ML081000217
| ML081000217 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 04/18/2008 |
| From: | David M NRC/NRR/ADRO/DORL/LPLIII-2 |
| To: | Pardee C Exelon Generation Co |
| marshall david 415-1547 | |
| References | |
| TAC MD7479, TAC MD7480, TAC MD7481, TAC MD7482 | |
| Download: ML081000217 (6) | |
Text
April 18, 2008 Mr. Charles G Pardee Chief Nuclear Officer and Senior Vice President Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNITS 1 AND 2 - WITHDRAWAL OF LICENSE AMENDMENT REQUEST RE: STEAM GENERATOR TUBE ALTERNATE REPAIR CRITERIA TECHNICAL SPECIFICATION (TAC NOS. MD7479, MD7480, MD7481, AND MD7482)
Dear Mr. Pardee:
By letter dated November 29, 2007 (Agencywide Documents Access Management System (ADAMS) Accession No. ML073450569), Exelon Generation Company, LLC (Exelon, the licensee) submitted a license amendment request for Braidwood Station, Units 1 and 2 (Braidwood), and Byron Station, Units 1 and 2 (Byron) to revise the technical specification (TS) requirements related to steam generator (SG) tube integrity. Specifically, the request proposed an Alternate Repair Criteria (ARC) that would revise TS 5.5.9, "Steam Generator (SG)
Program," to permanently exclude, from eddy current inspection, the portion of the tube below 17 inches from the top of the tubesheet in the Braidwood, Unit 2, and Byron, Unit 2, SGs.
Although the proposed changes only affect Braidwood, Unit 2, and Byron, Unit 2, this request is docketed for Braidwood, Units 1 and 2, and Byron, Units 1 and 2, because the TS are common at each station.
The Nuclear Regulatory Commission (NRC) staff performed an acceptance review of this amendment request. The acceptance review was performed to determine if there is sufficient technical information in scope and depth to allow the NRC staff to complete its detailed technical review. The acceptance review also ensures that the application adequately characterizes the regulatory requirements and licensing basis of the plant.
Consistent with Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), an amendment to the license (including the TSs) must fully describe the changes requested, following as far as applicable the form prescribed for original applications. Section 50.34 of 10 CFR addresses the content of technical information required. This section stipulates that the submittal address the design and operating characteristics, unusual or novel design features, and principal safety considerations.
The NRC staff reviewed your application and concluded that it did not provide technical information in sufficient detail to enable the NRC staff to complete its detailed review and make an independent assessment regarding the acceptability of the proposed amendment in terms of regulatory requirements and the protection of public health and safety. The NRC staff discussed the information that was insufficient in a call to your staff on March 17, 2008.
C.
By letter dated April 1, 2008 (ADAMS Accession No. ML080950368), you requested to withdraw the application from NRC staff review. The NRC staff acknowledges your request to withdraw the application. Activities on the review have ceased, and the associated Technical Assignment Control numbers have been closed.
The information that the NRC staff found to be insufficient is listed in the enclosure to this letter for your consideration in the event that you submit a SG ARC license amendment request in the future. If you have any questions, please contact me at (301) 415-1547.
Sincerely,
/RA/
Marshall J. David, Senior Project Manager Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456, STN 50-457, STN 50-454 and STN 50-455
Enclosure:
As stated cc w/encl: See next page
ML080950368), you requested to withdraw the application from NRC staff review. The NRC staff acknowledges your request to withdraw the application. Activities on the review have ceased, and the associated Technical Assignment Control numbers have been closed.
The information that the NRC staff found to be insufficient is listed in the enclosure to this letter for your consideration in the event that you submit a SG ARC license amendment request in the future. If you have any questions, please contact me at (301) 415-1547.
Sincerely,
/RA/
Marshall J. David, Senior Project Manager Plant Licensing Branch III-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456, STN 50-457, STN 50-454 and STN 50-455
Enclosure:
As stated cc w/encl: See next page DISTRIBUTION:
PUBLIC RidsNrrDorlLpl3-2 RidsRgn3MailCenter LPL3-2 R/F RidsNrrPMMDavid RidsNrrAcrsAcnw&mMailCenter RidsOgcRp RidsNrrLAEWhitt RidsNrrDciCsgb Accession Number: ML081000217 NRR-106 OFFICE LPL3-2/PM LPL3-2/LA DCI/CSGB/BC LPL3-2/BC NAME MDavid EWhitt AHiser RGibbs DATE 04/14/08 04/11/08 04/14/08 04/18/08 08
ENCLOSURE INSUFFICIENT INFORMATION BRAIDWOOD STATION, UNITS 1 AND 2 BYRON STATION, UNITS 1 AND 2 STEAM GENERATOR TUBE ALTERNATE REPAIR CRITERIA The Nuclear Regulatory Commission (NRC) staff performed its acceptance review of Exelon Generation Company, LLCs (Exelon, the licensee) November 29, 2007 (Agencywide Documents Access Management System Accession No. ML073450569), request to revise the technical specification (TS) requirements related to steam generator (SG) tube integrity.
