RA-09-015, Proposed Relief Request to the Requirements of 10 CFR 50.55a Concerning Reactor Pressure Vessel Circumferential Shell Welds

Proposed Relief Request to the Requirements of 10 CFR 50.55a Concerning Reactor Pressure Vessel Circumferential Shell Welds

ML090440230
Person / Time
Site:	Oyster Creek
Issue date:	02/13/2009
From:	Cowan P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-09-015
Download: ML090440230 (7)

Text

10 CFR 50.55a RA-09-015 February 13, 2009 U. S. Nuclear Regulatory Commission Attn: Document Control Desk 11555 Rockville Pike Rockville, MD 20852 Oyster Creek Nuclear Generating Station Facility Operating License No. DPR-16 NRC Docket No. 50-219

Subject:

Proposed Relief Request to the Requirements of 10 CFR 50.55a Concerning Reactor Pressure Vessel Circumferential Shell Welds Attached for your review and approval is a proposed alternative in accordance with 10 CFR 50.55a(a)(3)(i), associated with the extended period of operation (April 10, 2009 through April 9, 2029) for the Inservice Inspection Program for Oyster Creek Nuclear Generating Station (OCNGS). Based on a start date of October 15, 2002, the OCNGS lSI Program complies with the requirements of the 1995 Edition through 1996 Addenda of the ASME,Section XI Code.

The fourth ten-year Interval began on October 15, 2002 and concludes on October 14, 2012.

This relief request satisfies commitment 47 of the OCNGS license renewal Safety Evaluation Report, which requires resubmittal of the reactor vessel circumferential shell weld relief request prior to the period of extended operation (Le., prior to April 9, 2009).

We request your review and approval by February 13, 2010.

There are no regUlatory commitments contained in this letter.

If you have any questions or require additional information, please contact Tom Loomis (610-765-5510).

Pamela B. Cowan Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Oyster Creek Nuclear Generating Station RPV Shell Weld Relief Request February 13, 2009 Page 2 of 2 Attachment - Oyster Creek Nuclear Generating Station Relief Request OC-08-01 cc:

S. J. Collins, Administrator, USNRC, Region I (w/attachment)

M. S. Ferdas, USNRC Senior Resident Inspector, OCNGS (w/attachment)

G. E. Miller, USNRC Project Manager, OCNGS (w/attachment)

ATTACHMENT OYSTER CREEK NUCLEAR GENERATING STATION RELIEF REQUEST OC-08-01

Oyster Creek Nuclear Generating Station RPV Circumferential Shell Weld Relief Request Page 1 of 4 Exelon Generation Company, LLC Oyster Creek Nuclear Generating Station Extended Period of Operation Relief Request OC-08-01 ASME CODE COMPONENTS AFFECTED:

Code Class:

Reference:

Examination Category:

Item Number:

==

Description:==

Weld Number:

1 Table IWB-2500-1, Examination Category B-A, Item B1.11 (Circumferential Shell Welds)

B-A B1.11 Permanent relief from circumferential shell weld examinations for the extended period of operation NR02-1-563, NR02-1-572, NR02-3-564, and NR02-3-572 APPLICABLE CODE EDITION AND ADDENDA:

The current edition for the Inservice Inspection (lSI) interval is the ASME Section XI, 1995 Edition through 1996 Addenda.

APPLICABLE CODE REQUIREMENT:

ASME Section XI, 1995 Edition through 1996 Addenda, Table IWB-2500-1, Examination Category B-A, Item B1.11, requires a volumetric examination of the circumferential shell welds each interval.

REASON FOR REQUEST:

Exelon Generation Company, LLC (Exelon) is requesting a proposed alternative in accordance with 10 CFR 50.55a(a)(3)(i) on the basis that this alternative provides an acceptable level of quality and safety. This relief request would provide relief from circumferential weld examinations as currently required by the ASME Code for the extended period of operation.

A permanent relief (Le., "for the remaining term of operation under the initial, existing license")

from the requirements of Item B1.11 concerning examination of circumferential shell welds was approved for Oyster Creek Nuclear Generating Station (OCNGS) in an NRC Safety Evaluation Report dated September 14, 2000 (Reference 1).

The renewal of the OCNGS operating license is discussed in the "Safety Evaluation Report Related to the License Renewal of Oyster Creek Generating Station," NUREG-1875, dated April 2007 (Reference 2). As discussed in Section 4.2.4.4 ("Conclusion"):

Oyster Creek Nuclear Generating Station RPV Circumferential Shell Weld Relief Request Page 2 of 4 "On the basis of its review, as discussed above, the staff concludes that the applicant has demonstrated, pursuant to 10 CFR 54.21 (c)(1 )(ii), that, for the reactor vessel circumferential weld examination relief TLAA, the analyses have been projected to the end of the period of extended operation. However, even though the analyses have been projected to the end of the period of extended operation, the applicant will still have to request an extension of the relief for circumferential welds examination for the renewal period."

Additionally, commitment 47 of the OCNGS license renewal Safety Evaluation Report (Reference 2) commits to resubmitting the circumferential shell weld relief request prior to the period of extended operation.

Accordingly, this relief request is provided to extend the relief from circumferential weld examinations for the extended period of operation.

PROPOSED ALTERNATIVE AND BASIS FOR USE:

The following information from the Reference 2 Safety Evaluation Report (Section 4.2.4.2) is provided as the basis for use of the proposed alternative:

"The staff's final SER concerning the BWRVIP-05 report, dated July 28, 1998, discusses the technical basis for relief. In this letter, the staff concludes that, since the failure frequency for circumferential welds in BWR plants is significantly below the criterion specified in RG 1.154, "Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors," dated January 1987, and below the core damage frequency (CDF) of any BWR plant, the continued inspection would result in a negligible decrease in an already acceptably low reactor vessel failure probability.

