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Category:Memoranda
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2022-06-06
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Text
November 16, 2011 MEMORANDUM TO: FILE FROM: Mohan C. Thadani, Senior Project Manager /RA/
Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
SUBJECT:
CALLAWAY PLANT, UNIT 1 -
SUMMARY
OF TELEPHONE CONFERENCE RE: VERBAL AUTHORIZATION FOR RELIEF REQUEST I3R-13 (TAC NO. ME7504)
This memorandum summarizes the telephone discussion on November 1, 2011, between the U.S. Nuclear Regulatory Commission (NRC) staff and Union Electric Company (UE), the licensee for Callaway Plant, Unit 1, related to the licensee=s request for relief I3R-13 for Callaway Plant, Unit 1. Participants in the discussion included licensee representatives Scott Maglio and Thomas Elwood, and NRC representatives John Tsao, Timothy Lupold, Michael Markley, Mohan Thadani, and David Dumbacher.
By letter dated October 31, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML113050122), the licensee requested relief from paragraph IWB-2420(b) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, at Callaway Plant Unit 1. Relief Request I3R-13 proposes to defer the second successive examination from the current refueling outage (RF-18), in the fall 2011 to RF-19 in spring 2013, and the third successive examination be rescheduled to the first period of the fourth 10-year inservice inspection (ISI) interval.
The licensee detected an indication in the safe-end to elbow weld (Weld 2-BB-01-F302) of the
'C' cold leg nozzle at the reactor pressure vessel during RF-13 (spring 2004). The indication exceeded the acceptance standards of the ASME Code,Section XI, IWB-3514 and the licensee performed a flaw evaluation in accordance with the ASME Code,Section XI, IWB-3640 as shown in Westinghouse Electric Company, LLCs report WCAP-16280-P, "Flaw Evaluation Handbook for Callaway Unit 1 Reactor Vessel Inlet Nozzle Safe-end Weld Region," May 2004 (proprietary), which was submitted in letter dated December 13, 2004 (ADAMS Accession No. ML043650441).
The licensees flaw evaluation indicated that leaving the indication in service was acceptable. In accordance with the ASME Code,Section XI, IWB-2420(b), the licensee needs to examine the indication during each of the three subsequent, successive inspection periods. The licensee completed the first successive examination in 2007 during RF-15, and confirmed no growth in the flaw size. The second successive examination should be performed during the ongoing refueling outage (RF-I8) in fall 2011. In lieu of performing the second successive examination during the RF-18, the licensee requested to defer the second successive examination to the
next refueling outage (RF-19) in 2013. The licensees basis for the relief request is the flaw evaluation in WCAP-16280-P and the results of ultrasonic examinations performed in 2004 and 2007.
The NRC staff concludes that:
- 1. The licensee predicted slow crack growth for the indication in Weld 2-BB-01-F302 based on fatigue degradation mechanism.
- 2. Based on the examination results of 2007 that show no growth in the indication from the 2004 examination, the indication will most likely not propagate significantly between 2007 and 2013.
- 3. As reported in the relief request, the indication is approximately 37.5 percent to 40 percent throughwall (crack depth). The maximum allowable depth per IWB-3600 is 75 percent throughwall. Therefore, the indication has sufficient margin before it would reach the maximum allowable depth.
- 4. Based on the previous submittal, the licensee stated that Weld 2-BB-01-F302 is connected to the inside surface of the weld. This implies that the indication is in contact with primary coolant and thus may experience potential stress-corrosion cracking. The licensee stated in WCAP-16280-P that a stress-corrosion cracking degradation mechanism does not occur in pressurized-water reactors (PWR). The NRC staff notes that certain stainless steel piping in PWRs has experienced stress-corrosion cracking due to halogens (e.g., chlorides) and sensitization of thin wall stainless steel piping as reported in NRC Information Notice 2011-04, Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking In Stainless Steel Piping in Pressurized Water Reactors, dated February 23, 2011 (ADAMS Accession No. ML103410363). However, stress-corrosion cracking has not been reported in the cold leg of PWRs because the cold leg would not likely be in contact with chlorides. Also, the cold-leg pipe-wall thickness is thick (approximately 2.32 inches) which minimizes the potential for sensitization. The NRC staff concludes that if stress-corrosion cracking were occurring, crack growth would have been identified in 2007, but was not.
Based on the above, the NRC staff concludes that a limited period of deferral for the examination in 2011 to 2013 is acceptable because such a deferral will not affect the safety of the plant and the proposed alternative provides an acceptable level of quality and safety.
Effective November 1, 2011, pursuant to paragraph 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations, the NRC staff authorized the use of Relief Request I3R-13 to defer the second successive examination of Weld 2-BB-01-F302 in C cold leg piping to refueling outage RF-19 scheduled for the spring 2013 at Callaway Plant Unit 1, and the third successive examination rescheduled to the first period of the fourth 10-year ISI interval.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector. The NRC staff will prepare a written safety evaluation for Relief Request I3R-13 as part of the verbal authorization.
Docket No. 50-483
ML113120087 OFFICE NRR/LPL4/PM NRR/LPL4/LA NRR/DE/EVIB/BC NRR/LPL4/BC NRR/LPL4/PM TLupold (JTsao NAME MThadani JBurkhardt for) MMarkley MThadani DATE 11/15/11 11/10/11 11/16/11 11/16/11 11/16/11