Specifically, the proposed amendment would revise TS 5.5.9, Steam Generator Program, which would redefine the primary pressure boundary from the tube end weld to 17 inches below the top of the tubesheet. The proposed amendment employs the H*/B* methodology to determine the length of SG tube within the tubesheet that must be inspected.
As a result of the acceptance review, the NRC staff determined that the request does not provide technical information in sufficient detail to enable the NRC staff to complete its detailed review and make an independent assessment regarding the acceptability of the proposed amendment in terms of regulatory requirements and the protection of public health and safety. Specifically, the request does not provide sufficient information to support a conclusion that the tubes would have adequate structural and leakage integrity if flaws were found in the region of the tube not inspected. The information that is insufficient is summarized below.
The request does not adequately demonstrate that the proposed H*/B* distances make adequate allowance for uncertainties such as to ensure with a high probability (e.g., 0.99) that all SG tubes will exhibit pullout resistances consistent with the TS performance criteria. The stacked uncertainty model does not conservatively bound the uncertainties for an individual tube. The discussion of extreme value considerations lacks completeness, as it does not consider certain key variables and combinations of variables.
The request does not provide sufficient information to adequately characterize the potential range in values of residual contact pressure between the tube and tubesheet (due to the hydraulic expansion process), which may be encountered among the many thousands of tubes. Only limited pullout data exist upon which the residual contact pressures are estimated, and the data exhibit significant scatter. Residual contact pressure (and, thus, residual pullout load capacity) is highly sensitive to several parameters, including hydraulic expansion pressure, tube yield strength, tube material strain hardening properties, and initial radial gap (i.e., pre-expansion) between the tube and tubesheet. The request does not provide information to establish whether the pullout test specimens adequately envelop the range of values of these parameters that may be encountered in the SGs.
- The request does not provide sufficient information (e.g., literature search for relevant data, expert opinion and experience) to adequately characterize the range in potential values of the thermal expansion coefficients (TECs) for the tubesheet material and tube material that may be encountered in the field. In addition, the request does not provide sufficient information concerning potential changes to the tube material TEC due to the hydraulic expansion process (i.e., due to cold working).
The submitted analysis does not sufficiently demonstrate that incremental slippage of the tubing under cyclic operational loads will not occur, nor has compensatory monitoring for slippage been proposed.
The method for accounting for the pressure in the crevice between the tube and tubesheet (i.e., limiting median crevice pressure model) has not been demonstrated to be conservative for all conditions (e.g., normal operating conditions and steamline break) and for all tubes within the tubesheet.
During a meeting with the NRC staff on December 13, 2007 (closed meeting due to discussion of proprietary information), Westinghouse identified the use of a beta factor, which is embedded in the crevice pressure model for at least some of the H*/B*
analyses and which may have a significant influence on the H*/B* distances. This beta factor was only briefly discussed during the meeting, but apparently may significantly influence some of the H*/B* analyses. The request does not address the use of this beta factor, its purpose, and its technical basis.
During the December 13, 2007, meeting, Westinghouse stated that the method used to scale the results of tubesheet finite element analyses from normal operating conditions to steamline break (SLB) conditions was inappropriate, leading to over-conservative results for the SLB. The request does not address the nature of the problem or its resolution.
Byron/Braidwood Stations cc:
Corporate Distribution Exelon Generation Company, LLC Via e-mail Braidwood Distribution Exelon Generation Company, LLC Via e-mail Byron Distribution Exelon Generation Company, LLC Via e-mail Dwain W. Alexander, Project Manager Westinghouse Electric Company Via e-mail Howard A. Learner Environmental Law and Policy Center of the Midwest Via e-mail Byron Resident Inspector U.S. Nuclear Regulatory Commission Via e-mail Ms. Lorraine Creek RR 1, Box 182 Manteno, IL 60950 Chairman, Ogle County Board Post Office Box 357 Oregon, IL 61061 Mrs. Phillip B. Johnson 1907 Stratford Lane Rockford, IL 61107 Attorney General Via e-mail Illinois Emergency Management Agency Division of Nuclear Safety Via e-mail Braidwood Resident Inspector U.S. Nuclear Regulatory Commission Via e-mail County Executive Via e-mail Ms. Bridget Little Rorem Appleseed Coordinator Via e-mail Chairman Will County Board of Supervisors Will County Board Courthouse Joliet, IL 60434 Mr. Barry Quigley 3512 Louisiana Rockford, IL 61108