NL-04-0371, Clarification of the Proposed Technical Specifications Revisions to Primary Containment Leakage Rate Testing Program

Clarification of the Proposed Technical Specifications Revisions to Primary Containment Leakage Rate Testing Program

ML040720423
Person / Time
Site:	Hatch
Issue date:	03/10/2004
From:	Sumner H Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-04-0371
Download: ML040720423 (3)

Text

H.L Sumner, Jr. Southern Nuclear Vice President Operating Company, Inc.

Hatch Project Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7279 SOUTHERNAKuuu 4 March 10, 2004 COMPANY Energy to Serve YourWorld"'

Docket Nos.: 50-321 NL-04-0371 50-366 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin 1. Hatch Nuclear Plant Clarification of the Proposed Technical Specifications Revision to Primary Containment Leakage Rate Testing Program Ladies and Gentlemen:

By letter dated December 1, 2003 Southern Nuclear Operating Company (SNC) submitted to the NRC a proposed change to the Unit I and Unit 2 Technical Specifications (TS) for the Edwin 1. Hatch Nuclear Plant. The amendment request proposes a change in the post-accident peak primary containment pressure (Pa) listed in TS section 5.5.12, "Primary Containment Leakage Rate Testing Program." This proposed change supports other efforts by SNC to increase the reactor nominal operating pressure for the Hatch Units. The containment evaluation performed for the pressure increase effort resulted in slightly higher post-accident peak calculated containment pressure values. Since the peak calculated containment pressures are explicitly listed in the Administrative section of the TS, a TS change is required.

Through a teleconference conversation with the NRC/NRR Hatch Project Manager, a request was made for SNC to provide a correspondence describing the additional effort associated with the pressure increase project. Also, an electronic communication (E-mail) was received by SNC requesting responses to three questions from a staff reviewer pertaining to the license amendment request (LAR). This letter provides a response to question I of the E-mail, as well as a more detailed description of the pressure increase project scope as requested. The answers to questions 2 and 3 of the E-mail request will be provided under a separate cover as a Request for Additional Information (RAI) response.

NRC Question (1) While the revisedreactorsteam dome pressure is stated to be iithin tile TS 3.4.10 allowable of 1073 psia, it would appearthat changes to the reactorsteam dome nominal operatingpressure would also impact other safety analyses, for example ECCS performance, and well as, perhaps, other operatingconsiderationsbased on reduced marginsto safety set points. (a) How are these to be addressed? (b, Is the LAR requestingapproval of the change to the reactorsteam dome nominal operatingpressure value?

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U. S. Nuclear Regulatory Commission NL-04-0371 Page 2 SNC Response The December 1, 2003 submittal requests NRC review and approval of the change to the Primary Containment Leakage Rate Testing Program only. All other analyses and evaluations in support of the nominal operating pressure increase project are performed under the provisions of 10 CFR 50.59.

The purpose of the pressure increase project is to increase the nominal reactor steam dome pressure from 1035 psig to 1045 psig to allow for additional flow control margin for the high pressure turbine. This flow margin is needed to operate at the 100% rated thermal power level of 2804 MWt. In order to implement the increase in nominal reactor steam dome pressure, a review of the applicable BOP and NSSS systems, structures, analyses, transients and special events was performed. The impact of the pressure increase on the applicable plant programs (i.e., MOV testing, EQ, FAC, Appendix J testing, etc.) was also evaluated. The results of the analyses and evaluations indicate that, except for the subject proposed TS revision request, the 10 psi nominal operating pressure increase can be accomplished under the provision of 10 CFR 50.59 since:

• A nominal operating pressure of 1045 psig is within the existing Reactor Steam Dome Pressure TS LCO (3.4.10) requirement of 1058 psig. The basis for this TS (the main steam isolation valve closure with flux scram) is not affected by this change because it is performed with an initial dome pressure of 1058 psig.

• There is no change to any TS allowable values associated with instrument settings that initiate protective functions. This includes:

1. No change to the reactor pressure vessel (RPV) steam dome pressure trip Analytical Limit or TS Allowable Values

2. No change to the RPV water level trip Analytical Limits or TS Allowable Values

• There is no change in the rated thermal power of 2804 MWt.

• The 10 psi increase in nominal reactor dome pressure will not result in a change to any Safety Relief Valve's (SRV) setpoints.

• The Maximum Reactor Core Flow and Power-to-Flow map are unchanged with the 10 psi increase.

The scope on pressure increase engineering project evaluation effort included but was not limited to the following:

• Turbine and Reactor Heat Balances

• Core and Fuel Performance

• Thermal-Hydraulic Stability

• SRV Performance

• RPV Fracture Toughness

• Reactor Vessel Integrity-Stress Evaluation

• RPV Internals Mechanical Evaluation

• Reactor Internal Differential & Fuel Lift

• Containment System Response

• Transient Analyses

• Accident Radiological Analysis

• Anticipated Transients Without Scram (ATWS)

• Station Blackout

• NSSS System

-s U. S. Nuclear Regulatory Commission NL-04-037 1 Page 3

. BOP System U Instrumentation Review U

Grid Stability Plant implementation of the 10 psi nominal operating pressure increase will be accomplished through the SNC design change process. Implementing the pressure change via the design change process ensures that the necessary changes to the plant operating procedures, maintenance procedures, instrument calibrations, as well as operator training requirements are captured. Following completion of site implementation, the affected FSAR sections will be submitted under the provisions of 10 CFR 50.71(e) to reflect the changes due to the 10 psi nominal operating pressure increase.

Mr. H. L. Sumner, Jr. states lie is a Vice President of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.

This letter contains no NRC commitments. If you have any questions, please advise.

Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY

-H. L. Sumner, Jr.

-Sworn to andsubscribedbefore me this 10 day of 9M CACl'-, 2004.

KtIhc ALfOu,

','-Notary Pibl

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HLS/whc/daj cc: Southern Nuclear Operating Company Mr. J. B. Beasley, Jr., Executive Vice President Mr. G. R. Frederick, General Manager - Plant Hatch RType: CHAO2.004 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. C. Gratton, NRR Project Manager - Hatch Mr. D. S. Simpkins, Senior Resident Inspector - Hatch State of Georgia Mr. L. C. Barrett, Commissioner - Department of Natural Resources