ML20204B186
| ML20204B186 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/15/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20204B175 | List: |
| References | |
| NUDOCS 9903220020 | |
| Download: ML20204B186 (9) | |
Text
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g j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
.ON THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION l
PROPOSED ALTERNATIVE TO AUGMENTED EXAMINATION '
l FDB l ARKANSAS NUCLEAR ONE. UNIT 1 ENTERGY OPERATIONS INC i DOCKET NO. 50-313
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1.0 INTRODUCTION
The Technical Specifications for Arkansa,s Nuclear One, Unit 1 (ANO 1), state that the inservice inspection of the Amer.,an Society of Mechanical Engineers (ASME) Code Class 1,2 and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code (ASME Code) and applicable addenda as required by Title 10 of the Code of the i Federal Reaulations (10 CFR), Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Section 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the Nuclear Regulatory Commission (NRC), if (i) the proposed alternatives would provide an i acceptable level of quality and safety or (ii) compliance with the specified requirements would I result in hardship or unusual difficulty without a compensating increase in the level of quality j 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 decign and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, ' Rules for 1 inservice Inspection of Nuclear Power Plant Components," to the extent practical within the 1 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 ASME Code,Section XI, for the ANO-1, second 10-year inservice inspection (ISI) interval is the 1980 Edition through Winter 1981 Addendum. The components (including supports) may meet the requirements set forth in Enclosure 9903220020 990315 PDR ADOCK 05000313 P PDR i
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l subsequent editions and addenda of the ASME Code incorporated by reference in l 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to i Commission approval. j 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, i information shall be submitted to the Commission in support of that determination and a request j made for relief from the ASME Code requirement. After evaluation of the determination, i pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose i 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 upon the facility. i r
Additionally, pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), licensees that make a determination that f they are unable to completely satisfy the requirements for tes augmented reactor vessel shell i weld examination specified in 10 CFR 50.55a(g)(6)(ii)(A) shad submit information to the !
Commission to support the determination and shall propose ar, alternative to the examination l requirements that would provide an acceptable level of quality and safety.
s By letter dated October 23,1997. Entergy Oparations Inc., submitted to the NRC its attemadves ;
to the augmented examination of the reactor v* ssel shell welds01-002,01-005,01-009, and l 01-010 conducted pursuant to 10 CFR 50.55a(g)(6)(ii)(A) for ANO-1 during the second 10-year !
Interval. The licensee's proposed attemative to volumetric examination of " essentially 100%" of !
the subject welds in the reactor vessel is a best-effort examination resulting in limited !
examination coverage of the welds that provides an acceptable level of quality and safety. The j staff has reviewed and evaluated the licensee's proposed attemative and the supporting -
information, pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5) for ANO 1. j 2.0 EVALUATION i The staff, with technical assistance from its contractor, the Idaho National Engineering and l Environmental Laboratory, has evaluated the information provided by the licensee in support of '
its augmented examination of the reactor vessel shell welds performed during the second 10-year inservice inspection interval. Based on the results of the review, the staff has taken ;
exceptions to the contractor's conclusion, presented in the attached Technical Letter Report, that insufficient information was provided to support this request.
l Pursuant to 10 CFR 50.55a(g)(6)(ii)(A), all licensees must implement once, as part of the l inservice inspection interval in effect on September 8,1992, an augmented volumetric ;
examination of the reactor pressure vessel (RPV) welds specified in item B1.10 of Examination l Category B-A of the 1989 Edition of the ASME Code,Section XI. Examination Category B-A, ltems B1.11 and B1.12 require volumetric examination of essentially 100 percent of the RPV circumferential and longitudinal shell welds, as defined by Figures IWB-2500-1 and -2,
- respectively. Essentially 100 percent, as defined in 10 CFR 50.55a(g)(6)(ii)(A)(2), is greater than 90 percent of the examination volume of each weld. During the second 10-year inspection interval of ANO-1, the licensee performed the examination to the maximum extent possible for
l ear h RPV shell weld, but was unable to achieve essentially 100 percent coverage of the examination volume for the following RPV shell welds.
Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), the licensee has proposed an attemative to the examination coverage requirement for the augmented RPV examination of the welds listed in the following table.
l Code ISI EXAM COVERAGE NO. EXAM AREA DESCRIPTION ACHIEVED LIMITATIONS01-002 Shell-to-shell circumferential 74.60 % Limitations due to seam intersecting inlet and outlet nozzles01-005 Lower shell-to-transition 32.9;% Limitations evenly spaced piece circumferential weld around this horizontal weld due to 12 core guide lugs01-009 Lower shell-to-long seam 88.09 % Limitations at the bottom of (at Z-W) this vertical weld due to core guide lugs01-010 Lower shell-to-long seam 88.09 % Limitations at the bottom of (at X-Y) this vertical weld due to core guide lugs Three of the above welds were volumetrically examined between 74.6 to 88.09 percent due to constraints, which impose limitations to scanning. The licenses has made a reasonable effort to maximize examination coverage using state-of-the, trt equipment. The staff has previously reviewed comparable examination coverage for other ieactor vessels with the same type of constraints. However, weld 01-005 was volumetrically examined to 32.99 percent due to masking of this circumferential weld by 12 core guide lugs that are welded around the vessel periphery. The Ltaff has determined that no upgrade of tooling for the auto:nated ultrasonic equipment would change the volumetric coverage appreciably. Nevertheless, a manual ult:asonic examination from the vessel exterior will increase the volumetric coverage, with high radiation penalty. The staff, therefore, has evaluated the licensee's proposed attemative to augmented examination of the subject welds on the basis of the following: j The licensee has performed a best-effort examination of the subject welds using state-of-the art equipment commercially avaliable.
