NRC Generic Letter 1993-06
- -WUNITED STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON, D. C. 20555October 25, 1993TO: ALL HOLDERS OF OPERATING LICENSES OR CONSTRUCTION PERMITS FORNUCLEAR POWER REACTORS
SUBJECT: RESEARCH RESULTS ON GENERIC SAFETY ISSUE 106, 'PIPING AND THE USEOF HIGHLY COMBUSTIBLE GASES IN VITAL AREAS' (GENERIC LETTER 93-06)
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter toinform addressees about technical findings resulting from the NRC resolutionof Generic Safety Issue 106 (GSI-106), Piping and the Use of HighlyCombustible Gases in Vital Areas.' It is expected that recipients will reviewthe information for applicability to their facilities and consider actions, asappropriate, to avoid similar problems. However, suggestions contained inthis generic letter are not NRC requirements; therefore, no specific action orwritten response is required.DISCUSSIONThe basic regulatory requirement dealing with the storage, distribution, anduse of combustible gases at nuclear power plants is General Design Criterion(GDC) 3, wFire Protection,' Appendix A, Part 50, Title 10 of theCode of Federal Reaulations (10 CFR Part 50). This criterion states, in part,that 'structures, systems, and components important to safety shall bedesigned and located to minimize, consistent with other safety requirements,the probability and effect of fires and explosions.' Additional discussion ofthe regulation of this subject is provided in NUREG-1364, 'Regulatory Analysisfor the Resolution of Generic Safety Issue 106: Piping and the Use of HighlyCombustible Gases in Vital Areas,' Section 1.2 (Enclosure 1).*Reviews of plant literature, site visits, and discussions with licensees haveindicated large differences in individual plant characteristics that couldaffect risk from failures of hydrogen system lines or components. Thesedifferences include the hydrogen storage and distribution system designfeatures and relative locations of hydrogen components and safety-relatedequipment. On the basis of generic evaluations, the NRC staff has concludedthat several possible methods to reduce risk, involving equipmentmodifications and administrative controls, could provide cost-effective safetybenefits at some plants. However, the NRC staff also concludes, based on asmall sample of plants, that the safety benefit of recommended actions for*Copies of this document are enclosed for addressees. For other readers, acopy of this document is available for inspection and copying in the NRCPublic Document Room, 2120 L Street NW, Washington, DC 20037.9310200286
- - eGeneric Letter 93-06 -2 -October 25, 1993some or all licensees or applicants is marginal. The reviews indicated that anumber of plants have system design characteristics, operating procedures, andother mitigating features that would be responsive to some or all of theconcerns of this generic issue. While the staff analysis indicates that theindustry-wide risk is small, It cannot preclude the possibility of larger riskat some plants. The NRC is aware that information relevant to 10 CFR Part 21,has been made available to licensees and applicants with General Electricboiling-water reactor (BWR) plant designs, emphasizing the need for individuallicensees and applicants to determine the safety hazard of a postulatedgenerator coolant hydrogen explosion in their plants (Enclosure 2). Inaddition, in March 1993, a turbine fire, which may have been caused by turbineblade failure, vibration, and hydrogen seal leakage, occurred in a nuclearpower plant in India.In view of the observed large differences in plant-specific characteristicsaffecting the risk associated with the use of hydrogen, and the marginalgeneric safety benefit that can be achieved in a cost-effective manner, theNRC intends to resolve this generic issue by making these results available inthis generic letter. This information may help licensees in their plantevaluations recommended by Generic Letter 88-20, Supplement 4, "IndividualPlant Examination of External Events for Severe Accident Vulnerabilities,"June 28, 1991.As part of the NRC evaluation of GSI-106, the risk from potential hydrogensystem failures was analyzed by the Idaho National Engineering Laboratory(INEL). The technical findings are reported in NUREG/CR-5759, "Risk Analysisof Highly Combustible Gas Storage, Supply, and Distribution Systems inPressurized Water Reactor Plants," July 1991; EGG-SSRE-10198, 'Risk Analysisof Highly Combustible Gas Storage, Supply, and Distribution Systems inPressurized Water Reactor Plants--Supplementary Cost/Benefit Analysis,"March 1992; and EGG-NTA-9082, "Scoping Risk Analysis of Highly Combustible GasStorage, Supply, and Distribution Systems in Boiling Water Reactor Plants,"November 1991. In addition, the NRC staff evaluated the safety benefits andcosts of implementing various alternatives to reduce generic risk inNUREG-1364. This regulatory analysis includes discussion of several precursorevents involving the storage, distribution, and use of hydrogen (thecombustible gas of principal concern) at nuclear power plants.**The scope of GSI-106 included evaluation of the risk from (1) the storage anddistribution of hydrogen for the volume control tank (VCT) in PWRs and themain electric generator in BWRs and PWRs; (2) other sources of hydrogen suchas battery rooms, the waste gas system in PWRs and the offgas system in BWRs;and (3) small, portable bottles of combustible gases used in maintenance,testing, and calibration. The risk from large storage facilities outside thereactor, auxiliary, and turbine buildings is being addressed separately and isnot within the scope of GSI-106.**Copies of these reports are available for inspection and copying in the NRCPublic Document Room, 2120 L Street NW, Washington DC 20037.
