NRC 2013-0073, 10 CFR 50.55a Request. Relief Request RR-4L3 Lnservice Inspection Impracticality Examination Limitations Due to Configuration Fourth Ten-Year Lnservice Inspection Program Interval Response to Request for Additional Information

10 CFR 50.55a Request. Relief Request RR-4L3 Lnservice Inspection Impracticality Examination Limitations Due to Configuration Fourth Ten-Year Lnservice Inspection Program Interval Response to Request for Additional Information

ML13241A201
Person / Time
Site:	Point Beach
Issue date:	08/28/2013
From:	Meyer L Point Beach
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NRC 2013-0073, TAC MF1144, TAC MF1145
Download: ML13241A201 (4)

Text

NEXTera, ENERGY~

POINT BEACH August28,2013 NRC 2013-0073 10 CFR 54 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 10 CFR 50.55a Request. Relief Request RR-4L3 lnservice Inspection Impracticality Examination Limitations Due to Configuration Fourth Ten-Year lnservice Inspection Program Interval Response to Request for Additional Information

References:

(1) NextEra Energy Point Beach, LLC letter to NRC, dated March 19, 2013, 10 CFR 50 .55a Request, Relief Request RR-4L3 lnservice Inspection Impracticality Examination Limitations Due to Configuration Fourth Ten-Year lnservice Inspection Program Interval (ML13079A144)

(2) NRC electronic mail to NextEra Energy Point Beach, LLC, dated July 1, 2013, Point Beach Nuclear Plant Units 1 and 2- Draft RAI Regarding Relief Request RR-4L3 (TAC Nos. MF1144 and MF1145)

NextEra Energy Point Beach, LLC (NextEra) requested in Reference (1) that the Nuclear Regulatory Commission (NRC) grant relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code),Section XI, 1998 Edition through 2000 Addenda requirement for 100 percent coverage of the subject weld(s) due to geometric or design configuration, which limited the examination coverage which could be obtained. Relief was requested on the basis that alternative methods will provide an acceptable level of quality and safety.

Via Reference (2), the NRC determined additional information was required to enable the staff's continued review of the Relief Request RR-4L3. The Enclosure to this letter contains the response to the request for additional information in Reference (2).

NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241

Document Control Desk Page 2 This letter contains no new commitments and no changes to existing commitments.

Very truly yours, NextEra Energy Point Beach, LLC Enclosure cc: Administrator, Reg ion Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC Mr. Mike Verhagan, Department of Commerce, State of Wisconsin

ENCLOSURE 1 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 10 CFR 50.55a REQUEST, RELIEF REQUEST RR-4L3 INSERVICE INSPECTION IMPRACTICALITY EXAMINATION LIMITATIONS DUE TO CONFIGURATION FOURTH TEN-YEAR INSERVICE INSPECTION PROGRAM INTERVAL RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION The NRC staff determined that additional information was required (Reference 1) to enable the continued review of the Relief Request RR-4L3 (Reference 2). The following information is provided by NextEra Energy Point Beach, LLC (NextEra) in response to the NRC staff's request.

By letter dated March 19, 2013 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML13079A144), NextEra Energy Point Beach, LLC (the licensee) submitted relief request RR-4L3 covering two welds from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components for Point Beach Nuclear Power Plant, Units 1 and 2. The request for relief applies to the fourth 10-year inservice inspection (IS/) interval.

In accordance with Title 10 of the Federal Code of Regulations (10 CFR) 50.55a(g)(5)(iii), the licensee submitted the subject requests for relief for limited examinations. The ASME Code requires that 100 percent of the examination volumes, or surface areas, described in Tables /WB-2500 and /WC-2500 be performed during each interval. The licensee stated that 100 percent of the ASME Code-required volumes are impractical to obtain for the welds.

The NRC staff has reviewed and evaluated the information provided by the licensee and has determined that the following information is needed in order to complete its review of the relief request.

1. While PDI-UT-10 does not allow for coverage credit to be taken beyond the weld centerline, "best effort" coverage is often calculated and is useful in evaluating inspections where essentially 100 percent code coverage cannot be obtained.

Please discuss if the "best effort" coverage was calculated and, if so, what is the estimated "best effort" coverage for each weld?

Response to RAI 1 As noted on Table 4L3-1, the maximum amount of weld volume and base material on the far side of the weld was insonified utilizing guidance contained within NDE-173 (PDI -UT-2) for single-sided examinations. Due to the configuration of both welds and the inability to perform additional surface preparation on the valve side of the weld, the maximum area insonified using the recommended techniques are as follow:

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AC-1 0-SI -1001 -19: 80% coverage (circumferential flaw detection), 50% coverage (axial flaw detection).

AC-1 0-SI-2001-17: 56.25% coverage (circumferential flaw detection), 50% coverage (axial flaw detection).

2. The pipes, valves, and welds are all described as being austenitic steels.

Please discuss if any of the pipes or valves are composed of cast stainless steel?

Response to RAI 2 The piping is A376 TP 316 wrought austenitic stainless steel and the valves are ASTM A-351 Gr.

CF 8 cast austenitic stainless steel.

3. Both welds are described as being susceptible to stress corrosion cracking. As stress corrosion cracking initiation and propagation are strongly influenced by temperature, an important factor in evaluating welds that are subject to stress corrosion cracking is the operating temperature of the weld.

Please provide the operating temperatures for each of the welds?

Response to RAI 3 The system temperature I pressure for these welds are considered to be 543°F at 2250 psig .

References (1) NRC electronic mail to NextEra Energy Point Beach, LLC, dated July 1, 2013, Point Beach Nuclear Plant Units 1 and 2 - Draft RAI Regarding Relief Request RR-4L3 (TAC Nos. MF1144 and MF1145)

(2) NextEra Energy Point Beach, LLC letter to NRC, dated March 19, 2013, 10 CFR 50.55a Request, Relief Request RR-4L3 lnservice Inspection Impracticality Examination Limitations Due to Configuration Fourth Ten-Year lnservice Inspection Program Interval (ML13079A141)

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