ML20127F959
| ML20127F959 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 06/20/1985 |
| From: | Dixon O SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8506250310 | |
| Download: ML20127F959 (5) | |
Text
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,s SOUTH CAROLINA ELECTRIC & GAS COMPANY POST 07FeCE 76a CotuusrA. Soutw CAnoLINA 29218
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June 20, 1985 VICE P4fs40t NT NUCLEan oPanatioNs Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Removal of Boron Injection Tank
Dear Mr. Denton:
In a letter to Mr. H. R. Denton from Mr. O. W. Dixon, Jr., dated April 9,1985, South Carolina Electric and Gas Company (SCE&G) requested an amendment to the Virgil C. Summer Nuclear Station Technical Specifications allowing for the removal of the Boron Injection Tank (BIT).
In discussions with the NRC Staf f, additier.a1 information on the reanalysis of the main steam line break accident was requested. This letter and its Attachments are hereby provided to supply this information.
If you should have any further questions, please advise.
{
- yours, O. W.
xo,
l AMM/csw f
At tachments I
cc:
V. C. Summer C. A. Price l
T. C. Nichols, Jr./0. W. Dixon, Jr.
C. L. Ligon (NSRC) l E. H. Crews, Jr.
K. E. Nodland E. C. Roberts R. A. St ough W. A. Williams, Jr.
G. O. Percival D. A. Nauman C. W. Hehl J. Nelson Grace J. B. Knotts, Jr.
Group Managers H. G. Shealy O. S. Bradham NPCF File l
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8506250310 8506.70 PDR ADOCK 05000395 P
Attachment "A" Summary of the Analyses Performed Supporting Removal of the BIT NOTE: All analyses were performed utilizing end of life core conditions, minimum safety injection flow, and the highest worth control rod fully withd rawn from the core.
In addition, the BIT was assumed to be installed in the flow path and filled with water containing 0 ppm boric acid.
t I.
Accidental Depressurization of the Main Steam System - FSAR 15.2 Evaluation of this event considered the opening and sticking of one steam line safety valve at hot zero power. Of fsite power was available to drive the reactor coolant pumps. This evaluation indicated that a reactor restart occurred with the power reaching a maximum of 3%.
II.
Major Rupture of a Main Steam Line - FSAR 15.4 Two analyses were done for this event, one with offsite power available and one without power. The analysis assuming of fsite power availability was the bounding event because greater heat transfer was available with the Reactor Coolant Pumps in operation. These analyses assumed hot, zer o power conditions and the largest size postulated break. The analyses showed restart with power levels remaining below 15%.
See the attached Figures 15.4-59 and 15.4-60.
[II.
Containment Analysis for Peak Pressure - FSAR 6.2 The main steam line breaks that were analyzed and the f ailures associated with the analyzed breaks are shown on the attached Table 6.2-la.
The results of the analyses showgd peak reactor building pressure (45.8 psig) 1.4 f t double ended rupture from 102% power with was generated for the a diesel generator failing to start, a main steam isolation valve failing to close, and an emergency feedwater control valve falling to isolate.
IV.
Containnent Analysis for Peak Temperature - FSAR 6.2 The main steam break (see Table 6.2-la) which gave the highest 2
temperature in containment (321*F) was the.645 f t split break from 102% power with a diesel generator failing to start and an energency feedwater control valve f ailing to isolate. The analyses were combined (i.e.
multiple failures assumed in each break) to limit the time and cost of the modification analyses.
Results with multiple failures represent a conservative analysis methodology.
Attachmant "B" TABLE 6.2-la Main Steam Line. Breaks Analyzed Double Ended Ruptures Failure (s) Assumed ( }
Area / Power
'1.4.ft / 0%
EFWV
.1.4 ft / 0%
MSIV 1.4 ft / 0%
DG 1.4 ft / 30%
1.4 ft / 70%
EFW, MSIV, DG Small Double Ended Ruptures (with Entrainment)
Failures Assumed (2)
Ares / Power 0.2 ft / 0%
DG, MSIV, EFW Small Double Ended Ruptures (No Entrainnent)
Failures Assumed ( }
i Area / Power 2
0:1 ft / 0%
DG, MSIV, EFW Split Ruptures' (No Entrainment)
' Area /Fover'
' Failures' Assumed ( )
2 O.30 ft. / 0%
O.7065 f / 30%
' Note:
1.
Effective break area for broken loop is 1.4 ft.
2.
DG = Diesel' generator fails to start.
MSIV = Main steam isoir. tion valve fails to close.
EFW = Emergency feedwr.ter fails to isolate.
s.~q Attachment "C"
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VIRGIL C. SUMMER NUCLEAR STATION Transient Response to 5 team Line Break with Safetyinjection and Offsite Power (Case a) l Figure 15.4 59
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