Functionality Definition

CHAPTER

35

35.1 INTRODUCTION 35.1.1 Objectives

If plant systems, structures, or components (SSCs) are found tobe degraded, nonconforming, or subject to an unanalyzed condi-tion during operation, an evaluation of their functionality or oper-ability is necessary for the nuclear power generating station tocontinue operating safely. This evaluation establishes the abilityof the SSCs to perform their assigned safety functions. This chap-ter discusses the methodology and acceptance criteria applicableto this evaluation. It introduces typical SSCs that may requireoperability evaluations and also provides definitions related tooperability. Examples of the operating conditions and events toconsider are described, as are the evaluation methods and accep-tance criteria for short- and long-term operability.

Several references were used in the preparing the material pre-sented in this chapter; the most significant of which include:

(1) USNRC RIS-2005-20 [27](2) ASME B&PV Code, Section XI [7](3) ASME B&PV Code, Section III, Appendix F [8](4) ASME O&M Code [9] (5) USNRC Inspection Manual Part 9900 [15]This chapter discusses basic concepts, definitions, evaluation

methods, and acceptance criteria from these documents as theyrelate to mechanical systems, their components and structures.

35.1.2 Operability and FunctionalityIn U.S. regulatory space [15], operability determinations are

associated with SSCs described in the plant technical specifica-tions (TSs) and form the basis for compliance with regulatoryrequirements and limiting conditions for operation. The scope ofSSCs considered within the operability determination processinclude: SSCs required to be operable by TSs and SSCs that arenot explicitly required to be operable by TSs, but that performrequired support functions. Conversely, functionality assess-ments are performed for SSCs not described in the plant TSs.

From a practical standpoint these distinctions serve as a meansdifferentiating the evaluation processes employed to assess the fit-ness for service if safety related and non-safety related SSCs. Infact the fundamental basis for either operability or functionalityrests in the measure of the SSCs capability to perform its intend-ed function(s). For the most part the technical evaluation methods

and acceptances criteria employed to make these determinationare common for technical specification and non-technical specifi-cation SCCs.

35.1.2.1 Operable /Operability. The U.S. NRC StandardTechnical Specifications [28, 29, 30, 31, 32, 33] define operable/operability as follows:

A system, subsystem, train, component, or device shall beOPERABLE or have OPERABILITY when it is capable ofperforming its specified safety functions, and when all neces-sary attendant instrumentation, controls, normal or emergencyelectrical power, cooling and seal water, lubrication and otherauxiliary equipment that are required for the system, subsys-tem, train, component, or device to perform its function(s) arealso capable of performing their related support function(s).Several variations of the preceding definition exist in the plant

specific Technical Specifications. In all cases, however, a licenseesplant-specific definition should be accepted as governing how oneapplies the terms operable and operability. The specified functionsreferenced in the foregoing definition are the specified safetyfunctions described in the current licensing basis for the facility.The following are some examples of specified safety functions forseveral SSCs.

For piping:

(1) structural integrity where structural failure would interfere withother systems being able to perform their safety functions;

(2) pressure integrity to the extent that leakage is limited to lev-els permitted by the licensing basis and through-wall flawsremain stable when subjected to faulted loads; and

(3) ability to pass required flow rates.For supports:

(1) Motion control of the system or component within the lim-its required for the system or component to perform its safetyfunction or else to maintain the structural integrity of thesystem or component.

For valves:

(1) pressure integrity to the extent that leakage is limited to lev-els permitted by the licensing basis;

FUNCTIONALITY ANDOPERABILITY CRITERIAStephen R. Gosselin and Guy H. DeBoo

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(2) structural integrity; and (3) the ability to open or close as required in the licensing basis.In addition to providing the specified safety functions, a sys-

tem, subsystem, train, component, or device (referred to as systemin this section) is expected to perform as designed, tested, andmaintained. When system capability is so degraded that it cannotperform with reasonable certainty or reliability, the system shouldbe judged inoperable even if it is shown to provide the specifiedsafety functions [15]. Required action ranges included in theacceptance criteria of ASME O&M Code [9] or Section XI [7] areexamples of degraded capabilities for SSCs. Plant TechnicalSpecifications also contain limiting values, such as leakage rateand set point pressure, for component performance. These valuesconstitute the technical specification-based operability verificationcriteria that, if they are not met, necessitate the entering of theapplicable limiting Condition for Operation (LCO) and AllowedOutage Times (AOTs).

In some cases, the ASME O&M Coderequired or SectionXIrequired action ranges for certain components may be moreconservative than the plant technical specification limits. However,the component in question must be declared inoperable even if theexisting performance meets the technical specification safety limitbecause of the imposed ASME operability limit. An example is apump that is capable of delivering rated flow but exhibits vibrationin excess of the reference values and falls in the required actionrange.

The discussion in this section accentuates the often complex,sometimes inconsistent, nature of operability concepts and criteria.Common practice involves a process having a consensus of regula-tor viewpoints, plant-specific technical specification requirements,applicable Codes and Standards, regulatory commitments, andother licensing-basis compliance requirements.

The general scope of SSCs that might require operability deter-minations include:

(1) Safety-related SSCs that are relied upon to: (a) ensurereactor-coolant pressure boundary integrity; (b) shut downand/or maintain the reactor in a safe shutdown condition; or(c) prevent or mitigate the consequences of accidents thatcould result in potential offsite exposure comparable to theguidelines in 10 CFR 50.34(a)(1), 50.67(b)(2), or 100.11, asapplicable [11].

(2) SSCs whose failure could prevent any of the SSCs identifiedin (a), (b), and (c) from performing their require functions.

(3) SSCs that are required to function by safety analyses orplant evaluations that are a part of the current licensingbasis. These analyses and evaluations include those that aresubmitted to support any license amendment requests,exemption requests, or relief requests, and also those sub-mitted to demonstrate compliance with regulations such asfire protection in 10CFR50.48, environmental qualification(in 10CFR50.49, pressurized thermal shock (in 10CFR50.61),anticipated transients without scram in 10CFR50.62, andstation blackout in 10CFR50.63 [11].

(4) SSCs that are subject to: 10CFR50, Appendix B; 10CFR50,Appendix A, Criterion 1; plant facility Technical Specifi-cations; or plant Technical specifications through the defin-ition of operability (i.e., support SSCs outside technicalspecifications) [5]

(5) SSCs described in the Updated Final Safety AnalysisReport (UFSAR).

35.1.2.2 Functional /Functionality. As previously mentioned,the U.S. NRC [15] defines functionality as an attribute of SCCs notcontrolled by the plant technical specifications which warrantprogrammatic controls to ensure SSC availability and reliability(e.g., quality standards and records in Appendix B to 10 CFR 50,maintenance rule in 10 CFR 50.65, etc.) or other functionalityrequirements in the current license basis. A non-technical specifi-cation SSC is functional when it is capable of performing itsintended function(s) as specified in the plants current license basis.

A detailed discussion of the U.S. regulatory requirements asso-ciated with the scope of operability determinations and func-tionality assessments is contained in NRC Inspection ManualPart 9900 [15]. The NRCs distinction between these two assess-ments is summarized in Figures 35.1 and 35.2.

