NRC Generic Letter 1980-12

NRC Generic Letter 1980-012, Transmittal of IE Bulletin 1980-004: Analysis of a PWR Main Steam Line Break with Continued Feedwater Addition

ML031350294
Person / Time
Issue date:	02/08/1980
From:	Grier B H NRC Region 1
To:
References
BL-80-004 GL-80-012, NUDOCS 8002250445
Download: ML031350294 (6)

mec1FW74f4L-F"4rUNITED STATESNUCLEAR REGULATORY COMMISSIONREGION I631 PARK AVENUEKING OF PRUSSIA, PENNSYLVANIA 19406February 8, 1980Docket Nos. 50-0350-247Consolidated Edison Company ofNew York, Inc.ATTN: Mr. W. J. Cahill, Jr.Vice President4 Irving PlaceNew York, New York 10003Gentlemen:The enclosed IE Bulletin No. 80-04, is forwarded for action. A written responseis required. If you desire additional information regarding this matter,please contact this office.Sincerely,j Boyce H. GrierC/- DirectorEnclosures:1. IE Bulletin No. 80-042. List of Recently Issued IEBulletins

CONTACT

0. L. Caphton(215-337-5253)cc w/encls:L. 0. Brooks, Project Manager, IP NuclearW. Monti, Manager -Nuclear Power Generation DepartmentM. Shatkouski, Plant ManagerJ. M. Makepeace, Director, Technical EngineeringW. D. Hamlin, Assistant to Resident ManagerJ. D. Block, Esquire, Executive Vice President -AdministrationJoyce P. Davis, Esquire.7-800,8,8,60il-'116'

ENCLOSURE 1UNITED STATES SSINS No.: 6820NUCLEAR REGULATORY COMMISSION Accessions No.:OFFICE OF INSPECTION AND ENFORCEMENT 7910250504WASHINGTON, D.C. 20555IE Bulletin No. 80-04Date: February 8, 1980 ANALYSIS OF A PWR MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION

Description of Circumstances

Virginia Electric and Power Co. submitted a report to the Nuclear RegulatoryCommission dated September 7, 1979 that identified a deficiency in the originalanalysis of containment pressurization as a result of reanalysis of steam linebreak for North Anna Power Station, Units 3 and 4.Stone and Webster Engineering Corporation performed a reanalysis of containmentpressure following a main steam line break and determined that, if the auxiliaryfeedwater system continued to supply feedwater at runout conditions to thesteam generator that had experienced the steam line break, containment designpressure would be exceeded in approximately 10 minutes. The long term blowdownof the water supplied under runout conditions by the auxiliary feedwatersystem had not been considered in the earlier analysis.On October 1, 1979, the foregoing information was provided to all holders ofoperating licenses and construction permits in IE Information Notice No.79-24. The Palisades facility did an accident analysis review pursuant to theinformation in the notice and discovered that with offsite power available,the condensate pumps would feed the affected generator at an excessive rate.This excessive feed was not considered in the analysis for the steam linebreak accident.On January 30, 1980, Maine Yankee Atomic Power Company informed the NRC of anerror in the main steam line break analysis for the Maine Yankee plant.During a review of the main steam line break analysis, for zero or low powerat the end of core life, the licensee identified an incorrect postulation thatthe startup feedwater control valves would remain positioned "as is" duringthe transient. In reality, the startup feedwater control valves will ramp to80% full open due to an override signal resulting from the low steam generatorpressure reactor trip signal. Reanalysis of the event shows the opening ofthe startup valve and associated high feedwater addition to the affected steamgenerator would cause a rapid reactor cooldown and resultant return-to-power,a condition outside the plant design basis.Actions to be Taken by the Licensee:For all pressurized water power reactors with an operating license and thosereactors listed in Attachment 1:

