ML20217D259
| ML20217D259 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/21/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217D239 | List: |
| References | |
| NUDOCS 9804240251 | |
| Download: ML20217D259 (5) | |
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11 UNITED STATES g
j NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D.C. 20066-0001
'g.....,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PLAN PROPOSED ALTERNATIVE TO REACTOR VESSEL AUGMENTED EXAMINATION SOUTHERN NUCLEAR OPERATING COMPANY. INC.
JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 1 DOCKET NO. 50-348
1.0 INTRODUCTION
The Technical Specifications for Joseph M. Farley Nuclear Plant (FNP), Unit 1, state that the inservice inspection and testing 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 Boiler and Pressure Vessel Code (ASME Code) and applicable adder,da as required by Title 10 of the Code of 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 attematives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed alternatives 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 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 Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of the ASME Code,Section XI, for FNP, Unit 1, during the second 10-year inservice inspection (ISI) interval, is the 1983 Edition including the Summer 1983 Addendum. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission approval.
Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination ; aquirement of Section XI of the ASME Code is not practical for its facility, l
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 9804240251 900421 PDR ADOCK 05000348 P
2 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 on the facility.
As stated in 10 CFR 50.55a(g)(6)(ii)(A)(1), the Commission revoked all previous reliefs granted to licensees for the extent of volumetric examinations of reactor vessel shell welds, ac specified in Section XI, Division 1 of the ASME Code. The Commission further required that alllicensees augment their reactor vessel examination by implementing once, as part of the ISI interval in effect on September 8,1992, the item B1.10 requirements (examine essentially 100 percent of '
the volume of each shell weld) of the 1989 Edition of the ASME Code.
Under 10 CFR 50.55a(g)(6)(ii)(A)(4), licensees may satisfy the augmented requirements by performing the ASME Code,Section XI, reactor vessel shell weld examinations scheduled for implementation during ISI intervals in effect on September 8,1992. As a result, the licensee is required to submit both an attemative to 10 CFR 50.55a(g)(6)(ii)(A) and a request for relief pursuant to 10 CFR 50.55a(g)(5)(iii), or a proposed alternative pursuant to 10 CFR 50.55a(a)(3) for the same welds when the licensee obtains less than the required coverage (essentially 100 percent) during the examinations.
Additionally, pursuant to 10 CFR 50.55a(g)(S)(ii)(A)(5), licensees that make a determination that they are unable to completely satisfy the requirements for the augmented reactor vessel shell weld examination specified in 10 CFR 50.55a(g)(6)(ii)(A) shall submit information to the Commission to support the determination and shall propose an attemative to the examination i
requirements dat would provide an acceptable level of quality and safety. Licensees may use the proposed alternative when authorized by the Director of the Office of Nuclear Reactor Regulation.
In a letter dated September 8,1997, Southem Nuclear Operating Company, Inc. (SNC),
submitted to the NRC its alternatives to th6 augmented examination of the reactor vessel shell weld 1100-8 conducted pursuant to 10 CFR 50.55a(g)(6)(ii)(A) for FNP, Unit 1, during the second 10-year interval. SNC's proposed attemative to the examination of " essentially 100 percent" of the subject weld in the reactor vessel is a best-effort examination resulting in j
limited examination coverage of the welds that provide an acceptable level of quality and safety.
The staff has reviewed and evaluated SNC's proposed attemative and the supporting information provided, pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5) for FNP, Unit 1.
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l l 2.0 DISCUSSION l
Component identification i
Code Class:
ASME Code,Section XI, Class 1
Reference:
Table IWB-2500-1 Examination Category:
B-A i
item Number:
B1.11
==
Description:==
Limited volumetric examination of Lower Shell-to-Bottom Head Circumferential Weld with 89.2 percent coverage.
Examination Reauirement Section 50.55a(g)(6)(ii)(A)(2) states that all licensees shall augment their reactor vessel examinations by imp ementing the examination requirements for reactor pressure vessel (RPV) shell welds specified in item B1.10 of Examination Category B-A, " Pressure Retaining Welds in Reactor Vessel,"in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of Section XI, Division I of the ASME Code, subject to the conditions specified in 50.55a(g)(6)(ii)(A)(3) and (4).
l For the purpose of this augmented examination, essentially 100 percent as used in Table IWB-l 2500-1 means more than 90 percent of the examination volume for each weld. Additionally,10 CFR 50.55a(g)(6)(ii)(A)(5) requires licensees that are unable to completely satisfy the augmented RPV shell weld examination requirement to submit information to the Commission to supportihe determination, and propose an alternative to the examination requirements that l
would provide an acceptable level of quality and safety.
