ML20003D749
| ML20003D749 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 03/26/1981 |
| From: | Nichols T SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8103310458 | |
| Download: ML20003D749 (2) | |
Text
"
south CAROLINA ELECTRIC & GAS COMPANY CoLUMe:A, Sourn CAnoWNA 29218 vc..$i.7.*,".$N[i.em.
March 26, 1981 am... c.
Mr. Ilarold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Subject:
Virgil C. Summer Nuclear Station Docket No. 50/395 Decrease in Reactor Coolant Inventory
Dear Mr. Denton:
The attached revision to FSAR Section 15.3.1.2.1 is provided to address the Safety Evaluation Report Confirmatory Issue 1.7.20 regarding the decrease in reactor coolant inventory event. This should provide you with sufficient infor-mation to resolve this issue. It will be incorporated in the next FSAR amendment.
If you have any questions, please let us know.
Very truly yours,
//
tv T. C. Nichols, Jr.
RBC:TCNiglb Attachment ec:
V. C. Summer w/o att.
' ' 7 W ) [to N
, ^.
l -l' G. H. Fischer w/o att, s
T. C. Nichols, Jr. w/o att.
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C. A. Price 1
D. A. Nauman W. A. Williams, Jr.
,2 ~
f,,' fig n Whi-b R. B. Clary
\\,-( u.s,Mgsces A. R. Koon A. A. Smith A
H. N. Cyrus J. B. Knotts, Jr.
~
J. L. Skolds B. A. Bursey O. S. Bradham ISEG PRS NPCF File 8103310'/57 f
q l
einergowy core cooling system capability end operability has been 23 ass w.d ie_theae. analyses.
i 5
Peak clad temperature analyses are performed with the LOCTA-IV CodeI which determines the reactor coolant system pressure, fuel rod power f g 7 p. o d history, steam flow past the uncovered part of the core and mixture ya { ; S q}
n ~
P height hintory.
l (k.(g 3
~
V
% ?-
q e
so 3P 03 b
Figure 15.3-2 presents the hot rod power shape utilized to perform the.,C E 'd a
r-
- 1. mall break analysis presented here. This power shape was chosen "n
3
+
a o
($ Q g because it provides an appropriate distribution of power versus core (o g q.
,s P
\\, /
height and alr.o linear power is maximized in the upper regions of the >"o0-2.
%s,E 3 e reactor coce (10 to 12 foot).
This power shape is skewed to the top of g
the core with the peak linear power occurring at the 10 foot core 31 F A (4 elevation.
kQ ' %
R N
9 s
E 7 fs $%
%S j
Th;s i.4 limiting for the small break analysis because of the uncovering
- O
{4 process for the small break. As the core uncovers, thecladdinginthe@s tm $'
t>.
O S
upper elevation of the core heats up and is sensitive to the linear h4h p j power at that elevation. The cladding temperature in the lower eleva-
){ q ['
tions of the core, below the two phase mixture height, remains low. Thepn h n $u s
~
s-s ta peak clad temperature occurs above 10 feet.
\\$ $ DN P
- 2. k ss b
g 4
&1*
The small break analysis was performed with the October 1975 version of } d. iQ p-5 the W.:stinghouse ECCS Evaluati.on Model Sl'll'l l, and include the cor-23 r eted zirc/ water reaction rate calculaticn,.
.+
e
,4 l '). 3.1. 2. 2 Itesults
.N 4
B.ined on the results of the loss of coolant accident sensitivity studies (References [13 and 14]) the limiting small break was found to be less 23 than a 10 inch diameter rupture of the reactor coolant system cold Icg.
Therefore, a range of small break analyses are presented which estab-4 lashes the limiting break size.
metion presents results of the limii ng break size in terms of 4
- h. chest peak clad temperature. The worst break size (small break) is a 1 inch diameter break.
rhe depressurization transient for this break is shown in Figure 15.3-3.
The extent to which the core is uncovered is shown in Figure 15.3-4.
15.3-5
'