ML20195J313
| ML20195J313 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/12/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20195J299 | List: |
| References | |
| NUDOCS 9811240238 | |
| Download: ML20195J313 (17) | |
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4 , SW4 9 g k UNITED STATES
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30ee6 0001
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SAFETY EVALUATION P'/JHE OFFICE OF NUCLEAR REACTOR REGULATION BEQbRDING THE FIRST AND SECOND 10-YEAR INTERVAL INSERVICE INSPECTION l REQUESTS FOR RELIEF FOR
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DUQUESNE LIGHT COMPANY
! OHIO EDISON COMPANY
! PENNSYLVANIA POWER COMPANY THE CLEVELAND ELECTRIC ILLUMINATING COMPANY ,
THE TOLEDO EDISON COMPANY BEAVER VAtt FY POWER STATION. UNIT NOS.1 AND 2 j DOCKET NOS. 50-334 AND 50-412 f
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1.0 INTRODUCTION
The Technical Specifications (TSs) for Beaver Valley Power Station, Unit Nos.1 and 2 (BVPS 'l j
and BVPS-2) state that the inservice inspection (ISI) of the American Society of Mechar,ical i Engineers (ASME) Code Class 1,2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel (B&PV) Code and applicable addenda as required by the Code of Federal Reaulations (CFR),10 CFR 50.55a(g), except where written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). In accordance
- with 10 CFR 50.55a(a)(3), altematives to the requirements of paragrapn (g) may be used, when i
authorized by the Director, Office of Nuclear Reactor Regulation, if (i) the proposed altematives l would provide an acceptable level of quality and safety or (ii) compliance with the specified ,
requirements would result in hardship or unusual difficultly 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 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requaements, set forth in the ASME Code,Section XI, ' Mes for Inservice !
Inspection of Nuefear 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 examinatbn of components and system pressure terts 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) !
twelve months prior to the start of the 120-month interval, subject to the limitations and i modifications listed therein. The applicable edition of Section XI of the ASME Code for the l
BVPS-1 second 10-year ISI interval, which ended September 20,1997, and for the BVPS-2 first 1 10-year ISI interval, which ended November 16,1997, is the 1983 Edition through Summer 1983 Addenda.
By letter dated November 24,1997, as supplemented May 12,1998, the licensee submitted '
requests for relief and proposed attematives to the Code requirements for BVPS-1 and BVPS-2.
9811240238 981112 %
PDR ADOCK 05000334
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G PDR ENCLOSURE
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l 2.0 EVALUATION The Nuclear Re datory Commission (NRC) staff, with technical assistance from its contractor, l the Idaho National Engineering and Environmental Laboratory (INEEL), has evaluated the information provided by the licensee in support of its 10-year inservice inspection interval .
program plan requests for relief BVPS-1 and BVPS-2. Based on the results of the review, the NRC staff adopts the contractor's conclusions and recommendations presented in the Technical Letter Report (TLR) attached.
[ Raquest for Relief RR-BV1-RV-AUG-1 (Rev 1): The ASME Code,Section XI, Table IWB- 2500- !
i 1, Examination Category B-A, requires 100% volumetric examination of Shell Welds as defined by Figure IWB-2500-1 and IWD-2500-2.
1 in accordance with 10 CFR 50.55a(s)(g)(6)(ii)(A), alllicensees must implement once, as part of the inservice inspection interval in effect on September 8,1992, an augmented volumetric j examination of the reactor pressure vessel (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 Figures IWB-2500-1 and -2, )
respectively. Essentially 100%, as defined by 10 CFR E9 55a(g)(6)(ii)(A)(2), is greater than 90%
of the examination volume of each weld.
Under 10 CFR 50.55a(a)(3)(ii), the licensee proposed an attemative to the Code-required augmented RPV examination required by the regulations as 100% coverage could not be achieved for Weld RC-R-C-8. The licensee stated:
"The altemative to the ruleis to perform the examinations to the maximum extent possible.
These examinations are supplemented by the visual examination performed on the interior of the vessel. The four support keyways are included in this examination. Therefore, the UT examination coupled with the visual examination of the radial support keyways and the surrounding areas provide an adequate measure of assurance of the integrity of this weld."
To comply with the augmented reactor vessel examination requirements of 10 CFR 30.55a(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 characterized as l greater than 90% of the examined weld volume.
