Text

Leak-Before-Break Evaluation for Steam Generator Replacement
	ML030560703
Person / Time
Site:	Oconee
Issue date:	02/19/2003
From:	Rosalyn Jones; Duke Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML030560703 (12)
	v • d • e

SDuke R. A.JONES SPower. Vice President A Duke Energy Company Duke Power 29672 / Oconee Nuclear Site 7800 Rochester Highway Seneca, SC 29672 864 885 3158 864 885 3564 fax February 19, 2003 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk

Subject:

Duke Energy Corporation Oconee Nuclear Station, Units 1, 2, and 3 Docket Nos. 50-269, 50-270, 50-287 Leak-Before-Break Evaluation for Steam Generator Replacement By letter dated September 6, 2001, the NRC approved Duke Energy Corporation's (Duke) proposed methodology for the analysis of the reactor coolant loop (RCL) in support of steam generator replacement at Oconee Nuclear Station, Units 1, 2, and 3. As noted in that letter, one basis of the Staff's conclusion was that prior to finalizing the piping design, that Duke demonstrate that the leak-before-break (LBB) application previously approved by the NRC is valid for the extended period of operation.

An evaluation was performed by Babcock & Wilcox Canada to address changes due to steam generator replacement and plant life extension. Attachment 1 provides the requested summary results of this evaluation. Results show that the LBB scenario remains valid with the considered changes to the Oconee Reactor Coolant System.

Inquiries on this matter should be directed to Robert Sharpe at (704) 805-2007 or Robert Douglas at (864) 885-3073.

Very r y yours, VR1 tones, s Site-vice President www duke-energy com

U. S. NRC Document Control Desk Page 2 February 19, 2003 xc:

L. A. Reyes Regional Administrator, Region II U. S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303 M. C. Shannon Senior Resident Inspector Oconee Nuclear Station S. J. Vias U. S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303 L. N. Olshan Senior Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission R. Wingard Director, Division of Radioactive Waste Management Bureau of Land & Waste Management S. C. Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201

Attachment 1 Oconee Nuclear Station Leak-Before-Break Re-Evaluation For Steam Generator Replacement Backgqround In 1985, the B&W Owners Group (B&WOG) performed a generic Leak-Before-Break (LBB) evaluation for B&W designed plants (Reference 1). This evaluation reviewed the base metal and weld data for B&WOG plants and established a reasonable lower bound of material properties to be used in the evaluation. By letter dated December 12, 1985 (Reference 2), the NRC transmitted their Safety Evaluation Report and acceptance of the B&WOG LBB approach.

This SER was applicable to Oconee.

By letters dated August 28, 2000 and July 26, 2001, Duke requested NRC's approval of the Framatome-ANP methodology that would be used for re-analysis of the reactor coolant loop in support of steam generator replacement at Oconee. This re-analysis did not include breaks in the large bore primary system piping because of the NRC's previous approval of LBB for Oconee. In reviewing this request, the NRC requested that Duke provide information to demonstrate that the analyses and results of the previous LBB evaluation still bound the plant specific applications at Oconee after steam generator replacement.

The Once-Through Steam Generators (OTSGs) will be replaced in Fall 2003, Spring 2004, and Fall 2004, for Oconee Units 1, 2, and 3, respectively. Replacement of the OTSGs will involve changes to the piping material and welds at specific locations:

"* A short (221/2) section of cold leg piping attached to each of the two OTSG outlet nozzles will be replaced with a one piece SA-508 Cl. 3a forging. The original cold leg elbow was fabricated from formed and welded SA-516 Gr. 70 plate.

"* The hot leg elbow attached to the inlet nozzle of the OTSG will be replaced with identical specification elbows. This does not represent a material change from Reference 1.

"* The original installation welds for the OTSG used shielded metal arc welds (SMAW) or submerged arc welds (SAW). For steam generator replacement, narrow groove gas tungsten metal arc welding (GTAW) will be used.

Re-evaluation Babcock & Wilcox Canada was contracted to evaluate the changes in the Oconee Units 1, 2, and 3 reactor coolant systems due to steam generator replacement and plant life extension, and to determine if the B&WOG LBB Topical Report (Reference 1) needed to be updated for Oconee. This evaluation increased the number of heatup and cooldown cycles from 240 to 360

U.S. NRC Document Control Desk Page 2 February 19, 2003 cycles to be consistent with the current Oconee licensing bases. This evaluation resulted in the following findings:

Weld Evaluation The replacement steam generators will be welded into the reactor coolant loop with narrow groove GTAW with ER 70S-6 filler material. An assessment was performed to determine the full fracture toughness (Jlc) and tensile (true stress/strain) properties at the operating temperature. The new weldment and weld filler materials were shown to have fracture properties equal to or better than those of the materials covered in Reference 1.

Cold Legq Elbow Evaluation The 221/2"section of the cold leg elbow will be replaced with a one piece SA-508 Cl. 3a forging.

The new cold leg elbow section was fabricated as an integral part of the replacement steam generator. Due to differences in strength, P-number, and product form, a program was initiated to determine the full fracture toughness (J1 c) and tensile (true stress/strain) properties of the replacement SA 508 Cl. 3a material at the operating temperature. The partial cold leg elbow material was shown to have fracture properties equal to or better than those of the materials covered in Reference 1.

