ML18009A490
| ML18009A490 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 04/20/1990 |
| From: | Cutter A CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-REGGD-01.099 NLS-90-091, NUDOCS 9004300027 | |
| Download: ML18009A490 (6) | |
Text
ACCELERATED DKERIBUTION DEMONSTRATION SYSTEM, I
~
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9004300027 DOC.DATE: 90/04/20 NOTARIZED: YES FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1, Carolina AUTH.NAME AUTHOR AFFILIATION CUTTER,A.B.
Carolina Power
& Light Co.
RECIP.NAME RECIPIENT'FFILIATION Document Control Branch (Document Control Desk)
DOCKET 05000400 R
SUBJECT:
Forwards addi info re RCS pressure-temp limits Tech Spec change request,per NRC 900416 request.
l DISTRIBUTION CODE:
A001D COPIES RECEIVED LTR ENCL SIZE:
TITLE: OR Submittal:
General Distribution NOTES:Application for permit, renewal filed.
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A NOTE TO ALL"RIDS" RECIPIENTS:
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Carolina Power & Light Company P.O. Box 1551
~ Raleigh, N.C. 27602 APR 2O iSSO A. B CUTTER Vice President Nuclear Services Department SERIAL:
NLS-90-091 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO ~ 50-400/LICENSE NO ~ NPF-63 RCS PRESSURE-TEMPERATURE LIMITS SUPPLEMENTAL INFORMATION Gentlemen:
By letter dated March 16, 1990 the NRC staff requested additional information to support their review of the Harris Plant RCS Pressure-Temperature Limits Technical Specification Change Request submitted on June 30, 1989.
Attached is the information requested.
It provides additional detail and related information.
However, it has no impact on the significant hazards consideration contained in the June 30, 1989 submittal.
Please refer any questions regarding this submittal to Mr. Steven Chaplin at (919) 546-6623.
Yours very trul A.
B Cutter ABC/SDC Enclosure Becker H. Brown Ebneter Tedrow cc:
Mr. R. A.
Mr. Dayne Mr.
~
ED 900420 05000400 PDC 9004300027 PDR ADOCK P
My commission expires: /-0)- gK pot A.
B Cutter, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are
- officers, employees, contractors, and agents of Carolina Power
& Light Company.
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Attachment to NLS-90-091 Page 1 of 2 NRC Re uest:
Describe the thermal-hydraulic analysis performed in the Power Operated Relief Valve (PORV) setpoint study to prevent a potential violation of allowable PressureTemperature (P/T) limits during plant heatup 6 cooldown.
Include the following:
a) Major assumptions used in the current analysis as well as the analysis presented in the Final Safety Analysis Report (FSAR).
b) Brief discussion of the method/computer code used in the analysis.
c) Transient Reactor Coolant System (RCS) pressure at various PORV setpoints.
d) Comparison of the peak RCS transient pressure with the allowable P/T limits to confirm that Appendix G P/T limits will not be violated.
CP L Res onse'he revised Low Temperature Overpressure Protection (LTOP) setpoints, submitted as part of the June 30, 1989 Technical Specification Change
- Request, were derived using the same methodology employed in development of the original LTOP setpoints currently described in the FSAR and SER (References 1 through 6).
This methodology is documented in the Westinghouse Owner's Group (WOG) Report dated July 1977 and its Supplement of September 1977.
The WOG methodology utilized the LOFTRAN computer code to generate the PORV setpoint overshoot for a bounding envelope of mass and heat inputs.
The plant specific PORV setpoints and overshoot were then determined by the licensee performing hand calculations with plant specific parameters and algorithms provided in the WOG report.
In developing the revised LTOP setpoints for the new heatup and cooldown
- curves, the basic assumptions used and described in FSAR Section 5.2.2.11 were again utilized (References 4,
5 and 6).
These included a single failure (i.e., only 1 of 2 p~essurizer PORVs is assumed to operate),
most limiting heat or mass inputs
, and a water solid RCS.
Only the plant specific parameters, such as the safety injection flow rate, have changed.
The heatup and cooldown curves shown in Figures 3.4-3 and 3.4-2 of Technical Specifications 3.4.9.1 and 3.4.9.2 were derived as a result of applying the guidance of Regulatory Guide 1.99 Revision 2 and compensation factors of -60 psig and +10'F to the "Appendix G" curves.
These factors allow for instrument uncertainties.
The enable temperature for the LTOP system was also revised to 325'F as a result of utilizing Regulatory Guide 1.99 Revision 2.
Limiting mass input - inadvertent startup of one charging/safety injection pump.
Limiting heat input - inadvertent startup of one reactor coolant pump while the steam generator secondary side is 50'F higher than the primary side.
Attachment to NLS-90-091 Page 2 of 2 The results of the analysis indicate margin between the transient Reactor Coolant Syst: em (RCS) pressure developed during a postulated LTOP event and the "Appendix G" curves as shown at the following representative temperatures:
At RCS Temperature
('F) 100 125 155 200 325 Limiting HU or CD Rate, ('/hr)*
10 HU 10 CD 30 HU 50 HU 50 HU Limiting HU or CD Pressure (psig)**
474 495 496 539 1686 High PORV Setpoint (psig) 380 410 410 410 450 Transient RCS Pressure (psig) 463 490 490 490 562 Margin to Limiting HU or CD Curve (psi) 49 1124 The most limiting of the maximum allowable Heatup (HU) or Cooldown (CD) rate for the RCS temperature indicated, as allowed by proposed Technical Specification 3.4.9.2 and associated Table 4.4-6 for 3 EFPY.
This is the limiting pressure at a temperature 6'F lower than at the indicated temperature above for the PORV setpoint, to allow for additional temperature instrument uncertainties.
References'1)
CPSL letter, M. A. McDuffie to H. R. Denton (NRC), LAP-83-475, October 12, 1983 (2)
CP6L letter, M. A. McDuffie to H. K. Denton (NRC), LAP-83-507, October 27, 1983 (3)
(4)
(5)
(6)
CP&L letter, S.
R.
Zimmerman to H.
R. Denton (NRC), NLS-86-119, April 23, 1986 Safety Evaluation Report, NUREG-1038, November 1983 Safety Evaluation Report, NUREG-1038, Supplement No.4, October 1986 SHNPP Technical Specification Bases 3/4.4.9