ML063400204

RAI Related to Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity

ML063400204
Person / Time
Site:	Byron, Braidwood
Issue date:	12/13/2006
From:	Robert Kuntz NRC/NRR/ADRO/DORL/LPLIII-2
To:	Crane C Exelon Generation Co
kuntz, Robert , NRR/DORL, 415-3733
References
TAC MC8966, TAC MC8967, TAC MC8968, TAC MC8969
Download: ML063400204 (7)

Text

December 13, 2006 Mr. Christopher M. Crane President and Chief Nuclear Officer Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

BYRON STATION, UNIT NOS. 1 AND 2 AND BRAIDWOOD STATION, UNIT NOS. 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY (TAC NOS. MC8966, MC8967, MC8968, AND MC8969)

Dear Mr. Crane:

By letter to the Nuclear Regulatory Commission (NRC) dated November 18, 2005, Exelon Generation Company, LLC submitted an amendment request that would revise the technical specification requirements related to steam generator tube integrity, for the Byron Station, Unit Nos. 1 and 2, and Braidwood Station, Unit Nos. 1 and 2.

The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on December 5, 2006, it was agreed that you would provide a response within 60 days from the date of this letter.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRCs goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-3733.

Sincerely,

/RA/

Robert F. Kuntz, Project Manager Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-454, STN 50-455, STN 50-456 and STN 50-457

Enclosure:

Request for Additional Information cc w/encl: See next page

Mr. Christopher M. Crane December 13, 2006 President and Chief Nuclear Officer Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

BYRON STATION, UNIT NOS. 1 AND 2, AND BRAIDWOOD STATION, UNIT NOS. 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY (TAC NOS. MC8966, MC8967, MC8968, AND MC8969)

Dear Mr. Crane:

By letter to the Nuclear Regulatory Commission (NRC) dated November 18, 2005, Exelon Generation Company, LLC submitted an amendment request that would revise the technical specification requirements related to steam generator tube integrity, for the Byron Station, Unit Nos. 1 and 2, and Braidwood Station, Unit Nos. 1 and 2.

The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on December 5, 2006, it was agreed that you would provide a response within 60 days from the date of this letter.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRCs goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-3733.

Sincerely,

/RA/

Robert F. Kuntz, Project Manager Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-454, STN 50-455, STN 50-456 and STN 50-457

Enclosure:

Request for Additional Information cc w/encl: See next page DISTRIBUTION:

PUBLIC LPL3-2 R/F RidsNrrDorlLpl3-2 RidsNrrPMCGratton RidsNrrLAEWhitt RidsAcrsAcnwMailCenter RidsOgcRp RidsRgn3MailCenter RidsNrrDorlDpr RidsNrrPMRKuntz RidsNrrDciCsgb ADAMS Accession Number:ML063400204 NRR-088 OFFICE LPL3-2/PM LPL3-2/PM LPL3-2/LA DCI/CSGB/BC LPL3-2/BC NAME RKuntz:mw CGratton EWhitt AHiser MMarshall DATE 12/7/06 12/7/06 12/7/06 12/11/06 12/13/06 OFFICIAL RECORD COPY

Byron/Braidwood Stations cc:

Dwain W. Alexander, Project Manager Plant Manager - Byron Station Westinghouse Electric Corporation Exelon Generation Company, LLC Energy Systems Business Unit 4450 N. German Church Road Post Office Box 355 Byron, IL 61010-9794 Pittsburgh, PA 15230-0355 Site Vice President - Byron Howard A. Learner Exelon Generation Company, LLC Environmental Law and Policy 4450 N. German Church Road Center of the Midwest Byron, IL 61010-9794 35 East Wacker Dr., Suite 1300 Chicago, IL 60601-2110 U.S. Nuclear Regulatory Commission Braidwood Resident Inspectors Office U.S. Nuclear Regulatory Commission 35100 S. Rt. 53, Suite 79 Byron Resident Inspectors Office Braceville, IL 60407 4448 N. German Church Road Byron, IL 61010-9750 County Executive Will County Office Building Regional Administrator, Region III 302 N. Chicago Street U.S. Nuclear Regulatory Commission Joliet, IL 60432 Suite 210 2443 Warrenville Road Plant Manager - Braidwood Station Lisle, IL 60532-4351 Exelon Generation Company, LLC 35100 S. Rt. 53, Suite 84 Ms. Lorraine Creek Braceville, IL 60407-9619 RR 1, Box 182 Manteno, IL 60950 Ms. Bridget Little Rorem Appleseed Coordinator Chairman, Ogle County Board 117 N. Linden Street Post Office Box 357 Essex, IL 60935 Oregon, IL 61061 Document Control Desk - Licensing Mrs. Phillip B. Johnson Exelon Generation Company, LLC 1907 Stratford Lane 4300 Winfield Road Rockford, IL 61107 Warrenville, IL 60555 Attorney General Site Vice President - Braidwood 500 S. Second Street Exelon Generation Company, LLC Springfield, IL 62701 35100 S. Rt. 53, Suite 84 Braceville, IL 60407-9619 Illinois Emergency Management Agency Senior Vice President - Operations Support Division of Disaster Assistance & Exelon Generation Company, LLC Preparedness 4300 Winfield Road 110 East Adams Street Warrenville, IL 60555 Springfield, IL 62701-1109

