ML20202H985

Safety Evaluation Accepting Rev 8 to First 10-year Interval Inservice Insp Program & Associated Requests for Relief for Vogtle Electric Generating Plant,Unit 1

ML20202H985
Person / Time
Site:	Vogtle
Issue date:	01/29/1999
From:	NRC (Affiliation Not Assigned)
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ML20202H975	List: ML20202H969 ML20202H985
References
NUDOCS 9902090018
Download: ML20202H985 (25)

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PM y i UNITED STATES g

o j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30666 0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REVISION 8. FIRST 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM AND 1

ASSOCIATED REOUESTS FOR REllEF SOUTHERN NUCLEAR OPERATING COMPANY. INC.. ET AL.

VOGTLE ELECTRIC GENERATING PLANT. UNIT 1 DOCKET NO. 50-424 l

1.0 INTRODUCTION

The Technical Specifications (TS) for Vogtle Electric Generating Plant, Unit 1 (VEGP-1), state that the inservice inspection (ISI) of the American Society of Mechanical Engineers (ASME)  !

Class 1,2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code (ASME Code or Code) and applicable addenda as required by Title 10 of the Code of Federal Reaulations (10 CFR) Section 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).

Section 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.  ;

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 cor.1ponents (including supports) shall 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," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for VEGP-1's first 10-year ISI interval is the 1983 Edition through Summer 1983 Addenda.

By letter dated December 1,1997, Southern Nuclear Operating Company, Inc. (the licensee),

submitted its First 10-Year Interval ISI Program, Revision 8, and associated requests for relief (RRs) for VEGP-1. Additional information was provided by the licensee in its letters dated May 21, May 26, and October 28,1998.

9902090018 990129 PDR ADOCK 05000424 0 PDR

2.0 EVALUATION The staff, with technical assistance from its contractor, the Idaho National Engineering and Environmentel Laboratory (INEEL), has evaluated the information provided by the licensee in support of its First 10-Year ISl Program, Revision 8, and associated RRs for VEGP-1. Based on the results of the review, the staff adopts the contractor 4 conclusions and recommendations presented in the Technical Letter Report, attached, with the exception of the conclusions regarding RR-5.

The Code of record for the VEGP-1 first 10-year ISI interval, which ended May 30,1997, is the 1983 Edition through Summer 1983 Addenda of Section XI of the ASME Code. Revision 8 of the VEGP-1 first 10-year intervalISI program modifies existing RRs, adds two new RRs, and makes minor editorial changes. Since changes to the ISI program are limited to acceptable, editorial changes, the ensuing evaluation is limited to the RRs. The information provided by the licensee in support of the RRs has been evaluated and the bases for disposition are documented below.

Proposed Attemative to 10 CFR 50.55afo)(6)(ii)(A) Auamented Reactor Pressure Vessel (RPV)

Examination:

In accordance with 10 CFR 50.55a(g)(6)(ii)(A), alllicensees must implement once, as part of the ISI intervalin effect on September 8,1992, an augmented volumetric examination of the RPV welds specified in item B1.10 of Examination Category B A of the 1989 Edition of the ASME Code,Section XI. Examination Category B-A, items B1.11 and 81.12 require volumetric examination of essentially 100 percent of the RPV circumferential and longitudinal shell welds, as defined by Figures IWB-2500-1 and -2, respectively. Essentially,100 percent, as defined by 10 CFR 50.55a(g)(6)(ii)(A)(2), is greater than 90 percent of the examination volume of each weld.

In acccidance with 10 CFR 50.55a(g)(6)(ii)(A), the licensee proposed an alternative to the examination coverage requirements of the augmented RPV examination for the welds listed in the table below.

Weld # '

ftern # Limitation Covera 0e WO4 B1.11 Main Loop Nozzles >90%

WOS NA 100 %

WO6 Core Support Lugs 62%

W12 B1.12 Main Loop Nozzles 75 %

W13 Ma'n Loop Nozzles 80%

W14 Main Loop Nozzles 85%

W15 NA 100 %

W16 NA 100 %

W17 NA 100 %

W18 Core Support Lugs 77%

W19 Core Support Lugs 77 %

W20 Core Support Lugs 77 %

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I I 1 To meet the augmented reactor vessel examination requirements of 10 CFR 50.55a(g)(6)(ii)(A),

licensees must volumetrically examine essentially 100 percent of each of the item B1.10 shell welds. In accordance with the regulations, essentially 100 percent is defined as greater than 90 percent of the examination volume of each weld.

At VEGP-1, the augmented examination coverage requirements cannot be met for seven shell 3

) welds due to physical restrictions, such as core barrel support lugs and main loop nozzles that l limit scan coverage. To achieve complete coverage for the subject welds, design modifications l would be required to increase access from the inside diameter (ID) surface.

l As a result of the augmented volumetric examination rule, licensees must make a reasonable effort to maximize examination coverage of their reactor vessels. In cases where examination coverage from the ID is inadequate, examination from the outside diameter (OD) surface using manual inspection techniques should be considered. This option was considered for VEGP-1.

However, access to the welds is either limited by the biological shield wall or by excessive radiation exposure. Therefore, it is concluded that examination from the OD surface is not a viable option.

The licensee has examined a considerable portion (a62 percent) of each RPV shell weld and has obtained cumulative coverage of all RPV shell welds of better than 85 percent. Based on the cumulative volumetric examination coverage obtained, the staff concludes that any significant patterns of degradation, if present, would have been detected and that the examinations performed provide an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5) and 10 CFR 50.55a(a)(3)(i).

Reouest for Relief RR-2:

The ASME Code,Section XI, Examination Categoiy B-A, item B1.11 requires 100 percent volumetric examination of RPV circumferential shell welds as defined in Figure IWB-2500-1. In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the examination coverage requirements of the Code for lower shell-to head Weld 11201-V6-001-WO6.

The Code requires 100 percent volumetric examination of the subject circumferential RPV weld.

However, access to this weld is restricted by core support lugs that preclude 100 percent volumetric examination. Therefore, the Code-required examination is impractical for this weld.

To meet the Code examination requirements, design modifications would be necessary to provide access for examination. Imposition of this requirement would cause a considerable burden on the licensee.

The licensee examined 62 percent of the subject weld. In addition, the licensee has examined a significant portion of the remaining RPV shell welds. The staff concludes that the examinations performed were sufficient to detect any existing patterns of degradation and that they provide reasonable assurance of the structuralintegrity of the subject RPV weld.

Therefore, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i).

Reauest for Relief RR-3:

The ASME Code,Section XI, Examination Category B-A, item B1.12 requires 100 percent volumetric examination of RPV longitudinal shell welds as defined in Figure IWB-2500-2. In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the examination coverage requirements of the Code for lower shell longitudinal Welds 11201-V6-001-W18, 11201-V6-001-W19, and 11201-V6-001-W20.

The licensee proposed no supplementary examination; however, an overall, general visual examination (VT 3) of the RPV was pedormed in accordance with the requirements of ASME Section XI, Category B-N-1, item No. B13.10, during the maintenance / refueling outage in which welds 11201-V6-001-W18,11201-V6-001-W19, and 11201-V6-001-W20 were examined volumetrically, i.e., during VEGP-1 Maintenance / Refueling Outage 1R6.

The Code requires 100 percent volumetric examination of the subject longitudinal RPV welds.

