ML073320020

Request for Additional Information Regarding Realistic Large Break LOCA Analysis Methods

ML073320020
Person / Time
Site:	Sequoyah
Issue date:	11/30/2007
From:	Moroney B NRC/NRR/ADRO/DORL/LPLII-2
To:	Campbell W Tennessee Valley Authority
Bowen, Jeremy NRR/ADRO/DORL/LPL2-2
References
TAC MD6259
Download: ML073320020 (6)

Text

November 30, 2007 Mr. William R. Campbell, Jr.

Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

SEQUOYAH NUCLEAR PLANT, UNIT 2 - REQUEST FOR ADDITIONAL INFORMATION REGARDING REALISTIC LARGE BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS METHODS (TAC NO. MD6259)

Dear Mr. Campbell:

By letter dated July 26, 2007, as supplemented on October 3, 2007, the Tennessee Valley Authority (TVA), requested the addition of topical report EMF-2103P-A, Realistic Large Break LOCA [Loss-of-Coolant Accident] Methodology for Pressurized Water Reactors, to the list of approved references in Technical Specification Section 6.9.1.14.a.

In order for the staff to complete its review of the information provided, we request that TVA provide responses to the enclosed request for additional information (RAI). Based on discussions with your staff, we understand that you plan to respond to the enclosed RAI by December 21, 2007. If you have any questions about this material, please contact me at (301) 415-3974.

Sincerely,

/RA/

Brendan T. Moroney, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-328

Enclosure:

RAI cc: See next page

ML073320020 NRR-088 OFFICE LPL2-2/PM LPL2-2/PM LPL2-2/LA SCVB/BC SRXB/BC LPL2-2/BC NAME JBowen BMoroney RSola RDennig by memo GCranston By memo TBoyce DATE 11 / 29 /07 11 / 21 /07 11 / 29 /07 10/16/07 11/08/07 11/ 30 /07

William R. Campbell, Jr.

SEQUOYAH NUCLEAR PLANT Tennessee Valley Authority cc:

Mr. Ashok S. Bhatnagar Senior Vice President Nuclear Generation Development and Construction Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. James R. Douet Senior Vice President Nuclear Support Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. H. Rick Rogers Vice President Nuclear Engineering & Technical Services Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Timothy P. Cleary, Site Vice President Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, TN 37384-2000 General Counsel Tennessee Valley Authority 6A West Tower 400 West Summit Hill Drive Knoxville, TN 37902 Mr. John C. Fornicola, Manager Nuclear Assurance Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Ms. Beth A. Wetzel, Manager Corporate Nuclear Licensing and Industry Affairs Tennessee Valley Authority 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801 Mr. James D. Smith, Manager Licensing and Industry Affairs Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, TN 37384-2000 Mr. Christopher R. Church, Plant Manager Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, TN 37384-2000 Senior Resident Inspector Sequoyah Nuclear Plant U.S. Nuclear Regulatory Commission 2600 Igou Ferry Road Soddy Daisy, TN 37379 Mr. Lawrence E. Nanney, Director TN Dept. of Environment & Conservation Division of Radiological Health Third Floor, L and C Annex 401 Church Street Nashville, TN 37243-1532 County Mayor Hamilton County Courthouse Chattanooga, TN 37402-2801 Mr. Larry E. Nicholson, General Manager Performance Improvement Tennessee Valley Authority 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801 Ms. Ann P. Harris 341 Swing Loop Road Rockwood, TN 37854

Enclosure REQUEST FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATION REVISION TO COLR REFERENCE TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-328 By Tennessee Valley Authority (TVA) letter dated July 26, 200T, as supplemented on October 3, 2007, the licensee for Sequoyah Nuclear Unit Plant 2 (SQN-2) submitted a license amendment request proposing to perform SQN-2 large break loss-of-coolant accident (LBLOCA) analyses using the AREVA best estimate (BE) LBLOCA methodology described in EMF-2103, (P)(A), Revision 0, and to include that LBLOCA methodology in the SQN-2 Core Operating Limits Report (COLR).