At BVPS-1, the augmented coverage requirements cannot be met as examination of the lower l head to shell weld has phyrL al encumbrances that limit scan coverage. Adjacent radial support !
keyways welded to the RPV intemal surface limit coverage to 85% of the required examination volume. To obtain an additional 5% increase in coverage fc: the subject welds, design modifications would be required to allow access from the inside diameter (ID) surface.
Therefore, the licensee has shown, consistent with 10 CFR 50.55a(g)(6)(ii)(A)(5), an inability to completely satisfy the requirements of augmented examination.
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3-4 As a result of the augmented volumetric examination rule, licensees must rrske a reasonable effort to maximize examination coverage of their reactor vessels. In cases where coverage from i
the ID is insufficient, examination from the odtside diameter (OD) surface using manual inspection techniques may be an option. However, at BVPS-1, the neutron shield tank prevents access to the OD. Therefore, the licensee cannot enhance coverage by examining from the OD.
The licensee has achieved 85% coverage of RC 1-C-8, and has met the augmented examination ,
requirements for the remaining welds. In addition to the volumetric examination, the licensee has supplemented the volumetric examination with a visual examination of the vesselinterior. Based on the volumetric examination coverage attained and the supplemental visual examination, the NRC staff concludes that imposition of the augmented reactor vessel examination requirements of 10 CFR 50.55a(g)(6)(ii)(A) would result in a hardship for the licensee, and any significant pattems of degradat: ion, if present, would be detected under the licensee's proposed attemative.
The volumetric and supplemental visual examination performed provide a reasonable assurance of the continued structuralintegrity of the subject lower head to shell weld. Therefore, pursuant to 10 CFR 50.55a(g)(ii)(A)(5) and 10 CFR 50.55a(a)(3)(ii), the licensee's proposed altemative is authorized in that compliance with the specified requirement would result in hardship without a compensating increase in the level of quality and safety.
Request For Relief RR-BV1-RV-WELDS (Rev 1): ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A, items 81.11 and B1.30 require 100% volumetric examination of the Class 1 reactor vessel circumferential shell weld and shell-to-flange weld as defined by Figures IWB-2500-1, and -4.
The Code requires that the subject reactor pressure vessel shell-to-flange weld and lower head to shell weld be 100% volumetrically examined during the inspection interval. Due to the i geometric configuration of the shell-to-flange weld (extreme taper of flange), the examination coverage was limited to 82.5%. Due to four radial support keyways, equally spaced around the :
circumference of the reactor vessel, the examinauon coverage of the lower head to shell weld l
was 'imited to 85%. Based on the information provided in this request for relief, it is impractical i to examine the subject welds to the extent required by the Code. For complete examination coverage, redesign and modification of the reactor vessel would be necessary, imposition of this requirement would cause a burden on the licensee.
The licensee proposes to perform the volumetric examinations to the maximum extent practical on the subject welds. The volumetric examinations are supplemented by visua! examination performed on the vessel interior. The visual examination includes the four support lugs and flange to shell weld. These examinations., in addition to the surface examination of the interior of the vessel, provide reasonable assurance that the structuralintegrity 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 granted pursuant to 10 CFR 50.55a(g)(6)(i).
Request for Relief RR BV2-RV-AUG-1 (Rev 1): ASME Code,Section XI, Examination Category B-A, item B1.10, Augmented Reactor Pressure Vessel Examination in ac.cordance with 10 CFR 50.55a(g)(6)(ii)(A).
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l' In accordance with 10 CFR 50.55a(g)(6)(ii)(A), alllicensees must implement once, as part of the inservice inspection interval in effect on September 8,1992, an augmented volumetric examination of the RPV welds specified in it6m 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 Figures IWB-2500-1 and -2, respectively. Essentially 100%, as defined by 10 CFR 50.55a(g)(6)(ii)(A)(2), is greater than 90% of the examination volume of each weld.
, At BVPS-2, the augmented coverage requirements cannot be met for the lower head to shell
! weld due to physical encumbrances that limit scan coverage. Adjacent core supports that are welded to the RPV intemal surface limit coverage to 89% of the required volume. To obtain an additional 1% coverage for the subject welds, design modifications would be required to increase access from the ID surface. Without a design modification, imposition of this requirernent would i i
result in a hardship without a compensating increase in the level of quality and safet).
