ML19290C122
| ML19290C122 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley, Surry, North Anna  |
| Issue date: | 11/13/1979 |
| From: | Grimes B Office of Nuclear Reactor Regulation |
| To: | Jordan E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| TAC-12263, NUDOCS 8001090509 | |
| Download: ML19290C122 (5) | |
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h, UNITED STATES NUCLEAR REGULATORY COMMISSION e
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a wAsmmoTow.u. c.2 eses NOV 131979 MEMORANDUM FOR:
E. Jordan, Assistant Director, Division of Reactor Operations, OI&E FROM:
Brian K. Grimes, Acting Assistant Director for Systems Engineering, DOR
SUBJECT:
OVERPRESSURIZATION OF CONTAINMENT IN PWR PLANT AFTER A MAIN STEAM LINE BREAK
REFERENCE:
Memorandum, dated October 29, 1979, from D. Shtmi to G. Lainas, "VEPCO Part 21 Notification on Containment Overpressurization."
In our discussions of your proposed transfer of lead responsibility on the subject of containment overpressurization following a MSLS inside containment, we agreed to the following courses of action:
1.
Audit some MSLB analyses to determine the effects of emergency feedwater systems on contair.:::ent pressure; and 2 -. Review the generic aspects on PWRs of emergency feedwater systems supplying the MSLB at runout flows.
With regard to the first course of action, we have reviewed the analyses for the following plants: North Anna 1. Beaver Valley 1, and Surry 1 and 2 (see Reference).
The results of these reviews are discussed in Enclosure 1.
We have concluded that these plants do not exceed containment design pressure if credit is given for operator action within 10 minutes.
With regard to the second course of action, we have enclosed (Enclosure 2) a draft bulletin which should be sent to all PWRs.
Please contact us if you have any further questions on this matter.
Brian K. Grimes, Acting Assistant Director for Systems Engineering Division of Operatir.g Reactors Enclosures and cc:
See page 2 1728 267 wg 3, y la.
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Enclosures:
As stated
Contact:
D. Shum X-27058 cc w/ enclosures:
D. Eisenhut R. Tedesco D. Ross T. Novak W. Gamill DOR 8/C's E. Adensam D. Shum J. Kerrfgan E. Butcher R. Denise M. Wilber, IE V. Noonan B.D. Liaw
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ENCLOSURE 1 RESULTS'0F AUDITING MAIN STEAMLINE BREAK ANALYSES FOR SOME PLANTS TO DtitININE THE EFFECTS OF AUXILIARY FEEDWATER ON CONTAINMENT OVERPRESSURIZATION On September 7,1979, VEPC0 filed a Part:21 Notification stating that the current containment pressure response analysis perfomed by Stone & Webster (S&W) for North Anna 3 and 4 showed that the containment may be overpressurized resulting from a main steamline break accident inside containment if auxiliary feedwater flow to the damaged steam generator were not isolated within ten minutes following the accident. PSB met with VEPCO, S&W, and Duquesne (Beaver Valley) a couple of times (September 13 and 21,1979) and talked to S&W on the phone several times to find out if the containment over-pressurization were due to peculiarities in the North Anna 3/4 design.
Other S&W designed plants were re-examined to provide a comparison with the North Anna 3/4 analysis.
Based on S&W's re-examination, North Anna 1 does not have a containment overpressurization problem because they have flow restricting orifices in their auxiliary feedwater lines (AF). The Beaver Valley containment is not predicted to exceed design pressure because the auxiliary feedwater pump lines are manifolded, such that the header backpressure prevents AF pump runout. The design of Surry 1/2 is similar to Beaver Valley.
The problem at North Anna 3/4 arou because North Anna 3/4 has larger AF pumps than other S&W designs. In addition, each pump at North Anna 3/4 delivers flow directly to each of the steam generators.
Other issues have arisen on the validity of previously accepted MSLB analyses.
IE issued an information notice on the Part 21 notification (79-24) on October 1,1979.
In response to that notice, Palisades took a look at theirMSLB analysis and notified IE that the analysis may not be conservative due to the fact that condensate pumps continue to feed the damaged steam generator unless loss-of-offsite power is assumed to occur concurrently with a MSLB. On October 30, 1979, they filed an LER to this effect which addresses their proposed fixes.
DSS has reconnended (Ref. E-1) that the NA 3/4 overpressurization be a board notification item and has sent questions to DPM for transmittal to the "near tem" OL applications. As a result of our review, however, we have concluded the following:
1.
Should the containment become overpressurized, it would not be expected to exceed design pressure by greater than a factor of 2 (2.5 times design is generally considered to be the ultimate failure load of the containment structures).
2.
The North Anna 3/4 analysis conducted by S&W showed that the contain-ment was just exceeding design pressure at 10.6 minutes.
For other plants, such as Beaver Valley 1, the calculations, assuning pump runaut, show design pressure being exceeded at about 12 minutes but reaching less than 1.3 times design at 30 minutes. We believe there would be operator action taken to isolate the damaged steam generator prior to reaching excessive loads.
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3.
At pressures greater than design pressure, leakage from the contain-ment might exceed La. However, the radiological consequences would not exceed those calculated for the DBA-LOCA because of the differences in the source tenn between the MSLB (secondary coolant inventory) and the LOCA (TID-14844).
4.
The calculations conducted by PSB (Ref. E-2) to predict containment pressura as a function of auxiliary feed flow (AFF) assumed that the AFF was transported to the containment with an enthalphy of 1200 Btu /lb. This is a conservative assumption as it assumes the reactor coolant pump is running to effect rapid heat transfer from the core to the secondary side of the damaged SG. At AFF ptznp runout flow rates, this assumption removes more energy from the primary system over an extended period than would be available from it.
REFERENCES E-1.
Memorandum from F. Schroeder to S. A. Varga, "Reconrnendation for Board Notification: Auxiliary Feedwater Runout Flow Following MSLB and Containment Pressurization," dated October 19, 1979.
E-2.
Memorandum from D. Shum to G. Lainas, "VEPCD Part 21 Notification on Containment Overpressurization," dated October 29, 1979.
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ENCLOSURE 2 IE Bulletin No.
CONTAI! MENT OVERPRESSURIZATION RESULTING FRCH A MAIN STEAMLINE BilEAK ACCIDENT Description of Ciretznstances:
On September 7,1979, Virginia Electric and PoNr Company (VEPCO) filed a notification to NRC under 10 CFR 21 stating that the current containment pressure analysis performed by Stone and Webster (S&W) for North Anna Units 3 and 4 showed that the containment will be overpressurized following a main steamline break accident inside containment if auxiliary )
feedwater flow (assumed to be at auxiliary feedwater pump runout condition to the damaged steam generator were not isolated. A similar problem may exist at other PWR operating plants.
On October 1,1979, IE Infomation Notice No. 79-24 was issued to all PWR licensees describing this event. Since receiving this infomation notice, one licensee, Palisades, has indicated that they may exceed containment desfgn pressure due to continuation of condensate flow. On October 30, 1979, the licensee filed an LER addressing this problem and their proposed action.
Actions to tie Taken by Licensees:
All pressurized water power reactor licensees are requested to review their containment pressure response analyses to detennine if the potential for containment overpressure for a main steamline break inside containment included the impact of runout flow from the auxiliary feedwater system or the impact of other energy sources, such as continuation of feedwater or condensate flow. The results of these reviews should be submitted within 90 days of the receipt of this bulletin to the Regional Director
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