ML020380576
| ML020380576 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 02/07/2002 |
| From: | Herman R NRC/NRR/DLPM |
| To: | Scalice J Tennessee Valley Authority |
| Hernan R W, NRR/NRC/DLPM,301-415-2010 | |
| References | |
| TAC MB3435, TAC MB3436 | |
| Download: ML020380576 (4) | |
Text
February 7, 2002 Mr. J. A. Scalice President, TVA Nuclear and Chief Nuclear Officer Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801
SUBJECT:
SEQUOYAH NUCLEAR PLANT (SQN), UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION ON TECHNICAL SPECIFICATION CHANGE NO.
01-08, INCREASE MAXIMUM ALLOWED REACTOR POWER LEVEL TO 3455 MEGA-WATT THERMAL (MWt) (TAC NOS. MB3435 AND MB3436)
Dear Mr. Scalice:
The subject Technical Specification Amendment Request was submitted to the U.S. Nuclear Regulatory Commission (NRC) for review and approval on November 15, 2001, by the Tennessee Valley Authority (TVA). The proposed license amendment would increase the full core thermal power rating by 1.3 percent from 3411 MWt to 3455 MWt, based on planned installation of the improved Caldon, Incorporated Leading Edge Flow Meter feedwater flow measurement instrumentation. The NRC staff is in the process of reviewing TVAs submittal.
As discussed during a conference call on February 7, 2001, the NRC staff requires responses to the enclosed Request for Additional Information to proceed with its review. Following the call, Mr. Kieth Weller of the SQN Licensing Staff stated that TVA would respond to this request by March 11, 2001.
Please have your staff contact me at (301) 415-2010 if there are any questions regarding the enclosed request.
Sincerely,
/RA/
Ronald W. Hernan, Senior Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328
Enclosure:
Request for Additional Information cc w/enclosure: See next page
Mr. J. A. Scalice February 7, 2002 President, TVA Nuclear and Chief Nuclear Officer Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801
SUBJECT:
SEQUOYAH NUCLEAR PLANT (SQN), UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION ON TECHNICAL SPECIFICATION CHANGE NO.
01-08, INCREASE MAXIMUM ALLOWED REACTOR POWER LEVEL TO 3455 MEGA-WATT THERMAL (MWt) (TAC NOS. MB3435 AND MB3436)
Dear Mr. Scalice:
The subject Technical Specification Amendment Request was submitted to the U.S. Nuclear Regulatory Commission (NRC) for review and approval on November 15, 2001, by the Tennessee Valley Authority (TVA). The proposed license amendment would increase the full core thermal power rating by 1.3 percent from 3411 MWt to 3455 MWt, based on planned installation of the improved Caldon, Incorporated, Leading Edge Flow Meter feedwater flow measurement instrumentation. The NRC staff is in the process of reviewing TVAs submittal.
As discussed during a conference call on February 7, 2001, the NRC staff requires responses to the enclosed Request for Additional Information to proceed with its review. Following the call, Mr. Kieth Weller of the SQN Licensing Staff stated that TVA would respond to this request by March 11, 2001.
Please have your staff contact me at (301) 415-2010 if there are any questions regarding the enclosed request.
Sincerely,
/RA/
Ronald W. Hernan, Senior Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328
Enclosure:
Request for Additional Information cc w/enclosure: See next page DISTRIBUTION:
PDII-2 Reading File OGC J. Calvo PUBLIC R. Correia J. Wermiel RHernan B. Clayton M. Shuaibi PFredrickson W. Bateman ACRS GGeorgiev SSun NTrehan ADAMS Acc. No. ML020380576 OFFICE PDII-2\PM PDII-2\LA PDII-2\SC NAME RHernan BClayton RCorreia DATE 02/07/02 02/07/02 02/07/02 OFFICIAL RECORD COPY
REQUEST FOR ADDITIONAL INFORMATION 1.3 PERCENT RATED POWER INCREASE SEQUOYAH NUCLEAR PLANT DOCKET NOS. 50-327 and 50-328 Request for Additional Information for the Review of Sequoyah Power Uprate Applications
- 1. Provide results of an Anticipated Transient Without Scram (ATWS) analysis demonstrating that the plant at power uprate conditions is within the bounds considered by the staff during your documentation of compliance with the ATWS rule (Code of Federal Regulations at 10 CFR 50.62.) For your power uprating discuss and justify that the assumptions for the ATWS analysis are adequate as they relate to input parameters such as the initial power level, moderator reactivity feedback, safety relief valves capacity and auxliary feedwater supply. The submittal should include a discussion and applicable values of the unfavorable exposure time, if any, and ATWS core damage frequency.