However, access to these welds is restricted by core support lugs that are welded over the longitudinal welds and preclude 100 percent volumetric examination. Therefore, the Code-required examination is impractical for these welds. To meet the Code examination coverage requirements, design modifications would be necessary to provide access for examination. Imposition of this requirement would cause a considerable burden on the licensee.

The licensee examined a significant portion (77 percent) of each weld. In addition, the licensee has examined a significant portion of the remaining RPV shell welds. Based on the examinations performed, the staff concludes that any patterns of degradation,if present, would have been detected, and that the licensee's examinations performed provide reasonable assurance of the structuralintegrity of the RPV welds. Therefore, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i).

Reauest for Relief RR-4:

The licensee was able to meet the ASME Code,Section XI, Examination Category B- A, Itelt B1.20, Reactor Pressure Vessel Meridional Welds coverage requirements for RPV meridionO Welds 11201-V6-001-W21,11201-V6-001-W22,11201-V6-001-W23, and 11201-V6-001-W20 RR-4 was withdrawn by the licensee in its letter dated December 1,1997.

Reauest for Relief RR-5:

The ASME Code,Section XI, Examination Category B-A, item B1.21, requires 100 percent volumetric examination of the RPV circumferential head welds as defined by Figure IWB-2500-3. RR 5 was originally evaluated and granted in an NRC Safety Evaluation (SE) dated November 26,1991. ,

in accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed an alternative to the Code examination requirements for circumferential head Weld 11201-V6-001-WO7. The licensee stated:

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Because of the physicalinterference presented by the instrumentation tubes in the proximity of the subject weld, the examination volume coverage is limited to approximately 29% of the weld length during inservice inspection.

No supplementary examination is proposed. It should be noted however that l

an overall, general visual examination (VT-3) of the RPV was performed in accordance with the requirements of ASME Section XI, Category B-N-1, item No. B13.10, during the maintenance / refuel;ng outage in which weld 11201-V6-001-WO7 was volumetrically examined, i.e., during VEGP-1 Maintenance / Refueling Outage 1R6.

The Code requires 100 percent volumetric examination of RPV circumferential head welds.

,However, as stated by the licensee, the subject RPV head weld could not be completely examined due to in-core instrumentation tubes that restrict access to the weld. Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee contends that the examinations performed, in conjunction with the Code-required VT-3 visual examination of RPV internals, provide an acceptable level of quality and safety.

This request was originally evaluated and granted in an NRC SE dated November 26,1991, based on the 74 percent examination coverage achieved during preservice examinations. In the NRC's request for additional information, the licensee was requested to provide technical justification for the significant reduction in coverage. in response, the licensee attributed the reduction in coverage to differences in coverage calculation methods. Review of the coverage obtained for other RPV welds at VEGP-1 indicates that coverage obtained during the first interval examinations were comparable to previous preservice inspection results. The staff determined that repeating previous results exactly is difficult, especially when tooling changes occur. Howeve~r, the examination coverage obtained for Weld 11201-V6-001-WO7 is net comparable to previous results and since the coverage for other RPV welds did not change appreciably, the significant reduction in coverage had not been justified. Furthermore, the licensee had not provided adequate information to justify an acceptable level of quality and j safety. Therefore, the licensee's proposed alternative was denied. l By letter dated October 28,1998, the acensee submitted a revision to the subject relief request that provided clarification to staff's concern on the difference in the volumetric coverage between the preservice inspection and the first 10-year ISI with reasonable assurance of structuralintegrity based on the extent of examination coverage. The licensee futher i demonstrated that compliance to the Wde requirement would result in hardship without a compensating increase in the level of quality and safety, pursuant to 10 CFR 50.55a(a)(3)(ii).

The staff noted from the submittal that the volumetric examination coverage by ultrasonics was greater during preservice inspection with the " immersion type" examination than that of the

" contact type" examination used during the first 10-year ISI. Because of physical interference from the instrumentation tubes in the proximty of Weld 11201-V6-001-WO7, the " contact type" examination resulted in a lesser volumetric coverage than that of the " immersion type" examination. If there were no physical constraints in scanning of the weld, both types of examination would have given comparable volumetric coverage. The staff has evaluated the feasibility of substantiating the volumetric coverage as stated in the licensee's submittal and has determined that scanning the weld mannually from outside the surface of the vessel may

i 6-significant radiation penalty and outage time. This is a hardship to the licensee without a compensating increase in the level of quality and safety.

The staff has assessed the structuralintegrity of the subject weld from the information provided by the licensee in its submittal dated October 28,1998, and has concluded that there is reasonable assurance of structuralintegrity of the weld for the reasons stated below.

The preservice inspection of the weld did not identify any recordable indication in the weld 1 or in the base metal.

4 There is no degradation mechanism known to exist in the weld, which has a cladding on the inside surface.

l The fatigue usage factor for the weld was calculated to be 0.01 for the design life of the vessel, being well below 1.

The weld is located far outside the vessel beltline in a low neutron fluence region, j

. precluding any harmful embrittlement of the weld and the basemetal due to neutron l irradiation.

The staff, therefore, has determined that supplementing the present examination coverage with '

examination from outside the surface of the vessel will result in high man-rem exposure to personnel and, thus, defeat the "as low as reasonably achievable" principles. Thus, Code compliance would result in hardship to the licensee without a compensating increase in the level of quality and safety given the assurance of structural integrity that is provided by the alternative.

Pursuant to 10 CFR 50.55a(a)(3)(ii), the revised RR-5 !s, therefore, authorized for the first 10-year ISI interval for VEGP-1.

! Reouest for Relief RR-30. Revision 8:

ASME Code,Section XI, Examination Categories C-A, C-8, and C-C, items C1.10, C1.20, C1.30, C2.21, C3.10, and C3.10, Class 2 Pressure-Retaining Welds and Integral Attachment

Welds.

RR-30 was previously evaluated and granted in an NRC SE dated November 26,1991. It was subsequently revised to include two Examination Category C-C welds; the revision was evaluated and relief granted pursuant to 10 CFR 50.55a(g)(6)(i) in an NRC SE dated March 8, 1996. In its December 1,1997, letter, the licensee added Boron injection Tank integral attachment Weld 11204-V6-001-WO5 to RR-30. The limitation and examination coverage obtained and the impracticality of conducting an examination for full coverage is similar to those of the previously evaluated welds and imposition of the Code would require replacement of the i affected component. Therefore, the addition of Weld 11204-V6-001-WO5 is consistent with the conclusions of the previous evaluations and RR-30, Revision 8, is granted pursuant to 10 CFR 50.55a(g)(G)(i).

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Reauest for Relief RR-63:

The' ASME Code,Section XI, Examination Category B-A, Item B1.12 requires 100 percent volumetric examination of RPV longitudinal shell welds as defined in Figure IWB-2500-2. In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code's examination requirements for RPV longitudinal shell Welds 11201-V6-001-W12,11201-V6-001-  !

W13, and 11201-V6-001-W14.

The licensee did not propose a supplemental examination; however, it should be noted that an ,

overall, general VT-3 of the RPV was performed in accordance with the requirements of ASME ] '

Section XI, Category B-N-1, item No. B13.10, during the maintenance / refueling outage in which the subject welds were volumetrically examined, i.e., during VEGP-1 Maintenance / Refueling Outage 1R6.