In cases where examination coverage from ID is inadequate, exa@ ation from the OD surface ,
- using manual inspection techniques may be an option. However, t. BVPS-2, the neutron shield !
- tank prevents access from the OD. Therefore the licensee cannot enhance coverage by
{ examining from the OD. )
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i The licensee has examined 89% of the subject RPV shell weld. Furthermore, the volumetric examination is supplemented with a visual examination performed on the interior of the vessel.
2 Based on the volumet ic examination coverage attained and the supplemental visual examination, the NRC staff concludes that imposition of the augmented reactor vessel examination requirements of 10 CFR 50.55a(g)(6)(ii)(A) would result in a hardship fo:' the
- licensee, and any significant pattems of degradation, if present, would be detected under the licensee's proposed attemative. The volumetric and supplemental visual examination performed provide a reasonable assurance of the continued structuralintegrity of the subject lower head to j shell weld. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee's proposed attemative is authorized.
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- Request for Relief RR-BV2-B1.11-1 (Rev 1)
- ASME Code,Section XI, IWS-2500-1, Examination
! Category B-A, item B1.11 requires 100 % volumetric examination of the Class i reactor vessel i circumferential shell welds as defined by Figure IWB 2500-1, 1
{ The licensee is requesting relief from examining 100% of the Code-required volume of the reactor pressure vessel circumferential weld. The licensee's request for relief is specifically for lower head-to-shell weld 2RCS-REV21-C-4.
J l The Code requires that the reactor pressure vessellower head to shall welds be 100%
volumetrically examined during the inspection interval. Due to four core support lugs, welded to the inside surface of the reactor vessel, the required volumetric examination of this lower head-to-shell weld was limited to an examination coverage of 89%. Based on the information provided in this request for relief, it is a hardship without a compensating increase in the level of quality and safety to examine the subject weld to the extent required by the Code. To obtain the required coverage, removal of one of the support lugs from the reactor vessel would be necessary and could damage the vessel.
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The licensee proposes to perform the volumetric examination to the maximum extent possible coupled with a visual examination of the vessel interior which includes the four core support lugs.
! The volumetric examination, in addition to the surface examination of the interior of the vessel, i provide reasonable assurance that the structuralintegrity will be maintained and the presence of degradation will be detected using the stated proposal. Based on the hardship 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 vesselinterior j completed, the attemative is authorized pursuant to 10 CFR 50.55a(a)(3)(ii).
Request for Relief RR-BV3-lWA-1 (Rev 1): ASME Code,Section XI, IWA-5242(a), Removal For j Visual Examination Of Bolting in Class 1 and 2, Borated Systems (BVPS Third interval, and ,
l BVPS Second Interval) requires that for systems borated for the purpose of controlling i reactivity, insulation shall be removed from pressure-retaining bolted connections for VT-2 visual
- examination. Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed an altemative to the i ASME Section XI requirements for removing insult tion from Class 1 and 2, pressure-retaining l bolted connections for VT-2 visual examination during VT-2 visual examinations.
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The licensee stated that Revision 1 of this relief requat was incided in the licensee's submittal, dated November 24,1997, for reference proposes only 4tief for Class 1 systems (Revision 0) was approved in an NRC Safety Evaluation (SE) dated October 8,1997. Revision 1 reflects additior6al statements made in Duquesne Light Cornpany's (DLC's) response to the NRC's Request for Additional Information (RAI) dated August 29,1997, and did not modify the allemative proposed. Based on the licensee's response to the RAI and the SE dated October 8, 1997, the NRC staff concluded that no additional evaluation was required. Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee's proposed Eltemative remains authorized for the current interval as approved in the SE dated October 8,1N7.
Request for Relief RR-BV3-lWA-2, (Re 1): ASME Code,Section XI, lWA-5250(a)(2),
Corrective Action Resulting from Leakaga M Bolted Connections (BVPS-1 -Third Interval, and BVPS Second Interval) requires that the s:nrce of leakages detected during a system pressure test be located and evaluated by the owner for corrective action. When the leakage is at a bolted connection, the botting shall be rl noved, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100 Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed an altemative to the ASME Ser/ ion XI requirements for removal of bolting at leaking connections for VT-3 visual examinatier.