- 2. Page 3-10 of Westinghouse Topical Report WCAP-15726 discusses use of the BWCMV-A critical heat flux (CHF) correlation and statistical core design (SCD) methodology for the departure from nucleate boiling ratio (DNBR) reanalysis. The licensee is requested to list the titles of topical reports that document the CHF correlation and SCD methodology, and reference the associated U.S. Nuclear Regulatory Commission (NRC) acceptance letters to confirm the acceptance of the correlation and SCD methodology used in the DNBR reanalysis for power uprate applications. Provide a discussion to address the compliance with each of limitations and restrictions specified in NRC safety evaluations for the applicable topical reports.
- 3. As stated in the NRC Standard Reivew Plan, one of the acceptance criteria for the transient analysis is related to the calculated DNBR. The staff finds that the information regarding the DNBR reanalysis for power uprate provided in Sections 3.3.7, 3.3.8 and 3.3.9 of WCAP-15726 involves qualitative discussion in nature. The licensee is requested to list the events with calculated DNBRs affected by power uprating and provide calculated results (such as figures showing calculated DNBRs, or margins to DBNR safety limits) for these events to show that the calculated DNBRs for power uprate conditions are within the acceptable safety limits. If the results of reanalysis are more limiting than the analysis of record, the reanalysis results should be included in the updated Final Safety Analysis Report.
- 4. Stress Corrosion Cracking of Reactor Internals Increased power is expected to increase the corrosion rates and speed up degradation of reactor internals. Identify the plant programs that are in place to periodically inspect reactor internals and discuss whether these programs are adequate to manage the projected increase of reactor internals degradation due to stress corrosion cracking and primary water stress corrosion cracking.
Enclosure
- 5. Flow Assisted Cracking (FAC)
Since the effects of FAC on degradation of carbon steel components are plant specific, the licensee needs to provide a predictive analysis methodology which must include the values of the parameters affecting FAC, such as velocity, and temperature before and after the power uprate (PU) and the corresponding changes in components wear rates due to FAC.
- 6. Indicate the degree of compliance with NRC Generic Letter 89-08, Erosion/Corrosion in Piping. This letter requires that an effective program be implemented to maintain structural integrity of high-energy carbon steel systems. Describe how this program was modified to account for the PU. If the computer code used in predicting wall thinning by FAC in this program is a generic code (e.g., CHECWORKS), specify it. However, if the code is plant specific provide its description.
- 7. Identify the predicted change of wear rates calculated by the revised code for the components most susceptible to FAC.
- 8. Will the PU have significant effect on FAC in balance of plant components? What is the value of the change in FAC wear rates?
- 9. The response to Question 2 under TXX-99105 (page E8-3 of Enclosure 8) addresses the acceptability of previously performed equipment qualification analyses for a 1.3 percent power increase. Please provide a statement that the previous analyses envelope conditions that will exist after the 1.3 percent power increase, if that is the case.
Mr. J. A. Scalice SEQUOYAH NUCLEAR PLANT Tennessee Valley Authority cc:
Mr. Karl W. Singer, Senior Vice President Mr. Pedro Salas, ManagerLicensing and Nuclear Operations Industry Affairs Tennessee Valley Authority Sequoyah Nuclear Plant 6A Lookout Place Tennessee Valley Authority 1101 Market Street P.O. Box 2000 Chattanooga, TN 37402-2801 Soddy Daisy, TN 37379 Mr. Jon R. Rupert, Vice President (Acting) Mr. D. L. Koehl, Plant Manager Engineering & Technical Services Sequoyah Nuclear Plant Tennessee Valley Authority Tennessee Valley Authority 6A Lookout Place P.O. Box 2000 1101 Market Street Soddy Daisy, TN 37379 Chattanooga, TN 37402-2801 Mr. Russell A. Gibbs Mr. Richard T. Purcell Senior Resident Inspector Site Vice President Sequoyah Nuclear Plant Sequoyah Nuclear Plant U.S. Nuclear Regulatory Commission Tennessee Valley Authority 2600 Igou Ferry Road P.O. Box 2000 Soddy Daisy, TN 37379 Soddy Daisy, TN 37379 Mr. Lawrence E. Nanney, Director General Counsel Division of Radiological Health Tennessee Valley Authority Dept. of Environment & Conservation ET 11A Third Floor, L and C Annex 400 West Summit Hill Drive 401 Church Street Knoxville, TN 37902 Nashville, TN 37243-1532 Mr. Robert J. Adney, General Manager County Executive Nuclear Assurance Hamilton County Courthouse Tennessee Valley Authority Chattanooga, TN 37402-2801 6A Lookout Place 1101 Market Street Ms. Ann P. Harris Chattanooga, TN 37402-2801 341 Swing Loop Road Rockwood, Tennessee 37854 Mr. Mark J. Burzynski, Manager Nuclear Licensing Tennessee Valley Authority 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801