The Code requires 100 percent volumetric examination of the subject RPV welds; however, the examination is restricted by adjacent nozzles that make the 100 percent volumatric examination impractical for these welds. To gain access for examination, the RPV would require design modifications to eliminate the nozzle obstructions. Imposition of this requirement would create an undue burden on the licensee.

The licensee has examined a significant portion (75-85 percent) of each of these welds. In addition, other RPV welds have been examined to the extent required by the Code. Therefore, any existing patterns of degradation would have been detected by the examinations that were completed and reasonable assurance of structuralintegrity has been provided. The staff ,

concludes that, based on the impracticality of meeting the Code examination coverage j requirements for the subject welds, and the reasonable assurance provided by the  :

examinations that were cornpleted on these and other welds, relief is granted pursuant to 1 10 CFR 50.55a(g)(6)(i).

Reauest for Relief RR-64: l The ASME Code,Section XI requires that repairs and replacements be performed in accordance with IWA-4000 and lWA-7000, respectively.

In accordance with 10 CFR 50.55a(a)(3)(ii), the licensee proposed to use the requirements of IWA-4130 of the 1995 Addenda of ASME Section XI to exempt items 1-inch nominal pipe size (NPS) and smaller.

The Code requires that repairs and replacements be performed in accordance with IWA-4000 and lWA 7000, respectively.' Pursuant to IWA-7400, piping, valves, and fittings NPS 1 inch and smaller are exempt from the requirements of Article IWA-7000. However, Article IWA-4000 has no exemption criteria for components NPS 1 inch and smaller. Therefore, some licensees may choose to replace rather than repair items to avoid the repair requirements of IWA-4000. As an attemative, the licensee proposed to use the requirements of IWA-4130 of the 1995 Addenda to exempt items 1-inch NPS and smaller from the repair and replacement requirements of the Code. This is equivalent to the alternative contained in Code Case N-544, Repair and

Replacement of Small items, which has not been approved for general use by the NRC in Regulatory Guide 1.147. The licensee, however, has not specifically asked that use of Code Case N-544 be authorized.

In accordance with the 1995 Addenda and Code Case N-544, piping, valves, and fittings NPS 1 inch and smaller, except for heat exchanger tubing and sleeves and welded plugs used for heat exchanger tubing, are exempt from both repair and replacement requirements of the Code. The exemption criteria used for the repair of items NPS 1 inch and smaller is comparable to ef = ting Code requirements for the replacement of similar items. Therefore, the staff concludes uat IWA-4130 of the 1995 Addenda provides reasonable assurance of operational readiness with one exception. The ASME Section XI Code differentiates between steam generators and heat exchangers by providing separate item numbers. As currently written, the 1995 Addenda does not address steam generator tubing, only heat exchanger tubing. Therefore,it appears that steam generator tubing could be exempt from repair and replacement requirements by Code Case N-544. To address this uncertainty, the licensee has stated that steam generators are considered heat exchangers and would be excluded from the alternative requirements of IWA-4130 w;th the heat exchanger tubing, sleeves, and welded plugs used for plugging heat exchanger tubing. The staff finds this acceptable.

As stated by the licensee, imposition of the Code requirements would necessitate a significant effort to review over 50,000 work orders to complete paperwork that is not required by later Code addenda. This effort could potentially divert plant personnel from other activities that could affect plant safety. Considering that the recordkeeping activities would require documentation of repairs that are not required to be documented by later Codes, the NRC staff concludes that imposition of the applicable Code requirements would result in an undue hardship without a compensating increase in the level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

3.0 CONCLUSION

The staff conc!udes that the Code requirements are impractical to meet for RR-2, RR 3, RR-63, and RR-30, Revision 8; therefore, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i). The relief granted is authorized by law and will not endanger life or property or the common defense and is otherwise in the public interested, given due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

The NRC staff concludes that the licensee's proposed alternative to the augmented RPV examination required by the regulations provides an acceptable level of quality and safety.

Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5) and 10 CFR 50.55a(a)(3)(i).

For RR-5 and RR-64, the staff concludes that the Code requirements result in a hardship without a compensating increase in the level of quality and safety, and the licensee's proposed i

.g.

. alternatives provide reasonable assurance of structuralintegrity of the subject components.

Therefore, the licensee's proposed alternatives are authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

Request for Relief RR-4 was withdrawn by the licensee.

Attachment:

Technical Letter Report Principal Contributors: T. McLellan D. Patnaik D. Jaffe Date: January 29, 1999 i

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TECHNICAL LETTER REPORT ON THE FIRST 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN. REVISION 8 AND REQUESTS FOR RELIEF E.QB 1 SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNIT 1 DOCKET NUMBER: 50-424

1.0 INTRODUCTION

By letter dated December 1,1997, the licensee, Southern Nuclear Operating Company (SNC) submitted Revision 8 to its first 10-year interval inservice inspection (ISI) program for Vogtlo Electric Generating Plant, Unit 1 (VEGP-1). Additional information was provided in letters dated May 21,1998, and May 26,1998, responding to a Nuclear Regulatory Commission (NRC) request for additionalinformation (RAI). The Idaho National Engineering and Environmental Laboratory (INEEL) staff has evaluated the information provided by the licensee in support of these requests for relief in the following section.

2.0 EVALUATION The Code of record for the Vogtle Electric Generating Plant, Unit 1, first 10-year inservice inspection interval, which ended May 30,1997, is the 1983 Edition through Summer 1983 Addenda of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. Revision 8 of the VEGP-1 first 10-year interval ISI Program modifies existing relief requests, adds two new relief requests, and makes minor editorial changes.

Since changes to the Program are limited to editorial changes, the ensuing evaluation is limited to the relief requests. The information provided by the licensee in support of the requests for relief has been evaluated and the bases for disposition are documented below.

1. Prooosed Alternative to 10 CFR 50.55a(a)(6)(ii)(A). Auamented Reactor Pressure Vessel (RPV) Examination Reaulatorv Reauirement: In accordance with 10 CFR 50.55a(g)(6)(ii)(A), all licensees must implement once, as part of the inservice inspection intervalin effect on September 8,1992, an augmented volumetric examination of the RPV welds specified in item B1.10 of Examination Category B-A of the 1989 Edition of the ASME Code,Section XI. Examination Category B-A, items B1.11 and B1.12 require volumetric examination of essentially 100% of the RPV circumferential and longitudinal shell welds, as defined by l Figures IWB-2500-1 and -2, respectively. Essentially 100%, as defined by l 10 CFR 50.55a(g)(6)(ii)(A)(2), is greater than 90% of the examination volume of each l weld.