35.2 MECHANICAL COMPONENTSAND FAILURE MODES

Typical failure modes for SSCs can be determined from exten-sive industry experience reported in the Nuclear Plant ReliabilityData System (NPRDS), Licensee Event Reports (LERs), andother industry studies.

35.2.1 Piping ComponentsSince the early 1990s, significant attention has been placed on

establishing comprehensive databases that compile all reported ser-vice induced degradation in operating nuclear power plant pipingcomponents. The most significant work in this area has been doneas part of an OECD Pipe Failure Data Exchange (OPDE) Projectwhich has established an international database of piping and com-ponent failures (service induced cracks/wall thinning, leaks, andlarge breaks) in commercial nuclear power plants worldwide [37].Pipe failure data has been collected from operating nuclear powerplants worldwide, representing approximately 11,000 commercialreactor operating years from 1970 thru 2007. Considering that thereare thousands of numbers of piping components in nuclear powerplants this represents tens of millions of component years of serviceexperience from which to derive insights.

Figure 35.3 shows the total number reported piping failuresreported from 1970 through 2007 as a function of the piping com-ponent in-service life at the time of failure [38]. The relativelyhigh incidence of failures for the interval 11-15 years is largelyattributed to IGSCC in BWR plants, FAC in PWR plants, and cor-rosion of raw water piping in all plant types.

In Figure 35.4, U.S. failure data is organized degradation mech-anisms. It shows that design/construction defects and thermalfatigue (which are the degradation mechanisms at the basis of thecurrent Section XI preservice and inservice inspection require-ments) account for approximately 8% of the failures in U.S. com-mercial light water reactor (LWR) plants since 1970. Of these on1.5% of the failures were primarily due thermal fatigue, thermalmixing and thermal stratification mechanisms and loading condi-tions not considered in the original design. Over 87% of the pip-ing failures resulted from flow accelerated corrosion (FAC), stresscorrosion cracking (SCC), vibration and fretting fatigue, and cor-rosion mechanisms including: crevice corrosion, microbiological-ly induced corrosion (MIC), and pitting.

Figure 35.5 shows all U.S. pipe failure data since 1970 orga-nized by pipe safety classification. Generally service induced fail-ures in safety related have been evenly distributed throughout the


COMPANION GUIDE TO THE ASME BOILER & PRESSURE VESSEL CODE 631

three safety classes. Over 1/3rd of the failures occurred in non-safety pipe which are normally not reported to the regulatoryauthorities. The Class 1 piping failures are dominated by stresscorrosion cracking in BWR piping and Class 2 piping failures areprimarily associated with vibration fatigue, thermal stratification

and FAC mechanisms. The Class 3 failures are dominated by cor-rosion mechanisms. In all causes piping reliability is not a func-tion of design safety class; but, is dominated by service loadingand degradation mechanisms not specifically addressed in theASME Section III design standards.

FIG. 35.1 NRC OPERABILITY DETERMINATION AND FUNCTIONALITY ASSESSMENT FLOW CHART [15].

FIG. 35.2 NRC SCOPE OF AN OPERABILITY DETERMINATION AS IT RELATES TO THESCOPE OF A FUNCTIONALITY ASSESSMENT [15].


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FIG. 35.3 PIPING FAILURE TRENDS IN COMMERCIAL NUCLEAR POWER PLANTS COVERINGTHE PERIOD BETWEEN 1970 THROUGH 2007 [38].

FIG. 35.4 U.S. PIPE FAILURE DATA BY DEGRADATION MECHANISMS AND OTHERCAUSES [36].

Generally service experience shows that failures (cracks, wallthinning, leaks, and breaks) in light water reactor (LWR) pipingdo not correlate with high stress or thermal fatigue usage valuesreported in plant Design Reports [35]; in fact, service inducedfailures typically result from either degradation mechanisms or

loading conditions not specifically considered in the originalplant design.

The Code stress calculations are used to qualify a design andprovide reasonable confidence that the plant will provide reliableservice throughout its design life. For the most part the peak



stress and fatigue usage values in the plants Design Reports donot necessarily provide an accurate indicator of failure potential[33]. The peak stress and fatigue usage values in the DesignReports are typically based on extremely conservative postulatedthermal transient loading conditions and numbers of cycles. Also,the sequencing of these design thermal transients in the fatigueanalyses adds significant conservatism. Finally, the high calculat-ed fatigue usage is dominated by those design transient load paircombinations that not only result in large thermal stresses at theinside surface thermal gradients but also large thermal stress gra-dients through the wall. The large thermal stresses would result ina low number of allowable stress cycles for crack initiation (highfatigue usage); however, because of the large through-wall ther-mal gradient, stresses drop-off rapidly as you move away from theinside surface. Consequently, if all anticipated loading conditionsare accounted for, the time to grow a crack through-wall can besignificantly longer then the original design and extended life ofthe plant [34].

35.2.2 SupportsEPRI reported failure information on standard supports in

nuclear power plants, obtained from the Institute of Nuclear PlantOperations (INPO) NPRDS database [39]. Figure 35.6 providesa breakdown of standard support service data by failure mode.Most of the reported degraded supports were associated withmissing/loose components or improper hanger settings. One thirdsupport problems involved snubbers. Snubbers frequently faileddrag and motion criteria and experienced wear, fatigue, lock-up,and seal leakage mainly due to mechanical vibration.

In Figure 35.7, the breakdown of the NPRDS data is shown bycause/degradation mechanism. We can see that the cause of nearly50% of the reported support failures was unknown. These caseswere typically associated with missing or loose components,improper hanger settings, and the failure of snubbers to pass func-tional tests [39].

35.2.3 PumpsTypical failure modes for pumps are the following:

(1) Fails to start: This failure mode is used to describe faultsinvolving pumps that do not start upon demand or that startonly to operate for a brief time period before trippingoffline.

FIG. 35.5 U.S. PIPE FAILURE DATA BY SAFETY CLASSIFICATION [36].

FIG. 35.6 STANDARD SUPPORT FAILURES BY FAILUREMODE [39].

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(2) Fails to run: This failure mode indicates that an operatingpump is automatically or manually tripped off-line to pre-vent damage to the pump. It also includes pumps that fail tomeet design and/or operating specifications.

(3) External leakage: This failure mode describes a fault inwhich the pump is operational but is removed from servicebecause of excessive leakage of the pumped medium. Acommon example is a seal or packing leak.

35.2.4 ValvesTypical failure modes for valves are the following:

(1) Fails to open: Valve fails to fully open on demand. (2) Fails to close: Valve fails to fully close on demand. Included

are safety/relief valves that fail to reseat. (3) External leakage: A leak or rupture of the valve that allows

the contained medium to escape from the component bound-ary. The most common examples are packing or flange leaks.

