ML081570674

Sequoyah, Units 1 and 2 - 10 CFR 50.46 - 30-Day Special Report of Significant Changes

ML081570674
Person / Time
Site:	Sequoyah
Issue date:	06/04/2008
From:	Smith J D Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML081570674 (3)

Text

June 4, 2008

10 CFR 50.46

U.S. Nuclear Regulatory Commission

ATTN: Document Control Desk

Washington, D.C. 20555-0001

Gentlemen:

In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority (TVA) ) 50-328

SEQUOYAH NUCLEAR PLANT (SQN) - 10 CFR 50.46 DAY SPECIAL REPORT OF SIGNIFICANT CHANGES

Reference:

TVA letter to NRC dated November 14, 2007, "Sequoyah Nuclear Plant (SQN) - 10 CFR 50.46 Annual Report of Non-Significant Changes"

The purpose of this letter is to provide changes to the calculated peak cladding

temperature (PCT) resulting from recent changes to the SQN emergency core cooling

system (ECCS) evaluation model. This submi ttal satisfies the reporting requirements in accordance with 10 CFR 50.46(a)(3)(ii). The enclosure contains a summary of the recent

changes to the SQN Units 1 and 2 ECCS evaluation model and the affect of these

changes on the calculated PCT. The changes result in an absolute calculated peak clad

temperature change in excess of 50 degrees Fahrenheit from that reported in the last

annual report.

There are no regulatory commitments in this letter. Please direct questions concerning

this issue to me at (423) 843-7170.

Sincerely, Original signed by:

James D. Smith

Manager, Site Licensing and

Industry Affairs

U.S. Nuclear Regulatory Commission Page 2 June 4, 2008

cc (Enclosure):

Mr. Brendan T. Moroney, Senior Project Manager

U.S. Nuclear Regulatory Commission

Mail Stop 08G-9a

One White Flint North

11555 Rockville Pike

Rockville, Maryland 20852-2739

E1 ENCLOSURE TENNESSEE VALLEY AUTHORITY (TVA)

SEQUOYAH NUCLEAR PLANT (SQN)

UNITS 1 AND 2 10 CFR 50.46 SPECIAL REPORT OF SIGNIFICANT CHANGES In accordance with the reporting requirements of 10 CFR 50.46 (a)(3)(ii), the following is a

summary of the limiting design basis accident (loss-of-coolant accident) (LOCA) analysis results

established using the current SQN emergency core cooling system (ECCS) evaluation model.

Small Break LOCA (SB LOCA)

PCT Previous Licensing Basis PCT 1162 degrees Fahrenheit (F)

(November 08, 2004)

Reanalysis for revised ECCS pump +241 degrees F

performance and core power peaking

analytical input assumptions.

Updated Licensing Basis PCT 1403 degrees F

Net Change +241 degrees F

The SQN large break LOCA (LB LOCA) has been recently analyzed using the realistic (LB LOCA) methodology described in Topical Report No. EMF-2103, Revision 00, "Realistic

Large Break LOCA Methodology for Pressurized Water Reactors." A number of modified

analytical input parameters were incorporated into the realistic LB LOCA analysis to support

improved fuel utilization and expand the operating margin for the ECCS pumps.

For consistency with the realistic large break LOCA analysis, the SQN SB LOCA analysis has recently been analyzed to apply similar changes to the SB LOCA analytical input parameters.

The analysis was performed using the same SQN plant-specific evaluation model with the same

evaluation methodology (i.e., Topical Report No. BAW-10168P-A, Revision 03, "BWNT Loss-of-

Coolant Accident Evaluation Model for Recirculat ing Steam Generator Plants - Volume II - Small Break") as the current analysis of record. Specific changes to the SB LOCA analytical input

parameters include 1) an increase in the core power peaking factor (Fq) from 2.5 to 2.65, 2) an

increase in the hot channel enthalpy factor (fh) from 1.70 to 1.89, and 3) a 5 percent reduction in the minimum developed head values for the ECCS charging (high head) and safety injection (intermediate head) pumps.

Results The SB LOCA analysis with the revised analytical input parameters discussed above meet the

10 CFR 50.46 acceptance criteria. The limiting calculated fuel cladding temperature was

determined to be 1403 degrees F for a 2.75-inch diameter break size. This result represents a

net increase in the calculated peak clad temperature from the previous analysis of record of

241 degrees F.