ML20203H988
| ML20203H988 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 02/18/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20203H984 | List: |
| References | |
| NUDOCS 9902230223 | |
| Download: ML20203H988 (10) | |
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n nrgg p k UNITED STATES
- ! j~ NUCLEAR REGULATORY COMMISSION
% e WASHINGTON, D.C. 30e86-0001 Q**@T)
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ALTERNATIVE TO 10 CFR 50.55a(a)(6)(ii)(A) AUGMENTED REACTOR VESSEL EXAMINATION '
VERMONT YANKEE NUCLEAR POWER STATION DOCKET NUMBER 50-271
1.0 INTRODUCTION
Inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Soiler and Pressure Vessel (B&PV) Code and applicable addenda as required by 10 CFR 50.55a(g),
except wnere specific written relief has been granted by the Commission pursuant to 10 CFR 50.55(a)(g)(6)(i).10 CFR 50.55a(a)(3) states that attematives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed attematives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, " Rules for Inservice inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for the Vermont Yankee Nuclear Power Station third 10-year inservice inspection (ISI) interval is the 1986 Edition, Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the Commission in support of that determination and a request made for relief from the ASME Code requirement. After evaluation of the determination pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose attemative requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed.
Enclosure 1 9902230223 990218 PDR ADOCK 05000271 G PDR
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- 2 The augmented RPV examination requirements of 10 CFR 50.55a(g)(6)(ii)(A), require that a
- IMensee must volumetrically examine essentially 100% (i.e. >90%) of each of the item B1.10 l shell welds. As an altamative to the regulations, the licensee proposed that the limited examinations that were performed be considered as acceptable to satisfy the augmented reactor vessel examination requirement. By letter dated January 23,1997, Vermont Yankee Nuclear Power Corporation (licensee) submitted an altamative to 10 CFR 50.55a(g)(6)(ii)(A) !
augmented reactor pressure vessel examination for Vermont Yankee Nuclear Power Station.- '
l 2.0 EVALUATION The statY, with technical assistance from its contractor, the Idaho National Engineering and Environmental Laboratory (lh'FEL), has evaluated the information provided by the licensee in support of its altamative to 10 C/,~R 50.55a(g)(6)(ii)(A) augmented reactor pressure vessel examination for Vermont Yankee Nuclear Power Sation. Based on the results of the review, the staff adopts the contractor's conclusions and recorrendations presented in the Technical Letter Report (TLR) enclosed.
l The licensee performed the augmented reactor pressure vessel weld examinations during the September 1996 refueling outage. The 1986 Edition of ASME Section XI is the Code of record for Vermont Yankee Nuclear Power Station's third 10-year ISI interval, which began September 1,1993, and ends August 31,2003. The information provided by the licensee in support of the proposed attemative has been evaluated and the basis for disposition is documented below.
Altemative to 10 CFR 50.55afo)(6)(ii)(A). Auamented Reactor Pressure Vesse! Examination Reaulatorv Reauirement: in accordance with 10 CFR 50.55a(g)(6)(ii)(A), all licensees must implement once, as part of the inservice inspection intervalin effect on September 8,1992, an augmented volumetric examination of the RPV welds specified in item B1.10 of Examination Category B-A of the 1989 Edition of the ASME Code,Section XI. Examination Category B-A, items B1.11 and B1.12 require volumetric examination of essentially 100% of the RPV
- circumferential and longitudinal shell welds, as defined by 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.
Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), the licensee proposed that the coverages obtained for the subject welds be considered to satisfy the requirement. The licensee indicates that the examinations were performed to the extent practical for each RPV shell weld, as required by 10 CFR 50.55a(g)(6)(ii)(A). Weld CH was completely inaccessible for examination. The coverages achieved for all the RPV welds are listed in Table A-1.
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3 Table A-1 Code Wold item Coverage Interference CD (Cire) 81.30 90 % Guide rods D1 (Vert) . 81.12 99 % Steam dryer support bracket D2 (Vert) 81.12 96 % Steam dryer support bracket DE (Circ) B1.11 90 % Guide rods E1 (Vert) - B1.12 51 % Feedwater sparger, core spray piping header, offset between upper and lower tool configurations E2 (Vert) 81.12 53 % Feedwater sparger, core spray piping header, offset between upper and lower tool configurations EF (Circ) 81.11 69% Guide rods, core spray piping down-comers, core shroud lifting iugs.