For welds01-002,01-009 and 01-010, the volumetric coverages between 74.6 and 88.09 percent provide reasonable assurance of structural integrity, since any pattom of degradation,if present, would have been detected. Therefore, the extent of the licensee's examination is acceptable.
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For weld 01-005, which was volumetrically examined to 32.99 percent, the access is i limited for examination from inside the vessel. The peservice inspection of 100 percent of weld volume did not reveal any flaw in this weld. Moreover, the weld is located ,
beyond the vessel beltline and, therefore, is not subject to irradiation ernbrittlement. t Hence, the probability of developing a flaw, which would eventually grow to cause a .
breach of pressure boundary, is extremely small. Therefore, the licensee's examination !
is acceptable as performed.
3.0 CONCLUSION
The staff has reviewed the licensee's submittal pursuant to the provisions of 10 CFR 50.55a(g)(6)(ii)(A)(5) for ANO-1 and concludes that the licensee has maximized the examination coverage for the reactor vessel shell welds01-002,01-005,01-009, and 01-010 ;
and that any service-induced degradation, if present, would have been detected. There were !
no unacceptable flaws found in the welds during the preservice inspection and the subsequent examinations conducted during the successive inspection intervals which, therefore, provides a reasonable assurance of structuralintegrity. The staff concludes that the extent of the licensee's preservice and the current best-effort examination would provide an acceptable level :
of quality and safety and, therefore, is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for ANO-J.
Attachment:
Technical Letter Report Principal Contributor: P. Patnaik Date: March 15, 1999 i l
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TECHNICAL LETTER REPORT ON THE ALTERNATIVE TO 10 CFR 50.55afo)(6)(ii)(A)
AUGMENTED REACTOR PRESSURE VESSEL EXAMINATION EQB ENTERGY OPERATIONS. INC.
ARKANSAS NUCLEAR ONE. UNIT 1 DOCKET NO. 50-313
1.0 INTRODUCTION
By letter dated October 23,1997, the licensee, Entergy Operations, Inc. (EOl), proposed an attemative to the augmented examination of the reactor pressure vessel (RPV) required by 10
. CFR 50.55a(g)(6)(ii)(A) for Arkansas Nuclear One, Unit 1 (ANO-1). The Idaho National Engineering and Environmental Laboratory (INEEL) staff's evaluation of the subject request for relief is in the following section.
2.0 EVALUATION The licensee performed the augmented RPV weld examinations as part of its second 10-year ISl interval examinations. The Code of record for the ANO-1 second 10-year ISIinterval, which ended June 1,1997, is the 1980 Edition through Winter 1981 Addenda of Section XI of the j ASME Boiler and Pressure Vessel Code. The information provided by Entergy Operations, Inc.
in support of the proposed alternative has been evaluated and the basis for disposition is documented below.
Attemative to 10 CFR 50.55afo)(6)(ii)(A). Auamented Reactor Pressure Vessel Examination i Reaulatorv Reauirement: In accordance with 10 CFR 50.55a(g)(6)(ii)(A), all licensees must implement once, as part of the inservice inspection interval in effect on j ATTACHMENT l
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2-September 8,1992, an augmented examination of the reactor pressure vessel 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 l 10 CFR 50.55a(g)(6)(ii)(A)(2), is greater than 90% of the examination volume of each weld.
Licensee's Proposed Alternative: Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), the licensee !
proposed that the coverages obtained for the subject welds be found acceptable. The I
licensee performed the examination to the maximum extent possible for each RPV shell '
weld, but was unable to achieve essentially 100% coverage of the examination volume for following RPV shell welds.
CODE ISIEXAM COVERAGE NO. EXAM AREA DESCRIPTION ACHIEVED LIMITATIONS01-002 Shell-to-shell circumferential 74.60 % Limitations due to seam intersecting inlet and outlet nozzles01-005 Lower shell-to-transition 32.99 % Limitations evenly spaced .
piece circumferential weld around this horizontal ws!d j due to 12 core guide lugs 1 1
01-009 Lower shell-to-long seam 88.09 % Limitations at the bottom of )
(at Z-W) this vertical weld due to core guide lugs01-010 Lower shell-to-long seam 88.09 % Limitations at the bottom of !