Byron/Braidwood Stations Director - Licensing and Regulatory Senior Vice President - Midwest Operations Affairs Exelon Generation Company, LLC Exelon Generation Company, LLC 4300 Winfield Road 4300 Winfield Road Warrenville, IL 60555 Warrenville, IL 60555 Manager Regulatory Assurance - Braidwood Exelon Generation Company, LLC 35100 S. Rt. 53, Suite 84 Braceville, IL 60407-9619 Manager Regulatory Assurance - Byron Exelon Generation Company, LLC 4450 N. German Church Road Byron, IL 61010-9794 Assistant General Counsel Exelon Generation Company, LLC 200 Exelon Way Kennett Square, PA 19348 Vice President - Regulatory & Legal Affairs Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Manager Licensing - Braidwood/Byron Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555

REQUEST FOR ADDITIONAL INFORMATION BYRON STATION, UNIT NOS. 1 AND 2, AND BRAIDWOOD STATION, UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-454, STN 50-455, STN 50-456 AND STN 50-457 In reviewing the Exelon Generation Companys (EGC) submittal dated November 18, 2005, as supplemented by letters dated August 18 and September 28, 2006, related to a proposed amendment to revise the technical specification requirements related to steam generator tube integrity, for the Byron Station, Unit Nos. 1 and 2 (Byron) and Braidwood Station, Unit Nos. 1 and 2 (Braidwood), the NRC staff has determined that the following information is needed in order to complete its review:

References:

1. EGC letter RS-05-129, Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity, dated November 18, 2005.

2. EGC letter RS-06-109, Response to Request for Additional Information Regarding Application for Steam Generator Tube Integrity Technical Specification, dated August 18, 2006.

Requested Information:

1. Regarding EGCs response to the request for information (RAI) question 8 (for Byron/Braidwood, question 5 for Seabrook) in Reference 2, Attachment 6, pages 8 to 16, provide a plot of crack-opening angle for circumferential cracks located 4 inches from the bottom of the tubesheet as a function of crack length for normal operating and main steam line break conditions. Also, provide a plot of leak rate as a function of the same parameters, neglecting the effect of crevice resistance.

2. Regarding EGCs response to RAI question 8 (for Byron/Braidwood, question 5 for Seabrook) in Reference 2, Attachment 6, pages 8 to 16, provide revised versions of Figures 3 and 4 to include the leak rate ratios for cracks in the range of 0.1 to 0.5 inches in length. It would seem from Figures 3 and 4, that if crack resistance dominates crevice resistance, then leakage ratios may exceed 2 for through wall crack lengths less than 0.5 inches for tubes near the periphery of the bundle, particularly for circumfrential cracks. Also, provide similar figures for the near radius and mid-radius locations.

3. The discussion accompanying Figures 3 and 4 states that cracks less than 0.5 inches in length are not expected to cause any relative significant leakage. Please explain basis for concluding the leakage contribution from the population of circumferential cracks of ENCLOSURE

through-wall length less than 0.5 inches is small, relative to the leakage contribution from the population of through-wall cracks greater than 0.5 inches in length such that the leakage ratio between normal operating and accident conditions is dominated by the leakage ratio (which is less than 2) exhibited by the population of cracks larger than 0.5 inches. This explanation should consider any relevant operating experience regarding the probability density function of 100 percent through wall crack lengths and, in addition, the plots provided in response to question 2 above.

