NRC Generic Letter 1979-05

NRC Generic Letter 1979-005: Information Relating to Categorization of Recent Regulatory Guides by Regulatory Requirements Review Committee - Perkins Nuclear Station

ML031320224
Person / Time
Issue date:	01/26/1979
From:	Boyd R S Office of Nuclear Reactor Regulation
To:	Dail L C Duke Energy Corp
References
NUDOCS 7902210145, GL-79-005
Download: ML031320224 (41)

rb..UNITED STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON, D. C. 20555CG<- 7S-January 26, 1979Docket No:STNSTNSTN50- 48850-48950 490NRC PDRLocal PDRLWR #4 FileC. MoonM. ServiceH. DentonE. CaseR. BoydD. Ross2 R. MattsonR. DeYoungD. VassalloF. WilliAmsS. VargaELDIE (3)ACRS (16)J. Buchanan,NSICT. Abernathy, TICMr. L. C. Dail, Vice PresidentDesign Engineering DepartmentDuke Power CompanyP. 0. Box 33189Charlotte, North Carolina 28242Dear Mr. Dail:SUBJECT: INFORMATION RELATING TO CATEGORIZATION OF RECENT REGULATORYGUIDES BY THE REGULATORY REQUIREMENTS REVIEW COMMITTEE -PERKINS NUCLEAR STATIONWe have recently advised utilities with plants in the post-CP phase of thereactor licensing process of the status of NRC staff review and use ofrecently-approved regulatory guides, and have indicated how these guideswould be used in the Operating License review of their Final SafetyAnalysis Reports. Such information, while not directly applicable toyou at this time, may nonetheless be useful to you for your futureplanning. The text of our letter to these utilities is the following:"SUBJECT:IMPLEMENTATION OF STAFF REVIEW REQUIREMENTS -(Name ofPlant) -OPERATING LICENSE REVIEWDuring the last several years, we have reviewed and approved severalnew regulatory guides and branch technical positions or othermodifications to existing staff positions. Our practice is thatsubstantive changes in staff positions be considered by the NRC'sRegulatory Requirements Review Committee (RRRC) which then recommendsa course of action to the Director, Office of Nuclear ReactorRegulation (NRR). The recommended action includes an implementationschedule. The Director's approval then is used by the NRR staffas review guidance on individual licensing matters. Some of theseactions will affect your application. This letter is intended tobring you up to date on these changes in staff positions so that youmay consider them in your Final Safety Analysis Report (FSAR)preparation.IT"cco .lb1/25/79a/79#4 DPM #'DVassallo1/4./79790 2210/q5'tI[ E F

-2-"The RRRC applies a categorization nomenclature to each of itsactions. (A copy of the summary of RRRC Meeting No. 31 concerningthis categorization is attached as Enclosure 1.) Category 1matters aare those to be applied to applications in accordance withthe implementation section of the published guide. We have enclosedlists of actions which are either Category 2 or Category 3, whichare defined as follows:Category 2: A new position whose applicability is to be determinedon a case-by-case basis. You should describe the extentto which your design conforms, or you should describean acceptable alternate, or you should demonstratewhy conformance is not necessary.Category 3: Conformance or an acceptable alternative is required.If you do not conform, or do not have an acceptablealternate, then staff-approved design revisions will berequired."We believe that providing you with a list of the Category 2 and 3matters approved to date will be useful in your FSAR preparation,and they will be an essential part of our operating license review.Enclosure 2 is a list of the Category 2 matters. Enclosure 3 is alist of the Catetory 3 matters."In addition to the RRRC categories, there also exists an NRRCategory 4 list which are those matters not yet reviewed by the RRRC,but which the Director, NRR, has deemed to have sufficient attributesto warrant their being addressed and considered in ongoing reviews.These matters will be treated like Category 2 matters until suchtime as they are reviewed by the RRRC, and a definite implementationprogram is developed. A current list of Category 4 matters isattached (Enclosure 4). These also should be considered in yourFSAR."In some instances the items in the enclosures may not be applicableto your application. Also, we recognize that your application may,in some instances, already conform to the stated staff positions.In your FSAR you should note such compliance."If you have any questions please let us know."

-3-For your information, I am enclosing a set of the enclosures that accompaniedthese individual letters. These enclosures list the present Category 1-4matters discussed in the letter.Sincerely,9oger .ctorDivision of ProjectOffice of Nuclear Reactor RegulationEnclosures:As statedcc: See next page Duke Power Companyccs:William L. Porter, Esq.Associate General CounselDuke Power CompanyCharlotte, North Carolina28242J. Michael McGarry, III, Esq.Debevoise & Liberman700 Shoreham Building806 Fifteenth Street, N. W.Washington, D. C. 20005William A. Raney, Jr.Special Deputy Attorney GeneralAttorney for the State ofNorth CarolinaDepartment of JusticeP. 0. Box 629Raleigh, North Carolina 27602Mary Apperson Davis, ChairmanYadkin River CommitteeRoute 4, Box 261Mocksville, North Carolina 27028Thomas S. Erwin, Esq.P. 0. Box 928Raleigh, North Carolina 27602David SpringerThe Point FarmRoute 4Mocksville, North Carolina 27028Elizabeth S. Bowers, Esq.ChairmanAtomic Safety and Licensing BoardU. S. Nuclear Regulatory CommissionWashington, D. C. 20555Dr. Donald P. deSylvaAssociate Professor of Marine ScienceRosenstiel School of Marine andAtmospheric ScienceUniversity of MiamaMiami, Florida 33149Dr. Walter H. Jordan881 W. Outer DriveOak Ridge, Tennessee37830Allan S. Rosenthal, ChiarmanAtomic Safety and LicensingAppeal BoardU. S. Nuclear Regulatory CommissionWashington, D. C. 20555Dr. John H. BuckAtomic Safety and Licensing Appeal BoardU. S. Nuclear Regulatory CommissionWashington, D. C. 20555Richard S. Salzman, Esq.Atomic Safety and Licensing Appeal BoardU. S. Nuclear Regulatory CommissionWashington, D. C. 20555William G. Pfefferkorn, Esq.2124 Wachovia BuildingWinston-Salem, North Carolina27101Richard P. Wilson, Esq.Assistant Attorney GeneralS. C. Attorney General's OfficeP. 0. Box 11549Columbia, South Carolina 29211 UNITED STATESNU-tEAR REGULATORY COMMISS..WASHINGTON. D. C. 20555SEP 2 4 175Lee V. Gossick.Executive Director for OperationsREGULATORY REQUIREMENTS REVIEW COMMITTEE MEETING NO. 31,JULY 11, 19751. The Committee discussed issues related to the implementation ofRegulatory Guides on existing plants and the concerns expressedin the June 24, 1974 memorandum, A. Giambusso to E. G. Case,subject: REGULATORY GUIDE IMPLEMIENTATION, and made the followingrecommendations and observations:a. Approval of new Regulatory Guides and approval of revisionsof existing guides should move forward expeditiously in orderthat the provisions of these regulatory guides be availablefor use as soon as possible in on-going or future staff reviewsof license applications. The Committee noted that over therecent past, the approval of proposed regulatory guides whosecontent is acceptable for these purposes has experiencedsignificant delays in RRRC review pending the determinationof the applicability of the guide to existing plants, oftenrequiring significant staff effort. To avoid these delays,the Comnittee concluded that, henceforth, approval of proposedregulatory guides should be uncoupled from the considerationof their backfit applicability.b. The implementation section of new regulatory guides shouldaddress, in general, only the applicability of the guide toapplications in the licensing review process using, in so faras possible, a standard approach of applying the guide tothose applications docketed 8 months after the issuance dateof the guide for comment. Exceptions to this general approachwill be handled on a case-by-case basis.c. The regulatory position of each approved proposed guide (orproposed guide revision) will be characterized by the Committeeas to its backfitting potential, by placing it in one of threecategories:Category 1 -Clearly forward fit only. No further staffconsideration of possible backfitting is required.ENCLOSURE 1 Lee V. Gossick-2-Category 2 -Further staff consideration of the need for back-fitting appears to be required for certain identified items ofthe regulatory position--these individual issues are such thatexisting plants need to be evaluated to determine their statuswith regard to these safety issues in order to determine theneed for backfitting.ategory 3 -Clearly backfit. Existing plants should beevaluated to determine whether identified items of theregulatory position are resolved in accordance with theguide or by some equivalent alternative.From time to time, for a specific guide, there will probably besome variation among these categories or even within a category,and these three broad category characterizations will bequalified as required to meet a particular situation.d. It is not intended that the Committee categorization appearin the guide itself. The purpose of the categorization isto indicate those items of the reculatory position for whichthe Committee can make a specific backfit recommendationwithout additional staff work (Categories 1 and 3), and toindicate those items for which additional staff work isrequired in order to determine backfit considerations(Category 2).e. The Committee recommends that for approved guides in Category 2,staff efforts be initiated in parallel with the process leadingto publication of the guide in order that specific backfitrequirements for existing plants be determined within areasonable period of time after publication of the guide.f. The Committee observed that more attention needs to be givento the identification of acceptable alternatives to thepositions outlined in the guides in order to provide additionaloptions and flexibility to applicants and licensees, with thepossible benefits of additional innovation and explorationin the solution of safety issues.2. The Committee reviewed the proposed Regulatory Guide l.XX: THERMALOVERLOAD PROTECTION FOR MOTORS OM HrOTOR-OPERATED VALVES andrecommended approval. This guide was characterized by the Committeeas Category 1 -no backfitting, with the stipulation that as anappropriate occasion presented itself in conjunction with thereview of some particular aspect of existing plants, the Lhermaloverload protection provisions be audited.ENCLOSURE 1 (CONT'D)

