ML20011E286
| ML20011E286 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 02/02/1990 |
| From: | Cockfield D PORTLAND GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9002130114 | |
| Download: ML20011E286 (3) | |
Text
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David W. Cockfield Vice President, Nuclear L
February 2, 1990 Trojan Nuclear Plant Docket 50-344 License NPF-1 U.S. Nuclear Regulatory Commission Attn Document Control Desk Washington DC 20555 Dear Sirst Annual Report of Emergency Core Cooling System (ECCS) Model and Application changes The reporting requirement of Titic 30, Code of Federal Regulations.
Part 50 Section 46 (10 CFR 50.46) (a)(3)(1) requires the holder of an operating license to "ostimato the offect of any change to or error in an acceptabic evaluation model or in the application of such a model to determine if the change or error is significant". The attached tabic summarizes the offects of model changes, errors, and plant changes on the Peak Cladding Temperature (PCT) of the limiting large and small break Loss-of-Coolant Accidents (LOCAs).
The magnitudo of large break LOCA changes total less than 50'F and are not significant. The magnitude of the small break LOCA changes total to more than 50*F above the limiting PCT using the past acceptable model (large break), however the overall PCT is still below the PCT limit of 2200*F.
Thus the small break LOCA change is significant, but within acceptr.ble limits.
For significant changes or errors, Item (a)(3)(11) of 10 CFR 50.46 requires submittal of a proposed schedule (within 30 days) for providing reunalyces or taking other actions. This submittal also satisfies this requirement.
Reanalysis is not required because the attached results demonstrate that the requirements of 10 CFR 50.46 are satisfied.
The small break LOCA analysis is now limiting at Trojan. As described in the Final Safety Analysis Report (FSAR) this event is au equivalent 3-inch cold-leg pipe break and is bounded by analyson of 2-and 4 inch brouks.
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i-W M MC0lTINV Document Control Desk February 2, 1990 Pano 2 To sunenarise, the sinall break LOCA results have changed significantly but remain within the acceptable limits of 10 CFR 50.46 acceptance criteria.
This letter satisfies the reporting requirements of 10 CFR 50.46.
Sincerely,
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I Attachment I
c Mr. John B. Martin Regional Administrator, Region V U.S. Nuclear Regulatory Commission Mr. David Stewart-Smith State of Oregon
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Department of Energy f
Mr. R. C. Barr WRC Resident Inspector Trojan Nuclear Plant i
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l o-4 Trojan Nucloac Plant Document Control. Dest Docket 50-344 February 2,1990 License NPP-1 Attachment Page 1 of 1 1989 EMERGENCY CORE COOLING SYSTFM MODEL AND APPLICATION _CBANilE A.
Large Break Loss-of-Coolant Accident (LOCA)
Touperature Final Safety Analysis Report (FSAR) Limiting Paak Cladding Temperature (PCT):
1983*F Modifications to Evaluation Model 10'F Pressuriser Pressure Trip Detpoint(1) 0'F
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Reactor Coolant System (RCS) Loop Temperature imbalance (2) 0*F Reduced Residual Heat Removal.(RHR) Flow (3) 18'F t
Total 2011'F B.
Small Break LOCA FSAR Limiting Break (3-inch) PCT 1925'F Modifications t-o Evaluation Model 18'F Pressuriser Pressure Trip Setpoint(1) 4*F RCS Loop Temperature Imbalance (2) 40*F Reduced RHR Flow (3)
O'F Auxiliary Feedwater Water Purge Delay (4) 86*F Total (New Limiting LOCA PCT) 2073*F Notest (1) Pressurizer Low Pressure Trip Setpoint in LOCA analysis reduced to be consistent with transient analysis.
(1855 + 1825 psia)
(2) An evaluation of the worst case RCS loop temperature imbalance, with one loop 4*F hotter than the others.
(3) This change covers the possibility of a miniflow valve remaining open on an inoperable RHR pump.
f (4) Effects of purging hot feedwater from feedlines not previously considered for small break LOCA.
Reference:
Letter POR-89-640 to A. N. Roller, PGE, from R. G. Perez,
)
Westinghouse, " Reporting of Emergency Core Cooling System j
Evaluation Model Revisions", dated December 5, 1989.
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