ML061280407
| ML061280407 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/05/2006 |
| From: | Pace P Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML061280407 (6) | |
Text
May 5, 2006 10 CFR 50.46 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of
)
Docket Nos. 50-327 Tennessee Valley Authority
)
50-328 SEQUOYAH NUCLEAR PLANT (SQN) - 10 CFR 50.46 30-DAY SPECIAL REPORT OF SIGNIFICANT CHANGES
Reference:
TVA letter to NRC dated November 8, 2004, 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 fuel cladding temperature resulting from recent changes to the SQN Emergency Core Cooling System (ECCS) evaluation model. This submittal satisfies the 30-day special reporting requirements in accordance with 10 CFR 50.46(a)(3)(ii). Enclosure 1 contains a summary of the recent changes to the SQN ECCS evaluation model and the effect of these changes on the calculated peak fuel cladding temperature. Enclosure 2 provides a detailed discussion of the changes to the large break loss-of-coolant accident summarized in Enclosure 1.
These changes have occurred since the last annual report that was submitted by the above reference.
U.S. Nuclear Regulatory Commission Page 2 May 5, 2006 As one of the changes to the ECCS evaluation model resulting in the predicted PCT exceeding 2200 degrees F, reportability was evaluated in accordance with 10 CFR 50.46 (a)(3)(ii). An evaluation reveled that neither unit was in an unanalyzed condition that significantly compromised plant safety at the time of discovery or in past plant operation. Therefore, no 10 CFR 50.72 or 10 CFR 50.73 reports were made.
As detailed in the enclosures, immediate steps were taken to demonstrate compliance with 10 CFR 50.46. Long-term plans are in place to perform a reanalysis of the large break loss-of-coolant accident to support the spring 2008 Unit 2 refueling outage.
There are no regulatory commitments in this letter.
Please direct questions concerning this issue to J. D. Smith at (423) 843-6672.
Sincerely, Original signed by James D. Smith for:
P. L. Pace Manager, Site Licensing and Industry Affairs Enclosures cc (Enclosures):
Mr. Edgar D. Hux 94 Ridgetree Lane Marietta, Georgia 30068 Mr. Douglas V. Pickett, Senior Project Manager U.S. Nuclear Regulatory Commission Mail Stop 08G-9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739 Mr. William T. Russell 400 Plantation Lane Stevensville, Maryland 21666
E1-1 ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY (TVA)
SEQUOYAH NUCLEAR PLANT (SQN)
UNITS 1 AND 2 SEQUOYAH NUCLEAR PLANT (SQN) - 10 CFR 50.46 30-DAY SPECIAL REPORT OF SIGNIFICANT CHANGES In accordance with the 30-day special reporting requirements of 10 CFR 50.46 (a)(3)(ii), the following is a summary of recent changes to the Sequoyah Emergency Core Cooling System (ECCS) evaluation model and the effect of these changes on the calculated peak fuel cladding temperature.
Large Break Loss-of-Coolant Accident (LB LOCA)
PCT Enclosure Previous Licensing Basis Peak Cladding 2157°F Temperature (PCT)
(Reported November 8, 2004)
- 1. Reactor Coolant Pump Internals
+183°F 2 Flow Resistance Model Error
- 2. Maximum Core Power Peaking Factor
-145°F 2 Reduction from 2.50 to 2.39 Updated Licensing Basis PCT 2195°F Net Change
+38°F Small Break Loss-of-Coolant Accident (SB LOCA)
PCT Previous Licensing Basis PCT 1162°F (Reported November 8, 2004)
Updated Licensing Basis PCT 1162°F Net Change None
E2-1 ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY (TVA)
SEQUOYAH NUCLEAR PLANT (SQN)
UNITS 1 AND 2 SEQUOYAH NUCLEAR PLANT (SQN) - 10 CFR 50.46 30-DAY SPECIAL REPORT OF SIGNIFICANT CHANGES REACTOR COOLANT PUMP INTERNALS FLOW RESISTANCE MODELING ERROR
=
Background===
During a recent review of the SQN plant-specific emergency core cooling system (ECCS) evaluation model, an error was discovered in the modeled reactor coolant pump characteristics. As indicated in Section 4.3.4.5 of Topical Report No. BAW-10168P-A, Revision 03, BWNT Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants, reactor coolant pump characteristics included in the model are based on homologous pump relationships established by pump testing. The RELAP5/MOD2-B&W computer code used in the SQN evaluation model contains a standard set of reactor coolant pump homologous curves for various reactor coolant pump models. A review of the curves in the standard computer code data package found that the curves for the Westinghouse Model 93 pump were incorrectly identified as being applicable to the Model 93A pumps. Due to this data mislabel, homologous curves for a Westinghouse Model 93 reactor coolant pump were used in the SQN plant specific evaluation model. The actual pumps installed at SQN Units 1 and 2 are Westinghouse Model 93A pumps.