surveillance specimen holders and breckets F1 (Vert) B1.12 65% Jet pump riser brackets F2 (Vert) B1.12 65% Jet pump riser brackets FG (Circ) B1.11 68 % Guide rods, core spray piping down-comers, core snroud lifti1g lugs, surveillance specimen holders and brackets G1 (Vert) B1.12 45% Recire inlet pipe, jet pump instrumentation L-bracket G2 (Vert) B1.12 45% Recirc inlet pipe, jet pump instrumentation L-bracket GH (Circ) B1.11 0% Jet pump instrumentation L-bracket All(Vert) 65%
All(Circ) 63%
To coraply with the augmented RPV examination requirements of 10 CFR 50.55a(g)(6)(ii)(A),
licensees must volumetr;cally examine essentially 100% (i.e. >90%) of each cf the item B1.10 shell welds. As an alten3ative to the regulations, the licensee proposed that the limited examir.ations that were performed be considered to satisfy the augmented reactor vessel examination requirement.
The ultrasonic examinations of th* RPV at the Vermont Yankee Nuclear Power Station have to be performed from the inside of ine vesu due to the biological shield wall. The total coverage obtained was 63% for the circumkrential .. ids and 65% for the longitudinal welds.
Limitations associated with the circumfarelitial weld examinations included: Guide Rods, Surveillance Specimen Holders and Breckets, Core Spray Piping Down-Comers, Core Shroud Lifting Lugs, and Jet Pump Instrumentation L-bracket. Limitations associated with the longitudinal weld examinations included: Steam D ver Support Brackets, Feedwater Sparger, Core Spray Piping Header, Jet Punip Riser Brackets, Jet Pump instrumentation L-bracket, Offttt Botween Upper and Lower Tool Configurations, and the Recire inlet Pipe.
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As a result of the augmented volumetric examination rule, licensees must make a reasonable f effort to maximize examination coverage of their reactor vessels. In cases where examination i coverage from the inside is inadequate, examination from the outside using manual inspection i techniques is a potential option. However, at Vermont Yankee, the bioshield is only 12 inches !
from the uninsulated OD of the reactor vessel. Therefore, the close proximity of the bioshield 4
. severely limits access for equipment and personnel to the outside surface. Although limited
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access is available for four welds (above the bioshield or through nozzle windows), i examination at these locations would only increase the total examination volume by an !
estimated 1%. Therefore, it is concluded that the licensee cannot significantly enhance i coverage by examining from the outside surface.
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The licensee stated that only two vendors provide the service of ID volumetric examinations.
The licensee contracted ABB Combustion Engineering Nuclear Operations in conjunction with Southwest Research Institute to perform the augmented volumetric examinations. The licensee and the inspection vendor have worked continuously, including modifying the ;
inspection tool, to increase examination coverage. The examination revealed seven recordable indications, one of which exceeded the ASME Section XI, IWB-3511, acceptance criteria. The flaw was reported to the NRC and was accepted in a Safety Evaluation Report dated October 11,1996.
The licensee has examined a significant portion of the reactor pressure vessel shell welds (greater than 63%). Furthermore, it appears that the licensee, short of disassembly of the reactor pressure vessel intemal components, has worked to increase coverage to the maximum extent practical. Therefore, based on the cumulative volumetric examination coverage obtained, the staff concluded that any significant pattems of degradation, if present, would have been detected and that the examinations performed provide an acceptable level of quality and safety and the licensee's proposed alt 6tr,ative to the regulations is authorized pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), and 10 CFR 50.55a(a)(3)(i).
3.0 CONCLUSION
The staff concluded that the licensee's proposed alternative provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative to the regulations is authorized pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), and 10 CFR 50.55a(a)(3)(i).
Principal Contributor: Tom McLellan l Date: Fehn.u f 18, 1999 i
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TECHNICAL LETTER REPORT i
ALTERNATIVE TO 10 CFR 50.55afo)(6)(ii)(A) !
i AUGMENTED REACTOR PRESSURE VESSEL EXAMINATION VERMONT YANKEE NUCLEAR POWER CORPORATION
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VERMONT YANKEE NUCLEAR POWER STA' TION l
DOCKET NUMBER 50-271 i
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1.0 INTRODUCTION
By letter dated January 23,1997, the licensee, Vermont Yankee Nuclear Power Station, 3 proposed an attemative to the augmented examination of the reactor pressure vessel (RPV) required by 10 CFR 50.55a(g)(6)(ii)(A). The Idaho National Engineering and Environmental )'
I Laboratory (INEEL) staff has evaluated the information provided by the licensee regarding this attemative in the following section.