Lee V. Gossick -3-3. JThe Committee reviewed the proposed Regulatory Guide 1.XX:INSTRUMENT SPANS AND SETPOINTS and recommended approvalsubject to the following comment:Paragraph 5 of Section C (page 4 of the proposed Guide)should be reworded in light of Committee comments, tothe satisfaction of the Director, Office of StandardsDevelopment. This guide was characterized by theCommittee as Category 1 -no backfit.4. The Committee reviewed Proposed Regulatory Guide 1.97:INSTRUWIENTATION FOR LIGHT WATER COOLED NUCLEAR POWER PLANTSTO ASSESS PLANT CONDITIONS DURING AND FOLLOWING AN ACCIDENTand deferred further consideration to a later meeting inorder to permit incorporation of recent comments by theDivision of Technical Review.Edson G. ase, ChairmanRegulatory Requirements ReviewCommitteeENCLOSURE 1 (CONT'D)

SeptembL... 15, 1978CATEGORY 2 MATTERSDocumentNumber Revision Date TitleRG 1.27 2 1/76 Ultimate Heat Sink for NuclearPower PlantsRG 1.52 1 7/76 Design, Testing, and MaintenanceCriteria for Engineered-Safety-Feature Atmosphere Cleanup SystemAir Filtration and Adsorption Unitsof Light Water Cooled Nuclear PowerPlants (Revision 2 has been publishedbut the changes from Revision 1 toRevision 2 may, but need not,be considered.RG 1.59 2 8/77 Design Basis Floods for NuclearPower PlantsRG 1.63 2 7/78 Electric Penetration Assemblies inContainment Structures for LightWater Cooled Nuclear Power PlantsRG 1.91 1 2/78 Evaluation of Explosions Postulatedto Occur on Transportation RoutesNear Nuclear Power Plant SitesRG 1.102 1 9/76 Flood Protection for Nuclear PowerPlantsRG 1.105 1 11/76 Instrument SetpointsRG 1.108 1 8/77 Periodic Testing of DieselGenerator Units Used as OnsiteElectric Power Systems at NuclearPower PlantsRG 1.115 1 7/77 Protection Against Low-TrajectoryTurbine MissilesRG 1.117 1 4/78 Tornado Design ClassificationRG 1.124 1 1/78 Service Limits and LoadingCombinations for Class 1Linear Type Component SupportsRG 1.130 0 7/77 Design Limits and Loading Combinationsfor Class 1 Plate- and Shell-TypeComponent Supports(Continued)ENCLOSURE 2 CATEGORY 2 MATTERS (CONT'D)ContinuedDocument.,--- --kfr.nc; ae nna taTitleNumrer FVI v 1%*11 .U_RG 1.13701/78Fuel Oil Systems for StandbyDiesel Generators (Paragraph C.2)RG 8.823/77Information Relevant to Ensuringthat Occupational RadiationExposures at Nuclear Power StationsWill be as Low as is ReasonablyAchievable (Nuclear Power Reactors)BTP ASB9.5-1BTP MTEB 5-7RG 1.1411Guidelines for Fire Protection forNuclear Power Plants (See ImplementationSection, Section D)4/77Material Selection and ProcessingGuidelines for BWR Coolant PressureBoundary PipingContainment Isolation Provisionsfor Fluid Systems04/78-2-ENCLOSURE 2 (CONT'D)

September 15, 1978CATEGORY 3 MATTERSDocumentNumber Revision Date TitleRG 1.99 1 4/77 Effects of Residual Elements onPredicted Radiation Damage to-. Reactor Vessel Materials (ParagraphsC.1 and C.2.RG 1.101 1 3/77 Emergency Planning 'for NuclearPower PlantsRG 1.114 1 11/76 Guidance on Being Operator at theControls of a Nuclear Power PlantRG 1.121 0 8/76 Bases for Plugging Degraded PWRSteam Generator TubesRG 1.127 1 3/78 Inspection of Water-Control StructuresAssociated with Nuclear Power PlantsRSB 5-1 1 1/78 Branch Technical Position: Design Require-ments of the Residual Heat Removal SystemRSB 5-2 0 3/78 Branch Technical Position: ReactorCoolant System OverpressurizationProtection (Draft copy attached)RG 1.97 1 8/77 Instrumentation for Light WaterCooled Nuclear Power Plants toAssess Plant Conditions Duringand Following an Accident(Paragraph C.3 -with additionalguidance on paragraph C.3.d tobe provided later)RG 1.68.2 1 7/78 Initial Startup Test Program toDemonstrate Remote ShutdownCapability for Water-CooledNuclear Power PlantsRG 1.56 1 7/78 Maintenance of Water Purity inBoiling Water ReactorsAttachment:BTP RSB 5-2 (Draft)ENCLOSURE 3 P- RAFTBRANCH TECHNICAL POSITION RSB 5-2'OVERPRESSURIZATION PROTECTION OF PRESSURIZED WATER REACTORSWHILE OPERATING AT LOW TEMPERATURESA. BackgroundGeneral Design Criterion 15 of Appendix A, 10 CFR 50, requires that "theReactor Coolant System and associated auxiliary, control, and protectionsystems shall be designed with sufficient margin to assure that thedesign conditions of the reactor coolant pressure boundary are notexceeded during any condition of normal operation, including anticipatedoperational occurrences."Anticipated operational occurrences, as defined in Appendix A of 10 CFR 50,are "those conditions of normal operation which are expected to occur oneor more times during the life of the nuclear power unit and include butare not limited to loss of power to all recirculation pumps, tripping ofthe turbine generator set, isolation of the main condenser, and loss ofall offsite power."Appendix G of 10 CFR 50 provides the fracture toughness requirements forreactor pressure vessels under all conditions. To assure that theAppendix G limits of the reactor coolant pressure boundary are notexceeded during any anticipated operational occurrences, TechnicalSpecification pressure-temperature limits are provided for operatingthe plant.The primary concern of this position is that during startup and shutdownconditions at low temperature, especially in a water-solid condition,the reactor coolant system pressure might exceed the reactor vesselpressure-temperature limitations in the Technical Specificationsestablished for protection against brittle fracture. This inadvertentoverpressurization could be generated by any one of a variety of mal-functions or operator errors. Many incidents have occurred in operatingplants as described in Reference 1.Additional discussion on the background of this position is containedin Reference 1.ENCL 3 (CONT)