Enclosure 1 IE Bulletin No. 80-04Date: February 8, 1980 . Review the containment pressure response analysis to determine if thepotential for containment overpressure for a main steam line break insidecontainment included the impact of runout flow from the auxiliary feedwatersystem and the impact of other'energy sources, such as continuation offeedwater or condensate flow. In your review, consider your ability todetect and isolate the damaged steam generator from these sources and theability of the pumps to remain operable after extended operation atrunout flow.2. Review your analysis of the reactivity increase which results from a mainsteam line break inside or outside containment. This review shouldconsider the reactor cooldown rate and the potential for the reactor toreturn to power with the'most reactive control rod in the fully withdrawnposition. If your previous analysis did not consider all potential watersources (such as those listed in 1 above) and if the reactivity increaseis greater than previous analysis indicated the report of this reviewshould include:a. The boundary conditions for the analysis, e.g., the end oflife shutdown margin, the moderator temperature coefficient,power level and the net effect of the associated steamgenerator water inventory on the reactor system cooling, etc.,b. The most restrictive single active failure in the safetyinjection system and the effect of that failure on delayingthe delivery of high concentration boric acid solution tothe reactor coolant system,c. The effect of extended water supply to the affected steamgenerator on the core criticality and return to power,d. The hot channel factors corresponding to the most reactiverod in the fully withdrawn position at the end of life, andthe Minimum Departure from Nucleate Boiling Ratio (MDNBR)values for the analyzed transient.3. If the potential for containment overpressure exists or the reactor-re-turn-to-power response worsens, provide a proposed corrective action anda schedule for completion of the corrective action. If the unit isoperating, provide a description of any interim action that will be takenuntil the proposed corrective action is completed.4. Within 90 days of the date of this Bulletin, complete the review andevaluation required by this Bulletin and provide a written responsedescribing your reviews and actions taken in response to each item.Reports should be submitted to the Director of the appropriate NRC RegionalOffice and a copy should be forwarded to the NRC Office of Inspection andEnforcement, Division of Reactor Operations Inspection, Washington, D.C.20555.

Enclosure 1.IE Bulletin No. 80-04Date: February 8, 1980 For boiling water reactors with anand all pressurized water reactorsAttachment 1, this Bulletin is forresponse is required.operating license or a construction permitwith a construction permit, not listed ininformation purposes only and no writtenApproved by GAO, B180225 (R0072); clearance expires 7/31/80. Approval wasgiven under a blanket clearance specifically for identified generic problems.

Attachment No. 1 to IE Bulletin No. 80-04Plants with construction permits that are required to respond to the bulletin:Diablo CanyonMcGuireNorth Anna 2Salem 2SequoyahIf the permit holders have responded to earlier requests from the NRC on someof the items presented in the bulletin, they may respond to the bulletin byreference to the response to the earlier request.

-ENCLOSURE 2IE Bulletin No. 80-04Date: February 8, 1980 RECENTLY ISSUED IE BULLETINSBulletinNo.79-2579-02(Rev. 2)79-2679-2779-2879-01B80-0180-0280-03SubjectDate IssuedFailures of Westinghouse 11/2/79BFD Relays in Safety-Related SystemsPipe Base Plate Designs 11/8/79Using Concrete ExpansionBoltsIssued ToAll Power ReactorFacilities with anOperating License (OL)or Construction Permit(CP) (for Action)All Power ReactorFacilities with anOL or CPAll BWR Power ReactorFacilities with anOLAll Power ReactorFacilities with an OLand those nearingLicensing (for Action)All Power ReactorFacilities with a CP(for Information).Boron Loss From BWRControl BladesLoss of Non-Class-1-EInstrumentation and Con-trol Power System BusDuring OperationPossible Malfunctionof NAMCO Model EA180Limit Switches atElevated TemperaturesEnvironmental Quali-fication of Class IEEquipment1l20/7911/30/79127n9All Power ReactorFacilities with anOL or CP1/14/80All Power Reactorswith an OL exceptSEP PlantsOperability of ADS Valve 1/14/80Pneumatic SupplyAllOLBWRs with anInadequate QualityAssurance for NuclearSupplied EquipmentLoss of Charcoal FromStandard Type II, 2 Inch,Tray Adsorber Cells1/21/80All BWRs with anOL or CP2/6/80All Power ReactorFacilities with anOL or CP