ML20198A448
| ML20198A448 | |
| Person / Time | |
|---|---|
| Site: | McGuire  |
| Issue date: | 12/11/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20198A427 | List: |
| References | |
| NUDOCS 9812160290 | |
| Download: ML20198A448 (12) | |
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30006-0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST FROM ASME CODE. SECTION XI REQUIREMENT DUKE ENERGY CORPORATION MCGUIRE NUCLEAR STATION. UNIT 1 DOCKET NO. 50-369
1.0 INTRODUCTION
Title 10 of the Code of Federal Reaulations (10 CFR) Section 50.65a(g) requires that inservice inspection (ISI) of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Class 1,2, and 3, components shall be performed in accordance with ASME Code,Section XI, 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 NRC if (i) the proposed alternative provides 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(b) and (g), inservice inspection of Class CC (concrete containments), and Class MC (metallic containments) of light water cooled power plants must meet the requirements of the 1992 Edition with 1992 Addenda of Subsections IWE and IWL of
.Section XI of the ASME Code. Pursuant to 10 CFR 50.55a(g)(6)(ii)(B), licensees must incorporate the new requirements into their ISI plans and complete the first contalnment inspection by September 9,2001. However, a licensee can submit a request for relief from one or more requirements of the regulation (or the endorsed Code requirements) with proper justification.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 components (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 the Section XI of the ASME Code incorporated by reference in the 10 CFR 50.55a(b) 12 months prior to the start of the 120-month inten/al, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for the McGuire Nuclear Station first 10-year ISI interval,is the 1980 Edition.
By letter dated January 15,1998, as supplemented by letter dated July 1,1998, Duke Energy Corporation (DEC or the licensee), submitted to the NRC Relief Request No.97-004 for the 9812160290 981211 DR ADOCK 050003 9
McGuire Nuclear Station, Unit 1. The January 15,1998, letter requested relief under the provisions of 10 CFR 50.55a(g)(6)(i). The licensee is seeking relief from the ASME Code,Section XI,1992 Edition, IWE-2500, Table IWE-2500-1, Examination Category B A: Pressure Retaining Welds in Reactor Vessel, item B1.11, Circumferential Shell Welds, item B1.12, Longitudinal Shell Welds, item B1.21, Circumferential Head Welds, item B1.21, and Shell to Flange Weld, item B1.30; and Examination Category B-D: Full Penetration Welds of Nozzles in Vessels, Item B3.90, Nozzle to Reactor Vessel Welds, item B3.90, and Inside Radius Sections, item B3.100. The licensee believes that the above requirements are impractical and that compliance with the specified requirement of Subsection IWE would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
2.0 EVALUATION The staff, with technical assistance from its contractor, the Idaho National Engineering and ;
Environmental Laboratory, has evaluated the information provided by the licensee in support of its first 10-year ISI interval program plan requests for relief for the McGuire Nuclear Station, Unit 1. Based on the results of the review, the staff adopts the contractor's conclusions and recommendations presented in the attached Technical Letter Report.
Reauest for Relief 97-004 (Part 11: ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A, Pressure Retaining Welds in Reactor Vessel, item B1.11, Circumferential Shell Welds, item B1.12, Longitudinal Shell Welds, item B1.21, Circumferential Head Welds, item B1.21, and Shell to Flange Weld, item B1.30.
As a result of the augmented volumetric examination rule and the Code-required examination s coverage, licensees must make a reasonable effort to maximize examination coverage of their reactor vessels, in cases where coverage from the inside surface (ID) is insufficient, manual inspection from the outside surface (OD) may be an option. However, at McGuire, the design configuration prevents examination from the OD, and further examination from the lower head OD is not possible due to radiation levels. Imposition of the requirement would require redesign or replacement to allow the examination and would result in a burden upon the licensee.
Therefore, coverage cannot be enhanced by examining from the OD. The licensee has examined the welds indicated in the above table to the maximum extent practicable. Further, a visual examination has been performed on the reactor ID and internals that identified no rejectable conditions in accordance with Section XI acceptance standards. These examinations, in addition to the surface examination of the interior of the vessel, provide reasonable assurance that the structural integrity will be maintained and the presence of degradation will be detected using the stated proposal. Based on the impracticality of meeting the Code coverage requirements for the subject welds, and reasonable assurance provided by the volumetric examination in conjunction with the visual examination of the vessel interior completed, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i).
