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Home       > Electronic Reading       Room       > Document       Collections       > Generic       Communications >       Generic       Letters >       1997       > GL97004 
				UNITED STATES                         NUCLEAR REGULATORY COMMISSION                     OFFICE OF NUCLEAR REACTOR REGULATION                         WASHINGTON, D.C.  20555-0001                                October 7, 1997NRC GENERIC LETTER 97-04:  ASSURANCE OF SUFFICIENT NET POSITIVE SUCTION HEAD                           FOR EMERGENCY CORE COOLING AND CONTAINMENT HEAT                           REMOVAL PUMPS Addressees All holders of operating licenses for nuclear power plants, except those whohave permanently ceased operations and have certified that fuel has beenpermanently removed from the reactor vessel.  PurposeThe U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter(GL) to request that addressees submit information necessary to confirm theadequacy of the net positive suction head (NPSH) available for emergency corecooling (including core spray and decay heat removal) and containment heatremoval pumps.  BackgroundAs a result of recent inspection activities, licensee notifications, andlicensee event reports (LER), the NRC has identified a safety-significantissue that has generic implications and warrants action by the NRC to ensurethat the issue is adequately addressed and resolved.  The issue is that theNPSH available for emergency core cooling system (ECCS) (including core sprayand decay heat removal) and containment heat removal pumps may not be adequateunder all design-basis accident scenarios.  In some cases, this inadequacy may be a result of changes in plantconfiguration, operating procedures, environmental conditions, or otheroperating parameters over the life of the plant.  In other cases, a plant'sNPSH analysis may not bound all postulated events for a sufficient time, orassumptions used in the analysis may be non-conservative or inconsistent withassumptions and methodologies traditionally considered acceptable by thestaff.  For example, some licensees have recently discovered that they musttake new or additional credit for containment overpressure to meet the NPSHrequirements of the emergency core cooling system and containment heat removalpumps.  In the examples the NRC staff is familiar with, the need for creditingthis overpressure in NPSH analyses has arisen because of changes in plantconfiguration and operating conditions, and/or errors in prior NPSHcalculations.  As a result, the overpressure being credited by licensees maybe inconsistent with the plant's respective licensing basis.  9710060324.                                                            GL 97-04                                                            October 7, 1997                                                            Page 2 of 10 Current NPSH analyses (including any corresponding containment pressureanalyses) may not be available to the staff in docketed material (such asfinal safety analysis reports) because some licensees have changed theiranalyses.  Consequently, this generic letter requests that addressees providecurrent information regarding the NPSH analyses for emergency core cooling andcontainment heat removal pumps.  This generic letter applies only to ECCS andcontainment heat removal pumps that meet the following criteria:      (1)   pumps that take suction from the containment sump or          suppression pool following a design-basis loss-of-coolant          accident (LOCA) or secondary line break, or     (2)   pumps used in "piggyback" operation that are necessary for          recirculation cooling of the reactor core and containment          (that is, pumps that are supplied by pumps which take suction          directly from the sump or suppression pool).New NPSH analyses are neither requested nor required to be performed torespond to this information request.  However, new NPSH analyses may bewarranted if an addressee determines that changes in plant design orprocedures have occurred which may have reduced the available NPSH.  In suchcases, each affected addressee must take appropriate corrective action torestore its facility to compliance, in accordance with the requirements statedin Appendix B to 10 CFR Part 50.  The following is a sample of the NRC staff's recent findings concerning theNPSH issues addressed by this generic letter.  Haddam NeckIn 1986 and 1995, the licensee identified conditions for which the NPSHavailable for residual heat removal (RHR) pumps may be insufficient when thepumps are operating in the emergency core cooling mode.  In 1986, the licenseedetermined that the only extant NPSH analysis, which was performed in 1979 aspart of the Systematic Evaluation Program, did not properly account forhydraulic losses in suction piping.  As a result, that analysis erroneouslyindicated that containment overpressure was not needed to satisfy NPSHrequirements for the pumps in the recirculation mode of operation.  A subsequent analysis showed that the licensee needed to take credit for 41.36kPa (6 psig) of containment overpressure.  In another analysis conducted in1995 using increased service water temperature, the licensee found thatadditional containment overpressure was necessary to meet NPSH requirementsfor the same pumps.  This additional overpressure constituted a significantfraction of the peak calculated containment accident pressure.  .                                                            