ML20248D029
| ML20248D029 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/26/1998 |
| From: | Woodard J SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LCV-1124-A, NUDOCS 9806020303 | |
| Download: ML20248D029 (31) | |
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'~- e J.D.Woodard Southern Nuclear
- Operating Company,Inc.
Executive Vee President 40 invemess Center Parkway Post Office Box 1295.
Birmingham Alabama 35201
. . Ief 205.992!086 SOUTHERN L COMPANY Energy to Serve YourWorld" May 26,1998 l
LCV-1124-A Docket No.: 50-424 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Ladies and Gentlemen:
VOGTLE ELECTRIC GENERATING PLANT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION -
REVISION 8 TO VEGP-1 FIRST TEN-YEAR INSERVICE INSPECTION PROGRAM in response to a request for additional information dated April 10,1998, for Vogtle Electric Generating Plant, Unit 1 (VEGP-1), please find enclosed our response to your request concerning Revision 8 to the VEGP-1 First Ten-Year Interval Inservice Inspection Program. Requests for Relief RR-2, RR-3, and RR-63 have been revised and are provided herein as Revision 9.
After additional reviews of results of reactor pressure vessel (RPV) examinations conducted during VEGP-1 Maintenance / Refueling Outage 1R6, it was determined that four additional welds had limitations which were overlooked and not submitted in Revision 8 to VEGP-1 First Ten-Year Interval ISI Program document ISI-P-006. The welds in question are the RPV outlet nozzle-to-shell welds. Similar limitations were experienced during the VEGP-2 ten-year RPV examinations conducted in March 1998. Please refer to the attached Request for Relief RR-65 which should be considered part of Revision 9 of the VEGP-1 First Ten-Year Interval ISI Program.
1 When the enclosed new and/or revised requests for relief, i.e., RR-2, RR-3, RR-63, and RR-65, are approved by the NRC, the VEGP-1 First Ten-Year ISI Program document ISI-P-006 will be updated internally to Revision 9 to reflect the NRC approval of those particular requests for relief and to mak.e any purely administrative changes, e.g., updating List of Effective Pages, updating
. List of Requests for Relief, resulting from that approval. After the subject document is updated, 1
PDR ADOCK 05000424 I G PDR poh]
p U. S. Nuclear Regulatory Commission LCV-1124-A -
Page Two =
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- it will be retired since examinations and tests required for the first ten-year inspection interval -
have been completed for VEOP-1. The internal update of VEGP-1 First Ten-Year ISI Program
- document ISI-P-006 to Revision 9 will not be submitted to the NRC as a result of the NRC prior review and anticipated approval of the requests for relief submitted herein. The purely administrative changes to the subject document will not change the overall intent of the document.
Should there be any questions in this regard, please contact this office.
Sincerely, I
. . toodard JDW/JAE/jae
Enclosures:
- 1. Response to Request for Additional Information - Revision 8 to VEGP-1 First Ten-Year Inservice Inspection Program (includes 3 attachments)
- 2. Revision to Existing Requests for Relief RR-2, RR-3, and RR-63 L 3. New Request for Relief RR-65 xc: Southern Nuclear Operating Comnany Mr. J. B. Beasley, Jr. (w/o enclosure) -
Mr. W. L' Burmeister (w/o enclosure)
Mr. M. Sheibani (w/ enclosure)
SNC Document Management (w/ enclosure)
U. Si Nuclear Regulainry Commission g Mr. D. H. Jaffe, Senior Project Manager, NRR (w/ enclosure)
Mr. L. A. Reyes, Regional Administrator (w/ enclosure)
Mr. J. Zeiler, Senior Resident inspector, Vogtle (w/ enclosure) ,
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'LWCLOSURE1 VOGTLE ELECTRIC GENERATING PLANT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION-REVISION 8 TO VEGP-1 FIRST TEN-YEAR INSERVICE INSPECTION PROGRAM
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- ENCLOSURE 1 TO SOUTIIERN NUCLEAR OPERATING COMPANY LETTER LCV-1124-A ;
VOGTLE ELECTRIC GENERATING PLANT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION -
REVISION 8 TO VEGP-1 FIRST TEN-YEAR INSERVICE INSPECTION PROGRAM In its request for additional information dated April 10,1998, the NRC requested that Southern Nuclear Operating Company (SNC) provide information for Vogtle Electric Generating Plant, Unit 1 (VEGP-1), in response to four (4) items which are quoted herein and which involve Revision 8 of the VEGP-1 First Ten-Year Inservice Inspection (ISI) Program. The SNC response follows each of the NRC items.
