NUREG-1061Volume 2Report of theU.S. Nuclear Regulatory CommissionPiping Review CommitteeEvaluation of Seismic Designs - A Review of Seismic DesignRequirements for Nuclear Power Plant PipingU.S. Nuclear RegulatoryCommissionPrepared by the Seismic Design Tasl( Group o XZ%iQ

DISCLAIMERThis report was prepared as an account of work sponsored by anagency of the United States Government. Neither the United StatesGovernment nor any agency Thereof, nor any of their employees,makes any warranty, express or implied, or assumes any legalliability or responsibility for the accuracy, completeness, orusefulness of any information, apparatus, product, or processdisclosed, or represents that its use would not infringe privatelyowned rights. Reference herein to any specific commercial product,process, or service by trade name, trademark, manufacturer, orotherwise does not necessarily constitute or imply its endorsement,recommendation, or favoring by the United States Government or anyagency thereof. The views and opinions of authors expressed hereindo not necessarily state or reflect those of the United StatesGovernment or any agency thereof.

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NUREG 1061-Vol.2TI85 901400Report of theU.S. Nuclear Regulatory CommissionPiping Review CommitteeEvaluation of Seismic Designs - A Review of Seismic DesignRequirements for Nuclear Power Plant PipingManuscript Completed: February 1985Date Published: April 1985Prepared byThe Seismic Design Task GroupU.S. Nuclear Regulatory CommissionWashington, D.C. 20555 t m "EOt,**** 1JJJ-Lii UlllilEU


TABLE OF CONTENTSPageLIST OF FIGURESvLIST OF TABLESviFOREWORDvi iSEISMIC DESIGN TASK GROUP MEMBERS AND CONSULTANTSvi i iACKNOWLEDGMENTSixEXECUTIVE SUMMARYxiLIST OF ACRONYMS AND INITIALISMSxix1.INTRODUCTION1-12.ISSUES REGARDING OVERLAPPING CONSERVATISM IN SEISMIC DESIGN2-12.1 Problem Description2.2 Dampi ng Val ues2.3 Spectra Modification2.4 Nozzle Load Design Limit2.5 Inelastic Piping Response and Modifications to ElasticAnalysis Criteria2.6 Strain Rate Effects2.7 Single-Envelope Spectrum vs. Multiple-Independent Spectra2.8 Suggestions for Rationalizing Overall Design Margins2-12-22-52-82-122-262-282-31ROLE OF THE OPERATING BASIS EARTHQUAKE VS. SAFE SHUTDOWN EARTHQUAKEGROUND MOTION3-13.1 Consultant Views and Information3.2 Consultant Suggestions for Regulatory Changes3.3 Task Group Recommendations3-13-33-5DESIGN PRACTICE FOR MORE RELIABLE PIPING SYSTEMS4-14.1 Basic Problems in Current Industry Practice4.2 Use of Snubbers for Piping Systems in Nuclear Power Plants .4.3 Piping System Design Responsibilities4.4 Considerations for Support Design4-14-14-94-10INTERFACING ISSUES WITH OTHER TASKS5-15.1 Dynamic Loads5.2 Degraded Piping5-15-13.4.5.REFERENCESR-1iii

TABLE OF CONTENTS (Continued)PageAPPENDICESAppendixABCDEFGSargent and Lundy Engineers Nuclear Piping DataBechtel Power Corporation Nuclear Piping DataDuke Power Company Nuclear Piping DataGeneral Electric Company Nuclear Piping DataComposite-System and Cross-Cross Floor Spectra MethodsSnubber Deficiency Data Report, Summary 1973-1983Report on Foreign PracticeivA-1B-1C-1D-1E-1F-1G-1