GL 97-04                                                            October 7, 1997                                                            Page 3 of 10On August 30, 1996, the licensee reported in LER 96-016 that calculationsrecently performed to determine the NPSH available for the RHR pumps may havebeen in error for the alternate, short-term recirculation flow path, becauseof insufficient containment overpressure for a period of pump operation.  Thelicensee attributed this error to its failure to fully analyze the containmentpressure and sump temperature responses under design-basis accidentconditions.  Maine YankeeIn July and August 1996, an NRC Independent Safety Assessment Team (ISAT)conducted an inspection to determine if Maine Yankee was operating inconformance with its design and licensing bases.  During that inspection, theISAT identified potential weaknesses in the NPSH analysis conducted by thelicensee for the containment spray pumps.  These potential weaknesses includedconcerns regarding the validity of the containment sump temperature analysis,incorrect calculation of bounding pump suction head losses, and use of a hot-fluid correction factor to reduce NPSH requirements.  The licensee's calculation of record, performed in 1995 for a power level of2700 thermal megawatts (MWt) and which does not include the hot-fluidcorrection factor, indicates that the available NPSH for the containment spraypumps would be below the required NPSH for the first 5 minutes after pumpsuction is switched from the refueling water storage tank to the recirculationsump.  When the licensee repeated the analysis using the hot-fluid correctionfactor (the use of which the ISAT viewed as a non-conservative assumption asimplemented by Maine Yankee), the available NPSH was only slightly greaterthan the required NPSH for the same 5-minute period.  For the remainder of thetransient, the licensee's analysis showed that NPSH available to thecontainment spray pumps would exceed the amount required.  As a basis for thecontention that the containment spray pumps were operable despite the 5-minuteperiod with available NPSH below the required NPSH, the licensee cited recentpump tests showing that the pumps could operate for a 15-minute period withNPSH below the required value without damage to the hydraulic performance ormechanical integrity of the pumps.  The licensee performed another analysis for a power level of 2440 MWt, whichshowed that adequate NPSH margin would be available for the containment spraypumps in the recirculation mode of operation.  This analysis did not includeuse of the hot-fluid correction factor.  The ISAT concluded that it wasappropriate to consider the containment spray pumps operable at a power levelof 2440 MWt.    PilgrimAs indicated in the NRC safety evaluation for licensing of the Pilgrim plant,and in documents referenced by that evaluation, containment overpressure wasnot necessary to satisfy RHR and core spray pump NPSH requirements at the timeof licensing.  When the plant was modified in 1984, the licensee's safetyevaluation related to the modification stated  that the available NPSH wasdetermined assuming (1) maximum debris loading conditions on the sumpstrainers for the RHR and core spray pumps and (2) no credit for containment .                                                            GL 97-04                                                            October 7, 1997                                                            Page 4 of 10overpressure.  The licensee reaffirmed this assumption on April 14, 1994, inits response to NRC Bulletin 93-02, "Debris Plugging of Emergency Core CoolingSuction Strainers," dated March 23, 1993, stating that the NPSH available tothe residual heat removal and core spray pumps was analyzed assuming nooverpressure condition in the torus.However, in an analysis conducted by the licensee in 1995 in support of aproposal to raise the design seawater injection temperature to 75øF, creditwas needed and taken for containment overpressure.  At the time of thisanalysis, the licensee also indicated that the assumption of no overpressurein the torus, stated in its response to Bulletin 93-02, was incorrect.  Thisexample illustrates that the potential exists that other licensees may havemade modifications to their plants that could be inconsistent with the plant'slicensing basis, and could reduce the NPSH available to the ECCS pumps.    Crystal River, Unit 3In July 1996, an NRC inspection team conducted an Integrated PerformanceAssessment of Crystal River, Unit 3.  As part of that assessment, the teamreviewed the licensee's calculation which established the minimum post-LOCAreactor building water level required to ensure that adequate NPSH would beavailable for the reactor building spray pumps.  When the team compared thislevel with the minimum predicted level, they found that for one of the pumps,there was only a slight difference between the available water level and thelevel required to ensure adequate NPSH during the post-accident recirculationphase of pump operation.  The team found that the licensee used non-conservative assumptions incalculating the available NPSH for the spray pump.  For example, the licenseefailed to account for uncertainty in data regarding the required NPSH, as wellas for uncertainties associated with the hydraulic resistance of check valvesin the spray lines.  In addition, the licensee  used a hot fluid correctionfactor to reduce the required NPSH without considering the effects of non-condensable gases in the pumped fluid.  