NRC ltem "2.A"
" Requests for Relief RR-2, RR-3, and RR-63. For each of the welds in these requests, the licensee has requested relief from both augmented and Code requirements. The NRC staff considers these separate issues that must be addressed separately. Regarding the augmented examination requirements, licensees that are unable to satisfy the requirements of the augmented RPV examination imposed by the Regulations must submit an alternative that provides an acceptable level of quality and safety. Such alternatives may be used when authorized by the NRC staff. To find the proposed alternative to the augmented RPV examination acceptable, the licensee must submit the following information: 1) provide a more detailed justification supporting the use of the alternative examination (e.g., exams from the exterior surface, if any, stresses on the vessel, equipment used, etc[.]),2) identify each Category B-A, Item Bl.10 welds and provide the coverage achieved for each weld,3) provide the history of the reactor vessel weld inspections and results, and 4) provide a technical discussion describing how examination coverage was maximized (address both internal and external examinations). Include in that discussion the burden associated with performing examinations from the exterior (OD) surface, the expected radiation exposure, and the increase in examination volume that could be obtained by performing examinations from the OD surface."
SNC Resnonse to NRC Item "2.A" The issue of augmented and Code requirements has been separated into Requests for Relief revisions, where necessary, and a letter (to be submitted to the NRC under separate cover) concerning the augmented examination requirements.
Item 1) Please refer to the revised attached Requests for Relief RR-2, RR-3, and RR-63 for the additional information requested and/or clarifications. The attached Requests for Relief RR-2, RR-3, and RR-63 supersede those same numbered requests for relief submitted to the NRC by SNC letter LCV-1124 dated December 1,1997.
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. VOGTLE ELECTRIC GENERATING PLANT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION -
REVISION 8 TO VEGP-1 FIRST TEN-YEAR INSERVICE INSPECTION PROGRAM (continued)
SNC Resnonse to NRC Item "2.A"(continued)
Item 2) For the summary of the Category B-A, item Bl.10 welds, including the examination coverage for each of the subject welds, please refer to Attachment I to this enclosure.
Item 3) For the history of the VEGP-1 reactor pressure vessel (RPV) weld indications, please refer to Attachment 2 to this enclosure.
Item 4) Please refer to Attachment 3 to this enclosure for a technical discussion describing how examination coverage was maximized for the subject welds. Similar information is being submitted to the NRC under separate cover (Refer to SNC letter LCV-1124-B dated May 21,1998) in order to address the augmented requirements since the NRC staff considers the Code and augmented examinations to be separate issues.
NRC ltem "2.B"
" Requests for Relief RR-2, RR-3, and RR-63. Regarding relief from Code requirements, based on the initial review of the licensec's submittal, the staff has concluded that appropriate paragraphs of the Regulations have not been referenced. For Requests for Relief RR-2, RR-3, and RR-63, cite the appropriate paragraph of the Regulations to ensure that the request is evaluated in accordance with the appropriate criteria for relief from Code requirements, as discussed below.
A licensee may propose an alternative to CFR or Code requirements with 10 CFR 50.55a(a)(3)(i) or 10 CFR 50.55a(a)(3)(ii). Under 10 CFR 50.55a(a)(3)(i), the proposed alternative must be shown to provide an acceptable level of quality and safety, i.e., essentially be equivalent to the original requirement in terms of quality and safety. Under 10 CFR 50.55a(a)(3)(ii), the licensee must show that compliance with the original requirement results in a hardship or unusual difficulty without a compensating increase in the level of quality and safety. Examples of hardship and/or unusual difficulty include, but are not limited to, excessive radiation exposure, disassembly of components solely to provide access for examination, and development of sophisticated tooling that would result in only minimal increases in examination coverage.
A licensee may also submit a request for relief from ASME requirements. In accordance with 10 CFR 50.55a(g)(5)(iii), if a licensee determines that conformance with certain Code requirements is impractical for its facility, the licensee shall notify the Commission and submit, as specified in s50A, information to support that determination. When a licensee determines that an inservice inspection requirement is impractical, e.g., the system would have to be redesigned, or a component would have to be replaced to enable inspection, the licensee should cite 10 CFR 50.55a(g)(5)(iii). The NRC may, giving due consideration to the burden placed on the licensee, impose an alternative examination requirement."
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o o VOGTLE ELECTRIC GENERATING PLANT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION -
REVISION 8 TO VEGP-1 FIRST TEN-YEAR INSERVICE INSPECTION PROGRAM (continued)
SNC R.esponse to NRC Item "2.B" Please refer to the attached Requests for Relief RR-2, RR-3, and RR-63 for the additional information requested and/or clarifications.
NRC ltem "2.C"
" Request for Relief RR-5 concerns limited examination of RPV circumferential head Weld 11201-V6-001-WO7. This request was originally evaluated and granted in an NRC Safety Evaluation Report dated November 26,1991. The original request estimated coverage of 74%
could be obtained. As revised in the December 1,1997, submittal, the actual coverage was only 29%. This represents a significant reduction in coverage from the initial request. Provide a technical discussion justifying the reduction in coverage."
SNC Response to NRC Item "2.C" The VEGP-1 preservice inspection (PSI) of the RPV was conducted during September 1985 by Combustion Engineering (CE) using " immersion" techniques for the 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 an 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 location reports provide for a clearer and more accurate result for RPV examinations.