-2- L[AF~B. Branch Position1. A system should be designed and installed which will preventexceeding the applicable Technical Specifications and Appendix Glimits for the reactor coolant system while operation at lowtemperatures. The system should be capable of relieving pressureduring all anticipated overpressurization events at a rate sufficientto satisfy the Technical Specification limits, particularly whilethe reactor coolant system is in a water-solid condition.2. The system must be able to perform its function assuming any singleactive component failure. Analyses using appropriate calculationaltechniques must be provided which demonstrate that the system willprovide the required pressure relief capacity assuming the mostlimiting single active failure. The cause for initiation of theevent, e.g., operator error, component malfunction, will not beconsidered as the single active failure. The analysis should assumethe most limiting allowable oDerating conditions and systemsconfiguration at the time of the postulatea cause of the overDressureevent. All potential overpressurization events must be consideredwhen establishing the worst case event. *Some events may beprevented by protective interlocks or by locking Out power.tihse events should be reviewed on an individual basis. If thelnLerlock/piower lockout is acceptable, it car, be excluded fromf lilp aialyses provided the controls to prevent the event arein the plant Technical Specifications.3. The system must meet the design requirements of IEEE 279 (seeImplementation). The system may be manually enabled, however,the electrical instrumentation and control system must providealarms to alert the operator to:a. properly enable the system at the correct plant conditionduring cooldown,b. indicate if a pressure transient is occurring.4. To assure operational readiness, the overpressure protection systemmust be tested in the following manner:a. A test must be performed to assure operability of the systemelectronics prior to each shutdown.b. A test for valve operability must, as a minimum be conductedas specified in the ASME Code Section XI.c. Subsequent to system, valve, or electronics maintenance, a teston that portion(s) of the system must be performed prior todeclaring the system operational.ENCL 3 (CONT)

-3-5. The system must meet the requirements of Regulatory Guide 1.26.'Quality Group Classifications and Standards for Water-, Steam-,and Radioactive-Waste-Containing Components of Nuclear Power Plants"and Section III of the ASME. Code.-6. The overpressure protection system must be designed to functionduring an Operating Basis Earthquake. It must not compromise thedesign criteria of any other safety-grade system with which itwould interface, such that the requirements of Regulatory Guide1.29, "Seismic Design Classification" are met.7. The overpressure protection system must not depend on theavailability of offsite power to perform its function.8. Overpressure protection systems which take credit for an activecomponent(s) to mitigate the consequences of an overpressurizationevent must include additional analyses considering inadvertentsystem initiation/actuation or provide justification to show thatexisting analyses bound such an event.C. ImplementationThe Branch Technical Position, as specified in Section B, will be usedin the review of all Preliminary Design Approval (PDA), Final DesignApproval (FDA), Manufacturing License (ML), Operating License (OL), andConstruction Permit (CP) applications involving plant designs incorporatingpressurized water reactors. All aspects of the position will be applicableto all applications, including CP applications utilizing the replicationoption of the Commission's standardization program, that are docketedafter March 14, 1978. All aspects of the position, with the exceptionof reasonable and justified deviations from IEEE 279 requirements, willbe applicable to CP, OL, ML, PDA, and FDA applications docketed priorto March 14, 1978 but for which the licensing action has not beencompleted as of March 14, 1978. Holders of appropriate PDA's will beinformed by letter that all aspects of the position with the exceptionof IEEE 279 will be applicable to their approved standard designs andthat such designs should be modified, as necessary, to conform to theposition. Staff approval of proposed modifications can be applied foreither by application by the PDA-holder on the PDA-docket or by eachCP applicant referencing the standard design on its docket.The following guidelines may be used, if necessary, to alleviate impactson licensing schedules for plants involved in licensing proceedingsnearing completion on March 14, 1978:ENCL 3 (CONT)

-4v LUA1. Those applicants issued an OL during the period between March 14,1978 and a date 12 months thereafter may merely commit to meetingthe position prior to OL issuance but shall, by license condition,be required to install all required staff-approved codificationsprior to plant startup following the first scheduled refuelingoutage.2. Those applicants issued an OL beyond March 14, 1979 shall installall required staff-approved modifications prior to initial plantstartup.3. Those applicants issued a CP, PDA, or ML during the period betweenMarch 14, 1978 and a date 6 months thereafter may merely committo meeting the position but shall, by license condition, berequired to amend the application, within 6 months of the date ofissuance of the CP, PDA, or ML, to include a description of theproposed modifications and the bases for their design, and arequest for staff approval.4. Those applicants issued a CP, PDA, or ML after September 14, 1978shall have staff approval of proposed modifications prior toissuance of the CP, PDA, or ML.D. References1. NUREG-0138, Staff Discussion of Fifteen Technical Issues Listedin Attachment to November 3, 1976 Memorandum from Director, NRR,to NRR Staff.ENCL 3 (CONT)

CATEGORY 4 MATTERSA. Regulatory GuidesIssueDate Number4/74 1.1212/75 1.13not categorizedRevision118/751/754/749/756/746/747/7511/7412/742/761.141.75111.7601.791TitleInstrumentation for EarthquakesSpent Fuel Storage Facility DesignBasisReactor Coolant Pump Flywheel IntegrityPhysical Independence of ElectricSystemsDesign Basis Tornado for Nuclear PowerPlantsPreoperational Testing of EmergencyCore Cooling Systems for PressurizedWater ReactorsPreoperational Testing of InstrumentAir SystemsSumps for Emergency Core Cooling andContainment Spray SystemsInservice Inspection of PressurizedWater Reactor Steam Generator TubesQualification of Class lE Equipmentfor Nuclear Power PlantsAvailability of Electric Power SourcesOverhead Crane Handling Systems forNuclear Power Plants1.8001.8201.8311.891.931.104000ENCLOSURE 4