Cast Stainless Steel Agqing Assessment Cast Austenitic stainless steel (CASS) is used for the reactor coolant pump casing. An assessment of the thermal embrittlement of CASS for the extended operating lifetime was performed. Using data from the literature, the cast stainless steel material of the primary pump housings was shown to have a fracture toughness J 1c in excess of the requirement at the most highly loaded location in the cold leg piping.

Hot Leg Elbow Replacement Assessment As previously noted, the hot leg elbows will be replaced. The elbow material will be the same as evaluated in Reference 1. Consideration was also given to the hot leg cut location. The hot leg elbow will be cut from the OTSG at the inlet nozzle. The second cut will be made at a cut location 180 degrees away or at a lesser angle, based on the final fit-up plan.

A new hot leg elbow section will be made from the same carbon steel material as the original piping material. Therefore, the use of the bounding properties of Reference 1 are adequate to cover this change. There is no effect if the new weld location is not at the original 180° weld location because the Reference 1 analysis used the bounding weld properties in conjunction with the worst stress at any location and thus conservatively covered any new weld location.

U.S. NRC Document Control Desk Page 3 February 19, 2003 Reconciliation of Piping Loads Piping moments, with the replacement steam generators installed, were determined by separate studies. These results were compared to the piping loads in Reference 1. The piping moment loads were bounded by those of Reference 1 with one exception where the flaw size needed to obtain 10 gpm leakage flow increased relative to Reference 1. Re-analysis showed that the fracture and limit load criteria were still satisfied.

Reconciliation of Fracture Mechanics Analysis The required flaw sizes that result in 10-gpm leakage flow were determined for the hot and cold leg pipes under pressure plus normal operating moment loads and compared to Reference 1.

The fracture and limit load evaluation of the piping concluded that all locations satisfy the allowables.

Reconciliation of Crack Growth Analysis The crack growth analyses performed in Reference 1 were redone using the number of cyclic events anticipated for the extended plant life of 60 years.

Crack growth was shown in Reference 1 to occur predominately in the radial direction. This conclusion from Reference 1 is not affected by the planned changes.

The evaluation of crack growth due to the application of cyclic loads concluded that a substantial initial flaw size is required for the crack to grow through-wall over the 60-year plant life. As this size considerably exceeds the maximum that would be permitted to remain in the piping under the construction code, the presence of such large flaws is extremely unlikely.

Conclusion The reconciliation of the changes that result from steam generator replacement and life extension performed by Babcock &Wilcox Canada in Reference 3 demonstrated that the Leak Before-Break evaluation previously performed by the B&W Owners Group in Reference 1 remains conservative for Oconee Units 1, 2, and 3.

References

1. B&W Owners Group Topical Report BAW 1847, Rev. 1, "Leak-Before-Break Evaluation of Margins Against Full Break for RCS Primary Piping of B&W Designed NSS", September 1985

2. Letter from D. E. LaBarge, NRC, to W. R. McCollum, Jr. dated September 6, 2001

U.S. NRC Document Control Desk Page 4 February 19, 2003

3. Oconee Unit 1,2,3 Reconciliation of RCS Piping Leak Before Break Evaluation for Steam Generator Replacement, B&W Report No. 006K-SR-05, Revision 0, Proprietary (Oconee Calculation No. OSC-7844)

Duke R.A.JONES Power Vice President A Duke Energy Company Duke Power 29672 / Oconee Nuclear Site 7800 Rochester Highway Seneca, SC 29672 864 885 3158 864 885 3564 fax February 19, 2003 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk

Subject:

Duke Energy Corporation Oconee Nuclear Station, Units 1, 2, and 3 Docket Nos. 50-269, 50-270, 50-287 Leak-Before-Break Evaluation for Steam Generator Replacement By letter dated September 6, 2001, the NRC approved Duke Energy Corporation's (Duke) proposed methodology for the analysis of the reactor coolant loop (RCL) in support of steam generator replacement at Oconee Nuclear Station, Units 1, 2, and 3. As noted in that letter, one basis of the Staff's conclusion was that prior to finalizing the piping design, that Duke demonstrate that the leak-before-break (LBB) application previously approved by the NRC is valid for the extended period of operation.

An evaluation was performed by Babcock & Wilcox Canada to address changes due to steam generator replacement and plant life extension. Attachment 1 provides the requested summary results of this evaluation. Results show that the LBB scenario remains valid with the considered changes to the Oconee Reactor Coolant System.

Inquiries on this matter should be directed to Robert Sharpe at (704) 805-2007 or Robert Douglas at (864) 885-3073.

Very r y yours, Rtones, u Site-vice President www duke-energy com

U. S. NRC Document Control Desk Page 2 February 19, 2003 xc:

L. A. Reyes Regional Administrator, Region II U. S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303 M. C. Shannon Senior Resident Inspector Oconee Nuclear Station S. J. Vias U. S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303 L. N. Olshan Senior Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission R. Wingard Director, Division of Radioactive Waste Management Bureau of Land & Waste Management S. C. Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201

Attachment 1 Oconee Nuclear Station Leak-Before-Break Re-Evaluation For Steam Generator Replacement

Background

In 1985, the B&W Owners Group (B&WOG) performed a generic Leak-Before-Break (LBB) evaluation for B&W designed plants (Reference 1). This evaluation reviewed the base metal and weld data for B&WOG plants and established a reasonable lower bound of material properties to be used in the evaluation. By letter dated December 12, 1985 (Reference 2), the NRC transmitted their Safety Evaluation Report and acceptance of the B&WOG LBB approach.