The VEGP-1 first ten-year interval ISI examinations were conducted in April 1996 by 'WesDyne, using " contact" techniques and the WesDyne Reactor Vessel inservice inspection (RVISI) tool.
Volume coverage calculations were documented from tooling dimensions from the 0 ,45 , and 60 examinations. Along with CAD drawings, the results were conservatively calculated and weighted with their respective scan requirements (up, down, clockwise, counter-clockwise) for the 45 and 60 transducers. Due to limitations with the WesDyne tool and the bottom- mounted l instrumentation tubes (BMis), volumetric examination coverage 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 movements in most areas. The best practical examination coverage without causing potential damage to both the RVISI tool and/or the BMis was obtained during VEGP-1 Maintenance / Refueling Outage 1R6. The method for calculating the coverage was simple and relatively repeatable.
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+ VOGTLE ELECTRIC GENERATING PLANT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION -
REVISION 8 TO VEGP-1 FIRST TEN-YEAR INSERVICE INSPECTION PROGRAM (continued)
SHC Resnonse to NRC Item "2.C"(continued)
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:
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 since most examination limitations 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 perfonned to satisfy the requirements of NRC Regulatory Guide (RG) 1.150. In general, the 70 acquired coverage was greater than that of other angles due to the smaller volume required (l" of the near surface), thus potentially raising the accumulative total.
1 Since the RPV examinations conducted during VEGP-1 Maintenance / Refueling Outage IR6, WesDyne has developed a new system called "SUPREEM" which uses ROSA mechanized i technology along with smaller, better designed transducer sleds which, in moss 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.
NRC ltem "lD" l
" Request for Relief RR-64 proposes to use Paragraph IWA-4130 of the 1995 Addenda, l Alternative Requirementsfbr SmallItems, in lieu of the repair requirements of the Code of ;
Record. This request covers the period from May 31,1987,(the date commercial operation i began) to October 1996, when it was discovered that VEGP, Unit I was not in compliance with the requirements ofIWA-4000 for items 1-inch NPS and smaller. It appears that the licensee is requesting relief from fulfilling the administrative requirement to document past repairs that were inadvertently exempted from the Code requirements. Cost is not considered a burden. However, this request could be authorized if an acceptable level of quality rmd safety can bejustified.
Using the guidance provided in Question B above, review Request for Relief RR-64 for the proper regulatory basis and provide a technical discussion describing how the proposed alternative provides an acceptable level of quality and safety compared to the repair requirements of the Code. The discussion should include a summary describing the scope of this request, i.e.,
the number and significance of the components inadwrtently exempted and any additional Cc de requirements that were not met."
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. VOGTLE ELECTRIC GENERATING PLANT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION -
REVISION 8 TO VEGP.1 FIRST TEN-YEAR INSERVICE INSPECTION PROGRAhi (continued)
SNC Resnonse to NRC Item "2.D" It is the opinion of SNC that Request for Relief RR-64 as currently written provides the proper l regulatory basis for requesting relief to use the regmrements ofIWA-4130 of the 1995 Addenda to ash 1E Section XI. Without reviewing each of the approximately 52,000 MWO packages that were generated for the period from May 31,1987 (date of commercial operation) until October 1996 when the non-compliance was confirmed and remedied, the exact number of potential non-compliances cannot be quantified. All repair and replacement requirements are believed to have been met except possibly the recordkeeping requirements, i.e., completion of an ASME Form ,
NIS-2," Owner's Report for Repairs and Replacements", for piping and components 1-inch NPS and smaller. It is our belief that the number of any such occurrences of repairs to items 1" NPS and smaller are relatively few and are of minor safety significance. There are no other areas of non-compliance associated with this particular issue to the best of our knowledge and belief. We 4 acknowledge that cost is not considered a burden in complying with the Code requirements.
Ilowever, we do not believe that recordkeeping requirements should be retroactively imposed in this instance since no similar recordkeeping requirement exists for replacements 1-inch NPS and smaller. No commensurate increase in the level of safety or quality would be achieved were we to administratively backfit any repairs to items 1-inch NPS and smaller that are within the scope of the ISI Program for VEGP-1. Nor wenld there be a decrease in the margin of public health and safety if this administrative function, i.e., completion of an ASME Form NIS-2 and Replacements", were not perfonned. ASME has concluded that the repair requirements for items 1-inch NPS and smaller should be similar to those for replacements which do not require similar actions ter small items 1-inch NPS and smaller. This is documented in ASME Section XI Code Case N-544 whose provisions were incorporated into the 1995 Addenda to ASME Section XI.
Any repairs would have been performed using various existing approved procedures and/or programs as required by 10 CFR 50, Appendix B, which established control on the planning, work control, quality assurance / quality control, and implementation of work packages.