Generic Letter 93-06-3 -October 25, 1993Screening studies described in NUREG/CR-5759 and EGG-NTA-9082 indicated smallrisk for the battery rooms, waste gas and offgas systems, and portablebottles. The assessment for the generic risk associated with the hydrogendistribution system to the electric generator at BWRs involved a vital areaanalysis for an actual plant configuration (a BWR-4 with a Mark Icontainment), supplemented by information obtained from visits to five otherplants. The scoping analysis based on this sample of BWRs (two BWR-3s, twoBWR-4s, and two BWR-5s) indicates a small generic risk, but cannot precludethe possibility of a larger plant specific risk because of the possiblepresence of safety-related equipment in the turbine building. Inaddition, this scoping analysis did not consider the effect of hydrogenexplosions on barrier walls and on penetrations such as doors between theturbine building and the adjoining reactor, control, or auxiliary buildingsfor these six BWR plants.The findings of a more detailed generic risk analysis for the distributionsystems for the VCT and electric generator at PWRs are reported inNUREG/CR-5759. The hydrogen distribution systems to the VCT and generator arenot located near the reactor and primary coolant system piping. Hence,hydrogen fires or explosions would not lead to such events as pipe break lossof coolant accidents (LOCAs), anticipated transients without scram, and steamgenerator tube ruptures. INEL divided the remaining transient-induced coredamage events into transients with failure of decay heat removal systems(T/DHR) and transient-induced loss of coolant accidents (T/LOCA). Theinitiating event is either a random or seismically induced leak or break inthe hydrogen system that releases hydrogen. This released hydrogen createsthe potential for a fire or explosion that could cause loss of equipment andlead to either a T/DHR or a T/LOCA. The T/DHR events involve scenarios withloss of all forms of core cooling and coolant release at high pressure fromthe pressurizer safety and relief valves. The T/LOCA events involve failureof reactor coolant makeup or recirculation systems following a loss of reactorcoolant pump seal cooling or stuck-open safety or relief valves. In itsgeneric analysis of GSI-106, INEL addressed risks associated with the T/DHRand T/LOCA events and considered such plant functional characteristics asfeed-and-bleed cooling capability and relative locations of hydrogendistribution systems and pertinent equipment (e.g., auxiliary feedwater,normal and emergency ac power, essential service water, and component coolingwater).For the auxiliary building, which may contain most of the safety-relatedsystems at the plant, the following alternatives were found to be costeffective: (1) use of restricting orifices or excess flow valves to limit themaximum flow rate from the storage facility to the postulated break and(2) use of a smaller storage facility normally connected to the VCT to limitthe maximum hydrogen release in a single event. An alternative involving useof a normally isolated supply with intermittent manual makeup was somewhatless cost-effective. These approaches include preoperational testing andsubsequent retesting of excess flow valves and measures to prevent buildup ofunacceptable amounts of trapped hydrogen and inadvertent operation with thesafety features bypassed.