(4) Plugged (i.e., fails to remain open): This failure mode refersto any event that would stop or limit the flow through a nor-mally open valve. However, the following are not consid-ered plugged valves: those that fail to open or those that areeither intentionally or unintentionally closed by humanactivity when they are required to be open. Two examples ofa plugging event are (a) a valve disc that separates from the stem and falls into

the closed position, and (b) the air supply to an air-operated valve that fails, allow-

ing the valve to close. (5) Fails to operate as required: The fails-to-operate-

as-required mode is to be used whenever (a) a valve fails to meet specific requirements such as

stroke time, or (b) a valve loses its ability to control system parameters.

(6) Fails to open/fails to close: This failure mode is used whenspecific information regarding whether the valve failed toopen or failed to close is not available.

(7) Internal leakage (reverse leakage): This failure mode is usedto describe internal leakage through a check valve.

(8) Opens (prematurely): This failure mode applies strictly torelief and safety valves. Valve opening before its pressuresetting is a typical example of this mode, but the cause is notalways a pressure transient.

35.3 OPERABILITY/FUNCTIONALITYEVALUATIONS

35.3.1 Conditions Requiring Assessment Conditions requiring functionality evaluations are usually iden-

tified through the inservice inspection and testing programs andaugmented inspection programs such as GL 88-01 intergranularstress-corrosion cracking (IGSCC) and flow-accelerated corrosion(FAC) inspection programs. Some examples of conditions identi-fied by the inservice inspection and testing programs are wall-thinning from an FAC mechanism, snubber failures and pipe ornozzle cracking from fatigue and IGSCC mechanisms, leakageexceeding TS limits for valves, and excessive pump vibrations. Insome instances, a system-operating excursion subjects a systemand its components to an unanalyzed condition that requires anevaluation.

Unanticipated operating events require operability evaluations.Examples of such events include fluid transients such assteam/water hammer and steam/vapor bubble collapse; tempera-ture and pressure excursions; and thermal stratification. Flow-induced system vibration from pump operation, cavitation, two-phase flow conditions and acoustic pressure waves are known tolead to snubber wear, piping erosion, and fatigue failure, especiallyin socket-welded fittings.

Functionality evaluations are required for temporary loadingconditions such as lead shielding, rigging for maintenance ormodification installation and scaffolding support. The loading tobe considered in conjunction with the temporary loads is depen-dent on the system operating status while the temporary conditionexists. The seismic load is not considered with the temporary con-dition if the affected system is declared out of service and thesystems seismic failure does not undermine the ability of othersystems or components to perform their safety functions.

A functionality evaluation is required for as-built/as-found con-ditions that are not consistent with the design-basis configuration.Examples of these conditions include support system discrepan-cies, which might consist of such missing, damaged, or failedsupports as weight supports that slide off their stanchions, a snub-ber failing to activate and provide the required restraint, or sup-port members found to have missing or undersized welds.Additional or unwanted restraints, such as snubbers that lock upor exceed the drag force restrictions, or a support design withinsufficient rotational clearance, are all examples of other sup-porting system discrepancies.

35.3.2 Scoping Operability Evaluations Determining the conditions to be evaluated requires a clearly

defined scope of the operability determination. To define thescope of the evaluation, the following actions are required:

(1) Identifying the equipment or SSC that is degraded, poten-tially nonconforming, or subject to the unanticipated event.

(2) Establishing the safety functions performed by theequipment.

FIG. 35.7 STANDARD SUPPORT FAILURES BY ROOTCAUSE [39].



(3) Assessing the possible failure mechanisms. (4) Defining the operating events or loading conditions concur-

rent with the evaluation period. (5) Determining the basis for declaring the affected system

operable through any of the following conditions:

(a) analysis; (b) test or partial test; (c) operating experience; or (d) engineering judgment.

Each of the conditions stated in sections 35.2.1 through 35.2.4require an operability evaluation to be performed. This evaluationshould consider the operational status of the SSC for the time periodduring which it is degraded or nonconforming. As one determinesthe operating events and loading conditions for the evaluation, it isimportant to distinguish between an SSC that was restored to itslicensing basis and one that is expected to operate in the degraded ornonconforming condition. The evaluation of an SSC that wasrestored to its current licensing basis is based on all operating eventsthat actually occurred while the SSC was in the nonconforming ordegraded condition. Included in such operating events are any of theunanticipated events (described previously) that were determined tohave occurred. However, the evaluation does not include design-basis loads that did not occur, such as a loss-of-coolant accident(LOCA), post-LOCA thermal modes, and seismic loads.

An example of a degraded and nonconforming SSC notrestored to its current licensing basis is a degraded snubber foundduring the Inservice Testing program to have exceeded its draglimit. An operability evaluation is required to determine theimpact of the excessive drag on the piping system. When seismicand LOCA events did not occur, only the thermal modes that didoccur since the last time the snubber was determined to haveoperated normally are considered in the evaluation. A root-causeanalysis would need to be performed to ensure that some unantic-ipated operating event, such as waterhammer or system vibration,

did not degrade the snubber. If the root-cause analysis determinesthat some unanticipated operating event did occur, the evaluationwould include this unanticipated event. The root-cause analysisshould be performed to establish that the degradation mechanismor nonconformance was eliminated and also to determine whichother SSCs may be impacted by it.

If the SSC is to continue operating while in the degraded condi-tion, the operability evaluation must demonstrate its functionality,as previously defined, for design-basis events required by plantlicensing commitments. If the SSC has been degraded by anunanticipated operating event, the operability evaluation mustdemonstrate its functionality for the design-basis events and theeffects of the unanticipated operating event. The effects of theunanticipated operating event would need to be included in theoperability evaluation until the cause of the event can be eliminat-ed or contingencies are put in place to significantly reduce thepossibility of the event occurring. Additionally, if an SSC that hadbeen degraded by an unanticipated operating event is restored toits design-basis condition, an operability evaluation must demon-strate the functionality of the SCC for its design-basis events andthe effects of the unanticipated operating event.

Following the guidance proposed by the ASME Section XITask Group on operability, only the failure mechanisms and loadsfrom the faulted or Service Level D condition are required for theevaluation. In keeping with this philosophy, displacement-controlled loading is not considered for piping or pressure bound-ary items. (Displacement-controlled loading refers to self-limitingloads such as thermal expansion, thermal-anchor movement, andseismic-anchor movement.) For support designs susceptible tononductile failure, such as buckling and anchor-bolt failure, orsupport designs that deform so severely that an active componentmight be unable to perform its required safety function, the ther-mal-anchor motion and seismic-anchor motion loadings must beconsidered. Table 35.1 presents this load consideration philoso-phy in a matrix form.


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35.4 ASME CODE REQUIREMENTSThe ASME B&PV Code Section III [8], Section XI [7], and the

ASME O&M Code [9] all contain requirements directly and indi-rectly related to the operability of components. The Section IIIrules encompass requirements for the design, construction, stamp-ing, and overpressure protection for nuclear plant items. Theserules are for new construction and give consideration to mechani-cal and thermal stresses caused by cyclic operation; they typicallydo not address deterioration that might occur in service becauseof radiation effects, corrosion, erosion, or other material degrada-tion mechanisms. The rules of this section are not applicable tovalve operators, controllers, position indicators, pump impellers,and other nonpressure-retaining items (except for those within thescope of Subsection NF), as well as motor drives and instruments.