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. ' ATTACHMENT 1 VEGP-1 1R6 RPV EXAMINATION LIMITATIONS (Category B-A / Item No. Bl.10)
EE YY WO4 B1.11 MAIN LOOP NOZZLES - 100 % >90% CODE CASE N-460 WO5 Bl.11 N/A - 100 % 100 % N/A WO6 B 1.11 CORE SUPPORT LUGS RR 2 66 % 62 % RESUBAfl77ED A T62%
W12 Bl.12 MAIN LOOP NOZZLES RR-63 100 % 75 % SUBAf/77ED A T 75%
l W13 Bl.12 MAIN LOOP NOZZLES RK-63 100 % 1 80 % SUBAflT7EDAT80%
! W14 B l.12 MAIN LOOP NOZZLES RR-63 100 % 85 % SUBAflTTED AT85%
W15 Bl.12 N/A - 100 % 100 % N/A W16 Bl.12 N/A - 100 % 100 % N/A W17 Bl.12 N/A - 100 % 100 % N/A W18 Bl.12 CORE SUPPORT LUGS RR-3 71 % 77 % RESUBAflITED A T 77%
W19 Bl.12 CORE SUPPORT LUGS RR-3 71 % 77 % RESUBAfl77ED AT 77%
W20 B1.12 CORE SUPPORT LUGS RR-3 71% 77 % RESUBAf/TTED A T 77%
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- ATTACllMENT 2
SUMMARY
OF KNOWN RPV WELD ULTRASONIC NON-GEOMETRIC REFLECTORS
- VEGP-1 " Post Hydro" Fabrication (I.D. and O.D.) /Ji976; Weld ID Weld Description Comment 107-121 A (OD Exam) Nonle to Shell Code acceptable laniinations.
10712]C (OD Exam) Nonle to Shell Code acceptable indications.
101 124A (OD Exam) Middle Shell Longitudinal Code acceptable indications.
101-124C (OD Exam) Middle Shell Longitudinal Code acceptable indications.
101 14213 (OD Exam) Lower Shell Longitudinal Code acceptable laminations.
101-142C (OD Exam) Lower Shell Longitudinal Code acceptable laminations.
105-121 A (ID Exam) Nonle to Shell Code acceptable indications.
105-121D (ID Exam) Nonle to Shell Code acceptable indications.
101-12213 (lD Exam) Upper Shell Longitudinal Code acceptable indications. (3.9",7.3" & 8.1")
101 142C (ID Exam) Upper Shell Longitudinal Code acceptable indications. (25.9")
.VEGP-1 " PSI" Examination (Immersion) fl985.
Weld ID Weld Description Comment i 1201-V6-001-W!$ Intermediate Shell Code acceptable near surf ace " spot" indication (s).
I 1201-V6-001-W19 Lower Shell Longitudinal Code acceptable near surf ace " spot" indication (s).
Note:
Several welds contained "mid-plate segregates" observed with the zero degree transducer which did not meet the recording criteria.
VEGP-1 "40-Month". Nozzle Examinations (1R2)- / 190n :
Weld ID Weld Description Comment i1201-V6-001 W28 Nonle to Shell Code acceptable non-geometric indication (s).
I1201 V6-001 W36 Nonle to Safe-End Code acceptable non-geometric indication (s).
I 1201-V6-001-W37 Nonle to Safe-End Code acceptable non-geometric indication (s).
I 1201-V6-001-W40 Nonle to Safe-End Code acceptable non-geometric indication (s).
VEGP-1 "10-Year" Examin'ations (1R6) / 1996-Weld ID Weld Description Comment 11201-V6-001-WO3 1 lange to Upper Shell Code acceptable non-geometric indication (s).
I1201 V6-001-WO6 Lower Shell to Lower llead Code acceptable non-geometric indication (s). !
I1201-V6 001-W15 Intermediate Longitudinal Code acceptable non-geometric indication (s). j ll201-V6-001 W16 Intermediate Longitudinal (. ode acceptable non-geometric indication (s). I i1201 V6-001-W20 Lower Longitudinal Code acceptable non-geometric indication (s).
I1201 V6-001-W22 Meridional Seam Code acceptable non-geometric indication (s).
I1201-V6-001 W28 Nonle to Shell Code acceptable non-geometric indication (s).
I1201 V6-001-W35 Safe-End to tilbow Code acceptable non-geometric indication (s).
I1201 V6-001 W38 Safe-End to Elbow Code acceptable non-geometric indication (s-).
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+ ATTACHMENT 3 VOGTLE ELECTRIC GENERATING PLANT REOUEST FOR ALTERNATIVE TO 10 CFR 50.SSa(g)(6kii)(A) l l Southern Nuclear Operating Company (SNC) has determined that the augmented examinations l of the Vogtle Electric Generating Plant, Unit 1 (VEGP-1) reactor pressure vessel (RPV) cannot l be performed to the extent required by 10 CFR 50.55a(g)(6)(ii)(A) without undue hardship. In l accordance with the provisions of 10 CFR 50.55a(g)(6)(ii)(A)(5), SNC requests NRC authorization of an alternative examination based on other pertinent examinations performed to date which provide an acceptable level of quality and safety.