The SQN-2 LBLOCA analyses are presented in ANP-2655 (P), Revision 0. In order for the Nuclear Regulatory Commission (NRC) staff to complete its review, some portions of ANP-2655 (P) require clarification. Please address the following issues.

1. Reactor Power - Table 3.3, Item 2.1, and its associated Footnote 1 indicate that the assumed reactor core power Aincludes uncertainties.@ The use of a reactor power assumption other than 102 percent, regardless of BE or Appendix K methodology, is permitted by Title10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix K.I.A, Required and Acceptable Features of The Evaluation Models, Sources of Heat During a LOCA.@ However, the paragraph also states: A... An assumed power level lower than the level specified in this paragraph [1.02 times the licensed power level], (but not less than the licensed power level) may be used provided...@

Please clarify what is meant by Aincludes uncertainties.@ What are the uncertainties, and how are the uncertainties included?

2. Does the version of SRELAP used to perform the computer runs assure that the void fraction is less than 95 percent, and the fuel cladding temperature is less than 900 degrees Fahrenheit (°F) before it allows rod quench?

3. Provide justification that the SRELAP rod-to-rod thermal radiation model applies to the SQN-2 core.

4. In the SQN-2 calculations, is the Forslund-Rohsenow model contribution to the heat transfer coefficient limited to less than or equal to 15 percent when the void fraction is greater than or equal to 0.9?

5. Was the downcomer model for the SQN-2 design rebenchmarked, performing sensitivity studies, assuming adequate noding in the downcomer, the water volume, the vessel wall, and other heat structures, with all heat structures= initial temperature at or greater than 1800 °F, and containment pressure less than 90 pounds per square inch absolute?

6. Were all break sizes assumed greater than or equal to 1.0 square foot?

7. Verify that the SQN-2 ICECON model is that shown in Figure 5.1 of EMF-CC-39(P)

Revision 2, ICECON: A Computer Program Used to Calculate Containment Back Pressure for LOCA Analysis (Including Ice Condenser Plants).

8. In order to conduct the review of the SQN-2 application of AREVAs realistic LBLOCA methods in an efficient manner, the NRC would like to make reference to the responses to the staffs requests for additional information that were developed for the application of the AREVA methods to the North Anna Power Station Unit Nos. 1 and 2, and found acceptable during that review (the NRC Staff safety evaluation report was published on April 1, 2004.)

The staff would like to make use of the information that was provided by the North Anna licensee that is not applicable only to North Anna or only to subatmospheric containments.

This information is contained in letters to the NRC from the North Anna licensee (Docket numbers 50-338 and 50- 339) dated September 26, 2003 and November 10, 2003. The specific responses that the staff would like to reference are:

September 26, 2003 letter: NRC Question 1 NRC Question 2 NRC Question 4 NRC Question 6 November 10, 2003 letter: NRC Question 1 Please verify that the information in these letters is applicable to the AREVA model applied to SQN-2, except for that information related specifically to North Anna, and to sub-atmospheric containments.

9. ANP-2655(P) shows that the containment parameters treated statistically are: (1) upper compartment containment volume, (2) upper compartment containment temperature, and (3) lower compartment containment temperature. ANP-2655(P) states that in many instances the guidance of NRC Branch Technical Position CSB 6-1 was used in determining the other containment parameters.

(a)

How is the mixing of containment steam and ice melt modeled so as to minimize the containment pressure?

(b)

Verify that all containment spray and fan coolers are assumed operating at maximum heat removal capacity.

(c)

Describe how the limits on the volume of the upper containment are determined.

(d)

How are the containment air return fans modeled; and what is the effect of this modeling on the containment pressure?

(e)

Describe how passive heat sink areas and heat capacities are modeled so as to minimize containment pressure.

(f)

What value is used for the initial ice mass? Does this value result in minimizing containment pressure?

10. Provide a statement confirming that TVA and its LBLOCA analyses vendor have ongoing processes that assure that the input values and ranges of parameters for SQN-2 LBLOCA analyses conservatively bound the values and ranges of those parameters for the as operated SQN-2 plant. This statement addresses certain of the programmatic requirements of 10 CFR 50.46, Section (c).