LIST OF FIGURESFiguresPage2-12-22-272-3Effect of Strain Rate on Yield Stress, SA-106 MaterialEffect of Strain Rate on Flow Strength, Type 304 AusteniticStainless SteelLimit Stress for Combined Membrane and Bending, RectangularSection, from Reference 2-402-272-37Figures in Appendix EE-1E-2E-3E-4E-5E-6Oscillator-Structure Models Used in Generating Cross-CrossFl oor SpectraExample Piping-Structure SystemExample Equipment-Structure SystemMean Pseudovelocity Spectra for 20 Artificially GeneratedGround AccelerogramsFloor Spectra at Tenth Floor of Example BuildingCross-Cross Floor Spectra for Example Piping SystemE-7E-8E-8E-9E-9E-10Figures in Appendix FF-1F-2Reported Snubber Failure Incidents 1973-1983Estimate of Snubber Population in Nuclear Power Plants andNormalized Fai1ure RatesF-25F-27Figures in Appendix GG-1G-2G-3G-4G-5Nuclear Piping Design Process, Ontario Hydro Determination ofStatic LoadsNuclear Piping Design Process, Ontario Hydro Determination ofDynami c LoadsNuclear Piping Design Process, Ontario Hydro Final Stress Analysisand ReportsSchematic Representation of Ansaldo Impianti Design Process forReportsTypical German Schematic Representation of Design Process forComponentsVG-36G-37G-38G-39G-40

LIST OF TABLESTables2-12-22-32-42-52-62-72-82-9PageLimits on Stresses in Standard Weight Pipe Attached toRotating Equipment NozzlesStandard Review Plan Load CombinationsCode Equation (9) Stress LimitsInclusion/Noninclusion of Seismic Moments in PipingPressure Boundary EvaluationProposed Changes to Level D Allowable StressesMaximum Strain Rates for Selected Values of Maximum Strainand FrequencyConsiderations Involved in Establishing Design Margins forPiping SystemsCode Factors Used in Establishing Allowable Stresses inTension for Pressure Boundary EvaluationDesign Margins for Tensile Loadings Other Than Bolting2-102-142-152-152-242-262-332-352-35Tables in Appendix AA-1A-2A-3Typical BWR Piping*Summary of Fundamental Frequency Ranges for AllCategory I Subsystems in Typical BWRSample of Modal Frequencies in Typical PWRA-1A-3A-4Tables in Appendix BB-1Frequencies from 28 Subsystems of PWRsB-1B-2Frequencies from 25 Subsystems of BWRsB-2Table In Appendix EE-1Accelerations of Piping System in Units of GravityAccelerationE-7Tables in Appendix FF-1Estimated Snubber Population for Nuclear Power PlantsF-26Tables in Appendix GG-1G-2G-3G-4G-5G-6G-7Canadian Load Group Formation SummaryCanadian Class 2 Stress EquationsItalian Class 1 Piping Other Than Main Steam and ConnectedPiping: Load Combinations and Acceptance CriteriaCanadian Nozzle Load Structure Level Used by ContainerCriteriaGerman Classification of Loading CasesGerman Classification of StressesCanadian Relationship Between Support-Type and RestraintAssumptionsVIG-29G-30G-31G-33G-34G-34G-35

FOREWORDThe Executive Director for Operations of the U. S. Nuclear Regulatory Commission(NRC) requested that a comprehensive review be made of NRC requirements in thearea of nuclear power plant piping. In response to this request, an NRC PipingReview Committee was formed. The activities of this Conwiittee were dividedinto four tasks handled by appropriate Task Groups. These were:Pipe Crack Task GroupSeismic Design Task GroupPipe Break Task GroupDynamic Loads and Load Combination Task GroupEach Task Group has prepared a report appropriate to its scope. In addition,the Piping Review Committee will prepare an overview document rationalizingareas of overlap between the Task Groups. This will be released as a separatereport.The project titles of the five volumes are:Volume 1 - Investigation and Evaluation of Stress-Corrosion Cracking in Pipingof Boiling Water Reactor PlantsVolume 2 - Evaluation of Seismic Designs - A Review of Seismic Design Requirements for Nuclear Power Plant PipingVolume 3 - Evaluation of Potential for Pipe BreaksVolume 4 - Evaluation of Other Dynamic Loads and Load CombinationsVolume 5 - Summary - Piping Review Committee Conclusions and Recommendationsvii