-2-B. SRP CriteriaImplementa-tion Date1. 1.1/24/75BranchMTEB2. 11/24/75 CSB3. 11/24/75 CSB4. 11/24/75 CSB5. 11/24/75 CSB6. 11/24/75 ASBApplicableSRP Section5.4.2.16.2.16.2.1A6.2.1B6.2.1.26.2.1.36.2.1.46.2.1.56.2.56.2.36.2.49.1.410.4.93.5.3TitleBTP MTEB-5-3,-Monitoringof Secondary Side WaterChemistry in PWR SteamGeneratorsBTP CSB-6-1, MinimumContainment Pressure Modelfor PWR ECCS PerformanceEvaluationBTP CSB-6-2, Control ofCombustible Gas Concentra-tions in Containment Followinga Loss-of-Coolant AccidentBTP CSB-6-3, Determination ofBypass Leakage Path in DualContainment PlantsBTP CSB-6-4, ContainmentPurging During Normal PlantOperationsBTP ASB-9.l, Overhead HandlingSystems for Nuclear Power PlantsBTP ASB-10.1, Design Guidelinesfor Auxiliary Feedwater SystemPump Drive and Power SupplyDiversity for PWR'sProcedures for Composite SectionLocal Damage Prediction (SRPSection 3.5.3, par. Il.l.C)7. 11/24/75ASB8. 11/24/75 SEBENCLOSURE 4 (CONT)

-3-Implementa-tion DateBranch9. 11/24/75 SEB10. 11/24/75 SEB11. 11/24/75 SEB12. 11/24/75 SEB13. 11/24/75 SEB14. 11/24/75 SEB15. 11/24/75 SEB16. 11/24/75 SEB17. 11/24/75 SEBApplicableSRP Section3.7.13.7.23.7.33.8.13.8.23.8.33.8.43.8.53.711.211.311.4TitleDevelopment of Design TimeHistory for Soil-StructureInteraction Analysis (SRPSection 3.7.1, par. 11.2)Procedures for Seismic SystemAnalysis (SRP Section 3.7.2par. II)Procedures for Seismic Sub-system Analysis (SRP Section 3.7.3,par. II)Design and Construction ofConcrete Containments) SRPSection 3.8.1, par. II)Design and Construction ofSteel Containments (SRP Section3.8.2, par. II)Structural Design Criteria forCategory I Structures InsideContainment (SRP Section 3.8.3,par. II)Structural Design Criteria forOther Seismic Category I Structures(SRP Section 3.8.4, par. II)Structural Design Criteria forFoundations (SRP Section 3.8.5,par. II)Seismic Design Requirements forRadwaste Sysems and Their HousingStructures (SRP Section 11.2, BTPETSB 11-1,par. B.v)ENCLOSURE 4 (CONT)

-4-Implementa-tion Date18. 11/24/75ApplicableSRP SectionBranchSEBTitle19. 11/24/75 SEB20. 10/0l/75 ASB21. 11/24/75 AB22. 11/24/75 RSB23. 11/24/75 RSB3.3.23.4.210.4.74.45.2.53.2.2Tornado Load Effect Combi-nations (SRP Section 3.3.2,par. II.2.d)Dynamic Efects of Wave Action(SRP Section 3.4.2, par. II)Water Hammer for SteamGenerators with Preheaters (SRPSection 10.4.7 par. I.2.b)Thermal-Hydraulic Stability (SRPSection 4.4, par. II.5)Intersystem Leakage Detection (SRPSection 5.2.5 par. 1I.4) and R.G. 1.45Main Steam Isolation Valve LeakageControl System (SRP Section 10.3par. 11I.3 and BTP RSB-3.2)C. Other PositionsImplementa-tion Date1. 12/1/76ApplicableSRP SectionBranchTitleSEB3.5.3Ductilityand SteelSubjectedLoadsof Reinforced ConcreteStructural Elementsto Impactive or Impulsive2. 8/01/763. 4/01/764. 9/01/765. 10/01/76SEBSEBSEBSEB3.7.13.8.13.8.23.8.43.5.36.3Response Spectra in VerticalDirectionBWR Mark III Containment PoolDynamicsAir Blast LoadsTornado Missile ImpactPassive Failures During Long-Term Cooling Following LOCA6. 6/01/77 RSBENCLOSURE 4 (CONT)

-5-* Implementa-tion Date7. 9/01/77BranchRSBApplicableSRP Section6.3TitleControl Room Position Indica-tion of Manual (Handwheel) Valvesin the ECCS8. 4/01/77RSB9. 12/01/77 RSB15.1.55.4.65.4.76.33.5.110. 3/28/7811. 1/01/7712. 1/01/7813. 6/01/7614. 9/01/7715. 1/01/77RSBAB4.4PSBCSBCSBCSB8.36.2.1.26.2.66.2.1.43.6.13.6.29.2.210.4.73.11-Long-Term Recovery from SteamlineBreak: Operator Action to PreventOverpressurizationPump Operability RequirementsGravity Missiles, Vessel SealRing Missiles Inside ContainmentCore Thermal-Hydraulic AnalysisDegraded Grid Voltage ConditionsAsymmetric Loads on ComponentsLocated Within Containment Sub-compartmentsContainment Leak Testing ProgramContainment Response Due to MainSteam Line Break and Failure ofMSLIV to CloseMain Steam and Feedwater PipeFailuresDesign Requirements for CoolingWater to Reactor Coolant PumpsDesign Guidelines for Water Hammerin Steam Generators with TopFeedring Design (BTP ASB-10.2)Environmental Control Systems-forSafety-Related Equipment16. 11/01/77 ASB17. 1/01/7718. 8/01/7619. 1/01/76ASBASBICSBENCLOSURE 4 (CONT)

DESCRIPTION OF POSITIONS IDENTIFIED AS NRR CATEGORY 4MATTERS IN ENCLOSURE 4, PARAGRAPH CNumbering scheme corresponds to that used in Item C of Enclosure 4.ENCLOSURE 4 (CONT)

C.1 DUCTILITY OF REINFORCED CONCRETE AND STEEL STRUCTURAL ELEMENTSSUBJECTED TO IMPACTIVE OR IMPULSIVE LOADSINTRODUCTIONIn the evaluation of overall response of reinforced concrete structuralelements (e.g., missile barriers, columns, slabs, etc.) subjected toimpactive or impulsive loads, such as impacts due to missiles, assumptionof non-linear response (i.e., ductility ratios greater than unity) ofthe structural elements is generally acceptable provided that the safetyfunctions of the structural elements and those of safety-related systemsand components supported or protected by the elements are maintained.The following summarizes specific SEB interim positions for review andacceptance of ductility ratios for reinforced concrete and steelstructural elements subjected to impactive and impulsive loads.SPECIFIC POSITIONS1. REINFORCED CONCRETE MEMBERS1.1 For beams, slabs, and walls where flexure controls design, thepermissible ductility ratio ( U ) under impactive and impulsiveloads should be taken as= 0.05 for p -> ' .X0050 P10 for p -' < .005where p and P'are the ratios of tensile and compressivereinforcing as defined in ACI-318-71 Code.1.2 If use of a ductility ratio greater than 10 (i.e., P> 100)is required to demonstrate design adequacy of structuralelements against impactive or impulsive loads, e.g., missileimpact, such a usage should be identified in the plant SAR.Information justifying the use of this relatively high ductilityvalue shall be provided for SEB staff review.ENCLOSURE 4 (CONT)

