ML080080406

Request for Additional Info for the Review of the Vogtle Electric Generating Plant, Units 1 and 2, LRA

ML080080406
Person / Time
Site:	Vogtle
Issue date:	01/22/2008
From:	Ashley D NRC/NRR/ADRO/DLR
To:	Tynan T Southern Nuclear Operating Co
ASHLEY D NRR/DLR/RLRA 415-3191
References
Download: ML080080406 (7)

Text

January 22, 2008 Mr. Tom E. Tynan Vice President - Vogtle Southern Nuclear Operating Company, Inc.

7821 River Road Waynesboro, GA 30830

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION

Dear Mr. Tynan:

By letter dated June 28, 2007, Southern Nuclear Operating Company, Inc., submitted an application pursuant to 10 CFR Part 54, to renew the operating licenses for Vogtle Electric Generating Plant, Units 1 and 2, for review by the U.S. Nuclear Regulatory Commission (NRC or the staff). The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review. Further requests for additional information may be issued in the future.

Items in the enclosure were discussed with Chalmer Myer, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me at 301-415-3191 or e-mail DJA1@nrc.gov.

Sincerely,

/RA/

Donnie J. Ashley, Sr. Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket Nos. 50-424 and 50-425

Enclosure:

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

ML080080406 OFFICE LA:DLR PM:RPB1:DLR BC:RPB1:DLR NAME IKing DAshley LLund DATE 01/16/08 01/16/08 01/22/08 VOGTLE ELECTRIC GENERATING PLANT (VEGP), UNITS 1 AND 2 LICENSE RENEWAL APPLICATION (LRA)

REQUEST FOR ADDITIONAL INFORMATION (RAI)

RAI-4.2.3-1 We have reviewed the information on page 4.2-10 of the Application for License Renewal and have noted a number of errors, the most significant of which is the numerical temperature value

(-226°F) for the RTPTS for the Inlet Nozzle to Nozzle Shell Course Weld 105-121D shown in Table 4.2.3-1. This value appears to be inconsistent with the other data provided for the material. In addition, notes B and E to this Table 4.2.3-1 appear to be incorrect and, in particular, inconsistent with Note F to the same table. Please correct the information on page 4.2-10 and resubmit this page.

RAI 4.3-1 LRA Section 4.3.1 states, In addition to the original design transients, fatigue loading transients and issues subsequently identified are not parts of the original fatigue analyses. For the lower pressurizer head and surge line, thermal stratification and insurge/outsurge transients are evaluated (IEB 88-11).

a.) The evaluation result for the pressurizer lower head was not discussed in LRA Section 4.3.1.4, Thermal Stratification of the Surge Line and Lower Pressurizer Head (IEB 88-11). Please provide the limiting 60-year projected cumulative usage factor (CUF) value for the pressurizer lower head.

b.) An NRC safety evaluation entitled Vogtle Unit 1 Safety Evaluation on Pressurizer Surge Line Thermal Stratification, dated April 12, 1990, states, Applicant committed to revise applicable operating procedures to limit the system delta T (between the pressurizer head and the reactor coolant loop) for reactor coolant system (RCS) heatup (HU) to 320°F and RCS cool down (CD) to 300°F. The revised heatup and cooldown procedures ensure consistency between actual plant operation and the surge line analysis assumption.

Please discuss the procedures that have been used by VEGP and demonstrate the consistency between the recorded plant operational transient data and the assumptions (delta T limits of 320°F & 300°F during HU/CD) that were made and used in the surge line and pressurizer lower head thermal stratification analyses.

RAI 4.3-2 Section 4.3.1.5.3 of the LRA includes the following assessment for the environmentally-assisted fatigue analysis for the surge line hot leg nozzle:

The maximum design CUF determined from the ASME Code fatigue analysis for the VEGP surge line hot leg nozzle is 0.95. Applying the maximum Fen for stainless steel from ENCLOSURE

NUREG/CR-5704 (Ref. 10) of 15.35 increases the maximum CUF value to an environmental fatigue adjusted value of 14.58, which is greater than 1.0. Therefore, a different demonstration method is used.

Using fatigue monitoring software, cooldown/heatup cycles for Unit 1 and Unit 2 from 6/30/95 through 10/9/05 were analyzed to determine the average CUF per HU/CD cycle. Using this data, SNC has shown that at the surge line hot leg nozzle the projected CUF for 200 HU/CD cycles is 0.00534 for Unit 1 and 0.00628 for Unit 2.

Please discuss the changes in the heatup and cooldown procedures, specifically the implementation of the modified operating procedure (MOP), to mitigate pressurizer insurge/outsurge transients. Please explain how the impacts of MOP were factored into the calculation of the average CUF per HU/CD in the environmentally assisted fatigue analysis.

RAI 4.3-3 LRA Sections 4.3.1.5.4 and 4.3.1.5.5 state that the average Fen for the charging nozzle and safety injection nozzle is 7.6 and 5.535, respectively. In response to the audit question 4.3-5, VEGP stated that Fen values for normal charging, alternative charging and safety injection nozzles were computed from the actual plant events using an integrated strain rate (ISR) method as defined in the EPRI Report TR-1003083, Guidelines for Assessing Fatigue Environmental Effects in a License Renewal Application (MRP-47 Revision 1). ISR method calculates one Fen for one transient pair. Both the charging nozzle and safety injection nozzle were designed to several thermal transients. Please justify how one average Fen value per nozzle could be used for more than one transient pair having significant contribution to the CUF.