The licensee stated that Revision 1 of this relief request was included in the licensee's submittal, dated November 24,1997, for reference purposes only. Relief for Class 1 systems (Revision 0) was addressed in an NRC SE dated Oct@r 8,1997. Revision 1 reflects additional statements ,
made in DLC's response to the NRC% RAI dated August 29,1997. Based on the licensee's ;
response to the RAI and the SE of October 8,1997, the NRC staff concluded that no additional evaluation was warranted. The licensee's prr.sposed altemative as approved in the NRC GE dated October 8,1997, remains valid and is authorized for Class 1 systems for the current interval.
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3.0 CONCLUSION
The NRC staff has reviewed the information provided by the licensee for requests for relief for BVPS-1 and BVPS-2. Based on the evaluation of relief requests BV1-RV-AUG (Rev 1) and BV2-RV-AUG 1 (Rev 1), the NRC staff concluded that imposing the Code requirements on the ;
licensee results in a burden without a compensating increase in quality and safety. Therefore, i the licensee's proposed attematives are authorized for the current interval pursuant to 10 CFR 50.55a(a)(3)(ii). In addition, pursuant te 10 CFR 50.55a(a)(3)(ii) and NRC's SE dated October 8, I 1997, the licensee's attemative containe i in relief request BV3-lWA-2 (Rev 1), remains authorized for the interval approved in tha October 8,1997 SE.
For relief request BV3-lWA-1 (Rev 1), the NRC staff concluded that, pursuant to 10 CFR 50.55a(a)(3)(i) and the NRC's SE dated October 8,1997, the licensee's proposed altemative provides an acceptable level of quality and safety for Class 1 systems and remains authorized for the interval approved in the October 8,1997 SE. ;
For relief requests BV1 RV-WELDS (Rev 1) and BV2-B1.11.1 (Rev 1), the NRC staff concluded that the ASME Code examination seguirements are impractical. Therefore, pursuant to 10 CFR
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50.55a(g)(6)(i) the licensee's request for relief is granted. These relief requests are authorized '
by law and will not endanger life or property or the common defense and security and are i otherwise in the public interest giving due consideration to the burden upon the licensee that
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could result if the requirements were imposed on the facility.
Attachment:
Technical Letter Report Principal Contributor: G. Hatchett Date: November 12, 1998 1
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d TECHNICAL LETTER REPORT ON THE FIRST AND SECOND 10-YEAR INTERVAL REQUESTS FOR REllEF DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION UNITS 1 AND 2 l DOCKET NUMBERS 50 334 AND 50 412 !
1.0 INTRODUCTION
l By letters dated November 24,1997, and May 12,1998, the licensee, Duquesne Light Company, submitted requests for relief for the Beaver Valley Power Station, Units 1 and I
- 2. These requests for relief ar6 for the first 10-year inservice inspection (ISI) interval at l
Unit 2 and second 10 year interval for Unit 1. The Idaho National Engineering and Environmental Laboratory (INEEL) staff has evaluated the information provided by the
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licensee regarding the requests for relief in the following section.
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2.0 EVALUATION l
l The information provided by Duquesne Light Company in support of the requests for relief from Code requirements and the proposed alternative to Augmented Reactor Vessel Examinations has been evaluated, and the bases for disposition are documented below. The Code of record for Beaver Valley Power Station, Unit 1, second 10-year ISI interval, which ended September 20,1997, and Unit 2, first 10 year interval, which ended November 16,1997,is the 1983 Edition through Summer 1983 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code.
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ATIACHMEST
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d-A.
Raouant for Relief BV1-RV-AUG-1. Rev.1 (Unit 11. Framination Cataoorv B-A ltam B1.10 Annmented Ramednr Pra==ure Va==el Fvaminatinn ner 10 CFR 50.55ataV8viiVA 3 J
Regulatorv Requirement: In accordance with 10 CFR 50.55a(g)(6)(li)(A), all l
licensees must implement once, as part of the inseWice inspection interval in effect i
on September 8,1992, an augmented volumetric examination of the RPV weids specified in item B1.10 of Examination Category B-A of the 1989 Edition of the
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ASME Code,Section XI. Examination Category B-A,' Items B1.11 and B1.12 i require volumetric examination of essentially 100% of the RPV circumferential and i longitudinal shell welds, as defined by Figures IWB-2500-1 and -2, respectively. l Essentially 100%, as defined by 10 CFR 50.55a(g)(6)(ii)(A)(2), is greater than 90% !
of the examination volume of each weld.
