Flange Flange Flange Flange Face Face Face Face Face A-50% of A-50% of A-50% of A-50% of weld weld weld weld length length length length from from from from Flange Flange Flange Flange j

Face Face Face Face C-100%

C-100%

l ofweld ofweld ofweld ofweld length length length length from from from from Vessel Vessel Vessel Vessel Wall Wall Wall Wall Legend A - Initial 50% of shell-to-flange weld length examined from the flange face.

B - Remaining 50% of shell-to-flange weld length examined from the flange face.

C - 100% of shell-to-flange weld length examined with mechanized equipment from the vessel wall.

7-131 Rev. 2

ML20209H395

Text

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