Conservative assumptions included inthe licensee's calculation were those detailed in Regulatory Guide (RG) 1.1,"Net Positive  Suction Head for Emergency Core Cooling and Containment HeatRemoval System Pumps," dated November 2, 1970 (originally Safety Guide 1),regarding the use of maximum reactor building fluid temperature and lack ofcredit for containment overpressure.  The team concluded that the non-conservative assumptions used in thelicensee's NPSH calculation raise questions concerning the cavitation-freeoperation of reactor building spray pump 1B during the recirculation phase ofoperation.  However, the team also concluded that this issue did notconstitute an immediate safety concern since the licensee's calculationsconservatively assumed no credit for containment overpressure and used themaximum expected reactor building water temperature.   DresdenBy letter dated January 13, 1997, the licensee for Dresden submitted a licenseamendment request for approval of 13 kPa (2 psig) of containment overpressurefor the first 10 minutes .                                                            GL 97-04                                                            October 7, 1997                                                            Page 5 of 10following a design-basis LOCA.  This overpressure is necessary to compensatefor an NPSH deficiency for the low pressure coolant injection (LPCI) and corespray pumps.  The licensee identified the need for overpressure afterdiscovering that an incorrect value for the ECCS suction strainer head losshad been used in the design-basis NPSH calculation.  As part of a design-basisreview, the licensee determined that the actual head loss across the suctionstrainers was 1.8 m (5.8 feet) for clean strainers, rather than the 0.30 m(1 foot) head loss assumed in Dresden's original design basis as documented inthe final safety analysis report and vendor drawings.  Because the licensee could not determine with certainty if overpressure waspart of the original Dresden licensing basis, the licensee concluded that theuse of overpressure constituted an unreviewed safety question and thereforerequested staff approval to credit overpressure.  In a license amendment datedJanuary 28, 1997, the staff approved the requested use of 13 kPa (2 psig) ofcontainment overpressure.  In a subsequent license amendment issued onApril 30, 1997, the staff approved the use of a maximum of 65 kPa (9.5 psig)of containment overpressure for NPSH, for the first 240 seconds following adesign-basis LOCA.  The need for this greater amount of overpressure aroseprimarily because of a higher calculated suppression pool temperature thanthat used in the analysis to support 13 kPa (2 psig) of overpressure.Monticello In a report submitted to the NRC on April 15, 1997, pursuant to 10 CFR 50.72,the licensee for Monticello reported that the NPSH available to its core spraypumps may not meet the required NPSH under all accident conditions.  Thelicensee discovered this possibility during a review of ECCS pump NPSHrequirements, when a higher head loss than had previously been assumed for theECCS suction strainers was calculated.  During discussions with the licensee,the staff learned that the head loss across the suction strainers isapproximately 3.57 m (11.7 feet) per 38,000 liters/minute (10,000 gpm), ratherthan the 0.3048 m (1 foot) per 38,000 liters/minute (10,000 gpm) assumed inthe original design-basis analysis.  The licensee determined that for a recirculation line break with a singlefailure of the LPCI loop select logic, and with credit for containmentoverpressure, the core spray pumps would have an NPSH deficit and the LPCIpumps would have approximately 0.15 m (0.5 feet) of margin in NPSH.  Followingdiscovery of the NPSH condition, the licensee conducted an operabilityevaluation of the LPCI and core spray pumps, and made this evaluationavailable to the staff for review.  Subsequently, on May 9, 1997, the licenseefor Monticello commenced a voluntary shutdown of the plant because of thepossible NPSH deficit for the ECCS pumps that would occur as a result ofpostulated clogging of the ECCS suction strainers under design-basis LOCAconditions.   Related Generic CommunicationsOn October 22, 1996, the staff issued Information Notice (IN) 96-55,"Inadequate Net  Positive Suction Head of Emergency Core Cooling andContainment Heat Removal Pumps Under Design Basis Accident Conditions," toalert addressees to recent discoveries by .                                                            GL 97-04                                                            October 7, 1997                                                            Page 6 of 10licensees of possible scenarios for which the NPSH available for ECCS andcontainment heat removal pumps is insufficient.  Earlier INs describingsimilar events include IN 87-63, "Inadequate Net Positive Suction Head in LowPressure Safety Systems," dated December 9, 1987, and IN 88-74, "PotentiallyInadequate Performance of ECCS in  PWRs During Recirculation OperationFollowing a LOCA," dated September 14, 1988.  DiscussionIt is important that the emergency core cooling (including core spray anddecay heat removal) and containment spray system pumps have adequate NPSHavailable to ensure that the systems can reliably perform their intendedfunctions under all design-basis LOCA conditions.  Inadequate NPSH couldcause voiding in the pumped fluid, resulting in pump cavitation.  While someECCS and containment heat removal pumps can operate for relatively shortperiods of time while cavitating, prolonged operation of any pump undercavitation conditions can cause pump damage with potential common-mode failureof the pumps.   