Functional acceptability aspects of some components are notmentioned in Section III; for other components, a disclaimerstatement is provided. Such a disclaimer is provided for valves,for example, stating that Code valve design acceptability require-ments are not intended to ensure functional adequacy of thevalves. However, rules are provided for pressure-relief valves thatcover set pressure, lift, blowdown, and closure (NX-7000) [8].

The rules of Section XI and the O&M Code, on the other hand,are more directly related to various aspects of component oper-ability. These Codes define inspection and testing requirements toidentify degraded and nonconforming conditions for SSCs. Eachdefines the acceptance criteria and evaluation methods.

The following paragraphs itemize some specific stipulationsapplicable to pumps and valves from the O&M Code and vesselsand piping from Section XI. The pumps and valves covered bythese stipulations are components required for the following con-ditions:

(1) shutting down the reactor to the cold shutdown condition; (2) maintaining the reactor in a cold shutdown condition; and (3) mitigating the consequences of an accident.

35.4.1 Valves Valves within the scope of the inservice testing program are

placed in at least one of the following four categories as definedin Subsection ISTC of the O&M Code [9]:

Category Avalves for which seat leakage is limited to a spe-cific maximum amount in the closed position to fulfill theirrequired functions.

Category Bvalves for which seat leakage in the closed posi-tion is inconsequential for fulfillment of the required functions.

Category Cvalves that are self-actuating in response to a sys-tem characteristic, such as pressure (relief valves) or flow-direction (check valves) to fulfill the required functions.

Category Dvalves, such as rupture discs or explosively actu-ated valves, that are actuated by an energy source capable of onlyone operation.

The test requirements for each of these valve categories arespecified in ISTC3500 [9] and summarized in Table 35.2. Theserequirements define test intervals, leakage tests, and stroke times.

Table 35.3 provides the leak-testing criteria defined in ISTC-3600 [9]. Valves or valve combinations exceeding the specifiedcriteria shall be declared inoperable and either repaired orreplaced. An example of typical leakage rates from an operatingpressurized water reactor (PWR) is provided in Table 35.4.

Stroke-time acceptance criteria for active Category A and Bvalves are defined in ISTC-5000 [9] and summarized in Table 35.5.

If a valve fails to exhibit the required change of obturator positionor exceeds the limiting values of full stroke-time, the valve shall bedeclared inoperable. Valves declared inoperable may be repairedor replaced; in some cases, if the data can be analyzed to deter-mine the cause of deviation, they may even be defined as operatingwithin limits.

For Category C pressure-relief devices, testing requirementsare defined in the O&M Code, Appendix I [9]. The general rulefor pressure-relief valves is that they shall not exceed the stampedset-pressure criteria by more than 3%. Non-reclosing pressure-relief devices are not required to be tested. They are required, by



the O&M Code, Appendix I [9], to be replaced at least every fiveyears, unless historical operating experience indicates more fre-quent replacement is required. Facility Owners may have testingor replacement criteria varying from what is given here, which isbased on system and valve design requirements or TS require-ments. In most cases, if such criteria are not met, valves arerepaired, refurbished, or replaced.

Category C check valves are exercised for obturator movement.If a check valve fails to exhibit the required change of obturatorposition, it shall be declared inoperable.

35.4.2 Pumps For the subject pumps, inservice test parameters and parameter

ranges are specified in Subsection ISTB of the O&M Code [9]and examples of the test parameters with acceptance limits aresummarized in Tables 35.6, 35.7, and 35.8. Figure 35.8 depictsthe pump vibration limits in graphical form.

As shown in Table 35.7, the test parameter ranges are groupedas acceptable, alert, and required-action for the various parame-ters considered. The acceptance criteria for these pumps aredefined as the alert range, which indicates incipient degradationof performance, and the required action range. If deviations fallwithin the alert range, the specified frequency of testing shall bedoubled until the cause of the deviation is determined and thecondition corrected; if they fall within the required action range,the pump shall be declared inoperable until the cause of the devia-tion is determined and the condition corrected.

35.4.3 Snubbers Inservice test requirements for snubbers are specified in O&M

Code, Subsection ISTD, Inservice Testing of Dynamic Restraints(Snubbers) in Light-Water Reactor Power Plants [9]. This sub-sectionspecifies the scope, test intervals, sample sizes, and operability test

requirements for snubbers. The operability test requirements forsnubbers include the following parameters:

(1) The breakaway and drag force; (2) The activation-velocity or acceleration; and (3) The release rate. The acceptance criteria for each of these tested parameters are

specified as part of the test program and are determined by therestraint and motion requirements of the restrained component.Guidelines for determining the generic acceptance criteriagreater than the manufacturers specified criteria are provided inEPRI Report NP-6443, Improved Criteria for Snubber Testing[10]. These improved criteria for snubber operability are basedon the restraint requirements of the piping system or componentbeing restrained. By evaluating the piping system or componentthermal flexibility, increased snubber drag or breakaway forcesmay be justified.

35.4.4 Piping The rules of Section XI [7] define requirements for the evalua-

tion and acceptance of degraded piping components. Theserequirements have been developed for piping that is flawed ordegraded from a corrosion mechanism and are used to demon-strate the piping fitness for service when subjected to design-basisloads. (More detailed information regarding the flaw evaluationrequirements for piping is found in Chapter 29 of this book.) Forthe specific case of through-wall flaws in Class 3 moderate-ener-gy piping in which the maximum operating temperature does notexceed 200F and the maximum operating pressure does notexceed 250 psig, Code Case N-513 [25] specifies flaw evaluationmethods and acceptance criteria to justify the temporary accep-tance of this condition. The USNRC has approved the use of thisCode Case in the revision of 10CRF50.55a [11].

35.4.5 Reactor Vessel The rules of Section XI [7] define requirements for the evalua-

tion and acceptance of flaws and radiation embrittlement the reac-tor vessel. These requirements are used to demonstrate the ves-sels fitness for service when it is subject to design-basis loads.(More detailed information regarding flaw evaluation require-ments is found in Section 29.) Appendix E of Section XI Code [7]provides evaluation methods and acceptance criteria for the reac-tor vessel when it is subject to pressure in excess of the pressure-temperature limits required by 10CFR50.60 and Appendix G of10CFR50 [11]. Meeting the requirements of this appendix ensuresadequate structural integrity for returning the vessel to service.The Appendix E evaluation method reduces some of the margins


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inherent in the development of pressure-temperature limits byreducing both the postulated flaw depth and the safety factor onthe pressure-stress intensity factor, and by increasing the vesselmaterial fracture-toughness. This increase in vessel material frac-ture-toughness provides a significant margin for this structuralintegrity evaluation. The Appendix E evaluation changes the ves-sel material fracture toughness from Section XI, Appendix G [7],limits of Kla to the Klc limit defined in Section XI, Appendix A,Fig. A-4200-1. Figure 35.9 presents the Kla and Klc fracture-toughness values used for vessel evaluations. It should be noted thatthe Section XI, Appendix G criteria for developing the vesselpressure-temperature limits have been modified in Code Case N640[26] to permit use of the Klc vessel material-fracture toughness, but

such vessel pressure-temperature limits do not have this addition-al margin for their structural integrity assessments.