SEISMIC DESIGN TASK GROUP MEMBERSShou-Nien Hou, ChairmanGoutam BagchiDaniel J. GuzyKamal A. ManolyJohn A. O'BrienThe conclusions and recommendations presented in the Executive Summary andother identified sections of this report are those of the Seismic Design TaskGroup itself and do not necessarily reflect the technical positions of theindividual consultants.Seismic Design Task Group ConsultantsC.K. Chou of Lawrence Livermore National Laboratory contributed information and suggestions on damping values, spectra modification, and overalldesign margins. He also provided overall logistic and technical assistanceto the Task Group.R.P. Kennedy of Structural Mechanics Associates contributed informationand suggestions on inelastic piping response and on issues concerning theoperating basis earthquake vs. the safe shutdown earthquake.M. Reich of Brookhaven National Laboratory contributed information andsuggestions on piping support design as well as on single-envelope spectrum vs. multiple-response spectra methods.E.C. Rodabaugh of E.C. Rodabaugh Associates contributed information andsuggestions on nozzle load design limits, strain rate effects, and overalldesign margins.J.D. Stevenson of Stevenson & Associates contributed information and suggestions on the use of snubbers, foreign seismic design criteria, and piping performance in actual earthquake events.These consultants are the primary authors of the technical information givenin Chapters 2, 3, and 4 and in the appendices to this report. This informationformed much of the basis for the Task Group's recommendations. However, itshould be noted that some of the consultants' suggestions were not adopted bythe Task Group. No immediate Task Group action will be taken on those consultants' suggestions that were not adopted, but they are presented in this reportfor general information only.viii

EXECUTIVE SUMMARYThe Code of Federal Regulations Title 10 Part 50 requires that structures,systems, and components important to the safety of nuclear power plants bedesigned to withstand individual and combined effects caused by normal operations, by extreme natural phenomena, and by postulated accident conditions.The U.S. Nuclear Regulatory Commission staff, through various standards such asregulatory guides, branch technical positions, and the standard review plan,has specified how these effects are to be considered in the design of safetyrelated structures, systems, and components. In the area of nuclear power plantpiping, several of the current regulatory criteria and industry design practiceshad to be developed without extensive supporting data. Information obtainedsince their development has indicated that current nuclear piping design couldbe improved to increase overall safety and to reduce unnecessary costs.This report presents the findings and recommendations of the Seismic DesignTask Group of the NRC Piping Review Committee and includes contributions fromconsultants. The Task Group was directed to review current seismic criteriapertaining to piping such as definition of seismic loads, construction offloor response spectra, calculation of piping seismic response, and state-ofthe-art design practice. The Task Group was further directed to evaluate andrecommend changes in current requirements for piping design, drawing on bothdomestic and foreign experience in the process.Current nuclear piping systems use significantly more snubbers and otherseismic supports than systems in older plant designs. These stiffer designsare the result of conservatisms added in both design criteria and designpractice to account for the large uncertainty inherent in predicting seismiceffects. Today, however, in light of new data and a more integrated view ofpiping system behavior, it seems that some of these conservatisms are excessive.Because stiff systems generate high stresses and nozzle loads resulting fromrestraint of thermal expansion and can be more adversely influenced by construction, maintenance, and inspection errors, many experts believe that theydiminish overall plant safety.This report identifies the major issues influencing seismic design and discussesthe effects of current regulatory criteria and design practices. Recommendationsare then presented for resolving these individual issues. In developing theserecommendations, the Task Group recognized that a fundamental resolution of alltechnical issues will require advancement in many technical areas. An optimumbalance among all factors affecting piping design may not be achievable in thenear future. Initial efforts by the Task Group therefore concentrated onidentifying key elements affecting design and understanding the effects of each.From this understanding, a practical near-term regulatory position can beestablished to modify design. This would yield an immediate improvement inreliability during normal operation even though a more integrated approach topiping design may still be several years away.Truly optimum design criteria will ultimately not only resolve technical issuesbut will also improve safety. Although any detailed value-impact analysis isoutside the charter of the Piping Review Committee, the Task Group has givenqualitative consideration to cost-benefit in developing its specific recommendations. For example, in those situations where it is clear that substantialxi

