ML19249E973
| ML19249E973 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/21/1979 |
| From: | Parr O Office of Nuclear Reactor Regulation |
| To: | Proffitt W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| NUDOCS 7910020805 | |
| Download: ML19249E973 (7) | |
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DISTRIBUTION:
I NRC POR ELD SEP % 1 SS Local PDR HBerkow Decket File DEisenhut LWR #3 File BGrimes DVassallo-TERA V) Buchanan, NSIC 3
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Docket No. 50-339 OParr ADromerick 6
MRushbrook (w/ extra copies)
RMattson l
Mr. W. L. Proffitt SHanauer Senior Vice President - Power JKnight i
Virginia Electric and Power Company RTedesco i
P. O. Box 26666 RDeYoung Richmond, Virginia 23261 VMoore MErnst
Dear Mr. Proffitt:
RDenise SU3 JECT: REQUEST FOR ADDITIONAL INFORMATION f
To continue our review of your application for a license to operate the North Anna Power Station, Unit 2, additional infctmation is required.
The i
information requested is described in the Enclosure.
i To maintain our licensing review schedule, we will need a completely adequate response to all enclosed requests by October 5,1979.
Please inform us after receipt of this letter of your confirmation of the i
above date or the date you will be able to meet.
Sincerely.
. Orfgtrul SJgned by Olan Parr Olan D. Parr, Chief Light Water Reactors, Branch No. 3 Division of Project Management
Enclosure:
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Mr. W. L. Proffitt-cc: Mr. Anthony Cambaradella Cl arence T. Ki pps, Jr., Esq.
Office of the Attorney General 1700 Pennsyl vania Avenue, N.W.
11 South 12th Street - Room 308 Washington, D. C.
20006 Richmond, Virginia 232 19 Carroll J. Savage, Esq.
Richard M. Foster, Esq.
1700 Pennsyl vania Avenue, N.W.
211 Stribling Avenue Wa shington, D. C.
20006 Charlottesville, Virginia 22903 Mr. James C. Dunstan Michael W. Naupin, Esq.
State Corporation Commission Hunton, Williams, Gay & Gibson Conmonwealth of Virginia P. O. Box 1535 Blandon Building Richmond, Virginia 23212 Richmond, Virginia 23209 Mrs. June Allen Alan S. Rosenthal, Esq.
412 Owens Drive Atomic Safety and Licensing Appeal Foard Hunt sville, Alaban1 35801 U.S. fluclear Regulatory Connission Washington,D. C.
20555 Mr. James Torson 501 Leroy Michael C. Farrar, Esq.
Socorro, New Mexico 87801 Atomic Safety and Licensing Appeal Paard U.S. Nuclear Regulatory Commission Mrs. Fergaret Dietrich Washington, D. C.
20555 Route 2, Box 568 Gordonsville, Virginia 22942 Dr. Jol H. Buck Atomir afety and Licensing Appeal Coard Willi an H. Rodgers, Jr., Eso.
U.S..iu ' ear Regulatory Comnission Georgetovo University Law Ce' iter Washington, D. C.
20555 600 New Jersey Avenue, N.W.
Washington, D. C.
20001 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Connission Fr. Peter S. Fepp Washington, D. C.
20555 Executive Vice President Sun Shipping & Dry Dock Company I,r. Michael S. Ki dd P. O. Box 540 U.S. Nuclear Reg ul atory Comnission Chester, Pennsylvania 19013 P. O. Box 128 Spotsivania, Virginia 22553 Mr. R. B. Origgs Associate Director Dr. Paul W. Purdor 110 Evans Lane Department of Civil Engineering Oak Ridge, Tennessee 37830 Drexel University Philadel phia, Pennsyl vania 19104 1074 068
Mr. W. L. 3rof fitt cc:
Dr. Lawrence R. Quarles Apartment No. 51 Kendal-at-Longwood Kennett Square, Pennsylvania 19348 Mr. Irwi n B. Kroot Citizens Energy Forum P. O. Box 138 McLean, Vi rginia 2210i James B. Dougherty, Esq.
Potomac Alliance 1416 S Street, N.W.
Washington, D. C.
20009 1074 069
ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION NORTH ANNA POWER STATION, UNIT 2 DOCKET NO. 50-339 6.0 Engineered Safeguards 6.158 In a letter dated September 30, 1979, the NRC was informed by you that overpressurization of the containment at North Anna 3 and 4 could occur as a result of.a main steam line break inside containment. T.is overpressurization resulted wher. auxiliary feedwater flow was included in the analysis.
NRC is currently assessing the generic implications of this letter.
To assist us in determining if a sir.:ilar circumstance could occur at your facility, you should take the following actions.
1)
Review your original analysis of this event, and provide.NRC with the assumptions used during this analysis.
Particular empnasis should be placed on describing how auxiliary feed. vater flow (AFF) was accounted for in your original analysis.
(Refer-ence to previously submitted information is acceptable if identified as to page number and date.) Any changes in your design which would impact the conclusions of your original analysis should be discussed. We are particularly concerned with design changes that could lead to an underestimation of the containment pressure following a MSLB inside containment.
1074 070
.. 2)
Speci fically, provide the following information for the analyses performed to determine tFe maximum containment pressure for a spectrum of postulated main steam line breaks for various reactor power levels:
a.
Specify the auxiliary feedwater flow rate that was used in your original containment pressurization analyses.
Provide the basis for this assumed flow rate.
b.
Provide the auxiliary feedwater rated flow rate, the run out flow rate, and the pump head capacity curve of your current design.
Provide schematic drawings to show the auxiliary feec, tater c.
system arrangement in your current design.
d.
Provide the time span over which it was assumed in your original analysis that AFF was added to the affected steam generator following a MSLB inside containment.
Discuss the design provisions in the auxiliary feedwater e.
system used to termi;; ate the auxiliary feedwater flow to the affected steam generator.
If operator action is required to perform this function, discuss the information that will be available to the operator to alert him of the need to isolate the au.~iliary feedwater to the af fected steam generator, the time when this information would be-come available, and the time it would take the operator 1074 071
. to complete this ac'. ion.
If teraination of auxiliary feca-water flow is dependent on automatic action, describe the basic operation of the auto-isolation systca.
Describe the failure modes of the system. Describe any annunciation devices associated with the system.
f.
Provide the single active failure analyses which specifically identifies those safety grade systems and components relied upon to limit the mass and energy release and the contain-ment pressure response.
The single failure analysis should include, but not necessarily be limited to:
partial loss of containment cooling systems and failure of the auxiliary feedwater isolation valve to close.
g.
For the single active failure case which results in :ne maximum containment atmosphere pressure, provide a chronology of events. Graphically, show the containment atmosphere pressure as a function of time for at least 30 minutes following the accident.
For this case, assume the auxiliary feedwater flow to the broken loop steam generator to be at the pump run out flow (if a run out control system is not part of the current cesign) for the entire transient if no automatic isolation to auxiliary feedwater is part of the current design.
h.
For the case identified in (g) above, provide tne mass and energy release data in tabular form.
Discuss and justi fy Ni k 0
4-the assumptions made regarding the time at which active containment heat rent] val systems become effective.
1074 073