-2-1.3 For beam-columns, walls, and slabs carrying axial compressionloads and subject to impulsive or impactive loads producingflexure, the permissible ductility ratio in flexure shouldbe as follows:(a) When compression controls the design, as defined by aninteraction diagram, the permissible ductility ratioshall be 1.3.(b) When the compression loads do not exceed O.lfcAg or one-third of that which would produce balanced conditions, which-ever is smaller, the permissible ductility ratio can be asgiven in Section 1.1.(c) The permissible dutility ratio shall vary linearly from 1.3to that given in Section 1.1 for conditions between thosespecified in (a) and (b). (See Fig 1.)1.4 For structural elements resisting axial compressive impulsive orimpactive loads only, without flexure, the permissible axialductility ratio shall be 1.3.1.5 For shear carried by concrete only= 1.0For shear carried by concrete and stirrups or bent bars= 1.3For shear carried entirely by stirrups= 3.02.0 STRUCTURAL STEEL MEMBERS2.1 For flexure compression and shearU = 10.02.2 For columns with slenderness ratio (l/r) equal to or less than 20U = 1.3ENCLOSURE 4 (CONT)

pa(ttes)z boo,° whi clev-eri Is the I..smaller MOl M i;T- k) 9Ctl>i'Dctr FlAT1011FOR~s

-3-where I -effective length of the memberr c the least radius of gyrationFor columns with slenderness ratio greater than 20= Z 1.02.3 For members subjected to tension.5 towhere cu= uniform ultimate strain of the materialcy = strain at yield of materialC.2 RESPONSE SPECTRA IN THE VERTICAL DIRECTIONSubsequent to the issuance of Regulatory Guide 1.60, the report"Statistical Studies of Vertical and Horizontal Earthquake Spectra"was issued in January 1976 by NRC as NUREG-0003. One of theimportant conclusions of this report is that the response spectrumfor vertical motion can be taken as 2/3 the response spectrum forhorizontal motion over the entire range of frequencies in the WesternUnited States. According to Regulatory Guide 1.60, the verticalresponse spectrum is equal to the horizontal response spectrum between3.5 cps and 33 cps. For the Western United States only, consistentwith the latest available data in NUREG-0003, the option of taking thevertical design design response spectrum as 2/3 the horizontal responsespectrum over the entire range of frequencies will be accepted.For other locations, the vertical response spectrum will be the sameas that given in Regulatory Guide 1.60.C.3 BWR MARK III CONTAINMENT POOL DYNAMICS1. POOL SWELLa. Bubble pressure, bulk swell and froth swell loads, dragpressure and other pool swell loads should be treated asabnormal pressure loads, Pa. Appropriate load combinationsand load factors should be applied accordingly.b. The pool swell loads and accident pressure may be combinedin accordance with their actual time histories of occurrence.ENCLOSURE 4 (CONT)

4 !2, SAFETY RELIEF VALVE (SRV) DISCHARGEa. The SRV loads should be treated as live loads in all loadcombinations 1.5Pa where a load factor of 1.25 should beapplied to the appropriate SRV loads.b. A single active failure causing one SRV discharge mustbe considered in combination with the Design BasisAccident (DBA).c. Appropriate multiple SRV discharge should be considered incombination with the Small Break Accident (SBA) and Inter-mediate Break Accident (IBA).d. Thermal loads due to SRV discharge should be treated as TOfor normal operation and Ta for accident conditions. 0e. The suppression pool liner should be designed in accordancewith the ASME Boiler and Pressure Vessel Code, Division 1Subsection NE to resist the SRV negative pressure, consideringstrength, buckling and low cycle fatigue.C.4 AIR BLAST LOADS (Pa, Ta, To as defined in ACI 359-740)The following interim position on air blast loadings on Nuclear PowerPlant Structures should be used as guidance in. evaluating analyses.1. An equivalent static pressure may be used for structural analysispurposes. The equivalent static pressure should be obtained fromthe air blast reflected pressure or the overpressure by multiplyingthese pressures by a factor of two. Any proposed use of a dynamicload factor less than two should be treated on a case by case basis.Whether the reflected pressure or the overpressure is to be used forindividual structural elements depends on whether an incident blastwave could strike the surface of the element..2. No load factor need be specified for the air blast loads, and theload combination should be:U = D + L + Bwhere, U is the strength capacity of a sectionD is dead loadL is live loadB is air blast load.3. Elastic analysis for air blast is required for concrete structuresof new plants. For steel structural elements, and also for rein-forced concrete elements in existing plants, some inelastic responsemay be permitted with appropriate limits on ductility ratios.ENCLOSURE 4 (CONT)

-5-4. Air blast generated ground shock and air blast wind pressure maybe ignored. Air blast generated missiles may be important insituations where explosions are postulated to occur in vesselswhich may fragment.5. Overturning and sliding stability should be assessed by multiplyingthe structure's full projected area by the equivalent staticpressure and assuming only the blast side of the structure isloaded. Justification for reducing the average equivalent staticpressure on curved surfaces should be considered on a case by casebasis.6. Internal supporting structures should also be analyzed for theeffects of air blast to determine their ability to carry loadsapplied directly to exterior panels and slabs. Moreover.invented structures, interior structures may require analysis even ifthey do not support exterior structures.7. The equivalent static pressure should be considered as potentiallyacting both inward and outward.C.5 TORNADO MISSILE PROTECTIONAs an interim measure,the minimum concrete wall and rooffor tornado missile protection will be as follows:thicknessConcrete Strength (psi)Wall Thickness(inches)Roof Thickness(inches)3000 27 24Region I 4000 24 215000 21 183000 24 21Region II 4000 21 185000 19 163000 21 18Region III 4000 18 165000 16 14These thicknesses are for protection against local effects only. Designersmust establish independently the thickness requirements for overall structuraresponse. Reinforcing steel should satisfy the provisions of Appendix C, ACI349 (that is, .2% minimum, EWEF). The regions are described in RegulatoryGuide 1.76.ENCLOSURE 4 (CONT)

-6 -C.6 PASSIVE ECCS FAILURES DURING LONG-TERM COOLING FOLLOWING A LOCAPassive failures in the ECCS, having leak rates equal to or less thanthose from the sudden failure of a pump seal and which may occur duringthe long-term cooling period following a postulated LOCAshould be con-sidered. To mitigate the effects of such leaks, a leak detection systemhaving design features and bases as described below should be Includedin the plant design.The leak detection system should Include detectors and alarms which wouldalert the operator of passive ECCS leaks in sufficient time so that appro-priate diagnostic and corrective actions may be taken on a timely basis.The diagnostic and corrective actions would include the identification andisolation of the faulted ECCS line before the performance of more than onesubsystem is degraded. The design bases of the leak detection system shouldinclude:(1) Identification and justification of the maximum leak rate;(2) Maximum allowable time for operator action and justification therefor;(3) Demostration that the leak detection system is sensitive enough toinitiate and alarm on a timely basis, i.e., with sufficient lead timeto allow the operator to identify and isolate the faulted line beforethe leak can create undesireable consequences such as flooding of re-dundant equipment. The minimum time to be considered is 30 minutes;(4) Demonstration that the leak detection system can identify the faultedECCS train and that the leak can be isolated; and(5) Alarms that conform with the criteria specified for the control roomalarms and a leak detection system that conforms with the require-ments of IEEE-279, except that the single failure criterion need notbe imposed.C.7 CONTROL ROOM POSITION INDICATION OF MANUAL (HANDWHEEL) VALVESRegulatory Guide 1.47 specifies automatic position indication of eachbypass or deliberately induced inoperable condition if the followingthree conditions are met:(1) The bypass or inoperable condition affects a system that isdesigned to perform an automatic safety function.ENCLOSURE 4 (CONT)