benefit could be realized at minimal implementation cost, the Task Group suggestsitems for immediate NRC action, including changes in current design criteria.In situations where potential technical benefit exists but the implementationcost of regulation change is uncertain, suggestions for further action, includingrelevant research programs, have been made.The specific technical issues studied by the Task Group are presented below,together with recommendations for (1) immediate NRC action, (2) changes inregulatory positions, and (3) research programs to improve understanding ofspecific technical issues.Damping ValuesBecause of a lack of understanding of the parameters affecting damping, lowerbound values are currently used for seismic design (Regulatory Guide 1.61).The use of higher allowable damping values would reduce calculated response andthus allow increased piping system flexibility. Experimental evidence to datesuggests a strong correlation between damping and frequency. In light of thisevidence, the Pressure Vessel Research Committee (PVRC) has recommended a technical position allowing 5% of critical damping for piping frequencies up to10 Hz, 2% for frequencies between 20 Hz and 33 Hz, with linearly varying dampingbetween 10 Hz and 20 Hz. This has been adopted by the ASME as Code Case N-411.The Task Group makes the following recommendations regarding damping values forpiping design:0Action ItemsImmediately endorse ASME Code Case N-411 for use in calculatingseismic response using spectra! analysis methods. (This limitedendorsement does not apply to damping values for time-historyanalysis.)0Change in Regulatory PositionRevise Regulatory Guide 1.61 and Standard Review Plan 3.9.2 toincorporate the new damping criteria.0Research Programs Complete the Idaho National Engineering Laboratory (INEL) dampingtests, which should establish more precisely the dependency ofdamping to both modal response frequency and load frequency.Investigate the possibility of applying the PVRC damping positionto dynamic loads other than seismic, and address proper damping forfrequencies above 33 Hz.xii

Spectra ModificationsRegulatory Guide 1.122 stipulates that the peaks of floor response spectra usedfor piping design be broadened by 15% to account for uncertain structural frequencies resulting from uncertainties in parameters such as material propertiesof the structure and soil damping values, soil-structure-interaction analysistechniques, and approximations in modeling techniques used in seismic analysis.In reality, a piping system will excite at only one peak frequency in theartificially broadened range. The required 15% peak broadening takes intoaccount the uncertainties associated with frequency estimation, but also substantially (and artificially) increases the total energy input to the pipingsystem. As a result, this requirement introduces additional conservatism intothe seismic piping design. The PVRC has recommended an alternative to peakbroadening whereby response spectra peaks would be shifted throughout the 15%range and the responses to various inputs, rather than the inputs themselves,would be enveloped. This has been adopted by the ASME as Code Case N-397 andwill also be included in Appendix N of the 1984 Summer Addendum to the ASMECode.The Task Group makes the following recommendations regarding spectra modification:0Action ItemsImmediately endorse ASME Code Case N-397.Initiate NRC internal review regarding the adequacy of the 15%range for treating uncertainties in spectral peak frequencies.0Change in Regulatory PositionRevise Regulatory Guide 1.122 to permit peak shifting as an alternative to peak broadening.0Research ProgramsAssess uncertainty range of spectral frequencies, including uncertainties in piping system modeling.Develop a simple spectra-broadening procedure based on equivalentenergy input.Nozzle LoadsPiping systems generally terminate at nozzles connected to piping, vessels, orrotating machinery. The design of piping branch connections and tank and vesselnozzles do not generally take credit for nozzle flexibility, resulting in highercalculated stresses. Also, equipment manufacturers often specify unnecessarilylow nozzle allowable loads. Improving nozzle design procedures could helpreduce the number of seismic restraints required in current piping design andcould lessen the restrictions of present nozzle load limits when new pipingcriteria (e.g., damping) are introduced.xiii

