ML20077J928
| ML20077J928 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 07/30/1991 |
| From: | William Cahill, Walker R TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TXX-91287, NUDOCS 9108050254 | |
| Download: ML20077J928 (2) | |
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Log # TXX-91287
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File # 10010-315.6 1.
Ref. # 10CFR50.36
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1UELECTRIC July 30, 1991 Wluiam J. Cabin. Jr.
E.arewM VM hearno U. S. Nuclear Regulatory Commission Attn: _ Documen_t Control Desk Washington, DC-20555
SUBJECT:
. COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) - UNIT 1 DOCKET ~NO.-50-445 LICENSE AMENDHENT REQUEST LAR-91 007 WESTINGHOUSE SAFETY EVALUATION FOR PHTC REF:
TU' Electric Letter logged TXX-91179, dated May 24, 1991, from, W. J. Cahill,'Jr. to the NRC Gentlemen:
The Westinghouse Safety Evaluation for Operation of_CPSES Unit I with a 1
Positive Moderator Temperature Coef ficient wa:: provided to the NRC as part of-j the-License Amendment. Request in the referenced-letter.
Subsequent to the-
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docketting of the referenced letter, Westinghouse determined that an error in the reporting of'the results-of the Reactor Coolant Pump Shaft Seisure (Locked Rotor)/ Shaft Break analysis existed.
Attached is a marked-up page from the referenced letter reflecting the correct results.
Please note that the conclusions reached for the subject transient in-the referenced-letter remain-
' unchanged._'The WestinghouseLSafety Evaluation for Operation of CPSES Unit 1 with-a Positive Moderator Temperature Coefficient, as corrected in the attachment. continues to support:the' conclusions reached in-the Significant Hazards Consideration for LAR-91-007.
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If Lyou have any questions, please contact Mr. J. D. Seawright: at (214) 812-4375.
. Sincerely, f,
William J. Cahill, Jr.
By:
Roger D. Walker Manager of Nuclear Licensing JDS/grp Attachment l
ic.- Mr. R. D._Hartin, Region IV Resident inspectors, CPSES_(2) i Hr. T.-A-Bergman, NRR kk
- hb[ bN bbN 45 400 North Olive Suret. L3. 81 Dallas, Texas 75201
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4 4 to TXX-91287
-Attachment
'ei-Page-1 of-l
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C.
Reactor Ceolant Pumo Shait Seiiure (Locked Rotor)) Shaft Break Introduction The case presented in the FSAR (Section 15.3.3) for this transient was analyzed.. Following a shaft seizure / locked rotor.
event, the reactor coolant system temperature rises until-shortly after~ reactor trip.
Becaure.ONB is conservatively assumed to
. occur at the beginning of the event, a positive MTC will r.ot affect the time to DNB.
The transient was analyzed, however, to assess the effect on the nu: lear power transient and thus on the peak reactor'cociant system pressure and fuel and clad
-temperatures.
Method of Analysis The digital computer codes ured in the analysis to evaluate' the pressure transient and thermal transient were the same as those i
usec in the FSAR.
The analysis employed a constant NTC of i
+5 pcm/*F. Other assumptions used were consistent with those employed.in the FSAA; b
-Results and Conclusiong Analysis of the locked rotor event with a +5 penV'T KTc shows-that the' peak reactor coolant tystem pressure remains below that which would cause stresses to exceed the faulted condition stress limits'.
The peak clad average temperature at the ' hot spot" was determined to b De*F which 1. much less than the l.
L 2700*F limit. The amount of ziiconium - water reaction at the ' hot spot"'was_ calculated tt be less than 173 which is less than the limit of 16%..Also, the peak RCS pressure was calculated to b Ae psia which is.less than 110% of the design g
RCS pressure.
Therefore. the co1clusions presented in the FSAR remain valid.
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