35.5 OPERABILITY EVALUATIONMETHODS

Design analysis methods and acceptance criteria applicable topiping systems, components and supports inherently containavailable margins to compensate for uncertainties in the opera-tion, fabrication, and construction of systems. Generally, designanalysis methods employ a conservative finite element model andsmall deformation linearelastic theory to estimate the system

FIG. 35.8 PUMP VIBRATION LIMITS (Source: Fig. ISTB-5200-1 of the ASME O&M Code)



response to the design-basis loads, an approach that generatesconservative system responses for the following reasons:

(1) piping system nonlinearity; (2) energy losses from localized plastic deformation; (3) load redistribution associated with items (1) and (2) of this

list; (4) conservatism of building structure response spectra; (5) conservatism of Enveloped/Uniform Response Spectra

analysis method; and (6) conservatism of critical damping values. These areas of conservatism have been substantiated by many

research and testing programs. Some of these test programs arethe Electric Power Research Institute (EPRI) multiple structurepiping system tests [1], the Heissdampf Reactor (HDR) tests [2],and the EPRI- and USNRC-sponsored Piping and FittingDynamic Reliability Program [20].

35.5.1 Piping Systems For piping systems, the evaluation method used to demonstrate

operability usually is to examine the design-basis analysis meth-ods for inherent conservative methods. The analytical methodsgenerally employed for the design basis of the piping are usuallybased on conservative models, load definitions, and solutionmethods. Refining the analytical model to include support andequipment nozzle stiffness values may reduce some of the conser-vatism inherent in the design basis, which is especially true whena time-history analysis is performed and support nonlinearities areincluded. Instead of using a uniform or enveloped response spec-tra method of analysis, the independent support motion methoddescribed in Welding Research Council (WRC) Bulletin 352 [3]together with an alternative system damping (such as Code CaseN-411 [4]) can be used to reduce the response of seismic and

other building-filtered loads evaluated by response spectra meth-ods. Section III, Appendix N [8] provides detailed descriptions ofthese alternative, more rigorous methods that may be used todetermine system responses to seismic and other building-filtereddynamic loads. Also, the combination of dynamic loadingresponses from different events permits the use of the squarerootsum-of-the-squares method. This combination method is jus-tified when the predominant frequencies of the events being com-bined are sufficiently separated to make the probability of simul-taneous maximum responses for each event very unlikely [6].

Finally, the load cases used in the evaluation should be consis-tentwith and limitedthose loads that can realistically occurduring the period of operation when the piping system is in thedegraded or nonconforming condition.

35.5.2 Supports For supports, the evaluation method usually used to demonstrate

operability is a structural analysis method. Although for supportsand especially expansion anchors, the operability evaluation may bebased on test results or a combination of test results and analysis.

Generally, support designs are based on conservative models,load definitions, and solution methods. Developing more rigorousanalytical models can be used to redistribute loads in the supportstructure. In many cases support operability is demonstrated byevaluating the system or component being supportedan evalua-tion that is performed to reduce the loads acting through thedegraded or nonconforming support. In other cases, the system orcomponent evaluation demonstrates that support functionality isnot needed for the system or component to perform as required.

35.5.3 Valves Operability evaluations of valves are generally based on the

determination of margins available in the original qualification

FIG. 35.9 LOWERBOUND KLA AND KLC TEST DATA REACTOR VESSELS (Source: Fig. A-4200-1, Appendix A of the ASMEB&PV Code)


640 Chapter 35

documents. In most cases, the typical parameter to consider is thedynamic acceleration limit used for the qualification. Valve quali-fications are often provided for generic specification values ofseismic demand, which are arbitrarily set limits. For plant-specificapplications, the actual seismic demand is compared with thegeneric qualification values for acceptance.

The most effective approach for valve operability evaluation isto verify that the available margins between the qualified demandand the demand determined for the conditions requiring operabilityverification are sufficient for accepting the critical operabilityparameter (e.g., stress level, deformation, and load). In caseswhere this direct comparison does not provide the required relief,efforts may be directed toward realistic determination of the actualdemand. Methods previously discussed for the reevaluation ofpiping and support systems may result in sufficiently reduceddemand at valve locations. Valves may also be analyzed by usingfinite element methods to assess stress and deformation at criticallocations for increased loads. These analyses often indicate onlylocalized high stress and deformation conditions that do not affectthe valves functionality.

A particular valve operability problem is related to theincreased torque or thrust requirements to meet certain systemdemands related to the assigned safety function. In most cases, itis possible to show analytically that the weakest link on the postu-lated load paths have sufficient capacity and integrity to meet theincreased demand. However, in extreme cases in which an analyt-ical approach does not provide the required verification, testingmay ensure the functionality. Care must be exercised in extendingthe applicability of the test results to untested configurations.

35.5.4 Pumps, Tanks, and Heat Exchangers Typically, equipment items such as pumps, tanks, and heat

exchangers have allowable nozzle load limits. As with valves, theVendor Equipment Qualification Report should be reviewed todetermine the basis of the allowable nozzle loads.

In a typical heat exchanger, the allowable nozzle loads are usedto check the following:

(1) the nozzle at the junction with the shell; (2) the shell in the region of the nozzle; (3) gross shell stresses; (4) localized shell stresses at the saddles; (5) gross stresses in the saddle supports; and (6) gross loads on the anchor bolts. Items (1) and (2) of the preceding list use the loads for their

respective nozzles. Items (3)(6), however, use the combinedeffects of weight, dynamic inertia, operational loads (e.g., asmotor torque in a pump), and all nozzles.

In many cases, particularly on Class 2 and 3 vessels, item (2) isthe limiting component; for this item, the local stress analysis ofthe shell should be reviewed. Such analyses are typically doneusing the methods of WRC Bulletin 107 [21] and WRC Bulletin297 [22]methods for which vendors occasionally use the totalthrough-wall stress (i.e., membrane bending, or using WRCBulletin 107 terminology N, membrane force terms M, andbending moment terms) when comparing the nozzle stress to theprimary local membrane allowable, rather than just the membrane(N, terms) component. For Service Level D conditions, local mem-brane effects should be checked in the shell at the shell nozzle junc-tion, but not local membrane plus local bending (PL Pb Q).The requirements for local checks are taken directly fromAppendix F of the ASME Boiler and Pressure Vessel Code [8].

Through-wall bending effects (the M terms) are secondary effectsand, as such, do not need to be considered, which can providesubstantial relief.

Alternate analyses may be performed by using finite elementmodels at critical areas to assess the stresses and deformations.