Correcting the error changes the performance of the reactor coolant pumps in the evaluation model. The bounding effect of the reactor coolant pump modeling error involves the calculated peak cladding temperature (PCT) for the limiting transient (large break loss-of-coolant accident) analyzed with the evaluation model. In accordance with the requirements of 10 CFR 50, Appendix K, the SQN evaluation model assumes a reactor coolant pump locked rotor condition whenever it is required to produce limiting analysis results. During the large break loss-of-coolant accident, limiting results are produced when the reactor coolant pumps are locked during the refill/reflood portion of the transient. For this limiting condition, the locked rotor acts as a flow restriction in the reactor coolant piping and is modeled as a form loss in the primary loop geometry. The form loss coefficient for the locked reactor coolant pump internals is generated from the homologous curves for the specific model pump. Using the correct homologous curve for the SQN reactor coolant pump model increased the form loss coefficient by approximately 160 percent. This change increased the flow resistance for steam
E2-2 exiting the core and increased the pressure in the reactor vessel during the refill/reflood portion of the loss-of-coolant transient. This higher pressure results in a reduction of ECCS flow to the core which increases the calculated PCT of the limiting fuel pin.
Results The SQN large break loss-of-coolant accident was reanalyzed for the limiting fuel elevation with the corrected reactor coolant pump characteristic data. The change in the reactor coolant pump characteristics resulted in a calculated fuel peak clad temperature increase of 183ºF.
As this increase resulted in a calculated maximum fuel peak clad temperature in excess of 2200ºF for the existing analysis limits, a second analysis of the limiting fuel elevation was performed with the corrected ECCS evaluation model. This analysis included a reduction in the maximum core power peaking limit (FQ) from 2.50 to 2.39. The change in the maximum core power peaking reduced the calculated peak clad temperature by 145ºF (to 2195ºF). Based on this result, SQN has adopted a revised maximum core power peaking limit of 2.39. Revised axial flux difference (AFD) limits have been established for present and future operating cycles to ensure that the 2.39 power peaking limit is met. Adequate core maneuvering margins exist to support the revised core design and operating limits.
Additional Information As a result of the ECCS evaluation model error described above, the calculated fuel peak clad temperature exceeded the established limit for the assumed analysis limits. While a change to the power peaking design and operating limits will support present and future operation under the existing ECCS evaluation model of record, the error in the evaluation model has existed since the implementation of a Technical Specification change for conversion to Framatome fuel beginning with Unit 1, Cycle 9 operation.
To evaluate past operation, a qualitative comparison of the SQN deterministic evaluation methodology was made to results obtained using the Framatome realistic evaluation methodology.
These qualitative evaluations have shown the current deterministic analysis results are significantly more conservative (200ºF to 400ºF in terms of calculated peak clad temperature) than results obtained with the realistic methodology for similar power levels and power peaking limits.
Based on this result, past operation with the original core design and operating limits for power peaking would not have resulted in calculated peak cladding temperatures in excess of the 2200ºF limit. TVA is currently working with Framatome to apply the realistic evaluation methodology to SQN as the
E2-3 analysis of record. It is anticipated that the plant specific application of the realistic evaluation methodology will be submitted on a schedule to support Unit 2, Cycle 16 operation in the spring of 2008.