Such common-mode failure would result in the inability of theECCS to provide adequate long-term core cooling and/or the inability of thecontainment spray system to maintain  the containment pressure and temperaturebelow design limits.  This generic letter addresses situations in which the NPSH available to theECCS and containment heat removal pumps may be inadequate as a result ofchanging plant conditions and/or errors and non-conservative assumptions inNPSH calculations.  In some cases, NPSH reanalyses conducted to support plantmodifications may result in a substantial reduction of margin in availableNPSH or a change in the original design basis of the plant.  In particular,recent examples indicate that licensees have credited containment overpressureto satisfy NPSH requirements in response to changing plant conditions anderrors discovered in earlier NPSH calculations.  RG 1.1 establishes the regulatory position that emergency core cooling andcontainment heat removal systems should be designed so that adequate NPSH isprovided to system pumps assuming maximum expected temperatures of pumpedfluids and no increase in containment pressure from that present before anypostulated LOCAs.  NRC Standard Review Plan (SRP) 6.2.2, "Containment HeatRemoval Systems" (NUREG-0800, Revision 4, dated October 1985) clarifies RG 1.1by stating that the NPSH analysis should be based on the assumption that thecontainment pressure equals the vapor pressure of the sump water, in order toensure that credit is not taken for containment pressurization during thetransient.  As part of licensing and Systematic Evaluation Plan reviews, theNRC staff has, in the past, selectively allowed limited credit for acontainment pressure that is above the vapor pressure of the sump fluid (i.e.,an overpressure) to satisfy NPSH requirements on a case-by-case basis.  Requested InformationOn the basis of the preceding discussion and examples, addressees arerequested to review, for each of their respective reactor facilities, thecurrent design-basis analyses used to .                                                            GL 97-04                                                            October 7, 1997                                                            Page 7 of 10determine the available NPSH for the emergency core cooling (including corespray and decay heat removal) and containment heat removal pumps that meeteither of the following criteria:     (1)   pumps that take suction from the containment sump or          suppression pool following a design-basis LOCA or secondary          line break, or     (2)   pumps used in "piggyback" operation that are necessary for          recirculation cooling of the reactor core and containment          (that is, pumps that are supplied by pumps which take suction          directly from the sump or suppression pool).Based on this review, within 90 days from the date of this generic letter,addressees are requested to provide the information outlined below for each oftheir facilities.  New NPSH analyses are neither requested nor required.  1.  Specify the general methodology used to calculate the head loss associated    with the ECCS suction strainers.2.  Identify the required NPSH and the available NPSH.3.  Specify whether the current design-basis NPSH analysis differs from the    most recent analysis reviewed and approved by the NRC for which a safety    evaluation was issued.     4.  Specify whether containment overpressure (i.e., containment pressure above    the vapor pressure of the sump or suppression pool fluid) was credited in    the calculation of available NPSH.  Specify the amount of overpressure    needed and the minimum overpressure available.5.  When containment overpressure is credited in the calculation of available    NPSH, confirm that an appropriate containment pressure analysis was done    to establish the minimum containment pressure. Required ResponseWithin 30 days from the date of this generic letter, each addressee isrequired to submit a written response indicating (a) whether or not therequested information will be submitted, and (b) whether or not the requestedinformation will be submitted within the requested time period.  Addresseeswho choose not to submit the requested information, or are unable to submitthe information within the requested period, must describe in their responsean alternative course of action that is proposed to be taken, including thebasis for the acceptability of the proposed alternative.After reviewing responses to this generic letter, the NRC staff will notifyindividual addressees if concerns are identified with regard to theirfacilities.  .                                                            GL 97-04                                                            October 7, 1997                                                            Page 8 of 10Addressees should submit the required written response to the U.S. NuclearRegulatory Commission, ATTN:  Document Control Desk, Washington, D.C.  20555-0001, under oath or affirmation under the provisions of Section 182a, AtomicEnergy Act of 1954, as amended, and 10 CFR 50.54(f).Backfit DiscussionThis generic letter only requests information from addressees under theprovisions of Section 182a of the Atomic Energy Act of 1954, as amended, and10 CFR 50.54(f).  