Reauest for Relief 97-004 (Part 2): ASME Code,Section XI, Table IWB-2500-1, Examination Category B-D, Full Penetration Welds of Nozzles in Vessels, item B3.90, Nozzle-to-Reactor Vessel Welds, item B3.90, and inside Radius Sections, item B3.100.
Due to the limits of design, the requirements of the Code are impractical. Imposition of the Code requirements would require redesign or replacement and result in a burden upon the i
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licensee. The licensee has examined these welds to the maximum extent practicable, obtaining the coverages outlined in the above mentioned table. The volumetric examinations are supplemented by visual examination performed on the vessel interior. These examinations, in addition to the surface examination of the interior of the vessel, provide reasonable assurance l that the structural integrity will be maintained and the presence of degradation will be detected using the stated proposal. Based on the impracticality of meeting the Code coverage .
requirement for the subject welds, and reasonable assurance provided by the volumetric examination in conjunction with the visual examination of the vessel interior completed, relief is l granted pursuant to 10 CFR 50.55a(g)(6)(i).
3.0 CONCLUSION
For Relief Request 97-004, Parts 1 and 2, the staff concludes that the ASME Code examination requirements are impractical. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), the licensee's
, request for relief is granted and the alternative imposed. The relief granted is authorized by law and will not endanger life or property or the common defense and security and is otherwise in {
j the public interest giving due consideration to the burden upon the licensee that could result if !
the requirements were imposed on the facility. 4
Attachment:
Technical Letter Report !
Principal Contributor: G. Hatchett Date: December 11, 1998 l
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l TECHNICAL LETTER REPORT DN THE FIRST 10-YEAR INTERVAL INSERVICE INSPECTION REQUEST FOR RELIEF NO.97-004 l EDB DUKE POWER COMPANY McGUIRE NUCLEAR STATION. UNIT 1 DOCKET NUMBER: 50 369
- 1. INTRODUCTION By letter dated January 15,1998, the licensee, Duke Power Company, submitted Request for Relief 97-004 seeking relief from the requirements of the American Society of Mechanical Engineers (ASME) Code,Section XI for the first 10-year inservice inspection (ISI) interval for McGuire Nuclear Station, Unit 1. In response to a Nuclear Regulatory Commission (NRC) request for additional information, the licensee submitted further information in a letter dated July 1,1998. The Idaho National Engineering and Environmental Laboratory (INEEL) staff has evaluated the information provided by the licensee in support of this request for relief in the following section.
- 2. EVALUATION The information provided by Duke Power Company in support of the request for relief from Code requirements has been evaluated and the basis for disposition is documented below. The Code of record for the McGuire Nuclear Station, Unit 1, first 10-year ISI interval, which ended November 30,1992, is the 1980 Edition through Winter 1980 Addenda of Section XI of the ASME Boiler and Pressure Vessel Codo.
A. Reauest for Relief 97-004 (Part 11 Examination Cateaorv B-A. Pressure Retainina Welds in Reactor Vessel. Item B1.11. Circumferential Shell Welds. Item B1.12.
Lonaitudinal Shell Welds Item B1.21. Circumferential Head Welds. and item B1.30 Shell to Flana_ e Weld Code Reauirement Examination Category B-A, item B1.11, requires 100%
volumetric examination of the circumferential shell welds, as defined by Figure IWB-2500-1. Item B1.12 requires 100% volumetric examination of the longitudinal shell welds, as defined by Figure IWB-2500-2. Item B1.21 requires 100% volumetric examination of the circumferential head welds, as defined by Figure IWB-2500-3.
Item B1.30 requires 100% volumetric examination of the circumferential shell-to-flange welds, as defined by Figure IWB-2500-4.
Licensee's Code Relief Reauest: In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code-required 100% volumetric examination coverage for the welds listed below.
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ATTACHMENT
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WELD ITEM DESCRIPTION COVERAG LIMITATION E
l l 1RPV10442 B1.11 Lower Shell-to- 59% Presence of six core Bottom Head Weld guide lugs l 1RPV1442A B1.12 Upper Shell Long. 90 % Nonle locations that .