Licensee's Proposed Alternative: In accordance with 10 CFR 50.55a(g)(6)(ii)(A), the licensee proposed an alternative to the examination coverage requirements of the augmented RPV examination for the welds listed in the table below.

i Attachment

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rgw io em watemY MMMLimitation;

• BA E Coverage

• i i

WO4 B1.11 Main Loop Nozzles >90%

WO5 NA 100 %

WO6 Core Support Lugs 62% I W12 B1.12 Main Loop Nozzles 75 %

W13 Main Loop Nozzles 80%

W14 Main Loop Nozzles 85 %

W15 NA 100 %

W16 NA 100 %

W17 NA 100 %

W18 Core Support Lugs 77 %

W19 Core Support Lugs 77 %

W20 Core Support Lugs 77 %

I Licensee's Basis for Prooosed Alternative (as stated). l Lower Shell to Bottom Head Weld (WO6) and Lonaitudinal Welds (W18. W19. & W20) {

"Six RPV core support lugs are located on the lower shell of the RPV adjacent to lower shell-to-bottom head welds 11201-V6-001-WO6. Three of these six lugs are welded l directly onto intersecting longitudinal welds W18, W19 and W20. i 1

"These core support lugs obstructed movement of the mechanized examination l equipment sled / transducer along the lower shell side (upper scan region) of circumferential weld WO6. As a result, examination coverage of this non-beltline weld from the inside diameter (ID) of the RPV was limited to approximately sixty-two (62%) of the weld length. This result is comparable to the sixty-six percent (66%) coverage reported during preservice examinations (PSI).

" Examination of the affected longitudinal welds undemeath the core support lugs from the  !

ID of the RPV is not physically possible, therefore, the examination volume coverage was limited to approximately seventy-seven (77%) of the weld length for each of the l longitudinal welds. This result is comparable to the seventy-one (71%) coverage i reported during preservice examinations.

" Maximum, practical coverage was obtained for the subject longitudinal welds from the ID; however, performance of supplemental examinations from the RPV outside diameter '

(OD) was evaluated as a possible means of increasing coverage for these welds. These evaluations concluded that supplemental OD examinations could increase the total

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i coverage to " greater than 90%"; however, such coverage was considered impractical due to the associated radiation exposure (estimated as approximately 9.625 REM (R)). This conclusion was based on the following:

General area dose rates at the bottom of the vessel (as measured for VEGP-2 during its sixth maintenance / refueling outage (2R6)) are estimated to be -

approximately 200 millirem / hour (mr/hr) with contact dose rates at the insulation surface approximately 1 Rem / hour (R/hr).

Nondestructive examination (NDE) personnel would need to perform thirteen UT ,

scans for each area receiving the supplemental examinations. It is calculated that the dose to the NDE personnelin performing these examinations would be opproximately SR.

1 Prior to performing examinations, personnel would need to erect any necessary j scaffolding, remove insulation, and perform any required weld preparation in the 1 high radiation field.

]

This effort is further exacerbated by the fact that much of the RPV insulation used I at VEGP was designed using rivets and screws and does not lend itself to easy removal and replacement. After examinations were completed, any scaffolding would need to be removed and insulation would need to be replaced. The actual number of person-hours spent in the vicinity of the RPV would not be known until such an effort was completed; however, the dose is estimated to be approximately 4.75 R.

NDE personnel would need to locate and maik the areas where the supplemental examinations need to be performed. When performing ID examinations, limitations are located in respect to the core support lugs and the RPV flange, using indexing i provided by the automated inspection tool. Translating these locations to the ID with a high degree of confidence would be an extremely difficult task while working l in a high radiation field.

Vocer Sheli Lonaitudinal Welds (W12. W13. W14)

" Physical obstructions, e.g., surface scan interference due to nozzle center bore configuration, created by the RPV nozzles in the proximity of the subject RPV upper shell longitudinal welds prevented 100% volumetric examination of their entire weld length from the ID of the RPV. As a result, the examination volume coverage was limited to approximately seventy-five percent (75%), eighty percent (80%) and eighty-five percent (85%) of the weld length for welds 11201-V6-001-W12,11201-V6-001-W13, and 11201-V6-001-W14, respectively during inservice inspection. Coverage reported during preservice examination was reported as one hundred percent (100%). Immersion techniques were used during preservice examinations versus the contact techniques generally used today by automated NDE vendors; however, for this configuration, the difference is considered to be primarily in the method used to calculated coverage.

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"The maximum, practical coverage was obtained for these welds from the ID.

Supplemental examinations from the OD of the RPV were evaluated but were considered to be impractical because the welds are located behind the biological shield wall.

Conclusion "The areas not receiving ID examinations are not located in the beltline region; therefore, concerns with radiation embritt!cment is not a factor, These welds had a complete ultrasonic examination performed from the OD in the fabrication shop, as a conservative measure, to ensure there were no unacceptable flaws that would need to be evaluated j during preservice examinations. A review of fabrication shop ID and OD data indicates that no indications were observed in the areas not receiving ID inservice coverage; therefore, there is little likelihood of a crack propagating from a fabrication defect in these areas.

"The examination of RPV shell welds provides an acceptable level of quality and safety even though all could not be fully examined. The average examination coverage of all Category B A, item No. B1.10 welds was greater than 85% and each weld (or portions of welds) located in the beltline region, i.e., welds WOS, W15, W16, and W17, received 100% coverage.

"These completed examinations provide reasonable assurance that unacceptable service-induced flaws have not developed in these welds and that RPV shell weld integrity is maintained. The examinations were performed to the extent practical using state-of the-art equipment and techniques within the limitations of design and access of the RPV. The evaluations and examinations performed meet the objectives of the augmented examinations defined in 10 CFR 50.55a(g)(6)(ii)(A), therefore, the proposed alternative should be authorized by the NRC. Based on the results of the examinations discussed above, SNC concludes that the public health and safety will not be endangered."

Evaluation: To meet the augmented reactor vessel examination requirements of 10 CFR 50.55a(g)(6)(ii)(A), licensees must volumetrically examine essentially 100% of each of the item B1.10 shell welds. In accordance with the regulations, essentially 100% is defined '

as greater than 90% of the examination volume of each weld.

At VEGP-1, the augmented examination coverage requirements cannot be met for seven shell welds due to physical restrictions, such as core barrel support lugs and main loop nozzles, that limit scan coverage. To achieve complete coverage for the subject welds, design modifications would be required to increase access from the inside diameter (lD) surface.

As a result of the augmented volumetric examination rule, licensees must make a reasonable effort to maximize examination coverage of their reactor vessels. In cases where examination coverage from the ID is inadequate, examination from the outside diameter (OD) surface using manualinspection techniques should be considered. This option was considered for VEGP-1. However, access to the walds is either limited by the biological shield wall or by excessive radiation exposure. Therefore, it is concluded that examination from the OD surface is not a viable option.

i l

The licensee has examined a considerable portion (262%) of each RPV shell weld and 1

has obtained cumulative coverage of all RPV shell welds of better than 85%. Based on i the cumulative volumetric examination coverage obtained, the INEEL staff concludes that any significant patterns of degradation, if present, would have been detected and that the examinations performed provide an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to

- 10 CFR 50.55a(g)(6)(ii)(A).

' B.

Reauest for Relief RR-2. Examination Cateaorv B-A. Item B1.11. Reactor Pressure Vescel (RPV) Circumferential Shell Weld 11201-V6-001-WO6 Code Reauirement: Examination Category B-A, item B1.11 requires 100% volumetric examination of RPV circumferential shell welds as defined in Figure IWB-2500-1.

Licensee's Code Relief Reauest: In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the examination coverage requirements of the Code for lower shell-to-head Weld 11201-V6-001 WO6.

Licensee's Basis for Proposed Alternative (as stated):

"Six RPV core support lugs are located on the lower shell of the RPV adjacent to RPV lower shell-to-bottom head welds 11201-V6-001-WO6. These core support lugs obstruct movement of the mechanized examination equipment sled / transducer along the lower shell side (upper scan region) of this weld. As a result, examination coverage of this non-beltline weld from the inside diameter (ID) of the RPV was limited to 62% of the weld length. Complete coverage from the ID of the RPV would necessitate redesign and modification of the RPV which is not practical.