RAI 4.3-4 LRA Section 4.3.1.5 stated that the environmentally-assisted fatigue on the surge line hot leg nozzle, and charging nozzle was evaluated using fatigue monitoring software.

Please provide the benchmarking of the software using relevant transient data, proper 3-D model (cylinder to cylinder), and NRC endorsed computer code ANSYS. Please justify the use of the fatigue monitoring software to update the CUF calculation by using the monitored or projected transient data (cycles) and discuss the conservatisms in the calculation on a plant-specific basis.

Steam Generator Program - Tube Integrity The nominal tube wall thickness is 0.040. The Steam Generator Program - Tube Integrity requires steam generator tubes be plugged if they have a 40 percent degradation from the nominal wall thickness (0.040*.4 = 0.016 or a wall thickness less than 0.024). Based on the results of specific analysis for allowable tube wall thinning for the VEGP Model F steam generator tubes under normal operating and accident loadings, a minimum wall thickness of 0.014 is necessary to satisfy the stress limits of Regulatory Guide 1.121. The minimum inspection-acceptable wall thickness for new tubes is 0.039.

The assumed general wall loss due to corrosion and erosion over 40 years is 3 mils, which reduces the tube wall thickness to 0.036. The corrosion rate of 3 mils is based on a conservative weight-loss rate for Inconel tubing in flowing 650°F primary side reactor coolant fluid. The weight loss, when equated to a thinning rate and projected over a 40-year design objective with appropriate reduction after initial hours, is equivalent to 0.083-mils thinning. The assumed corrosion rate of 3 mils allows a conservative 2.917 mils for general corrosion thinning on the secondary side. Increasing the assumed corrosion rate by 50 percent from 3 mils to 4.5 mils has no effect on tube plugging criteria.

RAI 4.7.3-1 The LRA Section 4.7.3 states that 3 mils is the corrosion rate, not the general wall loss over 40 years. Please confirm that the 3 mils being referenced is the total wall loss over the 40 year design life.

RAI 4.7.3-2 The LRA Section 4.7.3 implies that the entire 3 mils of wall loss will be consumed from the primary side of the tube only, which appears to conflict with the statement that there are 2.917 mils of allowance for general corrosion thinning on the secondary side. Please confirm that the 3 mils of wall loss accounts for wall loss on both the primary and secondary sides of the tubes.

If this is not the case, please indicate what the corrosion allowances are for the primary and secondary sides of the tubes. Also, please provide a basis for stating that a 2.917 mils allowance for general corrosion thinning on the secondary side is conservative (e.g., discuss the operating experience of the Vogtle Electric Generating Plant regarding secondary side general thinning of steam generator tubes).

RAI B.3.14-1 In the License Amendment Request, Appendix B, Section B.3.14, Nickel Alloy Management Program for Non-Reactor Vessel Closure Head Penetration Locations, it is stated that currently, management of primary water stress-corrosion cracking in nickel alloys is a rapidly evolving area and as a result, program attributes have not yet been finalized. Further, where industry guidance has been developed, there are ongoing efforts to reach acceptable resolution of NRC staff concerns which may alter program requirements. Therefore, assessments for each of the ten aging management program elements are not included for this program.

Please identify when the assessments of each of the ten aging management program elements will be provided.

Letter to T. Tynan from D. Ashley, dated January 22, 2008 DISTRIBUTION:

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION HARD COPY DLR RF E-MAIL:

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Vogtle Electric Generating Plant, Units 1 and 2 cc:

Mr. N. J. Stringfellow Arthur H. Domby, Esquire Manager, Licensing Troutman Sanders Southern Nuclear Operating Company, Inc. Nations Bank Plaza P.O. Box 1295 600 Peachtree Street, NE Birmingham, AL 35201-1295 Suite 5200 Atlanta, GA 30308-2216 Mr. Jeffrey T. Gasser Executive Vice President Resident Inspector Southern Nuclear Operating Company, Inc. Vogtle Plant P.O. Box 1295 8805 River Road Birmingham, AL 35201-1295 Waynesboro, GA 30830 Mr. Steven M. Jackson Office of the County Commissioner Senior Engineer - Power Supply Burke County Commission Municipal Electric Authority of Georgia Waynesboro, GA 30830 1470 Riveredge Parkway, NW Atlanta, GA 30328-4684 Mr. Stanford M. Blanton, Esq.

Balch & Bingham, LLP Mr. Reece McAlister P.O. Box 306 Executive Secretary Birmingham, AL 35201 Georgia Public Service Commission 244 Washington Street, SW Ms. Moanica M. Caston Atlanta, GA 30334 Vice President and General Counsel Southern Nuclear Operating Company, Inc.

Mr. Harold Reheis, Director 40 Inverness Center Parkway Department of Natural Resources P.O. Box 1295 205 Butler Street, SE, Suite 1252 Birmingham, AL 35201-1295 Atlanta, GA 30334 Chalmer R. Myer Attorney General Southern Nuclear Operating Company, Inc.

Law Department 40 Inverness Center Parkway 132 Judicial Building P.O. Box 1295 Atlanta, GA 30334 Birmingham, AL 35201-1295 Mr. Laurence Bergen Oglethorpe Power Corporation 2100 East Exchange Place P.O. Box 1349 Tucker, GA 30085-1349