Seam Weld split weld into to two sections 1RPV1-442B B1.12 Upper Shell Long. 31 % Nonle location on both l Seam Weld sides of weld 1RPV1-442C B1.12 - Upper Shell Lorg. 90 % Nonle locations that Seam Weld split weld into to two j sections i 1RPV3442A B1.12 Lower Shell Long. 84 % Core guide lugs Seam Weld 1RPV3-442B B1.12 Lower Shell Long. 84 % Core guide lugs Seam Weld 1RPV3 442C B1.12 Lower Shell Long. 84 % Core guide lugs Seam Weld i
1RPV4-469 B1.21 Bottom Head Weld 29% incore nonles 1RPV7-442 B1.30 Nonle Belt-to- 82 % Keyway and specimen Flange Weld tube cutouts, flange configuration Licensee's Basis for Relief Recutat (as stated):
l "During the ultrasonic examination of the Reactor Vessel welds during Refueling l
Outage 7, the minimum coverage requirement of ASME Section XI,1980 Edition through Winter 1980 Addenda, clarified by Code Case N-460, could not be met.
Due in part to geometry and actual physical barriers this~ coverage was l unattainable. A combination of multiple beam angles was used to obtain the
- maximum coverage possible. The attached examination reports document the
! actual amount of examination coverage obtained. Drawings showing details of the i affected welds including calculation methods are included as Attachment 2.'
1 Tables, Figures, Drawings and attachments furnished with the licensee's submittal are not included in this report.
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Justification "The Reactor Vessel Lower Shell to Bottom Head, Upper Shell Long Seam, Lower Shell Long Seams at 60,180 and 300 degrees, and the Bottom Head Welds were l
examined to the maximum extent practical using ultrasonic techniques in l l accordance with the requirements of Section V, Article 4 of the 1980 Edition thru l Winter 1980 Addenda of the ASME Boiler and Pressure Vessel Code.
" Duke Energy Corporation will continue to ultrasonically examine the welds, including inside radius sections, to the extent practical within the limits of original design and construction. This will provide reasonable assurance of weld / component integrity. Thus, an acceptable level of quality and safety will have been achieved and public health and safety will not be endangered by allowing relief from the aforementioned Code requirements. )
i "The Reactor Vessel Lower Shell to Bottom Head Weld (1RPV10-442) (item Number B01.011.003): The principal limitation for this weld is the presence of six core guide lugs welded to the vessel ID just above the weld on the lower shell section. These lugs are appioximately 19 inches tall,27 inches wide and extend 8 inches radially toward the center of the vessel, each positioned every 60 degrees around the vessel circumference. These dimensions include the attachment weld radius which physically limits the contact transducer head travel. The weld was scanned in two directions, perpendicular (axial) and paralle! (circumferential) to the weld between the six guide lugs. Some physicallimitations also exists due to the surface profile where the hemispherical head and lower shell cylinder meet.
Therefore, the 100% volumetric examination is impractical for this weld. The imposition of this requirement would create a considerable burden on Duke Energy Corporation. During the examination of this weld, the Utilized Wall Transducer Head process was used to obtain the maximum possible coverage.
"The Reactor Vessel Upper Shell Long Seam Welds (1RPV1-442A,1RPV1-442B, 1RPV1-442C) (item Number B01.012.001, B01.012.002, B01.012.003): The long ;
seam welds join sections of the upper reactor shell (nozzle belt) to each other. The principal limitation for these welds is the present nozzle geometry. Due to the nozzle geometry,100% volumetric examination is impractical for these welds. The imposition of this requirement would create a considerable burden on Duke Energy Corporation. During the examination of these welds, the Utilized Wall Transducer Head process was used to obtain the maximum pessible coverage. Reference drawing 1213926D for scan coverage.
"The Reactor Vessel Lower Shell Long Seam Welds at 60,180 and 300 degrees (1RPV3-442A,1RPV3-442B,1RPV3-442C) (Item Numbers B01.012.007, B01.012.008 and B01.012.009): These longitudinal seam welds join the lower circumferential shell sections in the core region of the reactor vessel. The principal limitations for these welds were the presence of the core guide lugs welded to the' vessel ID above the weld (weld 10-442) on the lower shell section. These lugs are approximately 19 inches tall,27 inches wide , and extend 8 inches radially toward
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' 1 the center of the vessel, and are positioned every 60 degrees around the vessel circumference. These dimensions include the attachment weld radius. These guide lugs interfere with the axial and circumferential scans.
i-l "These welds were scanned in two directions, perpendicular (axial) and parallel (circumferential) to the weld. Scan coverage from the perpendicular (axial) and parallel (circumferential) scans in the vicinity of the six guide lugs is limited due to '
i interference with the contact transducer head. Therefore, the 100% volumetric
! examination is impractical for these weids. The imposition of this requirement
! would create a considerable burden on Duke Energy Corporation. During the j examination of these welds the Utilized Wall Transducer Head process was used to obtain the maximum possible coverage.