"This weld is a non-beltline area weld; therefore, radiation embrittlement is not a factor.

This weld had a complete ultrasonic examination performed from the OD in the fabrication shop, as a conservative measure, to ensure that no unacceptable flaws were present that would required evaluation during preservice examinations. A review of data indicates that no indications were observed in the areas not receiving ID inservice  ;

examinations.

" Compliance with Code coverage requirements would necessitate refabrication of the RPV to perform complete Code examinations from the ID or it would necessitate performance of supplemental examinations from the OD. Refabrication of the RPV to perform the Code required examinations from the ID is not practical and supplemental OD examinations have be evaluated by VEGP as impractical due to radiation exposure considerations. Fabrications shop examinations indicated that no indications were observed in the areas not receiving ID inservice coverage; therefore, there is little ,

likelihood of a crack propagating from a fabrication defect. '

" Examinations performed from the ID, combined with good fabrication shop examination results and lower embrittlement rates (of a non-beltline area) should provide reasonable assurance of the operation readiness of this weld and the RPV. Denial of this relief request would cause and excessive burden on VEGP; therefore, approval should be l granted pursuant to 10 CFR 50.55a(g)(6)(i).

__ . . _ _ . - . _ _ ___ - _ _ _ . _ . = _ _ . _ . _ _ . _ _ _ . _ . . _ . ,

,= ..

Licensee's Prooosed Alternative (as stated):

"No supplemental examination is proposed. However, it should be noted that an overall, general visual examination (VT-3) of the RPV was performed in accordance with the requirements of ASME Section XI, Category B-N-1, Item No. B13.10, during the maintenance / refueling outage in which weld 11201-V6-001-WO6 was examined volumetrically, i.e., during VEGP-1 Maintenance / Refueling Outage 1R6.

Evaluation: The Code requires 100% volumetric examination of the subject circumferential RPV weld. However, access to this weld is restricted by core support lugs that preclude 100% volumetric examination. Therefore, the Code-required examination is impractical for this weld. To meet the Code examination requirements, design modifications would be necessary to provide access for examination. Imposition of this requirement would cause a considerable burden on the licensee.

The licensee examined 62% of the subject weld. In addition, the licensee has examined a significant portion of the remaining RPV shell welds. The INEEL staff concludes that the examinations performed were sufficient to detect any existing patterns of degradation and that they provide reasonable assurance of the structuralintegrity of the subject RPV weld. Therefore, based on the impracticality of meeting the Code examination coverage requirements and the reasonable assurance of structuralintegrity provided by the examinations that were completed, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).

C. Reauest for Relief RR-3. Examination Cateaorv B-A. Item B1.12. Reactor Pressure Vessel (RPV) Lonaitudinal Shell Welds Code Reauirement: Examination Category B A, Item B1.12 requires 100% volumetric examination of RPV longitudinal shell welds as defined in Fi0ure IWB-2500-2.

Licensee's Code Relief Reauest: In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the examination coverage requirements of the Code for lower shell longitudinal Welds 11201-V6-001-W18,11201 V6-001-W19, and 11201-V6-001-W20.

Licensee's Basis for Proposed Alternative (as stated):

" Core support lugs are welded over the subject longitudinal welds in the lower shell of the RPV (see Attachment 1 to this relief request), thereby preventing 100% volumetric examination of their entire weld length. Examination of the affected welds underneath the core support lugs from the inner radius (ID) is not physically possible. Therefore, the examination volume coverage was limited to seventy-seven percent (77%) of the weld length for each of the welds during inservice inspection. (NOTE: These welds intersect circumferential weld 11201-V6-WO6 for which examination is also restricted by core support lugs).

Licensee's attachments not included in this report.

,w . -. , .-m -.

i i

e }

l I

I i "The portions of these welds not receiving an inservice ID examination are located in the non-beltline area weld; therefore, radiation embrittlement is not a factor. These welds had a complete ultrasonic examination performed from the OD in the fabrication shop, as a conservative measure, to ensure that no unacceptable flaws were present that would required evaluation during preservice examinations. A review of data indicates that no indications were observed in the areas not receiving ID inservice examinations.

" Compliance with Code coverage requirements would necessitate refabrication of the RPV to perform complete Code examinations from the ID or it would necessitate i performance of supplemental examinations from the OD. Refabrication of the RPV to perform the Code required examinations from the ID is not practical and supplemental OD examinations have be evaluated by_VEGP as impractical due to radiation exposure considerations. Fabrications shop examinations indicated that no indications were observed in the areas not receipng 10 inservice coverage; therefore, there is little likelihood of a crack propagating from a fabrication defect.

" Examinations performed from the ID, combined with good fabrication shop examination results and lower embrittlement rates (of a non-beltline area) should provide reasonable assurance of the operation readiness of this weld and the RPV. Denial of this relief request would cause and excessive burden on VEGP; therefore, approval should be granted pursuant to 10 CFR 50.55a(g)(6)(i).

Licensee's Proposed Alternative (as stated).

"No supplementary examination is proposed. It should be noted however that an ow all, general visual examination (VT-3) of the RPV was performed in accordance with the requirements of ASME Section XI, Category B-N-1, item No. B13.10, during the maintenance / refueling outage in which welds 11201-V6-001-W18,11201-V6-001-W19, and 11201-V6-001-W20 were examined volumetrically, i.e., during VEGP-1 Maintenance / Refueling Outage 1R6."

Evaluation: The Code requires 100% volumetric examination of the subject longitudinal RPV welds. However, access to these welds is restricted by core support lugs that are welded over the longitudinal welds and preclude 100% volumetric exarnination.

Therefore, the Code-required examination is impractical for these welds. To meet the Code examination coverage requirements, design modifications would be necessary to provide access for examination. Imposition of this requirement would cause a considerable burden on the licensee.

The licensee examined a significant portion (77%) of each weld. In addition, the licensee has examined a significant portion of the remaining RPV shell welds. Based on the examinations performed, the INEEL staff concludes that any patterns of degradation, if present, would have been detected, and that reasonable assurance of the structural integrity of the RPV welds has been provided. Therefore, based on the impracticality'of meeting the Code examination coverage requirements and the reasonable assurance of structural integrity provided by the examinations that were completed, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i). i

D. Reauest for Relief RR-4. Examination Catgrv B-A. Item B1.20. Reactor Pressure  !

Vessel Meridional Weids '

t Note: The licensee was able to meet the Code examination coverage requirements for

. RPV meridional Welds 11201-V6-001-W21,11201-V6-001-W22,11201-V6-001-W23, i and 11201-V6-001 W24. As a result, Request for Relief RR-4 was withdrawn by the ,

licensee in the December 1,1997, letter.

E. Reauest for Relief RR-5. Examination Cateoorv B-A. Item B1.21. Reactor Pressure

. Vessel (RPV) Circumferential Head Weld 11901-V6-001-WO7 Note: Request for Relief RR 5 was originally evaluated and granted in an NRC SER  ;

dated November 26,1991.