-4;Generic Letter 93-06 -4 -October 25, 1993For the turbine building, which may also contain safety-related equipment, twocost-effective alternatives were found for protection against breaks in thehydrogen supply line up to the hydrogen control station below the generator,including any branch lines from this line to other buildings. These involvelimits on the maximum flow rate or operation with a normally isolated supply.Isolation of the large quantities of hydrogen (up to about 700 standard cubicmeters [25,000 standard cubic feet]) contained in the generator probably isnot possible for most breaks downstream of the hydrogen control station. Theonly alternative considered applicable to breaks at or near the generatorinvolved structural modifications to prevent fire or blast damage to affectedsafety-related equipment; this alternative was not found to be cost-effective.Additional general measures for risk reduction, such as the use of colorcoding, warning signs and training to handle events in the auxiliary andturbine buildings were considered; Of these, training to stop hydrogen flow(e.g., isolation of the storage facility or venting and purging of thegenerator) and training to prevent associated large oil fires in the turbinebuilding were deemed most important.
BACKFIT DISCUSSION
In this generic letter, the NRC is only communicating information on resultsof government-sponsored research to resolve a generic safety issue and is notrecommending that licensees or applicants take particular courses of action orrequesting that licensees communicate information back to the NRC on thismatter. Consequently, this generic letter does not represent a backfit.If you have any questions about this information, please call one of thetechnical contacts listed below or the appropriate Office of Nuclear ReactorRegulation (NRR) project manager.
Sincerely,PatwAssociate Director for ProjectsOffice of Nuclear Reactor Regulation
Enclosures:
1. NUREG-1364, "Regulatory Analysis forthe Resolution of Generic Issue 106:Piping and the Use of Highly CombustibleGases in Vital Areas" (for addressees)2. Letter from J. P. Riley, General ElectricCompany, to S. E. Scace, Millstone NuclearPower Station, on 'Postulated HydrogenExplosion in a Non-United States ReactorTurbine Building Mezzanine," December 23, 19923. List of Recently Issued NRC Generic LettersTechnical Contacts: Gerald Mazetis, RES Vern Hodge, NRR(301) 492-3906 (301) 504-1861
-aI -- ---JUN- 3-93 THU 9:40 Gen. Facil. Licensing FAX ^ 203 665 5896 P.02EN-IDRE 2GE Nuclear Energy.0&,G-EH-92-052 FO B t er *ef j' ^cd650December 23, 1992 4RSCEIEDMr. ULC 3 01992Vice President -Millstone StationMillstone Nuclear Power station PROJECT SERVICESDEPARTMENINortheast Utility Service Co. P.ISLYP.O. Box 128 % .Waterford, CT. 06385 -.Subject: P2s!tlatad Hvdroaen Explosion in a Non -United StatesReactor Turbine Building -Mezianine
Dear Mr. Scace:
A BWR/3 utility located outside of the U. S., recently informed GEthat a postulated hydrogen gas line break in the plant's TurbineBuildinmezzanne cUl sutin a hydrogen detonation. U. S. BWRutilities may have sizmilar turbine buildin-g configuration/equipmentarrangements which could result in a similar condition. However, th:portion of the plant is in the utility's Architect Engineer's scope (supply and therefore cannot be evaluated by GE within the context ofthe U.S. Code of Federal Regulations 10 CFR Part 21. GE is,therefore, obligated, under the requirementi=uf:Z=0 Part 21, topass this information on to the potentially affected BWR utilities scthat they can review the following information for applicability totheir unique plant design.At the subject plant, the pipe lines which carry hydrogen gas to coo]the generator are routed from the hydrogen bottles outside the turbirbuilding to the hydrogen controls in the turbine building mezzaninearea and up to the generator. It was postulated that if a breakoccurred in one of these lines, up to 16,000 cubic feet (459 cubicmeters) of hydrogen gas would rapidly blow down into the mezzanine (asemi-enclosed area). This room has a