"The Reactor Vessel Bottom Head Weld (1RPV4-469) (ltem Number B01.021.002):
This circumferential weld joins the reactor vessel transition ring (bwer head) to bottom head. The principal limitation for this weld is the presence of the incore nozzles. Therefore, the 100% volumetric examination is impractical for this weld.
The imposition of this requirement would create a considerable burden on Duke Energy Corporation. During tne examination of this weld, the Utilized Wall Transducer Head process was utilized to obtain the maximum possible co' erage.
Reference drawing 12139276D, for scan coverage.
"The Reactor Vessel Nozzle Belt to Flange Weld (1RPV7-442) (Item Number B01.030.002): This weld joins the reactor vessel flange to the upper shell (nozzle belt). The principal limitation for this weld is the presence of keyway and specimen tube cutouts and the flange configuration itself. Therefore, the 100% volumetric examination is impractical for this weld. The imposition of this requirement would ;
create a considerable burden on Duke Energy Corporation. During the examination, of this weld, the Utilized 5Nall Transducer Head process was used to obtain the maximum possible coverage. Reference drawing 1213923D for scan coverage.
" Pursuant to 10 CFR 55.55a(g)(6)(i) granting this relief for the Reactor Vessel Lower Shell to Bottom Head, Upper Shell Long Seam, Lower Shell Long Seams at 60,180 and 300 degrees, and the Bottom Head welds will provide reasonable assurance of weld / component integrity, and is authorized by law. In addition, the requested relief will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility."
Licensee's Proposed Attemative Examination (as statedh
- In addition to the volumetric examination that has been performed on the McGuire reactor vessel, Duke Energy has performed a visual examination of the intemals and the inside of the reactor vessel as required by ASME Section XI, Table IWB-
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2500-1. This Visual examination did not identify any rejectable conditions per ASME Section XI acceptance standards.
"The use of radiography as an attemate volumetric examination method is not feasible due to component thickness and restrictions from physical barriers which l prohibit access for the placement of source, image quality indicators, film, etc. In l addition, the background radiation levels would not allow for a radiographic examination that would render meaningful results.
" Performing the ultrasonic examination from the outside of the reactor vesselis not a viable option. The design of McGuire's reactor building prohibits access for the equipment and personnel from outside the vessel.
" Duke Energy Corporation will continue to perform ultrasonic examinations of all l reactor vessel welds tc. .he maximum extent practical in accordance with the requirements of ASME Section V, Article 4,1989 Edition and Regulatory Guide 1.150, Revision 1, Appendix A. Code Case N-460 will be utilized in all cases where less than 100% but greater than 90% weld coverage is obtained. In cases where weld coverGn of 90% or less is obtained, a request for relief from ASME Section XI Code requirements, will be submitted per the requirements of 10CFR50.55a.
" Duke Energy Corporation proposes in lieu of the greater than 90% coverage requirement of Code Case N-460 that the examination coverage obtained on the welds listed in Attachment 1 be corisidered to provide an acceptable level of quality and safety.
i "No additional examinations will be performed."
Evaluation: The Code requires 100% volumetric examination of the subject RPV welds. However, complete examination coverage is restricted by physical obstructions and geometric configuration that make 100% volumetric examination impractical for these welds. To gain access for examination, the RPV would require design modifications. Imposition of this requirement would impose a significant burden on the licensee.
Licensees must make a reasonable effort to maximize examination coverage of l ,. their reactor vessels. In cases where examination coverage from the ID is inadequate, examination from the outside surface (OD) using manual inspection techniques may be an option. However, at McGuire, Unit 1, the design of the reactor building prevents access from the OD. Therefore, the licensee cannot enhance coverage by examining from the OD.
The licensee has examined the subject welds to the extent practical. There are six core lugs in the lower shell of the vessel that limit scanning of the lower shell-to-bottom head weld to 59% of the examination volume and scanning of three longitudinal welds in the lower shell to 84%. The bottom head weld examination l was limited to 29% by incore instrumentation nozzles. In the upper shell of the 1
vessel, the proximity of the nozzles to the longitudinal welds restricts one weld's examination to 31% of the Code-required volume and two to 90%. The remaining longitudinal welds in the intermediate shell received complete examination. The nozzle belt-to-flange weld was limited to 82% coverage by keyway and specimen tube cutouts as well as flange geometric conditions. The licensee has examined greater than 82% of six of the welds requiring relief and greater than 90% of the remaining reactor vessel welds; furthermore, the licensee has visually examined the vesselinterior as required by the Code. Considering the significant volumetric examination coverage achieved, coupled with the Code-required visual examination, it is reasonable to conclude that any existing pattems 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).
