ML20052H753
| ML20052H753 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 05/14/1982 |
| From: | Nichols T SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20052H754 | List: |
| References | |
| NUDOCS 8205210405 | |
| Download: ML20052H753 (4) | |
Text
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SOUTH CAROLINA ELECTRIC & GAS COMPANY POST OFFICE BOX 764 COLUMBI A. S. C. 29218 May 14, 1982 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comnission Washington, D.C.
20555
Subject:
Virgil C. Summr Nuclear Station Docket No. 50/395 Cold Overpressurization Protection Systen
Dear Mr. Denton:
In Amendment 31 of the Virgil C. Sumer Nuclear Station Final Safety Analysis Report a setpoint curve for the pressurizer power operated relief valve (PORV) was provided. This curve applies when the PORVs are being used for cold overpressurization protection. The curve has been revised to reflect the results of an tpdated setpoint program. Selection of the setpoints was based on consideration of Appendix G reactor vessel NDP limits to 10 effective full power years (EFPY) of plant operation. Subsequent analysis of irradiated material specimens after 10 EFPY of plant operaticm will determine the need for any further setpoint refinements.
A copy of the affected FSAR and Technical Specification pages are attached. The FSAR changes will be incorporated on the next FSAR amendment.
If you have any questions, please let us know.
Very truly yours, Y
T. C. Nichols, Jr.
Senior Vice President Power Operation l
l RIE:'KN:rh g,od cc: Page '1%o V J tt L
l 8205210405 820514 PDR ADOCK 05000395 A
Mr. Harold R. Denton May 14, 1982 Page Two cc:
V. C. Sumer (w/o attach.)
G. H. Fischer (w/o attach.)
H. N. Cyrus T. C. Nichols, Jr.
(w/o attach.)
M. B. Whitaker, Jr.
J. P. O'Reilly H. T. Babb D. A. Nauman C. L. Ligon (NSIC)
W. A. Williams, Jr.
R. B. Clary O. S. Bradham A. R. Koon M. N. Browne G. J. Braddick J. L. Skolds 8
J. B. Knotts, Jr.
B. A. Bursey NPCF File 1
1 1
,r,,- -.
g
.-a-y
~~
-,e
,n-y.,-,w-,-,..,
- =---..y-
,--,,,v,r,,s--.-
w w.-w
,es
.-enn--ye n
, w w.--
(3 00, 3 Y60) 3 coo This curve is applicable for a service period up to et EFPY.
10 2 500 6
5b E
Dx 2 coo s
(Z.fo,10.3c)
A O
a M
5 8 1500 0
g t;
6
~ (L o o,12. t o) loco (150,720)
(70, 540)
(,z,,g3,,
500 (I00,5 70)
'2 o o 50 100
/50
- z. c o 2,5 o 3oo AVERAGE RFACTOR COOLANT SYSTEM TEMPERATURE (*F)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION 72 FIGUP.E 5.2-15
/0!ENDMENT vi LOW TEMP. RCS PRESS. CONTROL M"Se, 1982 MAXIMUM PORV LIFT SETPOINTS
i
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5.2.2.5.2
- Evaluation of Low Temperature Overpressura Transients Pressure Transient Analyces ASPE Section III, Appendix G, establishes guidelines and.imits for RCS Pressure primarily for low temperature conditions (<350 F). The
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relief system discussed in 5.2.2.5 satisfies these conditions as dis-cussed in the following paragraphs.
-~
Transient analyses were performed to establish a PORV lif t setpoint
[
program for the Overpressure Mitigating System (OMS) to be applied dur-ing shutdown operation of the plant.
This program maintains reactor coolant system pressure within acceptable limits f ollowing all credible overpressurization incidents occuring in, the plant during low tempera-ture, water solid operation.
The mass input transient analysis was performed assumiog the most severe event involving a single centrifugal charging pump.
Specifically, a loss of air incident is postulated, whereby the flow control valve on 31 the cha rging line f ails open and, simultaneously, the flow control valve on the letdown line fails closed.
The heat input mechanism considered f or analysis involved a RCS pump sta rtup in one loop with a water solid condition and temperature assymr metry in the reactor coolant system, whereby the steam generators were at,a higher temperature than the rema! ader of the system. A 50 F was rssoasd to 0
mismatch existed between the RCS (250 F) and the secondary side of the 31 A
steam generators (300 F).
(At lower temperatures, the mass input case g
is the limiting transient condition.)
Both analyses took into account the single failure criteria and there-fore, the operation of one PORV in the OMS was assumed to be available l
31 f or pressure relief.
The above events have been evaluated against the allowable pressure / temperature limits established in Figures 5.2-13 and 5.2-14.
The evaluation of the transient results conclude that the allowable limits will nct be exceeded and therefore will not constitute
'an impairment to vessel integrity and plant safety.
31.
5.2-4 5b AMENDMENT M.
N