volume about 56,000 cubic feet(1,600 cubic meterb) of oxygen so that even with uniform mixing, theaverage hyrtiren concentration could easily exceed 25% by volume,which is well above the lower lim etonatable mixture (13% byvolume).A scoping calculation indicated that hydrogen gas could enter the rocfrom a major pipe break at an initial rate of about 10,000 cubic feetper minute (cfm), and release all 16,000 cubic feet in about threeminutes. If all of the hydrogen were retained in the room, adetonatable concentration would exist. However, the mezzanine at thiplant is well ventilated with a ducted flow of 2,000 cfm of air and atotal flow of about 30,000 cfm coming from other sources. Thus, amore realistic calculation yielded an average hydrogen concentrationin the room of about 10%, which is equivalent to about 30 pounds ofhydrogen. This 10% concentration can be ignited and result in anexplosion.</
.( JUN- 3-93 THU 9:41 Gen. r iH. Licensing FAX NO. 203 r 5896 P.03At the subject plant, the room below the generator contains thehydrogen piping and generator output busses, their cooling and airconditioning systems, and the generator auxiliary equipment. Thisincludes the piping containing hydrogen and oil mixtures. An openingin the structure supporting the generator indirectly exposes some ofthe plant's IE and non-lE electrical busses to the effects of anexplosion or a fire. Based on the above estimated average hydrogengas concentration of lot by volume in a part of the mezzanine, aconservatively estimated blast pressure of 40 psi would have to becontained at the opening to prevent blast waves from exiting this areaand damaging the vital busses. To avoid the possibility ofconsequential damage, a recommendation will be made to the utility toprovide a blast shield or ignitors to mitigate or prevent a potentialblast.GE recommends that all BWR utilities perform an evaluation todetermine if a postulated generator coolant hydrogen explosionraubstantial sft? hazard at their facility.Northeast Utilities should evaluate this information as it relates toexisting or future plant equipment, conditions, procedures or plans.GE cannot determine if this information affects U. S. BWR utilities.GE is, therefore, notifying all U. S. BWR owners of this information.The owners of GE BWRs located outside the U. S. also will receive thisinformation.
Sincerely,J..RileyNuclear Services ManagerGE Nuclear Energycc: P.A. BlasioliL.D. DavisonE.A. DeBarbaR.T. HarrisH.F. HaynesH.P. RisleyW.D. RombergR.W. Tobin. GE SiteNoticeThis 10 CFR Part 21 information pertains only to GE BWRs. GE NuclearEnergy (GE-NE) has not considered or evaluated the applicability, ifany, of this information to any plant or facility other than GE BWRs.
IEnclosure 3GL 93-06October 25, 1993 LIST OF RECENTLY ISSUED GENERIC LETTERSGenericI +&&_Date ofT re Igala nrv'C..h;^tTecmiaA TnLeter L W .-UVAC%_&I aawuCIIR JUW_ IWV93-0589-10,SUPP. 593-0493-03LINE-ITEM TECHNICALSPECIFICATIONS IMPROVE-MENTS TO REDUCE SURVEILLANCEREQUIREMENTS FOR TESTINGDURING POWER OPERATIONINACCURACY OF MOTOR-OPERATED VALVEDIAGNOSTIC EQUIPMENTROD CONTROL SYSTEMFAILURE AND WITHDRAWALOF ROD CONTROL CLUSTERASSEMBLIES, 10 CFR 50.54(f)VERIFICATION OF PLANTRECORDSNRC PUBLIC WORKSHOP ONCOMMERCIAL GRADE PRO-CUREMENT AND DEDICATIONEMERGENCY RESPONSE DATASYSTEM TEST PROGRAMLIMITED PARTICIPATION BY NRCIN THE IAEA INTERNATIONALNUCLEAR EVENT SCALE09/27/9306/28/9306/21/93ALL HOLDERS OF OLs ORCPs FOR NPRsALL LICENSEES OFOPERATING NUCLEAR POWERPLANTS AND HOLDERS OFCONSTRUCTION PERMITS FORNUCLEAR POWER PLANTSALL HOLDERS OF OLs ORCPs FOR (W)-DESIGNEDNPRs EXCEPT HADDAM NECKALLCPsANDANDHOLDERS OF OLs ORFOR (CE)-DESIGNED(B&W)-DESIGN NPRsHADDAM NECK93-0293-0110/20/9303/23/9303/03/9312/31/92ALL HOLDERS OF OLs ORCPs FOR NPRsALL HOLDERS OF OLs ORCPs FOR NPRs AND ALLRECIPIENTS OF NUREG-0040,*LICENSEE CONTRACTOR ANDVENDOR INSPECTION STATUSREPORT" (WHITE BOOK)ALL HOLDERS OF OLs ORCPs FOR NPRs, EXCEPTFOR BIG ROCK POINT ANDFACILITIES PERMANENTLYOR INDEFINITELY SHUTDOWNALL HOLDERS OFOLs OR CPs FOR NPRs92-09OL -OPERATING LICENSECP -CONSTRUCTION PERMIT