-7 -(2) The bypass or inoperable condition can reasonably be expectedto occur more frequently than once per year.(3) The bypass or inoperable condition is expected to occur when thesystem is normally required to operate.Revision one of the Standard Review Plan in Section 6.3 requiresconformance with Regulatory Guide 1.47 with the intent being thatany manual (handwheel) valve Which could Jeopardize theoperation of the ECCS, if inadvertently left in the wrong position,must have position indication in the control room. In the PDA extensionreviews it is important to confirm that standard designs include thisdesign feature. Most standard designs do but this matter was probablynot specifically addressed in some of the first PDA reviews.C.8 LONG-TERM RECOVERY FROM STEAM LINE BREAK -OPERATOR ACTION TOPREVET OVERP SURAION PWR)A steam line break causes cooldown of the primary system, shrinkage ofRCS inventory and depletion of pressurizer fluid. Subsequent to planttrip, ECCS actuation, and main steam system isolation, the RCS inven-tory increases and expands, refilling the pressurizer. Without operatoraction, replenishment of RCS inventory by the ECCS and expansion at lowtemperature could repressurize the reactor to an unacceptable pressure-temperature region thereby compromising reactor vessel integrity. Anal-yses are required to show that following a main steam line break that(i) no additional fuel failures result from the accident, and (ii) thepressures following the initiation of the break will not compromise theintegrity of the reactor coolant pressure boundary giving due considera-tion to the changes in coolant and material temperatures. The analysesshould be based on the assumption that operator action will not be takenuntil ten minutes after initiation of the ECCS.C.9 PUMP OPERABILITY REQUIREMENTSIn some reviews, the staff has found reasonable doubt that some types ofengineered safety feature pumps would continue to perform their safetyfunction in the long term following an accident. In such instances therehas been followup, including pump redesign in some cases, to assurethat long term performance could be met. The following kinds of infor-mation may be sought on a case-by-case basis where such doubt arises.a. Describe the tests performed to demonstrate that the pumps arecapable of operating for extended periods under post-LOCA conditions,including the effects of debris. Discuss the damage to pump sealscaused by debris over an extended period of operation.ENCLOSURE 4 (CONT)

-8-b. Provide detailed diagrams of all water cooled seals and compo-nents in the pumps.c. Provide a description of the composition of the pump shaftseals and the shafts. Provide an evaluation of loss.of shaftseals.d. Discuss how debris and post-LOCA environmental conditions werefactored into the specificatfons and design of the pump.C.lO GRAVITY MISSILES, VESSEL SEAL RING MISSILES INSIDE CONTAINMENTSafety related systems should be protected against loss of function due tointernal missiles from sources such as those associated with pressurizedcomponents and rotating equipment. Such sources would include but not belimited to retaining bolts, control rod drive assemblies, the vessel sealring, valve bonnets, and valve stems. A description of the methods usedto afford protection against such potential missiles, including the basestherefor, should be provided (e.g., preferential orientation of the poten-tial missile sources, missile barriers, physical separation of redundantsafety systems and components). An analysis of the effects of such poten-tial missiles on safety related systems, including metastably supportedequipment which could fall upon impingement, should also be provided.ENCLOSURE 4 (CONT)

-9-C.Tl CORE THERMAL-HYDRAULIC ANALYSESIn evaluating the thermal-hydraulic performance of the reactorcore~the following additional areas should be addressed:1. The effect of radial pressure gradients at the exit-of openlattice cores.2. The effect of radial pressure gradients in the upper plenum.3. The effect of fuel rod bowing.In additiona commitment to perform tests to verify the transientanalysis methods and codes is required.C.12 DEGRADED GRID VOLTAGE CONDITIONSAs a result of the Millstone Unit Number 2 low grid voltage occurrence,the staff has developed additional requirements concerning (a) sustaineddegraded voltage conditions at the offsite power source, and (b) inter-action of the offsite and onsite emergency power systems. These additionalrequirements are defined in the following staff position.1. We require that a second level of voltage protection for the onsitepower system be provided and that this second level of voltage pro-tection satisfy the following requirements:a) The selection of voltage and time set points shall bedetermined from an analysis of the voltage requirements ofthe safety-related loads at all onsite system distributionlevels;b) The voltage protection shall include coincidence logicto preclude spurious trips of the offsite power source;ENCLOSURE 4 (CONT)

-10-c) The time delay selected shall be based on the followingconditions:(i) The allowable time delay, including margin, shallnot exceed the maximum time delay that is assumed inthe SAR accident analyses;(ii) The time delay shall minimize the effect of shortduration disturbances from reducing the availabilityof the offsite power source(s); and(iii) The allowable time duration of a degraded voltagecondition at all distribution system levels shall notresult in failure of safety systems or components;(iv) The voltage sensors shall automatically initiate thedisconnection of offsite power sources whenever thevoltage set point and time delay limits have been exceeded;(v) The voltage sensors shall be designed to satisfy theapplicable requirements of IEEE Std. 279-1971 "Criteriafor Protection Systems for Nuclear Power GeneratingStations"; and(vi) The Technical Specifications shall include limitingconditionsfor operation, surveillance requirements,trip set points with minimum and maximum limits, andallowable values for the second-level voltage protectionsensors and associated time delay devices.2. We require that the system design automatically prevent loadshedding of the emergency buses once the onsite sources aresupplying power to all sequenced loads on the emergency buses.The design shall also include the capability of the load sheddingfeature to be automatically reinstated if the onsite source supplybreakers are tripped. The automatic bypass and reinstatementfeature shall be verified during the periodic testing identifiedin Item 3 of this position.3. We require that the Technical Specifications include a test require-ment to demonstrate the full functional operability and independenceof the onsite power sources at least once per 18 months during shut-down. The Technical Specifications shall Include a requirement fortests: (a) simulating loss of offsite power; (b) simulating lossof offsite power in conjunction with a safety injection actuationsignal; and (c) simulating interruption and subsequent reconnectionof onsite power sources to their respective buses.ENCLOSURE 4 (CONT)

-11-4. The voltage levels at the safety-related buses should beoptimized for the full load and minimum load conditions thatare expected throughout the anticipated range of voltagevariations of the offsite power source by appropriate adjust-ment of the voltage tap settings of the intervening transformers.We require that the adequacy of the design in this regard beverified by actual measurement, and by correlation- of measuredvalues with analysis results.C.13 ASYMMETRIC LOADS ON COMPONENTSLOCATED WITHIN CONTAINMENT SUBCOMPARTMENTSIn the unlikely event of a pipe rupture inside a major component sub-compartment., the initial blowdown transient would lead to pressureloadings on both the structure and the enclosed component(s). Thestaff's generic Category A Task Action Plan A-2 is designed to developgeneric resolutions for this matter. Our present schedule calls forcompleting A-2 for PWR's during the first quarter, 1979. Pendingcompletion of A-2, the staff is implementing the following program:1. For PWRs at the CP/PDA stage of review, the staff requires appli-cants to commit to address the safety issue as part of their appli-cation for an operating license.2. For PWRs at the OL/FDA stage of review, the staff requires case-by-caseanalyses, including implementation of any indicated correctivemeasusres prior to the issuance of an operating license.3. For BWRs, for which this issue is expected to be of lesser safetysignificance, the asymmetric loading conditions will be evaluatedon a case-specific basis prior to the Issuance of an operating license.For those cases which analyses are required, we request the performanceof a subcompartment, multi-node pressure response analysis ofthe pressure transient resulting from postulated hot-leg and cold-leg(pump suction and discharge) reactor coolant system pipe ruptureswithin the reactor cavity, pipe penetrations, and steam generatorcompartments. Provide similar analyses for the pressurizer surgeand spray lines, and other high energy lines located in containmentcompartments that may be subject to pressurization. Show how theresults of these analyses are used in the design of structures andcomponent supports.ENCLOSURE 4 (CONT)