2.0 EVALUATION The licensee performed the augmented reactor pressure vessel weld examinations during the September,1996, refueling outage. The 1986 Edition of ASME Section XIis the Code of record for Vermont Yankee Nuclear Power Station's third 10-year ISI interval, which began September 1,1993, and ends August 31,2003. The information provided by the licensee in support of the proposed attemative has been evaluated and the basis for disposition is documented below.
. Attemative to 10 CFR 50.55afa)(6)(ii)(A). Auomented Reactor Pressure Vessel Examination Raoulatory Reauirement: 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 item 81.10 of Examination Category B-A of the 1989 Editien 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.
Licensee's Proposed Altemative: Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), the licensee proposed that the coverages obtained for the subject welds be considered to satisfy the requirement. The licensee indicates that the examinations were performed to the extent practical for each RPV shell weld, as required by 10 CFR 50.55a(g)(6)(ii)(A). Weld GH was :
completely inaccessible for examination. The coverages achieved for all the RPV welds are I listed in Table A-1. !
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Enclosure 2
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Table A-1 i
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! weld nem coversee erwerverene.
1 CD (Cire) B1.30 90 % Guide rods D1 (Vert) B1.12 99% Steam dryer support bracket D2 (Vert) 81.12 96 % Steam dryer support bracket l DE (Chc) - B1.11 90% Guide rods E1 (Vert) . B1.12 51 % Feisdwater sparger, cors spray piping header, offset between upper and lower toolconfigurations l
i E2 (Vert) B1.12 53 % Foodwater sparger, core spray piping header, offset between upper and lower tool configurations El-(Circ) 81.11 69 % Guide rods, core spray piping down-comers, core shroud lifting lugs, sorveillance specimen holders and brackets F1 (Vert) B1.12 ' 65 % Jet pump riser brackets F2 (Vert) B1.12 65 % Jet pump riser brackets i
FG (Cire) B1.11 68 % Gu;de iods, core spray piping down-comers, core shro.:J liftirag lugs, surveillance specimen holders and brackets G1 (Vert) 81.12 4);% i;;9c inlet pipe, jet pump instrumentation L-bracket G2 (Vert) B1.12 45% Recire inlet pipe, jet pump instrumentation L-bracket GH (Cire) B1.11 0% Jet pemp instrumentation L-bracket All(Vert) 65 %
All(Cire) 63 %
The licensee stated:
Tacause of restricted clearance between the biological shield wall and the vessel i outside diameter, it was necessary to perform the required ultrasonic examination from the insida of the reactor vessel. Only two vendors provide this service. Vermont Yankee contracted ABB Combustion Engineering Nuclear Operations in conjunction with Southwest Research Institute to perform the ultrasonic examination. The examination was performed during the Fall,1996, refueling outage in September,1996.
" Vermont Yankee used Appendix Vill qualified ultrasonic techniques rather than the l conventional ultrasonic technique that is specified in Regulatory Guide 1.150 and ASME Section V, Article 4, referenced by ASME Section XI, IWA-2232(a). Appendix Vill is not invoked in the 1986 Edition of Section XI. The superior alternative examination which was demonstrated in accordance with Appendix Vill was approved for use by the Vermont Yankee Authorized Nuclear Inservice inspector under the provisions of ASME I
Section XI, IWA-2240.
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"The examination revealed only seven recordable ind: cations, only one of which exceeded the ASME Section XI, IWB 3511, acceptance criteria. That flaw was reported to the NRC in Reference (c)' and was ace 3pted in Safcty Evaluation Report Refrence (d).
"10 CFR 50.55a(g)(6) states that " essentially 100%" d cach vessel shell weld shall be examined. " Essentially 100%"is defined as "moy than 90% of the examinatiol volume of each weld." The Vermont Yankee reactor vessel cc,ntains five circumferental and i eight vertical shell welds within the scope of 10 Cl R 50.55a(g)(6). For threr of th6 horizontal welds and six of the vertical welds,90'"- coverage was not achieved. Table A-1 gives examination coverage achieved for each cf the reactor vessel sheli welds.
"Since there are no vesso' featses in tha unexamined regions that are significantly different from those in the examind regions, and since the interferences which limited examination coverage do not afiect flaw susceptibii:ty, Veimont Yankee has concluded ,
that the examination resu!ts achieved are representative of the weld characteristics in l the unexam' wo recions. Therefore, Vermont Ycnkoc has determined that the i examina?.bn coverage achieved constitutes an alternative v.hich provides an acceptable '
leva! 6 quality and safety per 10 CFR 50.55a(g)(6)(ii)(A)(5)."