REOUIRED EXAMINATIONS 10 CFR 50.55a(g)(6)(ii)(A) requires that all licensees augment their RPV examinations by implementing once, as part of the inservice inspection interval in effect on September 8,1992, the examination requirements for reactor vessel shell welds specified in item No. Bl.10 of Examination Category B-A, " Pressure Retaining Welds in Reactor Vessel", in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of ASME Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code. To meet the requirements of 10 CFR 50.55a(g)(6)(ii)(A),"more than 90 % of the examination volume of each weld" shall be examined.
COMPLETED EXAMINATIONS Southern Nuclear Operating Company contracted the Nuclear Steam System Supply (NSSS) vendor to perform the examinations of the VEGP-1 RPV. The ultrasonic examinations (UT) were performed using a remote reactor vessel inspection tool to satisfy the requirements of tb 1983 Edition of ASME Section XI with Addenda through Summer 1983 and NRC Regulatory Guide 1.150. A total of three circumferential and nine longitudinal RPV shell welds were examined to satisfy the requirements of both the augmented RPV shell weld rule and ASME Code Section XI. The examination results for these welds revealed no recordable indications that exceeded the allowable standards of ASME Code,Section XI, Paragraph IWB-3500. The coverage achieved for each weld is listed in Table I.
ALTERNATE EXAMINATIONS Lower Shell To Bottom Head Weld 'WO6) and Longitudinal 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 weld 11201-V6-001-WO6. Three of these six lugs are welded directly onto intersecting longitudinal welds W18, W19, and W20.
These core support lugs obstructed movement of the mechanized examination 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 percent (62%) of the weld length. This result is A3-1 i
VOGTLE ELECTRIC GENERATING PLANT REOUEST FOR ALTERNATIVE TO 10 CFR 50.55afgV6ViiVA)
, (continued)
Lower Shell To Bottom Head Weld (WO6) and Longitudinal Welds (W18. W19. & W20)
(continued) comparable to the . ixty-six percent (66%) coverage reported during preservice examinations (PSI).
Examination of the affected longitudinal welds underneath 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 percent (77%) of the weld length for each of the longitudinal welds. This result is comparable to the seventy-one percent (71%) coverage 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 ofincreasing coverage for these welds. These evaluations concluded that supplemental OD examinations could increase the total 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 personnel in performing these examinations would be approximately 5 R.
. Prior to performing examinations, personnel would need to erect any necessary scaffolding, remove insulation, and perform any rer,uired weld preparation in the high radiation field.
This effort is further exacerbated by th. fact that much of the RPV insulation used at VEGP was designed using rivets and screws and does not lend itself' 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. ;
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- VOGTLE ELECTRIC GENERATING PLANT REOtIEST FOR ALTERNATIVE TO 10 CFR 50.55a(gV6ViiVA)
(continued)
Lower Shell To Bottom Head Weld (WO6) and Longitudinal Welds (W18. W19. & W20)
(continued) e NDE personnel wculd need to locate and mark 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 provided by the automated inspection tool. Translating these locations to the OD with a high degree of confidence would be an extremely difficult task while working in a high radiation field.
Upper Shell Longitudinal 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% volumetne 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. Coserage reported during preservice examinations 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 wnsidered to be primarily in the method used to calculate coverage.
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 embrittlement is not a factor. These welds had a complete ultrasonic exarnination 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 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.
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o VOGTLE ELECTRIC GENERATING PLANT REOUEST FOR ALTERNATIVE TO 10 CFR 50.55a(g)(6)(ii)(A)
(continued)
CONCLUSION (continued)
The examination of RPV shell welds provides an acceptable level of quality and safety even though all could riot be fully examined. The average examination coverage of all Category B-A, Item No. Bl.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.
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-. VOGTLE ELECTRIC GENERATING PLANT REOUEST FOR ALTERNATIVE TO 10 CFR 50.55a(gV6ViiVA)
(continued)
TABLE 1 VEGP-1 RPV WELDS EXAMINATION COVERAGE
SUMMARY
(Category B-A / Item No. Bl.10) sk WIiSM :
WO4 Bl.11 MAIN LOOP NOZZLES 100 % 100 % >90%
WOS . Bl.11 N/A 100 % 100 % 100 %
WO6 Bl.11 CORE SUPPORT LUGS 100 % 66 % 62 %
W12 Bl.12 MAIN LOOP NOZZLES 100 % 100 % 75 %
W13 Bl.12 MAIN LOOP NOZZLES 100 % 100 % 80 %
Wl4 Bl.12 MAIN LOOP NOZZLES 100 % 100 % 85 %
Wl5 B1.12 N/A 100 % 100 % 100 %
W16 Bl.12 N/A 100 % 100 % 100 %
W17 Bl.12 N/A 100 % 100 % 100 %
W18 Bl.12 CORE SUPPORT LUGS 100 % 71 % 77 %
W19 Bl.12 CORE SUPPORT LUGS 100 % 71 % 77 %
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W20 Bl.12 CORE SUPPORT LUGS 100 % l 71 % 77 %
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ENCLOSURE 2 VOGTLE ELECTRIC GENERATING PLANT REVISION TO EXISTING REQUESTS FOR RELIEF RR-2, RR-3, AND RR-63
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MJ j - VEGP-1 Ba-2 l System /Comnonent for Which Reliefis Reauested Reactor Pressure Vessel (RPV) lower shell-to-bottom head weld 11201-V6-001-WO6.
