ML18039A594
| ML18039A594 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/29/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML18039A593 | List: |
| References | |
| NUDOCS 9811040158 | |
| Download: ML18039A594 (14) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&4001 S
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0 The Technical Specifications (TS) for the Browns Ferry Nuclear Plant, Unit 2, state that the 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 Boiler and Pressure Vessel (B8PV) Code and applicable addenda as required by 10 CFR 50.55a(g),
except where specific written relief has been granted by the U.S. Nuclear Regulatory Commission (NRC or Commission) pursuant to 10 CFR 50.55a(g)(6)(i).
10 CFR 50.55a(a)(3) states that alternatives 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 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 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 Browns Ferry Nuclear Plant, Unit 2 second 10-year inservice inspection (ISI) interval is the 1986 Edition.
Pursuant to 10 CFR 50.55a(g)(5), ifthe 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
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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 alternative requirements that are determined to bejauthorized by law, will not endanger life, property, or the common defense and security, anB are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result ifthe requirements were imposed.
In letters dated November 12, 1996 and June 2, 1997, the Tennessee Valley Authority (licensee or TVA), submitted to the NRC its Second 10-Year Inservice Inspection Interval Program Plan, Revisions 7 and 8 respectively.
The licensee provided additional information in its letter dated July 7, 1997.
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The NRC staff, with technical assistance from its contractor, the Idaho National Engineering and Environmental Laboratory, has evaluated the information provided by the licensee in support of its Second 10-Year Inservice Inspection Interval Program Plan, Revisions 7 and 8, for the Browns Ferry Nuclear Plant, Unit 2. Based on the information submitted, the staff adopts the contractor's conclusions and recommendations presented in the Technical Letter Report attached.
The applicable edition of Section XI of the ASME Code for the Browns Ferry Nuclear Plant, Unit 2, second 10-year ISI interval is the 1986 Edition of the ASME Boiler and Pressure Vessel Code,Section XI. The changes and additions to the second 10-year Program, including information provided by the licensee in support of the requests for relief, have been evaluated and are documented below.
Revision 7 - incorporated organizational changes, various editorial corrections, changes in Section 8.1 to reflect the current examination schedule, and revised the coverage percentage in Relief Request ISI-2-4 (see section 2.4 below).
The above changes are editorial in nature and do not change the technical content of the Program.
Based on review of these revisions, no deviations from regulatory requirements or commitments have been identified.
Revision 8-changes the reporting of discrepancies from the "Inspection Report" to the "Notification of Indication" report. The licensee stated that disposition of indications willremain in accordance with the Browns Ferry Nuclear Plant site corrective action program.
Revision 8 also added a reference for the augmented enhanced visual examination of the Core Spray System piping creviced weld regions in accordance with the recommendations of General Electric Service Information Letter SIL-289. The creviced areas of the core spray piping welds inside the reactor pressure vessel shall be visually examined utilizing enhanced methods and/or ultrasonically examined in accordance with the recommendations of SIL-289.
The spargers require visual examination each refueling outage.
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The above revisions do not change the technical content of the Program.
Based on review of these revisions, no deviations from regulatory requirements or commitments have been identified.
Based on the review of the above documents the staff requested TVA(in a request for additional information dated May 6, 1997) to address their selection of substitute components when sufficient inspection coverage cannot be achieved for scheduled components.
In
- response, the licensee stated that "TVAmaintains the latitude to avert requests for relief during the interval by selecting welds that can be fullyexamined." The Code does not allow selection of an alternative examination area when the Code-required coverage of the scheduled examination areas cannot be obtained.
Substituting welds must be authorized by the NRC through the relief request process.
While justification for not obtaining complete coverage may be evident (on a case-by-case basis), a request for relief specifically addressing affected components should be submitted to allow for staff evaluation.