Code Reauirement: Examination Category B-A, item B1.21, requires 100% volumetric  ;

examination of the RPV circumferential head welds as defined by Figure IWB-2500-3. ,

t Licensee's Proposed Alternative: In accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed an alternative to the Code examination requirements for circumferential head Weld 11201-V6-001-WO7. The licensee stated:  ;

"Because of the physical interferences presented by the instrumentation tubes in the proximity of the subject weld, the examination volume coverage is limited to approximately 29% of the weld length during inservice inspection. No supplementary examination is proposed. It should be noted however that an overall, general visual examination (VT-3) of the RPV was performed in accordance with the requirements of ASME Section XI, Category B-N-1, item No. B13.10, during the maintenance / refueling outage in which weld 11201-V6-001-WO7 was volumetrically examined, i.e., during VEGP-1 Maintenance / Refueling Outage 1R6."

Licensee's Basis for Prooosed Alternative (as stated):

" Twenty-nine peripheral RPV in-core flux instrumentation tubes adjacent to the RPV bottom head torus-to-bottom head dome weld restrict movement of the mechanized examination equipment sled / transducer along the entire length of weld 11202-V6-001-WO7. As a result, obtaining the required examination coverage in the vicinity of the instrumentation tubes is not possible when performing examinations from the inside diameter (ID) of the RPV. See Attachment 18 to this relief request.

" Physical interferences presented by the twenty-nine RPV in-core flux instrumentation tubes in the vicinity of weld 11202-V6-001-WO7 prevent a full-Code examination, i.e.,

more than 90% of the required examination volume, as defined in ASME Section XI, Code Case N-460. A significant burden would be experienced if the required examination was attempted from the outside diameter (OD) of the RPV, particularly due to radiation dose and efforts to correlate the inner and outer diameter examination coverage plots.

Examination from the OD would not result in a compensating increase in the level of Attachments not included in this report.

. - - - . . . . - . _ - - _  :== - - - - - - - -. - -

I quality and safety. The volumetric examination of weld 11201-V6-WO6 from the ID of the RPV as performed, in conjunction with the overall, general visual examination, provides '

an acceptable level of quality and safety and is therefore justified per 10 CFR 50.55a(a)(3)(i)."

in the May 26,1998, letter, the licensee stated:

"The VEGP-1 preservice inspection (PSI) of the RPV was conducted during September 1985 by Combustion Engineering (CE) using ' immersion' techniques for mechanized examinations. According to the PSI RPV inspection plans and procedures, the reported Code examination coverage was apparently calculated by requiring two angles in the weld and only one in the base material, as a minimum, as allowed by ASME Section V.

Unless clear calculation methods are available, repeating the accumulative result is difficult, if not impossible. Only the interfering conditions, generic tooling movement, and '

en accumulative coverage result were recorded. At the time of the PSI examinations of the RPV, tooling device parameters were'not as advanced as today's applications. As the first ten-year ISI interval progressed, more accurate volume calculations were incorporated and documented for both piping and equipment welds. Current CAD technique drawings and computer-generated tool locations reports provide for a clearer l and more accurate result for RPV examinations, j

"The VEGP-1 first ten-year interval ISI examinations were conducted in April 1996 by l' WesDyne, using ' contact' techniques and the WesDyne Reactor Vessel Inservice Inspection (RVISI) tool. Volume calculations were documented from tooling dimensions  ;

from 0*,45*, and 60* examinations. Along with CAD drawings, the results were I conservatively calculated and weighted with their respective scan requirements (up, down clockwise, counter-clockwise for the 45' and 60* transducers. Due to the limitations with the WesDyne tool and the bottom-mounted instrumentation tubes (BMls), volumetric examination was conservatively calculated to be only twenty-nine percent (29%), as compared to the reported seventy-four percent (74%) from the PSI. The BMis interfered with both axial and circumferential movemonts in most areas. The best practical examination without causing potential damage to both the RVISI tool and/or the BMis was obtained during the VEGP-1 Maintenance / Refueling Outage 1R6. The method for calculating the coverage was simple and relatively repeatable.

"Other calculation methods could have been used to ' claim' additional credit, but were not. Other methods for calculating examination coverage include, but are not limited to, the following: i a) Single direction base material coverage. As allowed by ASME Section V, Article 4, Paragraph T-441.5.1, the base material portion of the examination could have been considered as meeting the Code provided that at least one beam direction passes through the base material. In general, vessel accrued percentages would increase i since most examination limitation are from limited examination coverage of the base material, b) Use of 70* results in the calculations. Although not a requirement of the 1983 )

Edition of ASME Section XI, the 70* examination was performed to satisfy the

.g.

4

=- -

= = = -- _ . - - - - - ,

i  !

requirements of NRC Regulatory Guide (RG) 1.150. In general, the 70' acquired coverage was greater than that of other angles due to smaller volume required (1" of the near surface), thus potentially raising the accumulative total. '

"Since the RPV examinations conducted during VEGP-1 Maintenance / Refueling Outage 1R6, WesDyne has developed a new system called 'SUPREEM' which uses ROSA mechanized technology along with smaller, better designed transducer sleds which, in most instances, increase examination coverage. This system was used during the RPV ten-year examinations conducted during VEGP-2 Maintenance / Refueling Outage 2R6 in  !

March 1998 with good results."

i Evaluation: The Code requires 100% volumetric examination of RPV circumferential head welds. However, as stated by the licensee, the subject RPV head weld cou!d not be completely examined due to in-core instrumentation tubes that restrict access to the weid.

Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee contends that the examinations -

performed, in conjunction with the Code-required VT-3 visual examination of RPV internals, provide an acceptable level of quality in safety.

This request was originally evaluated and granted in an NRC SER dated November 26, 1991, based on the 74% examination coverage achieved during preservice examinations. I in the NRC RAl, the licensee was requested to provide technical justification for the i significant reduction in coverage in response, the licensee attributed the reduction in j coverage to differences in coverage calculation methods. Review of the coverage i obtained for other RPV welds at VEGP-1 indicates that coverage obtained during first interval enminations were comparable to previous PSI results. The INEEL staff agrees that repeating previous results exactly is difficult, especially when tooling changes occur.

However, the examination coverage obtained for Weld 11201-V6-001 WO7 is not comparable to previous results and since the coverage for other RPV welds did not i change appreciably, the significant reduction in coverage has not been justified.

Furthermore, the licensee has not provided adequate information to justify an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative not be authorized.

F. Reauest for Relief RR-30. Revision 8. Examination Cateaories C-A. C-B. and C-C. Items C1.10. C1.20. C1.30. C2.21. C3.10 and C3.10. Class 2 Pressure Retainina Welds and Intearal Attachment Welds Note: Request for Relief RR-30 was previously evaluated and granted in an NRC SER dated November 26,1991. It was subsequently revised to include two Examination Category C-C welds; the revision was evaluated and relief granted pursuant to 10 CFR 50.55a(g)(6)(i) in an NRC SER dated March 8,1996. In the December 1,1997, letter, the licensee added BIT integral attachment Weld 11204-V6-001-WO5 to Request for Relief RR-30. The limitation and examination coverage obtained are similar to those of the previously evaluated welds. Therefore, the addition of Weld 11204-V6-001-WO5 does not alter the conclusions of the previous evaluations and relief should remain granted pursuant to 10 CFR 50.55a(g)(6)(i).

-_,,.m- _ = .. u- -- _ _ - -

7. _ - --_ _ _ _ ~ . _ .

i a i

i.