Code Requirement for Which Reliefis Reauested
- ASME Section XI, Category B-A, Item No. Bl.11, as found in ASME Section XI Table IWB-2500-1, requires that a volumetric examination of the pressure-retaining circumferential welds in the RPV be performed. The applicable examination volume is shown in ASME Section XI, Figure IWB-2500-1, and includes one hundred percent (100%) of the weld length. Reliefis being requested from the Code requirements because the examination volume coverage was limited.
Basis for Ilgjief Six RPV core support lugs are located cn the lower shell of the RPV adjacent to RPV lower shell-to-bottom head weld 11201-V6-001-WO6. These core support lugs (see Attachment I to this relief request) obstructed 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 approximately sixty-two percent (62%) of the weld length. Cornplete coverage from the inside diameter (lD) of the RPV would necessitate redesign and modification of the RPV which is not practical.
Performance of supplemental examinations from the RPV outside diameter (OD) was evaluated as a possible means ofincreasing coverage. These evaluations concluded that supplemental OD examinations could increase the total coverage to that required by ASME Section XI; however, such coverage was considered impractical due to the large radiation expor.ure (estimated as approximately 7.75 Rem (R)). This conclusion was based on the following:
(1) General area dose rates at the bottom of the vessel (as measured for VEGP-2 during Maintenance / Refueling Outage 2R6) are estimated to be approximately 200 mr/hr with contact dose rates at the insulation surface approximately 1 R/hr.
(2) Non-Destructive 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 personnel in performing these examinations would be approximately 3.25 R.
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- VEGP-1 l
RR-2 (continued)
Basis for Relief (continued 3 l
(3) Prior to performing examinations, personnel would need to erect any necessary scaffolding, remove insulation, and perform any required weld preparation in the high radiation field.
This effort is further exacerbated by the fact that much of the RPV insulation osed 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 tc be approximately 4.5 R.
(4) NDE personnel would need to locate and mark the areas where the supplemental examinations need to be performed. When perfctming ID examinations, limitations are located in respect to the core support lugs and the RPV tlange, using indexing providing by the automated inspection tool. Translating these locations to the OD with a high degree of confidence would be an extremely difficult task while working in a high radiation field.
Alternate Examination No supplemental examination is proposed. Ilowever, it should be noted that an overall, general j 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.
Justification for the Granting of Relief 1 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 require evaluation during preservice examinations. A review of data indicates that no indications were observed in the areas not receiving ID inservice coverage.
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VEGP-1 RR.2 (continued)
Justification for the Granting cf Relief (continued)
Compliance with' Code coverage requirements would necessitate prefabrication of the RPV to perform complete Code examinations from the ID or it would necessitate performance of-
- supplemental examinations from the OD. Prefabrication of the RPV to perform the Code
- required examinations from the ID is not practical and supplemental OD examinations have been evaluated by VEGP as impractical due to radiation exposure considerations. Fabrication shop examinations indicate 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 operational readiness of this weld and the RPV Denial of this relief request would cause an excessive burden to VEGP; therefore, approval should be granted pursuant to 10 CFR -
50.55a(g)(6)(i).
Implementation Schedule This relief request is applicable to the First Ten-Year ISI Interval on VEGP-1 which concluded l May 30,1997, exclusive of the one year period allowed by ASME Section XI, IWA-2400(c).
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.- VEGP-1 ER:2 (continued)
ATTACHMENT 1 Cross Sectional View Weld WO6 4
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- 1. 03 - 0' SCAN
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- VEGP-1 RR-3 System /Comnonent for Which Reliefis Requestea Reactor Pressure Vessel (RPV) lower shell longitudinal welds 11201-V6-001-W18,11201-V6-001-W19, and 11201-V6-001-W20.
Code Requirement for Which Reliefis Requested ASME Section XI, Category B-A, Item No. Bl.12, as found in ASME Section XI Table IWB-2500-1, requires that a volumetric examination of the pressure-retaining longitudinal welds in the RPV be performed. The applicable examination volume is shown in ASME Section XI, Figure IWB-2500-2, and includes one hundred percent (100%) of the weld length. Reliefis being requested from the Code requirements because the examination volume coverage was limited.
Basis for Relief Core support lugs are welded over the subject longitudinal welds in the lower shell of the RPV (see Attachment I to this relief request), thereby, preventing 100% volumetric examination of l 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 the core support lugs).