i B. Reauest for Relief 97-004 (Part 2L Examination Cateaorv B-D. Full Penetration Welds of Nozzles in Vessels. Item B3.90. Nozzle-to-Reactor Vessel Welds. and item B3.100. Inside Radius Sections Code Reauirement: Examination Category B-D, items B3.90 and r33.100, requires 100% volumetric examination of the Nozzle-to-Reactor Vessel Weld, and Nozzle Inside Radius Sections, as defined by Figure IWB-2500-7.
Licensee's Code Relief Reouest: In accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code-required 100% volumetric examination coverage for the reactor vessel nozzle welds listed below.
WELD rrEM DESCRIPTION COVERAGE LIMITATION 1RPVS-445A, B3.90 Inlet Nozzle-to- 85% Nozzle configuration 1RPV5-4458, ShellWeld 1RPVS-445C, 1RPV5-445D 1RPVS-445E, B3.90 Outlet Nozzle- 43% Nozzle configuration 1RPV5-445F, to-ShellWeld !
1RPV5445G, i 1RPV5-445H l
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1RPV5 445AR, B3.100 inner Radius 58 % Nozzle configuration I 1RPV5-445BR, Inlet Nozzle 1RPV5-445CR, 1RPV5-445DR I
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1RPV5-445ER, B3.100 Inner Radius 74 % Nozzle configuration 1RPVS-445FR, Outlet Nozzle 1RPVS-445GR, 1RPVS-445HR l Licensee's Basis for Relief Reauest (as stated):
"During the ultrasonic examination of the Reactor Vessel welds during Refueling Outage 7, the minimum coverage requirement of ASME Section XI,1980 Edition through Winter 1980 Addenda, clarified by Code Case N-460, could not be met.
Due in part to geometry and actual physical barriers this coverage was unattainable. A combination of multiple beam angles was used to obtain the maximum coverage possible. The attached examination reports document the actual amount of examination coverage obtained. Drawings showing details of the affected welds including calculation methods are included as Attachment 2.
Justificetion "The four Reactor Vessel Inlet Nozzle to Shell Welds and the Inner Radius Exams at 67,113,247 and 293 degrees, the four Reactor Vessel Outlet Nozzle to Shell Welds and the Inner Radius Exams at 22,158,202 and 338 degrees were examined to the maximum extent practical using ultrasonic techniques in accordance with the requirements of Section V, Article 4 of the 1980 Edition thru Winter 1980 Addenda of the ASME Boiler and Pressure Vessel Code, "The Reactor Vessel inlet Nozzle to Shell Welds at 67,113,247 and 293 degrees (1RPV5-445A,1RPV5-445B,1RPV5-445C and 1RPV5 4450) (Item Numbers B03.090.001, 803.090.001 A; B03.090.002, B03.090.002A; B03.090.003, B03.090.003A: and B03.090.004, B03.090.004A): The four inlet Nozzle to Shell Welds were limited due to the reactor vessel nozzle configuration. Therefore, the 100% volumetric examination is impractical for this weld. The imposition of this requirement would create a considerable burden on Duke Energy Corporation.
During the examination of this weld, techniques were utilized to obtain the maximum possible coverage. Reference drawing 1213931D for scan coverage.
"The Reactor Vessel Outlet Nozzle to Shell Welds at 22,158,202 and 338 degrees (1RPV5-445E,1RPV5-445F,1RPV5-445G and 1RPV5-445H) (Item Numbers B03.090.005, B03.090.005A; B03.090.006, B03.090.006A; B03.090.007, 803.090.007A; and B03.090.008, B03.090.008A): The four Outlet Nozzle to Shell Welds were limited due to the reactor vessel nozzle configuration Therefore, the 100% volumetric examination is impractical for this weld. The imposition of this requirement would create a considerable burden on Duke Energy Corporation.
During the examination of these welds, techniques were utilized to obtain the maximum possible coverage. Reference drawing,1213930D for scan coverage.
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"The Reactor Vessel Inlet Nozzle Inner Radius Exam (1RPV5-445AR, 1RPV5445BR,1RPV5-445CR and 1RPVS-445DR)(Item Number 803.100.001, ;
B03.100.002, B03.100.003 and B03.100004): The four Inlet Nozzle Inner Radius Sections are limited due to the reactor vessel nozzle configuration. Therefore, the 100% volumetric, examination is impractical for this weld. The imposition of this requirement would create a considerable burden on Duke Energy Corporation.