G. Reauest for Relief RR-63. Examination Cateaorv B-A. Item B1.12. RPV Lonaitudinal ,

Shell Welds Code Reauirement: Examination Category B-A, Item B1.12 requires 100% volumetric  ;

examination of RPV longitudinal shell welds as defined in Figure IWB-2500-2.

Licensee's Code Relief Reauest: In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code's examination requirements for RPV longitudinal L shell Welds 11201-V6-001-W12,11201-V6-001-W13, and 11201-V6-001-W14.

Licensee's Basis for Prooosed Alternative (as stated):

" Physical obstructions, e.g., surface scan interference due to nonle center bore configuration, created by the RPV nonles in the proximity of the subject RPV upper shell longitudinal welds prevented 100% volumetric examination of their entire weld length from i the inside diameter (ID). As a result, the examination volume coverage was limited to approximately 75%,80%, and 85% of the weld length for welds 11201-V6-001-W12,  !

11201-V6-001-W13, and 11201-V6-001-W14, respectively, during inservice inspection.

Supplemental outside diameter (OD) examinations are not practical because the welds are located behind the biological shield wall. i "These welds had a complete ultrasonic examination performed from the OD in the fabrication shop, as a conservative measure, to ensure that no unacceptable flaws were present that would require evaluation during preservice examinations. A review of data indicates that no indications were observed in the areas not receiving ID inservice coverage.

" Compliance with Code coverage requirements would necessitate refabrication of the RPV, which is not practical.. Fabrication shop examinations indicate that no indications were observed in the areas not receiving ID inservice coverage; therefore, there is little i likelihood of a crack propagating from a fabrication defect. Examinations performed from )

the ID, combined with good fabrication shop examination results should provide  !

reasonable assurance of the operation readiness of this weld and the RPV, Denial of this l

relief request would cause and excessive burden on VEGP; therefore, approval should be l granted pursuant to 10 CFR LO.55a(g)(6)(i).

l Licensee's Proposed Alternative (as stated):

"No supplemental examination is proposed. It should be noted however that an overall, general visual examination (VT-3) of the RPV was performed in accordance with the requirements of ASME Section XI, Category B-N-1, item No. B13.10, during the i maintenance / refueling outage in which the subject welds were volumetrically examined,  :

1.e., during VEGP-1 Maintenance / Refueling Outage 1R6."

Evaluation: The Code requires 100% volumetric examination of the subject RPV welds.

However, the examination is restricted by adjacent nonles that make the 100%

volumetric examination impractical for these welds. To gain access for examination, the RPV would require design modifications to eliminate the nonle obstructions. Imposition of this requirement would create an undue burden on the licensee.

The licensee has examined a significant portion (75-85%) of each of these welds. In addition, other RPV welds have been examined to the extent required by the Code. )

Therefore, any existing patterns of degradation would have been detected by the examinations that were completed and reasonable assurance of structuralintegrity has been provided. Based on the impracticality of meeting the Code examination coverage requirements for the subject welds, and the reasonable assurance provided by the examinations that were completed on these and other welds, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).

H. Reauest for Relief RR-64. Exemotion of items 1-inch and Smaller From the Reoair l Reauirements of IWA-4000 '

l Code Reauirement: The Code requires that repairs and replacements be performed in ,

accordance with IWA-4000 and IWA-7000, respectively.  !

4 Licensee's Proposed Alternative: In accordance with 10 CFR 50.55a(a)(3)(ii), the licensee proposed to use the requirements of IWA-4130 of the 1995 Addenda of ASME i Section XI to exempt items 1-inch NPS and smaller. The licensee stated:

"lWA-4130 of the 1995 Addenda to ASME Section XI allows the application of alternative requirements of replacement to al/repairand replacementactivities. Although after the )

fact, it is believed that these have been met for VEGP-1 except possibly for piping and components 1-inch NPS and smaller, ASME Code Class 1,2, and 3 heat exchangers, and steam generators, e.g., steam generator level taps. Because of the numerous MWOs that would have to be reviewed for possible applicability, GPC and its successor, SNC, do not wish to backfit the record keeping requirements of the 1983 Edition of ASME Section XI with Addenda through Summer 1983 for any past instances involving repairs to piping and components 1-inch NPS and smaller, particularly in light of the equalization of repair and replacement requirements as found in the 1995 Addenda to the Code."

Licensee's Basis for Proposed Alternative (as stated):

"Around October 1996, an Authorized Nuclear inspector (ANil) at VEGP identified a possible non-compliance with the requirements associated with record keeping for repairs of piping and components 1-inch NPS and smaller. Review of the VEGP Repair / Replacement Program during VEGP-2 Maintenance / Refueling Outage 2R5 in Fall 1996 confirmed the ANil's concern and revealed that repairs to piping and components 1-inch NPS and smaller were being treated similar to replacement of piping and components of that size, i.e., it was believed that repairs of piping and components 1-inch NPS and smaller were exempt, including the requirements associated with record keeping. As a result of this misunderstanding of the Code requirements for repairs, both VEGP units appear to have been in non-compliance with the repair requirements, including those associated with record keeping, e.g., use of ASME Form NIS-2 (Owner's Report for Repairs and Replacements), since the beginning of commercial operation ,

through October 1996 (time of discovery) for repairs of piping and components 1-inch NPS and smaller. The NRC was advised of this potential non-compliance with the requirements of the ASME Section XI Code in Georgia Power Company (GPC) letter LCV-0932 dated January 8,1997. Complicating the issue was direction reportedly given by an ANil who preceded the ANil who identified the potential noncompliance.

Reportedly, the previous ANil indicated to plant personnel responsible for repair activities

_.,___7 that he did not wish to witness and presumably otherwise verify repairs to piping and components 1-inch NPS and smaller. To the best of the knowledge and belief of GPC, the former licensee and operator of VEGP, and its sister company, Southern Nuclear Operating Company (SNC), the current licensee and operator of VEGP, any such repairs

• performed to piping and components 1-inch NPS and smaller were technically sound and i were performed in accordance with approved procedures.

t "Because a significant effort would be required to review approximately fifty-two thousand (52,000) Maintenance Work Orders (MWOs) generated for VEGP-1 for the period from May 31,1987 (date of commercial operation) through October 1996 (time of discovery) for possible noncompliance with a record keeping requirement, it is our position that ,

complying with the record keeping requirements, including use, of ASME Form NIS-2 for any such repairs of piping and componer'ts 1-inch NPS and smaller, would not provide a i commensurate increase in the level of safety were the Code requirements for record keeping imposed. As a result, relief is requested from the record keeping requirements for repairs to piping and components 1-inch NPS and smaller. Similar relief will be requested for VEGP-2 and will be submitted to the NRC for review and approval under separate cover as part of Revision 8 to VEGP-2 ISI Program document ISI-014. Since the identification of this potential non-compliance, plant personnel responsible for repair activities have been instructed that repairs, irrespective of the size of piping and component involved, are to be properly documented, in addition, the ASME Section XI Repair / Replacement Program was revised to require that repairs, irrespective of the size of the component involved, were to be properly documented. Further, in March 1997, a training course on ASME Section XI (with emphasis on repair and replacement requirements) was held and included personnel directly involved with repair and replacement activities at VEGP. To the best of our knowledge and belief, VEGP-1 has been in Code compliance with the repair requirements for piping and components 1 inch NPS and smaller since October 1996.