(at X-Y) this vertical weld due to core guide lugs The licensee stated:
"In addition to the ultrasonic examinations performed during 1R12, the reactor vessel interior surfaces and interior welded attachments also received visual (VT-1 and VT-3) inspections as required by ASME Code Section XI. A visual (VT-2) examination was performed on the exterior of the reactor vesselin 1R12 consistent with the visualinspections performed during the previous 11 refueling outages. No
r service-induced cracking or degradation has been found either with the ultrasonic examinations or with the visual inspections."
Licensee's Basis for Proposed Altemative (as stated):
"The ANO-1 reactor vessel contains eight welds that are categorized as either B1.11 or B1.12. Four of the eight welds received a complete examination.
"ANO-1 utilized automated, state-of-the-art, underwater, contact ultrasonic examination equipment to perform the examinations from the inside surface of the reactor vessel. The results of the examinations revealed that no service-induced flaws were found in any of the eight reactor vessel shell welds. Using this same sensitive equipment, no service-induced flaws were found in any of the nozzle or piping welds, including the nozzle to-vessel, nozzle-to-safe-end, nozzle-to-pipe, and safe-end-to-pipe welds.
"The likelihood of a significant flaw existing in any of these welds is very small.
When the vessel was originally fabricated, these full-penetration welds were l radiographed and found to be acceptable. Since that time, these welds have been l inspected ultrasonically once during the preservice inspection and once during the i
first interval. The preservice ultrasonic inspection achieved 100% coverage since the welds were then accessible from the outside of the vessel. All of the welds were found acceptable for service in this inspection. The first interval inservice inspection was performed on the inside of the vessel under the 1974 Edition of the Code (through the summer 1975 Addenda). Only one of the examinations of the four subject welds (01-005) failed to meet the 1974 Edition examination requirements (10% of the length of anylongitudinal weld,5% of the length of any circumferential weld, and 50% of the length of any weld exposed to a neutron flux in excess of 1 E19 nyt). Relief for the limited examination was granted by correspondence dated May 24,1990 (1CNA059002). The results of these inservice examinations concluded that the welds were satisfactory for continued service.
"In addition to the ultrasonic examinations required by the ASME Code Section XI and by 10CFR50, a suppleme".tal ultrasonic examination was performed on the eight welds in accordance with USNRC Regulatory Guide 1.150, Revision 1,
' Ultrasonic Testing of Reactor Vessel Welds During Preservice and inservice Examinations.' This supplemental examination was performed with special transducers which are designed and calibrated to detect flaws within the first inch of base material under the cladding. This special 'near surface' ultrasonic examination also found no service-induced flaws.-
"Since complete examination of reactor vessel shell welds is not practical, ANO-1 has examined the welds to the maximum extent possible using the technologies that are commercially available. Examinations of the accewible weld volume is sufficient to provide reasonable assurance of vessel integrity, especially since all past examinations of accessible shell welds have revealed no service-induced flaws. It is therefore reasonable to conclude that the same results would be
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obtained for the inaccessible portions of the welds if it were possible to inspect :
them. !
" Inspection of less that 100 percent of the weld volumes does not endanger the :
public health and safety since the reactor coolant system is designed and t constructed to have a low probability of gross rupture or significant leakage throughout its design life. In addition, any leakage that rnight occur would be ,
readily detected and contained within the reactor building.'
Ey.gluation: To comply with the augmented RPV examination requirements of ;
10 CFR 50.55a(g)(6)(ii)(A), licensees must volumetricelly examine essentially 100% of ,
i each of ths item B1.10 shell welds. Essentially 100% is defined as greater than 90% of l the examination volume of each weld, where the reduction in coverage is due to interference by another component or to part geometry. As an attemative to the .
I regulations, the licensee proposed that the examinations performed be deemed to satisfy the augmented reactor vessel examination requirement, j I r At ANO-1, the RPV enmination was performed from the vessel inside surface using j automated underwater ultrasonic testing techniques. The licensee was able to obtain '
essentially 100% examination coverage for four of the eight RPV B1.10 shell welds. The licensee obtained between approximately 33% and 88% coverage for the four subject RPV shell welds; the examinations were limited by intersecting nozzles and core guide lugs. j As a result of the augmented volumetric examination rule, licensees must make a reasonable effort to maximize the examination coverage of their reactor vessels.
Methods used to maximize RPV shell weld examination coverage may include performing a portion of the examination from outside the vessel and/or modifying the tooling to gain additional access to the welds. The licensee has not described efforts made at ANO-1 to maximize examination coverage, or provided information supporting the reason or decision not to maximize coverage. Weld 01-005, in particular, received only 32.99%
examination coverage with no explanation or description of efforts made to maximize coverage.
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3.0 CONCLUSION
Based on the lack of supporting information relating to efforts made by the licensee to maximize RPV examination coverage, it is recommended that the alternative to the 10 CFR 50.55a(g)(6)(ii)(A) augmented RPV examination not be authorized.
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