The Task Group makes the following recommendations with regard to nozzle loads:0Action ItemsRequest that the ASME Section III Working Group on Piping Designrevise ASME Code sections addressing pipe system flexibilitycalculation to also consider tank and vessel nozzle flexibility.(Completed)0Change in Regulatory PositionRevise Standard Review Plan 3.9.2 to consider nozzle flexibility inpiping analysis.0Research ProgramsDevelop improved design guidance on nozzle stress limits and flexibilities.Inelastic ResponseWell-designed piping systems are capable of absorbing and dissipating a considerable amount of energy when strained beyond their elastic limit. Also, anearthquake is capable of inputting only a limited amount of energy into suchsystems. Unless adjusted, a linear-elastic response analysis cannot accountfor the inelastic energy absorption capacity present at even the Service LevelC or D stress levels and therefore gives credit for only a fraction of the totalenergy absorption capability of the piping system. This conservatism leads tothe use of more pipe restraints than are actually necessary to ensure acceptablemargins against failure for dynamic loads that may occur. As a result, pipingsystem stiffness is increased causing a potential decrease in overall reliability.The Task Group makes the following recommendations regarding consideration ofinelastic piping system response:0Change in Regulatory PositionRevise Standard Review Plan 3.9.2 to state the goal of safe-shutdownearthquake performance criteria to be used in nonlinear piping analysis. Such performance criteria would establish a minimum marginagainst failure, where failure would be defined as (1) the onset ofplastic tensile instability, (2) low-cycle fatigue or plastic ratchetting, (3) the onset of local or system buckling, (4) excessive deformation (resulting in more than a 15% reduction in cross-sectionalflow area), or (5) functional failure of pipe-mounted equipment.0Research ProgramsDevelop pseudolinear-elastic estimation methods, and designprocedures to account for inelastic response.xiv

Uniform-Envelope vs. Multiple-Response Spectra MethodsBecause pipe supports are often attached to structural members located atdifferent elevations in one or more buildings, piping systems will experiencemultiple acceleration inputs during an earthquake. The total response of thepipe will depend on its own inertia as well as on the differential motionbetween support attachments. Time-history finite-element analyses are capableof capturing this response but are very costly. Therefore, most nuclear pipingsystems are qualified by response spectra methods using envelope spectra inputfrom Regulatory Guide 1.60 for the dynamic responses and a separate seismicanalysis for the pipe response due to support motion; the results of these twoanalyses are then combined by absolute sum.Brookhaven National Laboratory has recently completed (reported in NUREG/CR-3811)an NRC-funded evaluation of multiple-response spectra methods for evaluatingpiping systems. Multiple-response spectra methods, being closer to realitythan the uniform-envelope excitation assumption, could result in less designconservatism while still providing the required margin of safety. Furthermore,computer run times are comparable to those required for present response spectramethods.The Task Group makes the following recommendations regarding multiple Independentspectra input:0Change in Regulatory PositionsRevise Standard Review Plan 3.9.2 to permit and encourage the use ofmultiple-response spectra methods for more realistic seismic responseanalysis (for details, see Section 2 of Volume 4 of NUREG-1061).0Research ProgramsInvestigate phase correlation between floor responses, and recommendchanges to spectra combination methods, as appropriate.Overall Design MarginsA key element in the development of optimized design criteria Is that a properbalance be maintained among the margins associated with various individualeffects such as seismic and thermal loads. In balancing a design for thesevarious effects, it is important to achieve this balance at the level of actualfailure, rather than that of a code-defined standard.The most appropriate balance among effects Is difficult to define because ofa lack of real failure information for piping, particularly for seismic loads.The current ASME Code stress criteria were developed considering piping collapseas the chief failure mode for seismic inertlal loads. However, many expertsnow feel that low-cycle fatigue, or fatigue/ratchetting, is the likely failuremode. Our limited data base of piping failure (see the Addendum to Volume 2 ofNUREG-1061) seems to strengthen this belief. If this can be adequately proven,appropriate changes to the ASME Code could likely result in more flexible pipingsystem design. Development of a final position will require a research programto review behavior of piping in real strong-motion earthquakes and review of allXV