-12-C.14 CONTAINMENT LEAK TESTING PROGRAMTo avoid difficulties experienced in this area in recent OL reviews,the staff has increased its scope of inquiry at the CP/PDA stage ofreview. For this purpose, the following information with regard tothe containment leak testing program should be supplied.a. Those systems that will remain fluid filled for the Type A testshould be identified and justification given.b. Show the design provisions that will permit the personnel air-lock door seals and the entire air lock to be tested.c. For each penetration,i.e., fluid system piping, instrument,electrical, and equipment and personnel access penerations,identify the Type B and/or Type C local leak testing thatwill be done.d. Verify that containment penetrations fitted with expansionbellows will be tested at Pa. Identify any penetration fitted withexpansion bellows that does not have the design capabilityfor Type B testing and provide justification.C.15 CONTAINMENT RESPONSE DUE TO MAIN STEAM LINEBREAK AND MSLIV FAILUREIn recent CP and OL application reviews, the results ofanalyses for a postulated main steam line break accident (MSLB)for designs utilizing pressurized water reactors with conventionalcontainments show that the peak calculated containment temperaturecan exceed for a short time period the environmental qualificationtemperature-time envelope for safety related instruments andcomponents. This matter was also discussed in Issue No. 1 ofNUREG-0138 and Issue No. 25 of NUREG-0153. Thesignifiance of the matter is that it could result in a requirementfor requalifying safety-related equipment to higher time-temperatureenvelopes.The staff's generic Category A Task Action Plans A-21 and A-24 aredesigned to develop generic resolutions for these matters. Thepresently scheduled completion dates for A-21 and A-24 (Short TermPortion) are first quarter, 1979 and fourth quarter, 1978, respectively.Pending completion of A-21 and A-24, some interim guidance will beused as detailed below.We have developed and are implementing a plan in which all applicants forconstruction permits and operating licenses and those already issued con-struction permits must provide information to establish a conservativetemperature-time envelope.ENCLOSURE 4 (CONT)

-13-Therefore, describe and justify the analytical model used to conservativelydetermine the maximum containment temperature and pressure for a spectrum ofpostulated main steam line breaks for various reactor power levels. Includethe following in the discussion.(1) Provide single active failure analyses which specificallyidentify those safety grade systems and components relied uponto limit the mass and energy release and containment pressure/temperature response. The single failure analyses shouldinclude, but not necessarily be limited to: main steam andconnected systems isolation; feedwater auxiliary feedwater, andconnected systems isolation; feedwater, condensate, and auxiliaryfeedwater pump trip, and auxiliary feedwater run-out controlsystem; the loss of or availability of offsite power; dieselfailure when loss of offsite power is evaluated; and partial lossof containment cooling systems.(2) Discuss and justify the assumptions made regarding the time atwhich active containment heat removal systems become effective.(3) Discuss and justify the heat transfer correlation(s) (e.g., Tagami,Uchida) used to calculate the heat transfer from the containmentatmosphere to the passive heat sinks, and provide a plot of theheat transfer coefficient versus time for the most severe steam linebreak accident analyzed.(4) Specify and justify the temperature used in the calculationof condensing heat transfer to the passive heat sinks; i.e.,specify whether the saturation temperature corresponding to thepartial pressure of vapor, or the atmosphere temperature(whichmay be superheated)was used.(5) Discuss and justify the analytical model including the thermodynamicequations used to account for the removal of the condensed massfrom the containment atmosphere due to condensing heat transferto the passive heat sinks;(6) Provide a table of the peak values of containment atmosphere temperatureand pressure for the spectrum of break areas and power levels analyzed;(7) For the case which results in the maximum containment atmospheretemperature, graphically show the containment atmosphere temperature,the containment liner temperature, and the containment concretetemperature as a function of time. Compare the calculated contain-ment atmosphere temperature response to the temperature profileused in the environmental qualification program for those safetyrelated instruments and mechanical components needed to mitigatethe consequences of the assumed main steam line break and effectsafe reactor shutdown;ENCLOSURE 4 (CONT)

-14-(8) For the case which results in maximum containment atmospherepressure, graphically show the containment pressure as afunction of time; and(9) For the case which results in the maximum containment atmospherepressure and temperature, provide the mass and energy releasedata in tabular form.In order to demonstrate that safety-related equipment has been adequatelyqualified as described above, provide the following information regard-ing its environmental qualification.(1) Provide a comprehensive list of equipment required to be operationalin the event of a main steamline break (MSLB) accident. The listshould include, but not necessarily be limited to, the followingsafety related equipment:(a) Electrical containment penetrations;(b) Pressure transmitters;(c) Containment isolation valves;(d) Electrical power cables;(e) Electrical instrumentation cable; and(f) Level transmitters.Describe the qualification testing that was, or will be, done on this equipment.Include a discussion of the test environment, namely, thetemperature, pressure, moisture content, and chemical spray,as a function of time.(2) It is our position that the thermal analysis of safety relatedequipment which may be exposed to the containment atmospherefollowing a main steam line break accident should be based on thefollowing:(a) A condensing heat transfer coefficient based on therecommendations in Branch Technical Position CSB 6-1,"Minimum Containment Pressure Model for PWR ECCS PerformanceEvaluation,"should be used.(b) A convective heat transfer coefficient should be used whenthe condensing heat flux is calculated to be less than theconvective heat flux. During the blowdown period it isappropriate to use a conservatively evaluated forcedconvection heat transfer correlation. For example,ENCLOSURE 4 (CONT)

-15-Nu = C(Re)Where Nu -Nusselt No.Re = Reynolds No.C a empirical constants dependent ongeometry and Reynolds No.Since the Reynolds number is dependent on velocity, it isnecessary to evaluate the forced flow currents which will begenerated by the steam generaor blowdown. The CVTR experimentsprovide limited data in this regard. Convective currents offrom 10 ft/sec to 30 ft/sec were measured locally. We recommendthat the CVTR test results be extrapolated conservatively toobtain forced flow currents to determine the convective heattransfer coefficient during the blowdown period. After theblowdown has ceased or been reduced to a negligibly low value,a natural convection heat transfer correlation is acceptable.(3) For each component where thermal analysis is done in conjunctionwith an environmental test at a temperature lower than the peakcalculated temperature following a main steam line break accident,compare the test thermal response of the component with the accidentthermal analysis of the component. Provide the basis by which thecomponent thermal response was developed from the environmentalqualification test program. For instance, graphically show thethermocouple data and discuss the thermocouple locations, methodof attachment, and performance characteristics, or provide adetailed discussion of the analytical model used to evaluate thecomponent thermal response during the test. This evaluation shouldbe performed for the potential points of failure such as thincross-sections and temperature sensitive parts where thermal stressing,temperature-related degradation, steam or chemical interaction atelevated temperatures, or other thermal effects could result in thefailure of the component mechanically or electrically. If thecomponent thermal response comparison results in the prediction ofa more severe thermal transient for the accident conditions thanfor the qualification test, provide justification that the affectedcomponent will perform its intended function during a MSLB accident,or provide protection for the component whch would appropriatelylimit the thermal effects.ENCLOSURE 4 (CONT)