35.5.5 Specific Inspections For those operability evaluations on SSCs that have been

restored to their current licensing basis before being placed backinto service, or on SSCs subjected to a significant unanticipatedoperating event, an evaluation may be based on specific augment-ed inspections. Analyzing the degraded or nonconforming SSC,or analyzing the unanticipated event to identify those areas of theSSC that are most likely to sustain damage will guide these aug-mented inspections. The results of the inspection will be evaluat-ed to demonstrate the functionality of the SSC, after which theywill be documented. The forthcoming lists describe the probableinspection areas and the damage that may affect the performanceof the SSCs. These inspections will be visual and will includemanual testing for active components. For highly stressed areaswhere crack initiation may be expected, surface or volumetricexamination may be required as well.

Whenever damage is found, the results of the inspections andtests require an engineering evaluation to determine the function-ality of the affected SSC. The nature of this evaluation willdepend on the type and extent of the damage found.

35.5.5.1 Piping Systems. This will be a visual inspection of thepipe to determine whether any significant damage has occurred.The examination for damage will include the following:

(1) distortion of the piping cross section; (2) local denting or buckling of the wall; (3) leaks; (4) permanent distortion of piping; (5) interaction with adjacent SSCs; (6) broken or bent bolts; (7) distortion of flanged connections; (8) cracked socket welds; and (9) interferences or unintentional restraints.

35.5.5.2 Component Supports. This will be a visual inspectionof the supports to determine whether any significant damage hasoccurred. The examination for damage will include the following:

(1) buckling of members or rods; (2) local distortion of members; (3) cracks at welds; (4) pullout of concrete expansion anchors; (5) broken or bent bolts; (6) stripped-bolt threads; (7) spalled concrete; (8) distorted baseplate; (9) hydraulic snubber fluid leaks; and

(10) clearance at rigid supports. The examination for snubbers and other supports, which are

relied upon to change position, will include manually strokingthem through their expected range of operation to determinewhether any internal damage has occurred.

35.5.5.3 Valves. The valves will be visually examined for leak-age of and damage to bolts, as well as distortion to the yoke and



stem. Valves required to change position during the remainder ofthe operating cycle will be stroked to the extent practicable todetermine their functionality.

35.5.5.4 Pumps. Pumps that were operating during the eventwill be monitored for vibration and flow and will also be visuallyexamined for leakage. Standby pumps that were not operating dur-ing the event but that may be required to perform a safety functionwill require testing (to the extent practicable) for vibration and flowor pressure, and will also require a visual examination for leakage.

Passive mechanical components (e.g., vessels and heat exchang-ers) will require a visual examination for leakage as well as fordamage near the nozzles and anchorages. In addition, heat exchang-er performance requires monitoring for tube-to-shell leakage.

35.6 SHORT-TERM OPERABILITYACCEPTANCE CRITERIA

The acceptance criteria presented here are currently being con-sidered by the ASME Section XI Task Group on Operability forpublication as a Code Case [18]. These proposed criteria arebased on industry evaluations of the margins present in the analyt-ical methods and also in the acceptance criteria used for design ofthe SSC. For example, for piping systems these industry investi-gations have lead to alternative evaluation methods and accep-tance criteria for seismic and other building-filtered loads duringnormal and upset conditions.

The development of the criteria provided here was based onASME Section III, Appendix F [8] criteria and is augmented withindustry testing programs for application to specific piping systemcomponents.

35.6.1 Short-Term Operability Acceptance Criteriafor Piping

The short-term operability acceptance criteria presented hereare for piping systems designed to various ASME and ANSICodes and Code editions [8], [23][24]. The loads to consider inthe operability evaluations for piping are defined in Table 35.1.These criteria are intended for piping and typical piping compo-nents. If the evaluated system includes specialty piping compo-nents such as flanges and expansion joints, these specialty pipingcomponents are expected to meet the current licensing-basis ser-vice Level D/faulted condition limits.

35.6.1.1 Criteria for Piping Designed to ANSI B31.1, B31.7Classes 1 and 3, or ASME Section III, Classes 2 and 3(Prewinter 1981).

For Level A/normal loads:

(35.1)

For Level D/faulted loads:

(35.2)

(35.3) The criteria presented in the preceding equations were developed

from the work performed during the seismic upgrade program for

Pmax 2Pa

PmaxD4t

+ 0.75i (MA + MD)

Z6 2Sy

PD4t

+ 0.75i MAZ

6 Sy

the San Onofre station in California [12]. This program developedoperability criteria based on a 1% strain acceptance criterion forcarbon steel piping and 2% strain acceptance criterion for stainlesssteel piping, which were both approved by the USNRC [12]. Indeveloping the strain limit acceptance criterion, a stress acceptancecriterion of 2Sy was shown to be more limiting than that when thestress values were calculated using elastic methods.

35.6.1.2 Criteria for Piping Designed to ANSI B31.7, Class 1,or to ASME Section III, Class 1.


(35.4)


(35.5)

(35.6) The criteria presented in the preceding equations were founded

on the evaluations and acceptance criteria specified in AppendixF of Section III [8]. This stress limit has been specified as 2Sy ,not the lesser of 3Sm or 2Sy , based on prior acceptance of thislimit [17].

35.6.1.3 Criteria for Piping Designed to ASME Section III,Classes 2 and 3, Winter 1981 or Later Edition.


(35.7)


(35.8)

(35.9) The criteria presented in the preceding equations were founded

on the evaluations and acceptance criteria specified in Appendix Fof Section III [8].

35.6.2 Short-Term Operability Acceptance Criteriafor Supports

The proposed operability acceptance criteria are developed forcomponent standard supports, spring hangers, snubbers, linear-type supports, structural bolts, and concrete expansion anchors.These criteria were developed to address the load-carrying capa-bility of the supports. However, the other support failure modesmust also be addressed when applicable. Support design stability,including the potential for buckling, should be evaluated unlessthe system is shown to be functional without the support. Supportbinding should also be addressed in the support operability evalu-ation and the operability evaluation for the system. Table 35.1provides guidance to consider for the loadings in the support eval-uation. If the necessity of including seismic anchor motion isdetermined, it should be combined with the seismic inertia loadby a square rootsum-of-the-square method.

Pmax 2Pa

B1 PmaxD

2t+ B2

(MA + MD)Z

6 greater of 3 Sh or 2Sy

B1 PD2t

+ B2 MAZ

6 the greater of 1.5 Sh or Sy

Pmax 2Pa

B1 PmaxD

2t+ B2

(MA + MD)Z

6 2Sy

B1 PD2t

+ B2 MA

Z6 the greater of 1.5Sm or Sy


642 Chapter 35

35.6.2.1 Component Standard Supports. Component stan-dard supports are supplied by support manufacturers with definedload capacities. The following criteria will be used to determinethe operability of these components when they are subject toService Level D/faulted condition loads as defined in Table 35.1.

(a) The manufacturers ultimate tested load divided by a safetyfactor of 1.5 except for the side load on U-bolts, whichshould be 3.0. In determining this limit, the allowableshould be modified by the temperature effects on the mate-rial ultimate strength.