Licensee's Pronosed Altamative Pursuant to 10 CFR 50.55a(a)(3)(li), the licensee i has proposed an altamative to the coverage requirements of the augmented RPV !
examination required by the regulations because essentially 100% coverage could
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not be achieved for Weld RC-R-1-C-8. The licensee stated: I "The attemative to the rule is to perform the examinations to the maximum extent possible. These examinations are supplemented by the visual examination performed on the interior of the vessel. The four support keyways are included in j this examination. Therefore, the UT examination coupled with the visual examination of the radial support keyways and the surrounding areas provide an )
i adequate measure of assurance of the integrity of this weld." ;
Licensee's Basis for Raouesting Rehef(as stated)
"In accordance with 10 CFR 50.55a(a)(3)(ii), an attemative to the requirement is ;
proposed on the basis that compliance with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of i quality and safety. The lower head to shell weld (RC-R-1-C-8) examination is !
limited due to four radial support keyways.
l "This weld was examined using automated techniques during the 1R11 refueling !
outage. Eighty-five percent (85%) of the required volume was examined. The four radial support keyways are we'ded to the inside surface of the reactor vessel just above weld RC-R-1-C-8. Access to the outside diameter of this weld is precluded by the neutron shield tank. Supplemental scan angles were also considered. It l was determined that use of 45' and 60' shear waves, with a 70' refracted l longitudinal angle for the near surface and clad interface provided the maximum l examination coverage and efficiency of transducer movement. !
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"10 CFR 50.55a(g)(6)(ii)(A) defines " essentially 100%", for the purpose of the augmented examination, as more than 90 percent of the examination volume of l each weld, where the reduction in coverage is due to interference by another !
component, or part geometry. To obtain the required volume would necessitate l removal of one of the radial support keyways. Removal of a support keyway, for l
this purpose, is considered a hardship. A majority of the required volume was I examined (85%), and this amount of coverage provides an adequate level of quality and safety."
Evalomtinre To comply with the augmented reactor vessel examination requirements of 10 CFR 50.55a(g)(6)(il)(A), licensees must volumetrically examine essentially 100% of each of the item B1.10 shell welds. In accordance with the l regulations, essentially 100% is defined as greater than 90% of the examination volume of each weld.
At Beaver Valley Power Station, Unit 1, the augmented coverage requirements cannot be met for the lower head to shell weld due to physical restrictions that limit scan coverage. Adjacent radial support keyways that are welded to the RPV intemal surface lidt coverage to 85% of the required examination volume. To obtain an additional 5% increase in coverage for the subject welds, design modifications would be required to allow access from the inside surface (ID).
Therefore, imposition of this requirement would result in a hardship without a compensating increase in the oel of quality and safety.
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 surface (OD) using manual inspection techniques may be an option.
However, at Beaver Valley Power Station, Unit 1, the neutron shield tank prevents access from the OD. Therefore, the licensee cannot enhance coverage by examining from the OD.
The licensee has examined a significant portion of the subject weld, and has met the coverage requirements for the remaining welds. Furthermore, the volumetric examination is supplemented with a visual examination of the interior of the vessel.
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L Based on the volumetric examination coverage attained and with the visual examination, the INEEL staff concludes that any significant pattems of degradation, if present, would have been detected and that the examinations performed provide reasonable assurance of the continued structuralintegrity of the subject lower head to shell weld. Therefore, it is recommended that the licensee's proposed .
altemative be authorized pursuant to 10 CFR 50.55a(a)(3)(ii).
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Raauest for Relief BV1-RV-WFt r3S. Rev.1 (Unit 11 Examination Categorv B-A.
Item B1.11 and B1.30. Pressure Retaining Walds in Ranetnr Vammal Code Recuirement: Section XI, IWB-2500-1, Examination Category B-A, items B1.11 and B1.30 require 100% volumetric examination of the Class 1 reactor vessel circumferential shell weld and shell-to-flange weld as defined by Figures
- IWB-2500-1, and -4.
4 Licensee's Code Relief Raouant: Pursuant to 10 CFR 50.55a(g)(6)(i) , relief is requested from examining 100% of the Code-required volume of reactor vessel i circumferential Wald RC-R-1-C-8 and shell-to-flange Weld RC-R-1-C-1.