Request for Relief ISI-2-4 (revised): ASME Code',Section XI, Examination Categories B-F and B-J, Class 1 Piping Welds Request for Relief ISI-2-4 was previously evaluated and granted pursuant to 10 CFR 50.55a(g)(6)(i) in an NRC Safety Evaluation (SE) dated November 29, 1995. The only change made in the revised ISI-2-4 was to increase the examination coverage for residual heat removal (RHR) Weld DRHR-2-03 from 52% to 65%. This change provides for better coverage than previously documented.
Given that the requirements of the code are still impractical, the relief granted in the SE dated November 29, 1995, remains granted pursuant to 10 CFR 50.55a(g)(6)(i).
'I Request for Relief ISI-2-7: ASME Code,Section XI, Examination Category B-K-1, Item
. 810.10, Class 1 Integrally Welded Attachments In a letter dated June 4, 1997, TVArevised "Request for Relief ISI-2-7", and resubmitted it as "permanent Request for Relief 2-ISI-7 (revised)" 2-ISI-7 (previously identified as ISI-2-7) was revised to include Code Class 2 integrally welded attachments as well as Code Class 1
integrally welded attachments and identify the request for relief as being permanent.
The revised 2-ISI-7 has been previously evaluated by the NRC staff, and the results were documented in a letter dated October 16, 1997.
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Ls The staff has reviewed the licensee's submittals and has identified no deviations from regulatory requirements or commitments for the changes in Revisions 7 and 8 to the Brogans Ferry Nuclear Plant, Surveillance Instruction 2-SI-4.6.G, Inservice Inspection Program, Unit 2, for the second 10-year interval. However, the licensee's position regarding avoidance of requests for relief by selecting substitute components when the Code-required examination coverage cannot be met for scheduled items is not consistent with Code requirements and not acceptable to the staff.
The staff concludes that, in these instances, the licensee is required to submit requests for relief for evaluation (see Section 2.3 of this report). The Code does not allow for selection of an alternative examination area when the Code-required coverage of the selected examination areas cannot be obtained.
4 The staff concluded that for revised Request for Relief ISI-2-4, the only change made by the licensee was to increase the examination coverage for Weld DRHR-2-03 from 52% to 65%.
Therefore, relief remains granted, as previously reported in the NRC Safety Evaluation dated November 29, 1995, pursuant to 10 CFR 50.55a(g)(6)(i). The relief is authorized by law, will not endanger life, property, or the common defense and security, and is other wise in the public interest, giving due consideration to the burden upon the licensee that could result ifthe requirements were imposed.
Request. for Relief 2-ISI-7 (revised), as received in the June 4, 1997 submittal, was previously evaluated in a separate NRC SE dated October 16, 1997.
Principal Contributor: Tom McLellan, NRR Dated: October 29, 1998
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E'2 1.0 By letter dated November 12, 1996, the licensee, Tennessee Valley Authority (TVA), submitted Browns Ferry Nuclear Plant, Surveillance Instruction 2-Sl-4.6. G, Inservice Inspection Program, Unit 2, Revision 7, for the second 10-year interval.
Included in Revision 7 are incorporation of organizational changes, minor editorial corrections, one revised relief request and one new relief request.
A response to a May 6, 1997, Nuclear Regulatory Commission (NRC) request for additional information, regarding Revision 7 to the program plan, was submitted by letter dated July 7, 1997, By letter dated June 2, 1997, TVAsubmitted Browns Ferry Nuclear Plant, Surveillance Instruction 2-S/-4.6. G, lnservice Inspection Program, Unit 2, Revision 8, for the second 10-year interval. This revision changed reporting of discrepancies from the "Inspection Report" to the "Notification of Indication" report and incorporated General Electric Service Information Letter Number 289.
The Idaho National Engineering and Environmental Laboratory (INEEL) staff has evaluated the information provided by the licensee in the following section.
2.0 The applicable edition of Section XI of the American.Society of Mechanical Engineers (ASME)
Code for the Browns Ferry Nuclear Plant, Unit 2, second 10-year inservice inspection (ISI) interval is the 1986 Edition of the ASME Boiler and Pressure Vessel, Code,Section XI. The changes and additions to'the second 10-year Program, including information provided by the licensee in support of the requests for relief, hav,. been evaluated and are documented below.