Performance of supplemental examinations from the RPV outside diameter (OD) was evaluated as a possible means ofincreasing coverage. These evaluations concluded that supplemental OD examinations could increase the total coverage to that required by ASME Section XI; however, such coverage was considered impra"tical due to the large radiation exposure. This conclusion was based on the following:
(1) General area dose rates at the bottom of the vessel (as measured for VEGP-2 during Maintenance / Refueling Outage 2R6) are estimated to be approximately 200 mr/hr with contact dose rates at the insulation surface approximately 1 R/hr.
(2) NDE personnel would need to perform thirteen UT scans for each weld receiving the supplemental examinations. It is calculated that the dose to the NDE personnel in performing these examinations would be approximately 1.625 R.
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- VEGP-1 RR-3 (continued)
Basis for Relief (continued)
(3) Prior to performing examinations, personnel would need to erect any necessary scaffolding, remove insulation, and perform any required weld preparation in the high radiation field.
This effort is further exacerbated by the fact that much of the RPV insulation used 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 man-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 4.25 R. (If these activities were performed in conjunction with the examination ofintersecting circumferential weld 11201-V6-WO6 the incremental dose, above that shown in Request for Relief RR-2, is estimated to be approximately 1.875 R).
Alternate Examination No supplemental examination is proposed. It should be noted however that an overall, general l 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-Wl8,11201-V6-001-W19, and 11201-V6-001-W20 were examined volumetrically, i.e., during VEGP-1 Maintenance / Refueling Outage 1R6.
Justification for the Granting of Relief The portions of these welds not receiving an inservice ID examination are located in the non-beltline area; therefore, radiation embrittlement is not a concern for this portion of the subject welds. 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 prefabrication of the RPV to perform complete Code examinations from the ID or it would necessitate performance of supplemental examinations from the OD. Prefabrication of the RPV to perform the Code required examinations from the ID is not practical and supplemental OD examinations have been evaluated by VEGP as impractical due to radiation exposure considerations. Fabrication shop examinations indicate 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.
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+' VEGP-1 RR-3 (continued)
-Jncti fication for the Granti ng of Relief (continued) l - 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 operational readiness of these welds and the RPV. Denial of this relief request would cause an excessive burden to VEGP; therefore, approval should be granted pursuant to 10 CFR
- 50.55a(g)(6)(i).
Implementation Schedule
- This relief request is applicable to the First Ten-Year ISI Interval on VEGP-1 which concluded - l
- May 30,1997, exclusive of the one year period allowed by ASME Section XI, IWA-2400(c).
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6-7b - 006 Rev. 9 l
- VEGP-1 ER-l (continued)
ATTACHMENT I Crosa_Sutianni View Weld W18. W19. & W20 s
E" 9.63 BASE ' R 86.50 METAL TO CLAD
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. VEGP-1 RR-63 System / Component for Which Reliefis Reauested Reactor Pressure Vessel (RPV) upper shell longitudinal welds 11201-V6-001-W12,11201-V6-001-W13, and 11201-V6-001-W14.
Code Requirement for Which Reliefis Reauested ASME Section XI, Category B-A, Item No. Bl.12, as found in ASME Section XI Table IWB-2500-1, requires that a volumetric examination of the pressure-retainin2l longitudinal welds in the RPV be performed. The applicable examination volume is shown in ASME Section XI, Figure IWB-2500-2, and includes one hundred percent (100%) of the weld length. Reliefis being requested from the Code requirements because the examination volume coverage was limited.
Basis for Relief Physical obstructions (see Attachment I to this relief request), e.g., surface scan interference due l to nozzle 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 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-Wl4. respectively, during inservice inspection. Supplemental outside diameter (OD) examinations are not practical because the welds are located behind the biological shield wall.
Alternate Examination No supplemental examination is proposed. It should be noted however that an overall, general l 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 the subject welds were volumetrically examined, i.e., during VEGP-1 Maintenance / Refueling Outage 1R6.
Justification for the Granting of Relief 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 insenice coverage.
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+ VEGP-1 RR-63 (continued)
Justification for the Granting of Relief (continued)
Compliance with Code coverage requirements would necessitate prefabrication of the RPV, which is not practical. Fabrication rhop examinations indicate that no indications were observed in the areas not receiving ID inservice coverage; therefore, there is little likelihood of a crack l
propagating from a fabrication defect. Examinations performed from the ID combined with good fabrication shop examination results should provide reasonable assurance of the operational readiness of these welds and the RPV. Denial of this relief request would cause an excessive burden to VEGP; therefore, approval should be granted pursuant to 10 CFR 50.55a(g)(6)(i).
Imniementation Schedule This relief request is applicable to the First Ten-Year ISI Interval on VEGP-1 which concluded l May 30,1997, exclusive of the one year period allowed by ASME Section XI, IWA-2400(c).