Licensee's Pronosed Alternative (as stated):
' "The attemative to the examination requirement is to perform the examinations to the maximum extent possible. These examinations are supplemented by the visual examination performed on the interior of the vessel. The four support lugs and flange to shell weld are included in this examination. Therefore, the UT l
examinations coupled with the visual examination of the support lugs and the surrounding areas and the visual examination of the interior surface of the flange to shell weld provide an adequate measure of assurance of the integrity of these welds."
Licensee's Basis for Recuesting Relief (as stated):
"In accordance with 10CFR50.55a(g)(6)(i), relief is requested on the basis that compliance with the Code requirement is impractical. The flange to upper circumferential shell weld (RC-R-1-C-1) examination is limited due to component geometry. The lower head to shell weld (RC-R-1-C-8) examination is limited due to interference by the four radial support keyways.
"The flange to upper circumferential shell weld (RC-R-1-C-1) was examined using an automated technique during the 1R11 refueling outage. Additionally, a manual 4
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l UT examination was performed on weld RC-R-1-C-1 from the flange face surface in September 1989 (1" period of the current interval) in accordance with Note 5 of Examination Category B-A. The first period examination from the flange face 4
surface covered 62% of the weld circumference and used 5 different beam angles p (16' out,12' out,6' out, O*, and 6' in, as permitted by ASME V, Article 4, T441.4.2) to cover the required volume. There were no limitations for this s- examination.
"The automated examination of the flange to shell weld (RC-R-1-C-1) was limited to 54% of the required volume due to the inside and outside diameter taper of the flange (see sketch on next page for general configuration).' Scan angles of 0*,
45',60', and 70* were used. The movement of the transducer was limited on the inside surface above the mark noted due to the curvature of the flange taper (detailed proprietary vendor supplied drawings are available on site). Considering i
! both examinations (i.e., from the flange face and from the inside surface of the vessel wall) approximately 82.5% of the total required volume was examined during
{ the current interval, t
"The automated examination of the lower head to shell weld (C-8) was limited to 85% of the required volume. The four radial support keyways, equally spaced around the circumference, are located adjacent to the weld and preclude the examination of C-8 at these areas.
4 "Since the examination limitations are due to the geometry of the component or i interference by adjacent components, redesign of the reactor vessel would be necessary to completely accomplish the required examinations. It is impractical to i redesign the reactor vessel for this purpose. A majority of the required volume was
- covered in both cases."
Evaluation: The Code requires that the subject reactor pressure vessel shell-to-flange weld and lower head to shell weld be 100% volumetrically examined during
, the inspection interval. Due to the geometric configuration of the shell-to-flange l l
weld (extreme taper of flange), the examination coverage was limited to 82.5%. l j Due to four radial support keyways, equally spaced around the circumference of the reactor vessel, the examination coverage of the lower head to shell weld was l limited to 85%. Based on the information provided in this request for relief, it is l impractical to examine the subject welds to the extent req Ved by the Code. For 4
complete examination coverage, redesign and modification of the reactor vessel would be necessary. Imposition of this requirement would cause a considerable 1 Tables, Figures and attachments furnished with the licensee's submittal era not included in this report.
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i burden on the licensee.
The licensee proposes to perform the volumetric examinations to the maximum i extent practical on the subject welds. The volumetric examinations are also supplemented by the visual examination performed on the interior of the vessel.
The visual examination includes the four support lugs and the flange to shell weld. '
Based on the significant portion of the volumetric examinations completed, it is i
reasonable to conclude that a pattem of degradation, if present, would have been l
detected. As a result, reasonable assurance of continued structuralintegrity has been provided. Therefore, it is recommended that relief be grante., pursuant to '
I C. Reauest for Relief BV2-RV-AUG-1. Rev.1 (Unit 21 Examination LQegorv B-A.
!!em B1.10. Augmented Reactsr Pressure Vessel Examination oer 10 CFR 50.55alo)(8)(in(A)
Regulatorv Reauirement: In accordance with 10 CFR 50.55a(g)(6)(li)(A), all licensees must implement once, as part of the inservice inspection interval in effect on September 8,1992, an augmented volumetne examination of the RPV welds specified in item B1.10 of Examination Category B-A of the 1989 Edition of the l
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 I longitudinal shell welds, as defined by Figures IWB-2500-1 and -2, respectively.