(b) The manufacturers allowable load for Service Level D. (c) The manufacturers allowable load for Service Level A mul-

tiplied by either of the following: (i) the lesser of 2 or 1.167 (Su /Sy) when Su 1.2 Sy or

(ii) 1.4 when Su 1.2Sy. If the criteria in the preceding list cannot be met, the criteria in

the paragraphs that follow for linear-type supports can be used todemonstrate the operability of the standard component.

35.6.2.2 Linear-Type Supports. For linear-type supports,ASME Section III, Appendix F, paragraph F-1334 [8] provides thebasis for the operability criteria. The following limits summarizethe criteria of this subparagraph, although not all of the criteriaspecified in this subparagraph are presented here.

For Tension Stress, Ft min(1.2Sy, 0.7Su) except at pinholes,which use min(0.9Sy, 0.5Su).

For Shear Stress, Fv min(0.72Sy, 0.42Su).For Bending Stress, Fb (f ) Sy for compact sections where

(f ) the plastic shape factor. For noncompact sections, seeAppendix F, paragraph F-1334.4(c) [8].

For Compression, Fa min(Ft, 0.67Scr).For Combined Axial Tension and Bending, the stress of members

subjected to both axial tension and bending must be proportioned tosatisfy the requirements of equation (35.10), which follows.

(35.10)

WhereFa the smaller of 1.2Sy or 0.7SuFor Combined Axial Compression and Bending, the stresses of

members subjected to both axial compression and bending mustbe proportioned to satisfy the requirements of equations (20),(21), and (22) of ASME Section III, NF-3322.1(e)(1) [8]. TheNF-3322.1(e)(1) equations are modified to use the Fa, Fb, and definitions of Section III, Appendix F, paragraph F-1334.5 [8].

35.6.2.3 Structural Bolts. Structural bolting will meet therequirements of Section III, Appendix F, paragraph F-1335. For allstructural bolts, the average tensile stress computed on the basis ofthe average tensile stress area will not exceed the lesser of 1.0Sy and0.7Su. The average shear stress will not exceed the lesser of 0.6Sy and0.42Su. For high-strength structural bolts (Su 100 ksi at operatingtemperature), the maximum stress at the periphery of the cross sec-tion caused by direct tension plus bending but excluding stress con-centrationswill not exceed Su. If structural bolts are subject tocombined tensile and shear loads, the tensile and shear stresses mustbe proportioned so that the following equation is satisfied.

(35.11)f2t

F2tb+

f 2yF2yb

1

7

Fe

faFa

+fbxFbx

+fbyFby

1

67

35.6.2.4 Concrete Expansion Anchors. As previouslydescribed in Section 35.2.5, displacement-controlled loads shouldbe considered when evaluating the operability of concrete expan-sion anchors. An exception to this requirement is for undercutanchors if the failure mode is in the bolt, not in the concrete.

The operability limits for tension and shear loads acting onconcrete expansion anchors are obtained from the ultimate capac-ities determined by the manufacturer. A safety factor of 2 on theselimits is required by the NRC Inspection and EnforcementBulletin 79-02 [19]. For the bolt ductile failure mode of undercutanchors, a safety factor of 1.5 may be used. Also, anchors nearfree edges are subject to shear failure in the concrete and shoulduse a safety factor of 2. Anchors subject to combined tension andshear will be evaluated using the following interaction equation.

(35.12)

Reductions in the tensile and shear capacities of the expansionanchor caused by center-to-center and edge distance violationsmust be determined before multiplying by the required safety fac-tor of 2 or 1.5.

Some anchorages are installed with gaps between the plant andconcrete or plate-and-bolt head. In these cases, the prying forcetypically included in such analyses may be neglected. Anchorbolts with gaps up to in. are shown to develop full capacity and,therefore, are acceptable [16].

35.6.2.5 Integral Welded Attachments. For integral weldedattachments to straight pipe, the methodologies defined in CodeCases N-122 and N-318 will be used for rectangular attachments,and those defined in Code Cases N-391 and N-392 will be used forcircular attachments.

The stress limit to use for evaluations based on each of theseCode Case is 2.0Sy.

35.6.2.6 Spring Hangers. Spring hangers will be evaluated forloading clearances. The maximum pipe movement will be checkedagainst the stroke of the hanger, and where the spring hanger wassubjected to an excessive weight load, the individual componentswill be evaluated for the excess load.

35.6.2.7 Snubbers. Snubbers that are subject to an unanticipateddynamic load will be evaluated against the manufacturers ServiceLevel D allowable or against the one time allowable test load.

The primary failure mode for snubbers is excess drag and lock-up. In these cases the impact of the snubber failure on the pipingsystem requires evaluation, which includes a fatigue evaluation forpiping systems with fatigue design requirements. As previouslydiscussed in Section 2.5, the fatigue analysis considers the operat-ing events and their thermal transients known to occur while thesnubber is degraded. Using the guidance provided in EPRI ReportNP-6443 [10], increased snubber breakaway and running dragloads can be justified; for example, depending on the piping size,allowable drag loads of 5% of the snubber load rating are justified.For hydraulic snubbers, the common failure mode is leakage,resulting in a lack of restraint [14]. This failure mode requires thesystem or component to be evaluated without the snubber.

35.7 LONG-TERM OPERABILITY Generic Letter 91-18 provides specific guidance for long-term

operability by requiring that the degraded or nonconforming SSC be

18

a FtFtob5/3 + a Fy

Fyob 1.0



brought back into complete compliance with the licensing basisrequirements. This task requires the repair or replacement of the SSCin accordance with ASME Section XI requirements. Alternative tasksinclude a more detailed, sophisticated analysis to demonstrate thatthe SSC complies with licensing commitments, as well as revisingthe licensing basis to make it conform to the required USNRC reviewand approval to bring the SSC into compliance.

35.8 COMMON TERM DEFINITIONSCurrent Licensing Basis Current licensing basis (CLB) is the

set of USNRC requirements applicable to a specific plant com-posed of a licensees written commitments for ensuring compliancewith and operation within the applicable USNRC requirements andalso the plant-specific design basis (including all modifications andadditions to such commitments over the life of the license) that aredocketed and in effect. The CLB includes the following:

(1) The USNRC regulations contained in 10CFR Parts 2, 19,20, 21, 30, 40, 50, 51, 55, 72, 73, and 100, as well as appen-dices to those Parts.

(2) Orders, license conditions, exemptions, and the TechnicalSpecifications (TSs).

(3) The plant-specific design-basis information defined in10CFR50.2, as documented in the most recent Final SafetyAnalysis Report (FSAR) as required by 10CFR50.71.

(4) Docketed licensing correspondence such as licenseeresponses to USNRC bulletins, generic letters, and enforce-ment actions.

(5) Licensee commitments documented in USNRC safety eval-uations or licensee event reports.

Design Basis Design basis is the body of plant-specific designbases information defined by 10CFR50.2.

Design-Basis Events As defined in 10CFR50.49(b)(1)(ii),design-basis events include normal operating conditions, anticipatedoperating transients, design-basis accidents, external events, andnatural phenomena for which the plant was designed to with-stand.

Degraded Condition An SSC condition in which any loss ofquality or functional capability occurs.