"lWA-4130 in the 1995 Addenda to ASME Section XI allows application of alternative requirements for replacement to alt repair and replacement activities. The alternative requirements are specifically addressed in paragraphs IWA-4131 and IWA-4132 of IWA-4130 and exclude Class 1,2, and 3 heat exchanger tubes, sleeves, and welded plugs for ,

heat exchanger tubes. It is the position of SNC that steam generators would be similarly j excluded since they are considered at VEGP to be heat exchangers. Had the  !

requirements of this later addenda of the ASME Section XI Code been in effect for the period in question, the record keeping requirements for piping and components 1-inch NPS and smaller would not have been required except for tubes, sleeves, and welded plugs for Class 1,2, and 3 heat exchangers and steam generators. As a result, it is our position that compliance with the repair requirements of the 1983 Edition of ASME Section XI with Addenda through Summer 1983 for piping and components 1-inch NPS and smaller does not provide a commensurate increase in the level of safety were these repair requirements to be imposed.

"The 1983 Edition of ASME Section XI Code with Addenda through Summer 1983 provides an exemption for replacement items 1-inch NPS and smaller from the requirements of IWA-7000, but repairs to such items are not similarly exempted.

Therefore, a repair to an item is subject to more restrictive requirements than replacing

. them.

"lWA 4130 in the 1995 Addenda to ASME Section XI allows application of alternative requirements for replacement to g/Irepairand replacement activities. The attemative requirements are specifically addressed in paragraphs lWA-4131 and IWA-4132 of IWA-4130. Heat exchanger tubing, sleeves, and welded plugs used for plugging heat exchanger tubes for Class 1,2, and 3 systems are excluded from the alternative requirements of lWA-4130.

"It is the position of SNC that steam generators at VEGP would be similarly excluded since they are considered to be heat exchangers.

A "Except for steam generator tube plugs which were welded in place at the time the VEGP steam generators were manufactured, there have been no steam generator tube plugs installed by welding. All steam generator tube plugs currently installed (except for those welded plugs installed by the steam generator manufacturer) are of the mechanical type.

No steam generator tubes have been repaired using welding nor have any sleeves been installed using welding during plant operation for the period of time in question. While ASME Section XI provides guidance for repair activities with respect to the steam generators, plant Technical Specifications in effect at that time (i.e., pre-improved Technical Specification 3/4.4.5) provided requirements which were required to be met, including those for tube inspection and tube plugging. However, during the period of time in question there may have been repairs made to other piping 1-inch NPS and smaller associated with the steam generators, e.g. piping for level taps. Similarly, during the period in question, there may have been repairs conducted on piping and components 1-inch NPS and smaller in ASME Code Class 1,2 and 3 heat exchangers.

"In addition to the foregoing, corrective actions have occurred which will help prevent future problems in Code compliance with repairs to piping and components 1-inch NPS and smaller. These include the following:

a. Since identification of this potential non-compliance, plant personnel responsible for repair activities have been instructed that repa'rs, irrespective of the size of piping and components involved, are to be properly documented,

b. The plant ASME Section XI Repair / Replacement Program was revised to address repairs to piping and components irrespective of the size involved, and

c. A training session on ASME Section XI (with emphasis on repairs and replacements) was held. Participants included plant personnel directly responsible for repair / replacement activities and others, including, but not limited to, representatives from Quality Control, Quality Assurance, and other departments; as well as the current ANil.

"To the best of our knowledge and belief, we have been in compliance with the Code requirements for repairs to piping and components 1-inch NPS and smaller since the potential non-compliance was identified and corrective actions implemented.

i  !

" Based on the foregoing information, in addition to the significant effort which would be required to review the numerous MWOs generated for VEGP-l for the period from May 31, 1987 (date of commercial operation) through approximately October 1996 (time of discovery) for possible noncompliance with a record keeping requirement, it is our position that imposing the record keeping requirements after the fact for any such repairs of piping and components 1-inch NPS and smaller would not provide a commensurate increase in the level of safety particularly in light of changes to the Code. A significant hardship would result if the record keeping requirements were to be retroactively imposed. Accordingly, it is requested that the proposed alternative be authorized pursuant to IOCFR50.55a(a)(3)(ii).

Evaluation: The Code requires that repairs and replacements be performed in accordance with IWA-4000 and IWA-7000, respectively. Pursuant to IWA 7400, piping, valves, and fittings NPS 1-inch and smaller are exempt from the requirements of Article IWA-7000.

However, Article IWA-4000 has no exemption criteria for components NPS 1 inch and smaller. Therefore, some licensees may choose to replace rather than repair items to avoid the repair requirements of IWA-4000. As an alternative, the licensee proposed to use the requirements of IWA-4130 of the 1995 Addenda to exempt items 1-inch NPS and smaller from the repair and replacement requirements of the Code. This is equivalent to the alternative contained in Code Case N 544, Repair and Replacement of Small /tems, which has not been approved for general use by the NRC in Regulatory Guide 1.147.

In accordance with the 1995 Addenda and Code Case N-544, piping, valves, and fittings NPS 1-inch and smaller, except for heat exchanger tubing and sleeves and welded plugs used for heat exchanger tubing, are exempt from both repair and replacement requirements of the Code. The exemption criteria used for the repair of items NPS 1-inch and smaller is comparable to existing Code requirements for the replacement of similar items. Therefore, the INEEL staff believes that IWA-4130 of the 1995 Addenda provides reasonable assurance of operational readiness with one exception. The ASME Section XI Code differentiates between steam generators and heat exchangers by providing separate item numbers. As currently written, the 1995 Addenda does not address steam generator tubing, only heat exchanger tubing. Therefore, it appears that steam generator tubing could be exempted from repair and replacement requirements by the Code Case. To address this uncertainty, the licensee has stated that steam generators are considered heat exchangers and would be excluded from the alternative requirements of IWA-4130 with the heat exchanger tubing, sleeves, and welded plugs used for plugging heat exchanger tubing.

As stated by the licensee, imposition of the Code requirements would necessitate a significant effort to review over 50,000 work orders to complete paperwork that is not required by later Code Addenda. This effort could potentially divert plant personnel from other activities that could affect plant safety. Therefore, considering that the record keeping activities would document repairs that are not required to be documented by later Codes, it is concluded that imposition of the applicable Code requirements would result in an undue hardship without a compensating increase in the level of quality and safety.

Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

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3.0 CONCLUSION

The INEEL staff has reviewed the licensee's submittals and concludes that the Code requirements are impractical to meet for the issues contained in Requests for Relief RR-2, -3 and -63. Therefore, relief should be granted pursuant to 10 CFR 50.55a(g)(6)(i). In addition, the technical content of Request for Relief RR-30 has not changed. Therefore, relief should remain granted.

For the proposed alternative to the augmented RPV examination required by the regulations, it is concluded that the licensee's proposed alternative provides an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(g)(6)(ii)(A).

For Request for Relief RR-64, it is concluded that the Code requirements would result in hardship without a compensating increase in the level of quality and safety. Therefore, it is recommended that these proposed alternatives be authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

For Request for Relief RR 5, it is concluded that the licensee has not adequately justified the reduction in examination coverage and has not shown that the proposed alternative provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative should not be authorized.

Request for Relief RR-4 was withdrawn by the licensee.

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