Licensee's Basis for the Proposed Alternative (as stated):
"For all of the RPV shell welds, no significant addit;ona! cxamination coverage could be )
achieved without removing reactor vessel internats cr by cxtensivaly redesigning the j manipulator tool. It should be noted that Vermont Yankee and tho inspection vendor worked continuously to upgrade the tool to provide the maximum pssible coverage of the Vermont Yankee reactor vessel shell welds. It shou'd also be noted that the only ,
other tool performing RPV shell weld examinations from the insi0e of the vessel would !
not provide any sig"icant improvement in coverage. l "The amount of coverage achieved is directly affecu b; the proximity of the approach of the probe sleci dudng the scanning to voelintemalinteiferences. At the beginning l of the examinat'on, tnis offer tone was set at approximately one inch. Following an event where the :nanipulator tool collided with a specimen IWm Srait, the buffer zone was extended to three inches. The specimen holder b sckei was inspected and was verified to have suffered no damage, however, the marJpulator tool was extensively damaged and it required ten days to retrieve and ic pan w lower portion of the tool.
"Of special note is watd GH for which no examinction coverage was ach!?ved. Weld GH is the vessel she!' to bottom head weld, the centeriinc of which is located two inches above the inp of the jet pump baffle plate. ABB ucnsively modified their manipulator to access this we!d in preparation for the refueling outaje. However, when ABB attempted to fully extend the mast to reach this weld, the Jet pu r p instrumentation L-brackets would not allow the tool probe head to pass. There it. an extremely small envelope l
between those brackets and the vessel shell wat! The tool was pulled out of the vessel Figures, enclosures, and references furnished with the R mec's sobraittal are not included in this report.
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and an adjustment of shims at the top of the tool was performed to allow a better angle of approach through this envelope. This additional effort did not correct the interference situation. Rather than risk damaging reactor intemals or the inspection tool, it was 1 l decided to not make any further attempts to reach this weld. It should also be noted that the other vendor's inspection tool is also not capable of reaching this weld at l Vermont Yankee.
l "We also evaluated what additional ultrasonic coverage of the Vermont Yankee vessel i i could be obtained by supplementing the intemal tool scanning with manual examinations from the OD of the reactor vessel. As supported below, this would not be a significant amount.
"The only accessible areas to the reactor vessel OD are those areas above the biological shield wall and through the nozzle windows in the biological shield wall. There is only a twelve inch gap between the shield wall and the vessel OD. This gap is further obstructed by the vessel insulation, insulation brackets and clips, and thermocouple i
pads. The vesselinsulation is stacked from the bottom of the vessel to the top and ,
cannot be removed. In the following paragraphs, the vessel shell welds will be l l assessed individually for possible additional coverage.
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" Weld CD is above the top of the biological shield wall and could conceivably be I examined from the OD for the small percentage of weld that is not accessible behind the I guide rod from the intorior. Since 90% of this weld was already examined, it is !
concluded that the small additional amount of weld that could be examined would not provide significant benefit when weighed against the time and exposure necessary to complete this portion.
" Welds D1 and D2 are above the top of the biological shield wall for the portions of the welds that are in the vicinity of the steam dryer support lugs and could conceivable be examined for the small percentage of those welds that are inaccessible from the interior.
Since over 90% of these welds has already been examined, it is concluded that the small additional amount of weld that could be examined would not provide significant benefit when weighed against the time and exponure necessary to complete these portions.
l " Weld DE is below the top of the biological shield wall. There are no nozzle windows in the area of the guide rod interferences. Therefore, no additional examination could be performed on this weld. Also,90% of this weld was already examined.
i " Weld E1 and E2 are vertical welds at approximately 105 and 285, respectively. The l core spray nozzle window openings are centered on 90 and 270, but the edges of the windows only come to about 100 and 280 which are about 10 inches away from each of the mid m.7terlines. Therefore, no additional examination could be performed on these welds.
I l "There are no noa windows within the vicinity of the elevations of welds EF, F1 and l F2. Therefore, no additional examination could be performed on these welds.
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" Weld FG is a circumferential weld at elevation 286.5 feet. The recirc inlet nozzle openings are centered at elevation 282.4 feet, but the tops of the window openings only come to 283.9, which are about 30 inches away from the weld centerline. Therefore, no additional examination could be performed on this weld.
" Welds G1 and G2 are vertical welds at 75 and 225. Each is centered between jet pump pairs and, consequently, each are exactly centered between two recire inlet nozzles and their respective biological shield wall windows. The two jet pump instrumentation (N-8) nozzle windows are below and vertically aligned with two of the recire inlet nozzle windows and do not afford any access to welds G1 and G2 either.