Essentially 100%, as defined by 10 CFR 50.55a(g)(6)(ii)(A)(2), is greater than 90%
of the exarnination volume of each weld.
l Licensee's Proonsed Alternative: Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee l has proposed an alternative to the coverage requirements of the augmented RPV examination required by the regulations for Weld 2RCS-REV21-C-4. The licensee stated:
I "The attemative to the rule is to perform the examination to the maximum extent possible. This examination is supplemented by the visual examination p6rformed on the interior of the vessel. The four support lugs are included in this examination.
Therefore, the UT examination coupled with the visual examination of the support l
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e d
lugs and the surrounding areas provide an adequate measure of assurance of the integrity of this weld."
Licensaa's Basin for Raouesting Raliaf(as stated):
"In accordance with 10 CFR 50 55a(a)(3)(ll), an a4emative to the requirement is i
proposed on the basis that compliance with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The lower head to shell weld (2RC-REV21-C-4) examination is limited due to four core support lugs.
t i
i "This weld was examined using automated techniques during the 2R06 refueling l
' outage. Over eighty-nine percent (89%) of the required volume was examined.
l The four lugs are welded to the inside surface of the reactor vesseljust above weld '
i 2RC-REV21-C-4. Access to the outside diameter of this weld is precluded by the
- neutron chield tank. Supplemental scan angles were also considered. It was determined that use of 45' and 60' shear waves, with a 70' refracted longitudinal angle for the near surface and clad interface provided the maximum examination coverage and efficiency of transducer movement.
"10 CFR 50.55a(g)(6)(ii)(A) defines " essentially 100%", for the purpose of the augmented examination, as more than 90 percent of the examination volume of each weld, where the reduction in coverage is due to interference by another component, or part geometry. To obtain the last 1% of the required volume would necessitate removal of one of the support lugs. Removal of a support lug, for this purpose, is considered a hardship. A majority of the required volume was examined (89%), and this amount of coverage provides an adequate level of quality and safety."
Eyatuation To comply with the augmented reactor vessel examination requirements of 10 CFR 50.55a(g)(6)(ii)(A), licensees must volumetrically examine essentially 100% of eat.h 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 Beaver Valley Power Station Unit 2, the augmented coverage requirements cannot be met for the lower head to shell weld due to physical restrictions that limit scan coverage. Adjacent core support lugs that are welded to the RPV intemal surface limit coverage to 89% of the required examination volume. To obtain an additional 1% coverage for the subject welds, design modifications would be required to increase access from the inside surface (ID). Therefore, imposition of this requirement would result in a hardship without a compensating increase in the 7
0 level of quality ed safety.
i 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 surface (OD) using manual inspection techniques may be an option.
However, at Beaver Valley Power Station, Unit 2, the neutron shield tank prevents l
access from the OD. Therefore, it is concluded that the licensee cannot enhance coverage by examining from the OD.
l The licensee has examined a significant portion of the subject weld. Furthermore, 4
the volumetric examination is supplemented with a visual examination performed l on the interior of the vessel. Based on the volumetric examination coverage attained and the visual examination, the INEEL staff concludes that any significant !
patterns of degradation, if present, would have been detected and that the examinations performed provide reasonable assurance of the continued structural integrity of the subject lower head to shell weld. Therefore, it is recommended that the licensee's proposed altemative be authorized pursuant to 10 CFR 50.55a(a)(3)(ii).
D. Recuest for Relief BV2-B1.11-1. Rev.1 (II D?r 2t Examination Cateoorv B-A. Item B1.11 Pressure Retalnina Welds in Reactor Vessel Code Reauirement: Section XI, IWB-2500-1, Examination Category B-A, item B1.11 requires 100% volumetric examination of the Class 1 reactor vessel s
circumferential shell welds as defined by Figure IWB-2500-1.
Licensee's Code Relief Reauest: Pursuant to 10 CFR 50.55a(g)(6)(l) , relief is requested from examining 100% of the Code-required volume of the reactor vessel circumferential weld. Specifically the lower head-to-shell weld 2RCS-REV21-C-4.