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I RR-63 (continued) i ATTACHMENT 1 Plan View - RPV Upper Shell Longitudinal Welds W12, W13, and W14 Duft[1 N-1 INL[1 N P INL[ f N 3 {MLif h 4 DUT( LI h-D lNLLI N 6 INLIf N 7 DUTLCf N 6 PP00* 6700* 1*3 00* ISH 00* PDP 00* 247.00' P93 00* 33800*
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- - _ _ _ _ .,__.--,.-_________-______-____m__... _ _ , _ _ - _ _ _ _ , _ _ . , . , _ . , _ _ _ _ _ - _ , _ _ - - _ . , _ _ - _ _ _ - _ _ _ - - _ . - - _ _ _ . , _ _ . _ _ _ _ _ _ - __,____-_v-_____.-___ ___-,.7___.,,.__._____m___-_. _ _ _ , , _ _- , _ - - . _ _ ,___---__
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ENCLOSURE 3 VOGTLE ELECTRIC GENERATING PLANT NEW REQUEST FOR RELIEF RR-65
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- - VEGP-1 RR-65 System / Component for Which Reliefis Requested Reactor Pressure Vessel (RPV) outlet nozzle-to-shell welds 11201-V6-001-W25, W28, W29 and W32.
Code Requirement for Which Reliefis Reauested Item No. B3.90, Category B-D, Table IWB-2500-1 of ASME Section XI requires a volumetric examination of reactor pressure vessel (RPV) nozzle-to-vessel welds. The applicable examination volume is shown in Figure IWB-2500-7(a) for the outlet nozzles. The examination volume includes one hundred percent (100%) of the weld length. In addition, ASME Section XI, Paragraph IWA-2232, requires that ultrasonic examination of vessel welds greater than 2 inches in thickness be conducted in accordance with ASME Code,Section V, Article 4 which requires two-directional coverage wherever feasible. Typically, vessel welds are examined with a combination of 0 ,45 , and 60 transducers to meet the examination volume requirements, liowever, due to the special configurations involved in some welds, including the nozzle-to-vessel welds from the nozzle bore, ASME Section V, Article 4. Paragraph T-441.4.2 permits the use of other examination angles. Additional requirements are:
Reflectors Parallel to the Nozzle-To-Vessel Weld - T-441.5.1 requires that search units be directed perpendicular to the weld so that the angle beams pass through the entire volume of the weld metal. The adjacent base metal is required to be completely scanned by two angle beams; however, it is not necessary to scan from both directions.
Reflectors Transverse to the Nozzle-To-Vessel Weld - T-441.5.2 requires that search units be directed parallel to the axis of the weld such that the angle beams pass through all of the examination volume. Scanning is required to be done in two directions,180-degrees to each other, except that, for those areas blocked by geometric conditions scanning is required in at least one direction.
Code Requirement for Which Reliefis Requested Reliefis requested from meeting the required coverage for the subject welds.
Basis for Relief The required examination volume and associated weld configuration (barrel type nozzle with a protruding inner radius) for the outlet nozzles is shown in ASME Section XI, Figure IWB-2500-7(a). Coverage and limitations for this configuration are listed below:
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- VEGP-1 RR-65
! (continued)
Basis for Relief (continued)
(1) Reflectors Parallel to the Outlet No::le-To-Vessel Weld- Ultrasonic examinations will performed from the nozzle bore using scans as allowed by T-441.4.2 Coverage from this direction is 100%.
(2) Reflectors Transverse to the Outlet No::le-To-Vessel Weld- Ultrasonic examinations will be performed on the ID of the vessel wall and accessible portions of the adjoining nozzle using scans, directed clockwise and counterclockwise. The protruding inner radius (nozzle boss) limits scanning on the nozzle due to scanner interference's. Coverage from this direction is estimated at approximately forty percent (40%).
Composite Coverage - Composite coverage is calculated as seventy percent (70%) based on the average of the parallel and transverse scans listed above.
Alternate Examination Ultrasonic examination of these welds will be performed to the maximum extent practical from the nozzle bore and from the RPV ID surface. No other examination will be conducted.
Justification for Grantine Relief Various techniques have been evaluated including the use of additional angles and outside diameter (OD) examinations; however, it was concluded that the techniques described above permit the maximum practical coverage to be obtained. Compliance with Code coverage requirements weuld necessitate prefabrication of the RPV no721es, which is impractical. The 100% coverage from the nozzle bore will assure that circumferential cracking existing in the weld would have been detected. Coverage limitations existed only for the detection of axially oriented cracking, which has less safety significance. Therefore, the examinations performed will proside reasonable assurance of the operational readiness of the welds and the RPV. Denial of this relief request would cause an excessive burden to VEGP; therefore, approval of this relief request should be granted pursuant to 10 CFR 50.55a(g)(6)(i).
Implementation Schedule This relief request is applicable to the First Ten-Year ISI Interval on VEGP-1 which concluded May 30,1997, exclusive of the one year period allcwed by ASME Section XI, IWA-2400(c).
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'*- VEGP-1
. RIL-fil (continued)
ATTACHMENTS l
Cross Sectional View of Outlet Nazde to Shell Welds
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