ML20039D509
| ML20039D509 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 12/31/1981 |
| From: | Nichols T SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.13, TASK-2.K.2.17, TASK-2.K.2.19, TASK-2.K.3.05, TASK-2.K.3.25, TASK-2.K.3.30, TASK-TM NUDOCS 8201050081 | |
| Download: ML20039D509 (4) | |
Text
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SOUTH CAROLINA ELECTRIC a GAS COMPANY eost orrect som re4 CO COLUMBIA, SOUTH CAROLINA 29218 T. C. NicHO Ls. J n.
J RECEIVED
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Mr. Harold R. Denton, Director.
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U. S. Ibclear Regulatory Ccmnission 4
g Washington, D. C.
20555 6)
Subject:
Virgil C. Sununer Nuclear Station Docket No. 50/395 NUREG 0737
'IMI Itens II.K.2, II.K.3
Dear Mr. Denton:
South Carolina Electric Gas Chnpany (SCE&G) herein govides updated information required by NUREG 0737 implementation schedule for the subject 'IMI itens. Each iten is discussed in detail b310w.
1.
Iten II.K.2.13: Wermal Mechanical Report - Effect of High-Pressure Injection on Vessel Integrity for Small-Break Ioss-Of-Coolant-Accident with No Auxiliary Beedwater.
This iten requires a detailed analysis of the thermal-mechanical conditions in the reactor vessel during recovery fn i small breaks with an extended loss of all feedwater.
As mentioned in our January 6, 1981 letter, Westinghouse is performing an analysis for generic Westinghouse IMR plant groupings to address this issue. This analysis will be subnitted to the imC by the end of 1981 and will be applicable to the Virgil C. Sununer Nuclear Station.
If a3ditional efforts are necessary to completely address NRC concerns, this generic study can be referenced.
2.
Item II.K.2.17:
Potential for Voiding in the Beactor Coolant Systen During Transients.
Westinghouse has performed a study which addresses the rotential for void formation in Westinghouse designed FWRs during natural circulation cooldown/depressurization transients. 'Ihis study was subnitted to the NRC by Westinghouse Owners Group (WOG) letter OG-57, dated April 20, 1981, and is applicable to the Virgil C. Summer Nucimr Station. Ebrther, as discussed in our letter dated Decenber 2,1981, WOG subnitted an analysis (letter OG-62 to the NRC in July, 1981) showing the capability for Westinghouse EWR plants to cool down to cold shutdown conditions by natnral circulation and subsequent RHR operation without 'old formation in the upper head area.
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In addition, WOG has developed a natwd circulation cooldown 0
guideline that takes the results of the study into account to gg \\
preclude void formation in the upper head region during natural "8201050081 811231 PDR ADOCK 05000395 A
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A circulation cooldown/depressurization transients, and specifies those
- conditions under which upper head voiding may occur. These WOG.
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generic guidelines were subnitted to the NRC by letter OG-64,-dated f
November 30, 1981. As mentioned in our letter, dated nu am b r,2,
,w 1981, this information was govided to SCI %G in Septaber 1981, as_
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- Emergency Bes[onse Guideline ES-0.2 and will be used in the 3
implementation of Virgil C. Sumer plant specific operating procedures.
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Sequential Auxiliary 1%edwater Flow Analysis b,.
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It s II.K.2.19:
Subsequent to the issuance of NUREG 0737, the NRC has cmpleted.a_ _ m generic review on this subject and concluded that the concerns expressed in Item II.K.2.19 are not applicable to an NSSS with inverted U-tube steam generators, i.e., those designed.by A
Westinghouse. '1herefore, this item is not ' applicable ~ to the Virgil.
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s C. Sumer Nuclear Station and no further action is required.
4.
Item II.K.3.5:
Automatic Trip of Beactor Coolant Piznps During Ioss-of-Coolant Accident.
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Westinghouse has performed an analysis of delayed reactor coolant ~
p ep trip during small-break LOCAs.- 'This analysis.is documented-in iCAP-9584 (proprietary) and WCAP-9585.(non-proprietary), August 1979.