-16-C.16 ENVIRONMENTAL EFFECT OF PIPE FAILURESIdentify the "break exclusion" regions of the main steamand feedwater lines. Compartments that contain breakexclusion regions of main steam and feedwater lines and any safetyrelated equipment in these compartments should be designed to with-stand the environmental effects (pressure, temperature, humidity andflooding) of a crack with a break area equal to the cross sectionalarea of the'break excluded'pipe.C.17 DESIGN REQUIREMENTS FOR COOLING WATERTO REACTOR COOLANT PUMPSDemonstrate that the reactor coolant system (RCS) pump seal injectionflow will be automatically maintained for all transients and accidentsor that enough time and Information are availAhla t" perm'itcorrective action by an operator.We have established the following criteria for th-at portion of thecomponent cooling water (CCW) system which interfaces with the reactorcoolant pumps to supply cooling water to pump seals and bearingsduring normal operation, anticipated transients, and accidents.1. A single active failure in the component cooling water systemshall not result in fuel damage or a breach of the reactorcoolant pressure boundary (RCPB) caused by an extended lossof cooling to one or more pumps. Single active failuresinclude operator error, spurious actuation of motor-operatedvalves, and loss of CCW pumps.2. A pipe crack or other accident (unanticipated occurrence) shallnot result in either a breach of the RCPB or excessive fueldamage when an extended loss of cooling to two or more RCpumps occurs. A single active falure shall be considered whenevaluating the consequences of this accident. Moderate leakagecracks should be determined in accordance with Branch TechnicalPosition ASB 3-1.In order to meet the criteria established above, an NSSS inter-face requirement should be imposed on the balance-of-plant CCWsystem that provides cooling water to the RC pump seals and motorand pump bearings, so that the system will meet the following con-ditions:ENCLOSURE 4 (CONT)

-17-1. That portion of the component cooling water (CCW) system whichsupplies cooling water to the reactor coolant pumps and motorsmay be designed to non-seismic Category I requirements and QualityGroup 0 if it can be demonstrated that the reactor coolant pumpswill operate without component cooling water for at least 30minutes without loss of function or the need for operator pro-tective action. In addition, safety grade instrumentationincluding alarms should be provided to detect the loss ofcomponent cooling water to the reactor coolant pumps andmotors, and to notify the operator in the control room. Theentire instrumentation system, including audible and visual alarms,should meet the requirements of IEEE Std 279-1971.If it is not demonstrated that the reactor coolant pumps and motorswill operate at least 30 minutes without loss of function or operatorprotective action, then the design of the CCW system must meet thefollowing requirements:1. Safety grade instrumentation consistent with the criteria forthe reactor protection system shall be provided to initiateautomatic protection of the plant. For this case, thecomponent cooling water supply to the seals and pump andmotor bearings may be designed to non-seismic category I require-ments and Quality Group 0; or2. The component cooling water supply to the pumps and motorsshall be capable of withstanding a single active failure ora moderate energy line crack as defined in our BranchTechnical Position APCSB 3-1 and be designed to seismicCategory I, Quality Group 0 and ASME Section 1II, Class 3requirements.The reactor coolant (RC) pumps and motors are within the NSSS scopeof design. Therefore, in order to demonstrate that an RC pumpdesign can operate with loss of component cooling water for at least30 minutes without loss of function or the need for operator action,the following must be provided:1. A detailed description of the events following the loss ofcomponent cooling water to the RC pumps and an analysis demon-strating that no consequences important to safety may resultfrom this event. Include a discussion of the effect that theloss of cooling water to the seal coolers has on the RC pumpseals. Show that the loss of cooling water does not resultin a LOCA due to seal failure.ENCLOSURE 4 (CONT)

-18-2. A detailed analysis to show that loss of cooling water tothe RC pumps and motors will not cause a loss of the flowcoastdown characteristics or cause seizure of the pumps,assuming no administrative action is taken. The responseshould include a detailed description of the calculationprocedure including:a. The equations used.b. The parameters used in the equations, such as the designparameters for the motor bearings, motor, pump and anyother equipment entering into the calculations, andmaterial property values for the oil and metal parts.c. A discussion of the effects of possible variations inpart dimensions and material properties, such as bearingclearance tolerances and misalignment.d. A description of the cooling and lubricating systems (withappropriate figures) associated with the RC pump and motorand their design criteria and standards.e. Information to verify the applicability of the equationsand material properties chosen for the analysis (i.e.,references should be listed, and if empirical relationsare used, provide a comparison of their range of appli-cation to the range used in the analysis).Should an analysis be provided to demonstrate that loss ofcomponent cooling water to the RC pumps and motor assembly isacceptable, we will require certain modifications to the plantTechnical Specifications and an RC pump test conducted underoperating condtions and with component cooling water terminatedfor a specified period of time to verify the analysis.C.18 WATER HAMMER IN STEAM GENERATORS WITH TOP FEEDRING DESIGNEvents such as damage to the feedwater system piping at IndianPoint Unit No. 2, November 13, 1973, and at other plants, couldoriginate as a consequence of uncovering of the feedwater spargerin the steam generator or uncovering of the steam generatorfeedwater inlet nozzles. Subsequent events may in turn lead to-thegeneration of a pressure wave that is propagated through thepipes and could result in unacceptable damage.ENCLOSURE 4 (CONT)

-19-For CP/PDA and OL/FDA applications, provide the following for steamgenerators utilizing top feed:1. erevent or delay water draining from the feedring following adrop in steam generator water level by means such as J-Tubes;2. Minimize the volume of feedwater piping external to the steamgenerator whch could pocket steam using the shortest possible(less than seven feet) horizontal run of inlet piping to thesteam generator feedring; and3. Perform tests acceptable to the staff to verify that unacceptable feed-water hammer will not occur using the plant operating proceduresfor normal and emergency restoration of steam generator waterlevel following loss of normal feedwater and possible draining ofthe feedring. Provide the procedures for these tests for staff approvalbefore conducting the tests.Furthermore, we request that the following be provided:a. Describe normal operating occurrences of transients thatcould cause the water level in the steam generator todrop below the sparger or nozzles to cause uncovering andallow steam to enter the sparger and feedwater piping.b. Describe your criteria or show by isometric diagrams, therouting of the feedwater piping from the steam generatorsoutwards to beyond the containment structure up to the outerisolation valve and restraint.c. Describe any analysis on the piping system including anyforcing functions that will be performed or the resultsof test programs to verify that,either uncovering offeedwater lines could not occur or that, if it did occur,unacceptable damage such as the experience at the IndianPoint Unit No. 2 facility would not result with your design.ENCLOSURE 4 (CONT)

I-20-C.19 ENVIRONMENTAL CONTROL SYSTEMS FOR SAFETY RELATED EQtJIPMENTMost plant areas that contain safety related equipment depend on thecontinuous operation of environmental control systems to maintain theenvironment in those areas within the range of environmental qualificationof the safety related equipment installed in those areas. It appearsthat there are no requirements for maintaining these environmentalcontrol systems in operation while the plant is shutdown or in hot standbyconditions. During periods when these environmental control systems areshutdown, the safety related equipment could be exposed to environmentalconditions for which it has not been qualified. Therefore, the safetyrelated equipment should be qualified to the extreme environmentalconditions that could occur when the control equipment is shutdown orthese environmental control systems should operate continuously tomaintain the environmental conditions within the qualification limitsof the safety related equipment. In the second case an environmentalmonitoring system that will alarm when the environmental conditionsexceed those for which safety related equipment is qualified shallbe provided. This environmental monitoring system shall (1) be ofhigh quality, (2) be periodically tested and calibrated to verify itscontinued functioning, (3) be energized from continuous power sourcestand (4) provide a continuous record of the environmental parameters duringthe time the environmental conditions exceed the normal limits.ENCLOSURE 4 (CONT)