ML20113E279
| ML20113E279 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/26/1996 |
| From: | Shell R TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9607050212 | |
| Download: ML20113E279 (5) | |
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Tennessee Va ey Autcry Post othce Box 2000. Soddy Da'sv Tennessee 37379 n
June 26,1996 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of
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Docket Nos. 50-327 Tennessee Valley Authority
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50-328 SEQUOYAH NUCLEAR PLANT (SON)- 10 CFR 50.46 ANNUAL REPORT
References:
- 1. TVA letter to NRC dated July 12,1995
- 2. TVA letter to NRC dated September 6,1995,"Sequoyah Nuclear Plant (SON) - 10 CFR 50.46 Updated Annual Report" 10 CFR 50.46 requires reporting, at least on an annual basis, each change to or error discovered in an acceptable loss-of-coolant accident (LOCA) evaluation model or in the application of such a model that affects the peak clad temperature (PCT) calculation.
The purpose of this letter is to provide the annual report.
A detailed discussion of each of the two small break LOCA evaluation changes are contained in the attached enclosure. The first change is based on a review performed by Westinghouse Electric Corporation th.
entified an error in the NOTRUMP computer program. This resulted in a 2(
jree Fahrenheit (F) penalty. The second change was discovered due to a recent ador test of a flow orifice similar to one installed at SON in the charging / safety injection system. This resulted in a 12-degree F penalty. Additionally, it should be noted that the large break LOCA evaluation remains unchanged from that reported in Reference 2.
Reference 1 provided SON's 199510 CFR 50.46 Annual Report Reference 2 supplied an updated annual report as a result of L significant change identified in the PCT analytical result-s. To remain on the current update frequency, TVA is providing the subject report at this time.
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U.S. Nuclear Regulatory Commission Page 2 June 26,1996
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' Additionally potentialissues are under investigation by Westinghouse that may impact the PCT for both small and large break LOCA. The potentialissues have had l
PCT margin temporarily allocated to ensure that the cumulative efforts are tracked such that the 10 CFfi 50.46 PCT limit of 2200 degrees F is not exceeded. Upon their resolution, these issues w7tThontinue.to_be reported as appropriate.
Please direct questions concerning this issue to W. C. Ludwig at (423)-843-8460.
Sincerely, h /b-R. H. Shell Manager SON Site Licensing l
Enclosure cc (Enclosure):
Mr. R. W. Hernan, Project Manager Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike j
Rockville, Maryland 20852-2739 NRC Resident inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379-3624 Regional Administrator U.S. Nuclear Regulatory Commission Region 11 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323-2711
ENCLOSURE 10 CFR 50.46 REPORT DOCUMENTATION Laras Break Loss-of-Coolant Accident (LOCA)
ECI Attachment Previous Licensing Basis' Peak 2156oF Cladding Temperature (PCT)
(July 12,1995)
Revised Licensing Basis PCT _
1911 F (September 6,1995)
No Change identified 0F Updated Licensing Basis PCT 1911 F.
Net Change 0F Small Break LOCA ECI Previous Licensing Basis PCT 1716 F (July 12,1995)
NOTRUMP Specific Enthalpy Error
+20F 1
Charging / Safety injection
+12oF 2
Miniflow Assumption Updated Licensing Basis PCT 1748 F Net Change
+32 F A detailed discussion of the change outlined above is included in the indicated attachment.
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Attachment I-l NOTRUMP SPECIFIC ENTHALPY ERROR
Background
During.an internal review of the 1985 Westinghouse small break loss-of-cooling accident evaluation model, Westinghouse identified a typographical error in a line of computer coding for the NOTRUMP computer program. The line of code was intended to model the calculation found in Equation L-127 of Topical Report WCAP-10079-P-A. While the equation in the topical
- report is correct, the coding represented the last term of the equation as a partial derivative with respect to the fluid node mixture region total energy instead of the mixture region total mass.
Estimated Effect Based upon sensitivity calculations performed by Westinghouse, correction of the coding error results in a 20 F peak clad temperature increase for Sequoyah.
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CHARGING / SAFETY INJECTION MINIFLOW ASSUMPTION
Background
The charging / safety injection pump flow orifice installed in the minimum flow recirculation lines at Sequoyah was specified to allow 60 gpm flow at 6000 feet of differential pressure. The 60 gpm value was assumed to be the maximum charging pump recirculation flow loss in the Sequoyah small break loss-of-coolant accident analysis. Recent vendor testing of a flow orifice similar to that installed at Sequoyah indicated that the orifice assembly will allow flow in excess of 60 gpm under rated conditions. With a differential head of 6000 ft, the orifice design will allow as much as 67 gpm recirculation flow.
Estimated Effect Based upon sensitivity calculations performed by Westinghouse, the safety injection flow loss caused by the incre., sed charging pump recirculation flow will result in a 12 F peak clad temperature increase for Sequoyah.
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