«existing or planned test programs on seismic design to determine if theirresults are suitable for incorporating margins against failure into designregulations.The Task Group makes the following suggestions for rationalizing overalldesign margins:0Action ItemsMonitor and assess PVRC piping activities and ASME Code revisions asappropriate.Assess piping experience when a seismic event occurs.0Research ProgramsTest programs (e.g., EPRI's piping capacity tests) for verifyingpiping seismic design margins and identifying failure modes shouldbe supported. Evaluate test results and provide recommendationsfor criteria changes (e.g., reclassification of seismic inertialstresses as "secondary"), as appropriate.Encourage the nuclear industry to establish and justify anearthquake level that piping systems can sustain with sufficientconfidence that no seismic analysis is needed.Operating Basis Earthquake vs. Safe Shutdown EarthquakeOriginally, the seismic design of nuclear power plants followed the samebasic concept applied to other industrial facilities located in areas of highseismicity. Plant design was governed by the effects of an earthquake determined to have a reasonable probability of occurrence during the plant designlife and then was "safety checked" against a larger earthquake generallyassumed to induce ground motion twice that of the design earthquake. However,design emphasis eventually shifted so that an independently established safeshutdown earthquake (SSE) was that emphasized in plant design. The lesserearthquake, referred to as the operating basis earthquake (OBE), is nowgenerally specified as producing ground motion one-half that of the SSE. Asa result, the original concept of the OBE as having a "reasonable" probabilityof occurrence during plant life may have been lost.Although current design practice normally sets a plant's OBE at one-half theSSE ground motion, some plants have lower OBE to SSE ratios. The Atomic Safetyand Licensing Appeal Board's June 16, 1981 decision on Diablo Canyon clarifiesthat the OBE does not have to be directly coupled to the SSE value. Thepractice of setting OBE input to one-half the SSE input has several majorimpacts on the seismic design of piping systems:1.The design requirement to safely shut down the plant up to the SSElevel seems sufficient to ensure nuclear safety. It should be leftto the discretion of the plant owner to define a "reasonable" levelfor which the plant must be designed to remain functional.XV i

2.Current design criteria define different structure and pipe dampingvalues for the OBE than for the SSE, requiring that two separateseismic analyses be performed. This increases analytical effortwithout clearly adding to safety or reliability.3.In the absence of loss-of-coolant-accident (LOCA) or pipe-break loads,the OBE at Service Level B controls piping design over the SSE atService Level D.The Task Group makes the following recommendations regarding design for the OBEand the SSE:0Action ItemsInitiate an NRC internal review to investigate the feasibility ofusing uniform structural and piping damping values for evaluatingboth the OBE and SSE and thus permit scaling of a single earthquakeanalysis.Request ASME to consider effects of seismic anchor movement at ServiceLevel D (rather than at Level B) when the OBE becomes decoupled fromthe SSE.0Change in Regulatory PositionsRecommend that rulemaking be undertaken that would change the OBEdefinition in Appendix A to 10 CFR Part 100 to permit decoupling ofthe OBE and SSE.Use of SnubbersSnubbers are devices installed in piping systems to limit relatively rapiddynamic motion while permitting relatively slow motion. Operating experiencehas indicated that neither mechanical nor hydraulic snubbers have alwaysperformed reliably in service. In order to improve the overall reliability ofotherwise passive piping systems, the use of snubbers to meet piping systemdesign requirements should be limited.The Task Group suggests a program to limit the use of snubbers on safetyrelated piping in nuclear power plants, including the following specificrecommendations:0Action ItemsInitiate a nonmandatory snubber reassessment program for operatingplants and plants under construction.0Research ProgramsEncourage the nuclear industry to investigate methods and proceduresto limit the use of snubbers.Complete the Pacific Northwest Laboratories' study of Licensee EventReports related to snubber performance to identify failure causes andXV ii

effectiveness of various snubber types, and suggest methods of improving performance such as periodic testing or qualification testing.Support DesignCurrent design practice addresses independently seismic margins for componentsupports, pipe supports, and piping. A more integrated approach to piping systemdesign that considers the interaction between support and piping response andfailure could improve overall performance. It may be more desirable to establish separate seismic criteria for piping supports than for component supports.The Task Group suggests that such an integrated approach to pipe-support systemdesign be pursued and makes the following specific rec

4. DESIGN PRACTICE FOR MORE RELIABLE PIPING SYSTEMS 4-1 4.1 Basic Problems in Current Industry Practice 4-1 4.2 Use of Snubbers for Piping Systems in Nuclear Power Plants . 4-1 4.3 Piping System Design Responsibilities 4-9 4.4 Considerations for Support Design 4-10 5. INTERFACING ISSUES WITH OTHER TASKS 5-1 5.1 Dynamic Loads 5-1