Therefore, no additional examination could be performed on these welds.
Weld GH is a circumferential weld at elevation 275.5 feet. The jet pump instrumentation (N-8) nozzle windows are centered at elevation 277.7 feet and the bottoms of the windows reach to 275.6 feet, about 1.75 inches above the weld centerline. The bottoms of the two recire outlet nozzle windows only come to 276.8, still about 13.5 inches away from the weld centerline. An OD examination through the N-8 nozzle windows could conceivably examine a limited cross section of weld GH.
However, these nozzle v/ndow openings are very cluttered and, at best, the examination could on!y be performed m the length of the weld accessible through those two windows, about eight feel total (15% of the total length), and would achieve an estimated 50% of the weld cross-section at best la only one beam direction. An OD examination would be conducted in accordance with ASME Section V, Article 4 (this examination would not have been qualified in accordance with Appendix Vill).
Considering that four beam directions are required for scanning in accordance with ASME Section V, Article 4, the weld coverage would only be 50% divided by four, or 12% in cross-section for that eight feet, or less than 2% for the total length of the weld.
The jet pump instrumentation nozzles, and an ALARA estimate for manual ultrasonic examination at these windows, including insulation removal and restoration, protective framework removal and restoration and weld preparation is about 2 manRem. It is concluded that the small additional coverage that might be achieved would not provide significant benefit whten weighed against the time and exposure necessary to complete this small extra volume.
"In summary, the only welds where coverages could be marginally increased from supplemental OD examination would be welds CD, D1, D2, and GH. None of these I welds are in the vessel beltline where additional coverage might be more moaningful."
Evaluation: To comply with the augmented RPV examination requirements of 10 CFR 50.55a(g)(6)(ii)(A), licensees must volumetrically examine essentially 100% (i.e. >90%) of each of the item B1.10 shell welds. As an alternative to the regulations, the licensee proposed that the limited examinations that were performed be considered to satisfy the augmented reactor vessel examination requirement.
The ultrasonic axr,minations of the RPV at the Vermont Yankee Nuclear Power Station have to be performed from the inside of the vessel due to the biological shield wall. The total coverage i obtained was 63% for the circumferential welds and 65% for the longitudinal welds.
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Limitations associated with the circumferential weld examinations included: Guide Rods, l Surveillance Specimen Holders and Brackets, Core Spray Piping Down-Comers, Core Shroud Lifting Lugs, and Jet Pump Instrumentation L-bracket. Limitations associated with the longitudinal weld examinations included: Steam Dryer Support Brackets, Feedwater Sparger, )
Core Spray Piping Header, Jet Pump Riser Brackets, Jet Pump Instrumentation L-bracket, Offset Between Upper and Lower Tool Configurations, and the Recirc Inlet Pipe.
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 inside is inadequate, examination from the outside using manual inspection techniques is a potential option. However, at Vermont Yankee, the bioshield is only 12 inches from the uninsulated OD of the reactor vessel. Therefore, the close proximity of the bioshield severely limits access for equipment and personnel to the outside surface. Although limited access is available for four welds (above the bioshield or through nozzle windows), examination at these locations would only increase the total examination volume by an estimated 1%.
Therefore, it is concluded that the licensee cannot significantly enhance coverage by examining from the outside surface.
The licensee stated that only two vendors provide the service of ID volumetric examinations.
The licensee contracted ABB Combustion Engineering Nuclear Operations in conjunction with Southwest Research Institute to perform the augmented volumetric examinations. The licensee and the inspection vendor have worked continuously, including modifying the inspection tool, to increase examination coverage. The examination revealed seven recordable indications, one of which exceeded the ASME Section XI, IWB-3511, acceptance criteria. The flaw was reported to the NRC and was accepted in a Safety Evaluation Report dated October 11,1996. !
The licensee has examined a significant portion of the reactor pressure vessel shell welds (greater than 63%). Furthermore, it appears that the licensee, short of disassembly of the reactor pressure vessel intemal components, has worked to increase coverage to the maximum extent practical. Therefore, based on the cumulative volumetric examination coverage obtained, the INEEL staff concludes that any significant patterns of degradation, if present, would have been detected and that the examinations performed provide an acceptable level of quality and safety.
3.0 CONCLUSION
l The INEEL staff has reviewed the licensee's submittals and concludes that the licensee's !
proposed attemative provides an acceptable level of quality and safety. Therefore, it is l recommended that the licensee's proposed attemative to the regulations be authorized pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5).
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