4. Regarding a statement in the discussion underneath Figure 4 that reads, "The results from the crack-only analyses show that in the absence of the dent the resistance to flow is increased and each crack type produces a lower leak rate ratio, please clarify what is meant by the dent, and its impact on this statement. Additionally, please qualify what the increase in resistance to flow and lower leak ratios are relative to.

5. Reference 2, Attachment 6, page 12, Analysis of Circumferential Cracking, states that the circumferential crack model was developed in WCAP-15932-P, Revision 1, Improved Justification of Partial Length RPC inspection of Tube Joints of Model F Steam Generators of Ameren-UE Callaway Plant, dated May 2003. WCAP-15932P, Appendix C states that the main loadings on a circumferential crack below the H*

distance are the pressure loads acting on the crack face. It is also stated in the WCAP that the internal pressure end cap load is not transmitted below about 1/3 the H*

distance. Assuming that H* is determined correctly, the NRC staff agrees that this statement is true for normal operating pressure provided the tube is severed immediately below the 1/3 H* distance. Similarly, the 3 delta P end cap load does not extend below the full H* distance assuming the tube is severed immediately below the H* distance. If the tube is not severed, then much of the end cap load will be transmitted below the H* distance. Taking an extreme example, the calculated H*

distance is based in part on pull out tests (on specimens that were basically severed at the bottom) where the pull out criterion was an axial displacement of 0.25 inches at the bottom of the specimen. If the tube is intact below the H* distance, then the tube must be able to stretch by 0.25 inches between the weld and the H* location which means there must be considerable force transmitted below the H* distance. For smaller end cap loads where no slippage takes place, a severed tube end would be expected to displace upward due to the accumulated strains in the tube to tubesheet joint above the severed location. If the tube is not completely severed, the tube below the crack would be expected to resist this displacement and thus resist some of the pullout load. The tube to tubesheet joint (where the tube is not severed inside the tubesheet) is a redundant structure. How much of the end cap load that gets transmitted below the crack location (assumed to be 17 inches down from the top of the tubesheet) depends on the stiffness of the friction joint above the crack relative to the stiffness of the tube below the crack. It is not clear from the NRC staffs review of the model that this effect has been evaluated. Thus, it is not clear to the NRC staff that the axial load acting on the circumferentially-cracked cross section is limited solely to the pressure acting on the crack faces and that no portion of the internal pressure end cap load is acting on the cross section. Please address this concern, including how the stiffness of the tube to tubesheet friction joint above the crack relative to the stiffness of the tube below the crack have been specifically accounted for. Has a detailed analysis (e.g., finite element analysis) been performed to determine how much of the full internal pressure end cap

load is actually transmitted to the cracked cross section under normal operating and accident conditions? If so, describe the analysis and the results.

6. The Reference 1 application included the following provision in TS 5.5.9c: For Unit 2 only, degradation found in the portion of the tube from the top of the hot leg tubesheet to 17 inches below the top of the tubesheet shall be plugged or repaired upon detection.

In the example accompanying the NRC staffs draft RAI No. 4, the NRC staff inadvertently left this sentence out. It wasnt the NRC staffs intent to suggest this sentence should be deleted. Describe your plan for re-including this sentence as part of TS 5.5.9c. Also, as a point of consistency and clarification, the word degradation in the above sentence and in TS5.5.9c.4.i should be replaced by the word flaws consistent with the rest of the technical specifications. Please describe your plan for making this change as well.

7. Did any of the hydraulic expansions in the Model D5 SGs experience a stress relief during fabrication, directly or indirectly (e.g., as a result of stress relieving the shell to tubesheet welds)? If so, how was this reflected in the pullout and leakages tests in support of the tubesheet amendment requests?

8. The tubesheet bow analysis described in Westinghouse report, LTR-CDME-05-32-P, Rev 2, submitted as Attachment 7 to Reference 1, takes credit for resistance against bow provided by the divider plate. Cracks in the welds connecting the tubesheet and divider plate have been found by inspection at certain foreign steam generators. Please discuss how such cracks, if present at the Byron/Braidwood units, could affect the conservatism of the proposed 17-inch tubesheet inspection distance requirement for ensuring structural and leakage integrity.