SNC'S Attemative Examination (As stated)
The augmented examination of weld 1100-8 (the lower shell to bottom head circumferential weld) was limited based on physical configuration. Four core support lugs permanently attached to the inner surface of the vessel limit access to the weld, thereby, prohibiting greater than 90% examination. The weld received an 89.2% examination volume coverage.... Prior to the outage,' SNC and the NSSS vendor personnel originally estimated the examination coverage of weld 1100-8 to be 83%. As a result, SNC investigated options that could potentially maximize the examination coverage of this weld.
Ootion 1 Suoclemental Examination Of The Weld From The Outer Diameter (OD)
Of The Vessel. The evaluation of an OD examination showed that almost no increase in coverage would be obtained. Additionally, a significant burden would be involved with the radiation dose and the effort to correlate the inner and outer diameter coverage plots. Therefore, a significant burden would be placed on SNC without a compensating increase in safety.
I l Ootion 2 Sunnlemental Examination Of The Wald From The inner Diameter (ID)
Usina Multiole Scans. SNC estimated that coverage could be maximized to approximately 92% through the use of transducer repositioning and by performing j
additional scans beyond those required by the ASME Code.
It was decided that SNC would pursue Option 2; however, the fillet welds on the l
core support lugs were larger than expected, which resulted in the final coverage of l
89.2%.
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3.0 EVALUATION To comply with the augmented reactor vessel examination requirements of 10 CFR l
50.55a(g)(6)(ii)(A), licensees must volumetrically examine essentially 100 percent of each of the l
Item B1.10 shell welds. In accordance with the regulations, essentially 100 percent is defined I
as greater than 90 percent of the examination volume of each weld. As an attemative to the greater than 90 percent coverage requirement of the regulations, SNC proposes that the j
examination coverage obtained be considered to provide an acceptable level of quality and safety for the RPV welds.
At FNP, Unit 1, the volumetric coverage requirement cannot be met for weld 1100-8 due to physical restrictions that limit scanning of the weld. There are four core support lugs t
permanently welded in the lower shell of the vessel that limit scanning of the lower shell to the bottom head weld to 89.2 percent of the examination volume. All other welds in the intermediate shell and the upper shell received the Code examination coverage. To achieve complete coverage for the subject weld, design modifications would be required to increase access from the inside surface (lD).
1 As a result of the augmented volumetric examination rule, licensees must make a reasonable effort to maximize examination coverage of their reactor vessel. In cases where examination coverage from the ID is inadequate, examination f* om the outside surface (OD) using manual inspection techniques is a potential option. Hcwever, at FNP, Unit 1, the evaluation of the OD examination showed almost no increase in volumetric examination coverage. SNC has attempted to maximize coverage from the ID by optimizing transdacer arrangements for scanning close to obstructions, which resulted in obtaining an examination coverage of 89.2 percent. Therefore, it is concluded that SNC has made a reasonable effort to maximize examination coverage.
All welds in Examination Category B-A, item B1.11 except weld 1100-8, received 100 percent examination coverage. The examination did not reveal any unacceptable flaw. The staff has determined that if a pattem of degradation exists in the subject weld, a volumetric examination coverage of 89.2 percent would have detected it. Furthermore, the likelihood of an unacceptable flaw existing in the unexamined portion of the weld is extremely smati. Therefore, SNC's proposed attemative provides an acceptable level of quality and safety.
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4.0 CONCLUSION
The staff has reviewed SNC's submittal and concludes that SNC has maximized examination coverage for the reactor vessel weld 1100-8 and that service-induced degradation, if present, I
would have been detected. There were no unacceptable flaws found as a result of the examination. Thus, the staff concludes that SNC's proposed alternative provides an acceptable level of quality and safety. Therefcre, SNC's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for FNP, Unit 1.
Principal Contributor: P. Patnaik Date:
April 21, 1998 l
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