The requested information will enable the staff to determinewhether addressees' NPSH analyses for the emergency core cooling (includingthe core spray and decay heat removal) and containment heat removal systempumps conform with the current licensing basis for their respectivefacilities, including the licensing safety analyses and the principal designcriteria which require and/or commit that safety-related components andsystems be provided to mitigate the consequences of design-basis accidents. In particular, 10 CFR 50.46(a)(1)(i), which addresses the ECCS acceptancecriteria for light-water nuclear power reactors, requires in part that thecalculated cooling performance of the ECCS following a postulated LOCAconforms to the criteria set forth in 10 CFR 50.46, including provisions forpeak cladding temperature and long-term cooling.  The potential for loss ofadequate NPSH for ECCS pumps, and the cavitation that would result, raises theconcern that the ECCS would not be capable of maintaining the peak claddingtemperature below acceptable limits, and/or would not be capable of providingcore cooling over the duration of postulated accident conditions, as requiredby 10 CFR 50.46.  Furthermore, the licensing bases of some plants credit the operation ofcontainment sprays for pressure control as well as for fission productcontrol.  The potential for the loss of adequate NPSH for containment spraypumps, and the cavitation that would result, raises the concern thatcontainment spray would not be capable of reducing and maintaining thecontainment pressure and temperature below design values and would not becapable of reducing the radiological dose consequences consistent with plants'licensing bases. Considering the safety significance of removing heat from the containmentatmosphere and cooling the reactor core following a design-basis accident, therequested information is needed to verify addressee compliance with licensing-basis commitments regarding the performance of emergency core cooling(including core spray and decay heat removal) and containment heat removalsystem pumps.  The evaluation required by 10 CFR 50.54(f) to justify thisinformation request is included in the preceding discussion.Federal Register NotificationA notice of opportunity for public comment was published in the FederalRegister on February 20, 1997 (62 FR 7806) to solicit public comments on thedraft of this generic letter.  A total of 17 comments were received frominterested parties, including one industry group, one legal group affiliatedwith the nuclear power industry, and two licensees.  When .                                                            GL 97-04                                                            October 7, 1997                                                            Page 9 of 10redundant comments are considered, 12 distinct comments were identified by thestaff.  Copies of the staff evaluation of these comments have been madeavailable in the NRC Public Document Room.Paperwork Reduction Act StatementThis generic letter contains information collections that are subject to thePaperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.).  These informationcollections were approved by the Office of Management and Budget, approvalnumber 3150-0011, which expires on August 31, 2000.  The public reporting burden for this collection of information is estimated toaverage 200 hours per response, including the time for reviewing instructions,searching existing data sources, gathering and maintaining the data needed,and completing and reviewing the collection of information.  The NRC isseeking public comment on the potential impact of the collection ofinformation contained in the generic letter and on the following issues:1.  Is the proposed collection of information necessary for the proper    performance of the functions of the NRC, including whether the information    will have practical utility? 2.  Is the estimate of burden accurate?3.  Is there a way to enhance the quality, utility, and clarity of the    information to be collected?4.  How can the burden of the collection of information be minimized,    including the use of automated collection techniques?Send comments on any aspect of this collection of information, includingsuggestions for reducing this burden, to the Information and RecordsManagement Branch, T-6 F33, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001, and to the Desk Officer, Office of Information and RegulatoryAffairs, NEOB-10202 (3150-0011), Office of Management and Budget, Washington,DC  20503.The NRC may not conduct or sponsor, and a person is not required to respondto, a collection of information unless it displays a currently valid OMBcontrol number..                                                            GL 97-04                                                            October 7, 1997                                                            Page 10 of 10If you have any questions about this matter, please contact one of thetechnical contacts listed below or the appropriate Office of Nuclear ReactorRegulation (NRR) project manager.                                          signed by                                    Jack W. Roe, Acting Director                                    Division of Reactor Program Management                                    Office of Nuclear Reactor RegulationTechnical contacts:  William O. Long, NRR                     301-415-3026                     E-mail:  wol@nrc.gov                     Richard M. Lobel, NRR                     301-415-2865                     E-mail:  rml@nrc.govLead Project Manager:  T.J. Kim, NRR                       301-415-1392                       E-mail:  tjk3@nrc.govAttachment:  List of Recently Issued NRC Generic Letters