During the examination of these welds, techniques were done to utilized to obtain the maximum possible coverage. Reference drawing 1213931D for scan coverage.
l "The Reactor Vessel Outlet Nozzle Inner Radius Exams (1RPV5-445ER, 1RPV5445FR,1RPV5-445GR and 1RPV5-445HR) (Item Number B03.100.005, B03.100.006,803.100.007 and B03.100.008): The four Outlet Nozzle Inner Radius l Sections are limited due to the reactor vessel nozzle configuration. Therefore, the l 100% volumetric examination is impractical for this weld. The imposition of this requirement would create a considerable burden on Duke Energy Corporation.
During the examination of these welds, techniques were utilized to obtain the maximum possible coverage. Reference drawing 1213930D for scan coverage.
" Pursuant to 10 CFR 50.55a(g)(6)(i), granting this relief for the four Reactor Vessel inlet Nozzle to Shell Welds and the Inner Radius Exams at 67,113,247 and 293 degrees, the four Reactor Vessel Outlet Nozzle to Shell Welds and the Inner Radius Exams at 22,158,202 and 338 degrees will, provide reasonable assurance of weld / component integrity, and is authorized by law. In addition, the requested relief will not endanger life or property or the common defense and security and is '
l otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Licensee's Proposed Altemative Examination (as stated):
l "In addition to the volumetric examination that has been performed on the McGuire L reactor vessel, Duke Energy has performed a visual examination of the intemals l- and the inside of the reactor vessel as required by ASME Section XI, Table IWB-
- ' 2500-1. This Visual examination did not identify any rejectable conditions per ASME Section XI acceptance standards.
"The use of radiography as an attemate volumetric examination method is not l_ feasible due to component thickness and restrictions from physical barriers which prohibit access for the placement of source, image quality indicators, film, etc. In ;
addition, the background radiation levels would not allow for a radiographic i examination that would render meaningful results i f
l " Performing the ultrasonic examination from the outside of the reactor vesselis not )
l a viable option. The design of McGuire's reactor building prohibits access for the !
equipment and personnel from outside the vessel.
" Duke Energy Corporation will continue to perform ultrasonic examinations of all reactor vessel welds to the maximum extent practicalin accordance with the j
i requirements of ASME Section V, Article 4,1989 Edition and Regulatory Guide 1.150, Revision 1, Appendix A. Code Case N-460 will be utilized in all cases where less than 100% but greater than 90% weld coverage is obtained. In cases where weld coverage of 90% or less is obtained, a request for relief from ASME Section XI Code requirements, will be submitted per the requirements of 10CFR50.55a. l l
" Duke Energy Corporation proposes in lieu of the greater than 90% coverage requirement of Code Case N-460 that the examination coverage obtained on the welds listed in Attachment i be considered to provide an acceptable level of quality and safety.
"No additional examinations will be performed."
l Evaluation: The Code requires 100% volumetric examination of the subject RPV nozzle-to-vessel welds and inside radius sections. However, complete examination is restricted by geometric configuration that makes the 100% volumetric
! examination impractical to perform for these areas. To gain access for l examination, the RPV nozzles would require design modifications. Imposition of l
this requirement would create an undue burden on the licensee.
The licensee has examined these welds to the extent practical, obtaining 43 to 85%
coverage of each nozzle-to-vessel weld and 58 to 74% coverage for each nozzle inside radius section. In addition, other Class 1 nozzles are being examined as required by the Code. Therefore, any existing pattems of degradation would have
- been detected by the examinations that were completed and reasonable assurance l of the structuralintegrity has been provided.
I Based on the impracticality of meeting tfie Code coverage requirements for the l subject nozzle-to-vessel welds and inside radius sections, and the reasonable l assurance provided by the examinations that were completed on these and other l Class 1 nozzles, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
- 3. CONCLUSION l
The INEEL staff has reviewed the licensee's submittal and concludes that for Request for Relief 97-004, Parts 1 and 2, the licensee has demonstrated that the Code l examination coverage requirements are impractical for the subject welds at McGuire, Unit 1. Furthermore, reasonable assurance of the structuralintegrity of the subject components has been provided by the examinations that were performed. Therefore, it
- is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i) for Request l for Relief 97-004, Parts 1 and 2.
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