Licensee's Proposed Aftemative (as stated):
"The altemative to the examination requirement is to perform the examinations to i 1
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. . _ _ . _ - ~ _ _ . _ . . _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ . _ . _ . _
the maximum extent possible. This examination is supplement by the visual examination performed on the interior of the vessel. The four support lugs are included in this examination. Therefore, the UT examination coupled with the visual examination of the support lugs and the surrounding areas provides an adequate measure of assurance of the integrity of this weld."
Licensee's Ramle for Raouesting Raliaf(as stated):
"In accordance with 10 CFR 50.55a(g)(6)(1), reliefis requested on the basis that compliance with the Code requirement is impractical. The lower head to shell weld (2RCS-REV21-C-4) examination is limited due to the four core support lugs.
"This weld wac examined using automated techniques during the 2R06 refueling outage. Over eighty-nine percent (89%) of the required volume was examined.
The four lugs are welded to the inside surface of the reactor vesseljust above weld 2RCS-REV21-C-4.
"ASME XI defines ' essentially 100', as more than 90 percent of the examination volume. To obtain the last 1% of the required volume would necessitate removal of one of the support lugs. Removal of a support lug for this purpose is considered impractical."
Evaluation The Code requires that reactor pressure vessellower head to shell welds be 100% volumetrically examined during the inspection interval. Due to four core support lugs, welded to the inside surface of the reactor vessel, the required volumetric examination of this lower head-to-shell weld was limited to an examination coverage of 89%. Based on the information provided in this request for relief, it is impractical to examine the subject weld to the extent required by the Code. To obtain the required coverage, removal of one of the support lugs from the reactor vessel would be necessary. Imposition of this requirement would cause a considerable burden on the licensee.
The licensee proposes to perform the volumetric examination to the max l mum extent practical. The volumetric examination is supplemented by a visual examination of the interior of the vessel that includes the four core support lugs.
Based on the significant portion of the volumetric examination completed and the visual examination, it is reasonable to conclude that a pattem of degradation , if present, would have been detected. As a result, reasonable assurance of the continued structuralintegrity of this weld has been provided. Therefore,it is 9
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recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
E, Raouent for Relief BV3-lWA-1. Rev.1. Parnaraoh IWA-5242(a). Visual Framination of innolated Comnonents Note: The licensee stated that Revision 1 of this relief request was included in the licensee's submittal, dated November 24,1997, for reference purposes only. Relief for Class 1 systems (Revision 0) was approved in an NRC Safety Evaluation Report (SER) dated October 8,1997. Revision 1 reflects additional statements made in Duquesne Light Company's response to the NRC's Request for Additional Information dated August 29,1997. Therefore, an evaluation is not required, and it is recommended that the proposal remain authorized.
1 F. Recuent for Relief BV3-fWA-2. Rev.1. Paraaraoh IWA-5250(a)(2) I anknoe at Bolted Conner finna Note: The licensee stated that Revision 1 of this relief request was included in the licensee's submittal, dated November 24,1997, for reference purposes only. Relief for Class 1 systems (Revision 0) was addressed in an NRC Safety Evaluation Report (SER) dated October 8,1997. Revision 1 reflects additional statements made in Duquesne Light Company's response to the NRC's Request for Additional Information dated August 29,1997. Therefore, an evaluation is not required, and it is recommended that the proposal remain authorized.
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3.0 CONCLUSION
i The INEEL staff has reviewed the licensee's submittal and concludes that for Requests for Relief BV1-RV-AUG-1 and BV2-RV AUG-1, full compliance with the !
i augmented reactor pressure vessel examination requirements would result in a hardship without a compensating increase in the level of quality and safety. Therefore, it is recommended that the proposed attematives be authorized pursuant to 10 CFR 50.55a(3)(ii). For Requests for Relief BV1-RV-WELDS and BV2-B1.11-1, it is concluded that the Code requirements are impractical. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i). The Licensee included two Requests for Relief, BV3-lWA-1, Rev.1 and BV3-IWA-2, Rev.1 for reference purposes I only. Rev. O of these Requests for Relief was addressed in an NRC Safety Evaluation Report (SER) dated October 8,1997. Revision 1 reflects additional statements made in Duquesne Light Company's response to the NRC's Request for AdditionalInformation dated August 29,1997. Therefore, an evaluation was not performed on Rev.1 of these Requests for Relief, and it is recommended that the proposals remain authorized.
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