Therefore, elimination of the inservice inspection (lSI) requirements for reactor vessel circumferential welds is justified. The staff's letter indicated that BWR applicants may request relief from the lSI requirements of 10 CFR 50.55a(g) for volumetric examination of circumferential RPV welds by demonstrating that:

• at the expiration of the license, the circumferential welds satisfy the limiting conditional failure probability for circumferential welds in the SER dated July 28, 1998, and

• the applicants have implemented operator training and established procedures that limit the frequency of cold overpressure events to the frequency specified in the staff's SEA.

The letter indicated that the requirements for inspection of circumferential reactor vessel welds during an additional 20-year license renewal period would be reassessed, on a plant-specific basis, as part of any BWR LRA. Therefore, the applicant must request relief from inspection of circumferential welds during the license renewal period, pursuant to 10 CFR 50.55a.

Section A.4.5 of the BWRVIP-74 report indicates that the staff's SER of the BWRVIP-05 report conservatively evaluated the BWR reactor vessels to 64 EFPYs, which is 10 EFPYs greater than what is realistically expected for the end of the license renewal period. The NRC staff used the mean RTNOT value for materials to evaluate failure probability of BWR circumferential welds at 32 and 64 EFPYs in the staff SER dated July 28, 1998. The neutron fluence used in this evaluation was the neutron fluence at the clad-weld (inner) interface.

Oyster Creek Nuclear Generating Station RPV Circumferential Shell Weld Relief Request Page 3 of 4 Since the staff analysis discussed in the BWRVIP-74 report is a generic analysis, the applicant submitted plant-specific information to demonstrate that the OCGS beltline materials meet the criteria specified in the report. To demonstrate that the reactor vessel has not become embrittled beyond the basis for the relief, LRA Table 4.2.4-1 compares the 50-EFPY material data for the limiting circumferential welds with the 64-EFPY reference case in Appendix E to the staff's SER of the BWRVIP-05 report. The material data included amounts of copper and nickel, chemistry factor, the neutron fluence, ~RTNOT, initial RTNOT, and mean RTNOT for the limiting circumferential weld at the end of the period of extended operation. The staff verified the validity of the data for the copper and nickel contents and the initial RTNOT values for the reactor vessel beltline materials based on the evaluation in SER Section 4.2.2. The 50-EFPY mean RTNOT value is 9.8 OF. The staff reviewed the applicant's calculations for the 50-EFPY mean RTNOT values for the circumferential welds using the data presented in LRA Table 4.2.4-1 and found them to be accurate. These 50-EFPY mean RTNOT values are bounded by the 64-EFPY mean RTNOT value of 128.5 OF used by the NRC for determining the conditional failure probability of a circumferential weld. The 64-EFPY mean RTNOT value from the staff SER dated July 28, 1998, is representative of a CE weld because CE fabricated the circumferential welds in the OCGS reactor vessel.

Since the eCGS 50-EFPY mean RTNOT value is less than the 64-EFPY value used in the staff SER dated July 28, 1998, the staff concludes that the reactor vessel conditional failure probability is bounded by the NRC analysis."

In addition to the above discussion from the Reference 2 Safety Evaluation Report, procedures and training used to limit cold overpressure events remain in-place at OCNGS for the current and extended period of operation. As one example, OCNGS procedure 602.4.001 ("Nuclear Steam Supply System (NSSS) Leak Test"), which is used for Class 1 system leakage testing, contains controls for special management attention such as briefings (Section 3.1), temperature controls (Attachment 13) and pressure controls (Attachment 14). As also discussed in Section 4.2.4.2:

"The staff finds that the applicant's evaluation for this TLAA is acceptable because the eCGS 50-EFPY conditional failure probability for the reactor vessel circumferential welds is bounded by the NRC analysis in the staff SER dated July 28, 1998, and the applicant will be using procedures and training to limit cold overpressure events during the period of extended operation. This analysis satisfies the evaluation requirements of the staff SER dated July 28, 1998; however, the applicant is still required to request relief for the circumferential weld examination for the extended period of operation, in accordance with 10 CFR 50.55a."

SER Section 4.2.4.4 concludes:

"On the basis of its review, as discussed above, the staff concludes that the applicant has demonstrated, pursuant to 10 CFR 54.21 (c)(1 )(ii), that, for the reactor vessel circumferential weld examination relief TLAA, the analyses have been projected to the end of the period of extended operation. However, even though the analyses have been projected to the end of the period of extended operation, the applicant will still have to request an extension of the relief for circumferential welds examination for the renewal period."

Exelon has reviewed the above conclusions and has confirmed they are still valid for the extended period of operation. Therefore, the proposed alternative as discussed above, and as previously evaluated in the Reference 2 SER, provides an acceptable level of quality and safety for the extended period of operation.

Oyster Creek Nuclear Generating Station RPV Circumferential Shell Weld Relief Request Page 4 of 4 DURATION OF PROPOSED ALTERNATIVE:

This relief is requested for examinations of the circumferential shell welds for the extended period of operation at the OCNGS (April 10, 2009 through April 9, 2029).

REFRENCES:

1.

Letter from M. Gamberoni (NRC) to R. DeGregorio (AmerGen Energy Company, LLC),

"Safety Evaluation of the Request for Relief from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI Requirements for the Containment Inservice Inspection Program, Oyster Creek Nuclear Generating Station (TAC NO. MA8016)," dated September 14, 2000.

2.

"Safety Evaluation Report Related to the License Renewal of Oyster Creek Generating Station," NUREG-1875, dated April 2007.