2.1 Revision 7 - incorporated organizational changes, various editorial corrections,
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The above changes are editorial in nature and do not change the technical content of the Program.
Based on review of these revisions, no deviations from regulatory requirements or commitments have been identified.
2.2 Revision 8 - changes the reporting of discrepancies from the "Inspection Report" to the "Notification of Indication" report.
The licensee stated that disposition of indications will remain in accordance with the Browns Ferry Nuclear Plant site corrective action program.
Revision 8 also added a reference for the augmented enhanced visual examination of the Core, Spray System piping creviced weld regions in accordance with the recommendations of General Electric Service Information Letter SIL-289. The creviced areas of the core spray piping welds inside the RPV shall be visually examined utilizing enhanced methods and/or ultrasonically examined in accordance with the recommendations of SIL-289.
The spargers require visual examination each refueling outage.
The above changes do not change the technical content of the Program.
Based on review N
of these revisions, no deviations from regulatory requirements or commitments have been identified.
2.3 Based on the review of the above documents the staff requested TVA (in a request for additional information dated May 6, 1997) to address their selection of substitute components when sufficient inspection coverage cannot be achieved for scheduled components.
In response, the licensee stated that: "TVAmaintains the latitude to avert requests for relief during the interval by selecting welds that can be fully examined."
The Code does not allow selection of an alternative examination area when the Code-required coverage of the scheduled examination areas cannot be obtained.
Substituting welds must be authorized by the NRC through the relief request process.
While justification for not obtaining complete coverage may be evident (on a case-by-case basis),
a request for relief specifically addressing affected components should be submitted to allow for staff evaluation.
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NOTE: Request for Relief ISI-2-4 was previously evaluated and granted in an NRC Safety Evaluation Report (SER) dated November 29, 1995.
The only change made in the revised ISI-2-4 was to increase the examination coverage for residual heat removal (RHR) Weld DRHR-2-03 from 52% to 65%.
This change provides for better coverage than previously documented.
Therefore, it is recommended that, as previously reported in the SER dated November 29, 1995, relief remain granted pursuant to 10 CFR 50.55a(g)(6)(i).
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NOTE:
In a letter dated June 4, 1997, Tennessee Valley Authority revised "Request for Relief ISI-2-7", and resubmitted it as "permanent Request for Relief 2-ISI-7 (revised)".
2-ISI-7 (previously identified as ISI'-2-7) was revised to include Code Class 2 integrally welded attachments as well as Code Class 1
integrally welded attachments and identify the request for relief as being permanent.
Also included in the June 4, 1997, submittal were four (4) new "permanent requests for relief". Therefore, the revised 2-ISI-7 is currently being evaluated by the NRC staff along with the four (4) new requests for relief received June 4, 1997, and the results will be in a future report.
3.0 The INEEL staff has reviewed the licensee's submittals and has identified no deviations from regulatory requirements or commitments for the changes in Revisions 7 and 8 to the Browns Ferry Nuclear Plant, Surveillance Instruction 2-SI-4.G. G, Inservice Inspection Program, Unit 2, for the second 10-year interval. However, the licensee's position regarding avoidance of requests for relief by selecting substitute components when the Code-required examination coverage cannot be met for scheduled items is not consistent with Code requirements.
The INEEL believes that, in these instances, the licensee should be required to submit requests for
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relief for staff evaluation (see Section 2.3 of this report).
The Code does not allow for selection of an alternative examination area when the Code-required coverage of the selected examination areas cannot be obtained.
The INEEL staff concluded that for revis'ed Request for Relief ISI-2-4, the only change made by the licensee was to increase the examination coverage for Weld DRHR-2-03 from 52% to 65%.
Therefore, it is recommended that relief remain granted, as previously reported in the SER dated November 29, 1995, pursuant to 10 CFR 50.55a(g)(6)(i).
Request for Relief,2-ISI-7 (revised), as received in the June 4, 1997 submittal along with four new requests for relief, is currently being evaluated along with the four new requests for relief and the results will be reported in a later document.
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