In addition, Westinghouse has performed test predictions of l
LOET experiements L3-1 and L3-6.
The results 'of these p edictions are doc uented in Westinghouse Owner's Group letters OG-49,' March 3, 1981, OG-50, March 23,1981, and OG-60, June 15,1981.
Based.upon (a) the Westinghouse analysis, (b)'the excellent prediction of the LOET experienent L3-6 results using the,
Westinghouse analytical model, and (c) Westinghouse simulator data '
related to operator response time, the Westinghouse _ and SCI %G position is that automatic reactor coolant pump trip is not necessary since. sufficient time is available for manual tripping of the pumps.
Our understanding'of the. schedule for final resolution of this issue is:
a.
Once the NRC formally apg aves the Westinghouse model, a 3-month-study period'will ensue during ~which:the Westinghouse Owner's Group.
will attempt -to demonstrate empliance.with scme NBC acceptance
- , criteria for manual RCP trip. The NRC acceptance criteria will i
accompany their formal approval of the Westinghouse nodels.
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b.
If, at the end of the 3-nonth period, the Westinghouse Owner's Group cannot show c m pliance with the acceptance criteria, the NRCi N,. j,,
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sp' Wv N v will formally notify utilities that they must submit an autcmatic RCP s
(p' trip design.
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5.
Item II.K.3.25: Effect of loss of Alternating Current Power on
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3 Pump Seals e
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(s This item requires that the consequences of a loss of RCP seal cooling.due to loss of AC power (defined as loss of offsite power) s
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'Durbt3-normal operation, seal injection flow frm the chenical and t,
volme control system 'is~ pcovided to cool the RCP seals, and the g,.
s campanent cooling water system provides flow to the thermal barrier heat exchanger to limit the heat transfer frm the reactor coolant to the RCP internals.
In the event of loss of offsite power, the RCP notor is deenergized aM both of these cooling supplies are tenninated; however, the diesel generators are autanatically started ard seal injection flow to the RCP seals is autmatically restored within seconds. This cooling supply is adequate to provide seal cooling ard prevent seal failure due to loss of seal cooling during a loss'of offsite We for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
6.
Item II.K.3
_avised Small-Break LOCA Methods to Show Compliance with h CER Part 50, Appendix K.
m is item requires that the analysis methods used by NSSS vendors and/or fuel suppliers for small-break LOCA analysis for empliance with 10 CER 50, Appendix K, be revised, documented, and subnitted for NRC approval.
Westinghcuse feels very strongly, and SCE&G agrees, that the small-break LOCA analysis model currently approved by the NRC for use on Virgil C. Sumer Nuclear Station is conservative and in conformance with 10 CER 50, Appendix K.
However, as docunented in Westinghorse letter NS 'IMA-2318 to the NRC, dated September 26, -
1980, West!inghouse believes that improvement in the realism cf small-break calculations is a worthAile effort and has ocmnittal to revise its small-break LOCA analysis model to address NRC concerns (e.g., NUREG-0611, NUREG-0623, etc.). Bis revised Westinghouse model is currently scheduled for subnittal to the NRC by April 1, 1982, as docmented in Westinghouse letter NS-EPR-2524, dated ~
Novunber 25, 1981.
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i Mr. Harold R. Denton Decenber 31, 1981 Page 4 If you have any questions concerning these itens, please let us know.
Very truly yours, t
i T. C. Nichols, Jr.
AJ:TCN:lkb cc:
V. C. Sumer T. C. Nichols, Jr.
G. H. Fischer H. N. : Cyrus H. T. Babb (NSRC)
D. A. Nauman (NSRC) i M. B. Whitaker, Jr.
(NSRC)
W. A. Williams, Jr.
(NSRC)
O. S. Bradham (NSRC)
R. B. Clary M. N. Browne A. R. Koon G. J. Braddick l
J. L. Skolds J. B. Knotts, Jr.
B. A. Bursey J. C. Ruoff W. R. Baehr (NSRC)
S. R. Ross (NSRC) i M. E. Stern (NSRC) i NNF File
.