Nonconforming Condition An SSC condition in which thereis failure to meet requirements of licensee commitments. The fol-lowing are some examples of nonconforming conditions:

(1) A failure to conform to one or more applicable Codes orStandards specified in the FSAR.

(2) As-built or as-modified equipment that does not meet FSARdesign requirements.

(3) Operating experience or engineering reviews that demon-strate a design inadequacy.

(4) Documentation required by USNRC requirements, such as10CFR50.49, that is deficient or unavailable.

Full Qualification Full qualification constitutes conforming toall aspects of the current licensing basis, including Codes andStandards, Design Criteria, and commitments.

Active Components Those components that perform amechanical motion to accomplish their assigned safety functions.

35.9 NOMENCLATURE Ag area of gross section of linear support member Ap pipe cross-sectional (metal) area

b actual width of stiffened and unstiffened compressionelements

bf flange width of rolled beam or plate girder, in. B1 pressure stress index from NB-3683 [8]

B2, C2 moment stress indices from NB-3683 [8] Cm coefficient applied to the bending term in the interaction

equation and dependent upon column curvature causedby applied moments

D pipe outside diameter E modulus of elasticity of steel, ksi fa computed axial stress, ksi fb computed bending stress, ksi ft bolt tensile stress, ksifv bolt shear stress, ksi

Ftb allowable bolt tensile stress at temperaure, ksi Fvb allowable bolt shear stress at temperature, ksi

Euler stress divided by factor of safety, ksi Fa axial stress permitted in the absence of bending

moment Fb bending stress permitted in the absence of axial force Fs allowable related to strength in combined compression

bending Ft tensile stress

Fto allowable tensile stress in a concrete expansion anchor Fv shear stress

Fvo allowable shear stress in a concrete expansion anchor Fw stress in a fillet weld

i stress intensification factor K effective length factor

KIa material fracture toughness based on crack arrest KIc material fracture toughness based on fracture initiation

I for beams, distance between the cross section braced againsttwist or lateral displacement of the compression flange, in.

for columns, actual unbraced length of member, in., andunsupported length of lacing bar, in.

MA sustained moment MD amplitude of the moment for Level D (all dynamic loads) Msl amplitude of the moment for Level D/faulted condition

(reversing dynamic loads only) Msm range of the moment from Level D/faulted seismic

anchor motion P design pressure

Pcl maximum compressive allowable load of a linearsupport member

Pmax pressure for Level D/faulted condition that is coincidentwith loads being evaluated

Pa allowable working pressure or rated pressure for eachpiping component as determined by the pressuredesign section of the Construction Code or Code ofRecord

Psm amplitude of the axial force from Level D/faultedseismic anchor motion

rb radius of gyration about axis of concurrentbending, in.

Scr critical buckling loadSh ASME design stress allowable, Classes 2 and 3 Sm ASME design stress intensity allowable, Class 1 Su specified minimum tensile strength at temperature Sy specified minimum yield strength at temperature tf flange thicknesst wall-thickness

Z pipe section modulus

Fe


644 Chapter 35

35.10 REFERENCES 1. EPRI Report NP-4865, Experimental Studies of the Seismic

Response of Piping Systems Supported by Multiple Structures, FinalReport, Jan. 1987.

2. Kot, C. A., Srinivason, M. G., and Hsiek, B. J., Margins for the In-Plant Piping System Under Dynamic Cooling, Seismic Engineering,PVP-Vol. 220, The American Society of Mechanical Engineers, 1991.

3. WRC 352, (1992). Independent Support Motion (ISM) Method ofModal Spectra Seismic Analysis, Welding Research Council,Bulletin 352, New York.

4. ASME Boiler and Pressure Vessel Code, Section III, Division 1, CodeCase N-411-1, Alternative Damaging Valves for Response SpectraAnalysis of Class 1, 2, and 3 Piping; The American Society ofMechanical Engineers.

5. GL 91-18, Information to Licensees Regarding NRC InspectionManual Section on Resolution of Degraded and NonconformingConditions, Rev. 1, Oct. 8, 1999.

6. NUREG-484-1, Methodology for Combining Dynamic Responses,May 1980.

7. ASME Boiler and Pressure Vessel Code Section XI, Division 1; TheAmerican Society of Mechanical Engineers.

8. ASME Boiler and Pressure Vessel Code Section III, Division 1; TheAmerican Society of Mechanical Engineers.

9. ASME Operation and Maintenance Code, Code for Operation andMaintenance of Nuclear Power Plants; The American Society ofMechanical Engineers.

10. EPRI Report NP-6443, Improved Criteria for Snubber Testing, July1989.

11. U.S. CFR, Title 10, Energy, Part 50, Domestic Licensing ofProduction and Utilization Facilities, Regulations Effective Jan. 2000.

12. Safety Evaluation Report, Safety Evaluation for Return to ServicesPlanSeismic Reevaluation Program, San Onofre NuclearGenerating Station, Unit 1, Docket No. 50-206. USNRC Letter fromH. R. Denton to K. P. Baskin, Feb. 8, 1984.

13. TCG ReportFOAKE Task E-1, ASME Piping for AdvancedReactor Corporation.

14. NUREG/CR-5416, Technical Evaluation of Generic Issue 113:Dynamic Qualification and Testing of Large Bore HydraulicSnubbers, Sept. 1992.

15. USNRC Inspection Manual, Part 9900: Technical Guidance,Operable/ Operability, STS10OP.STS.

16. EPRI Report TR-101968, Guidelines and Criteria for Nuclear Pipingand Support Evaluation and Design, Vols. 18, AprilMay 1993.

17. USNRC Letter to Wisconsin Electric Company, Docket 50-266 and50-301, Nov. 8, 1989, Interim Operability Criteria for Safety-RelatedPiping and Associated Supports.

18. Minichiello, J. C., Tang, H. T., and Williams, H. L., A ProposedPiping System Short-Term Operability Criteria, Codes andStandards for Quality Engineering, PVP-Vol. 285, The AmericanSociety of Mechanical Engineers, 1994.

19. IE Bulletin No. 79-20, Pipe Support Base Plate Designs UsingConcrete Expansion Anchor Bolts, Rev. 1, Suppl. 1, Aug. 20, 1979.

20. EPRI/NRC Test Program, Piping and Fitting Dynamic ReliabilityProgram, TR-102792, Vols. 15, Oct. 1994.

21. WRC 107, Wichman, K. R., Hopper, A. G., and Mershon, J. L. (1979).Local Stresses in Spherical and Cylindrical Shells Due to ExternalLoadings, Welding Research Council, Bulletin 107, New York.

22. WRC 297, Mershon, J. L., Mokhtarian, K., Ranjan, G. V., andRodabaugh, E. C. (1987). Local Stresses in Cylindrical Shells Due toExternal Loadings on NozzlesSupplement to WRC Bulletin 107(Revision 1), Welding Research Council, Bulletin 297, New York.

23. ANSI B31.1, Power Piping; The American National Standards Institute.

24. ANSI B31.7, Power Piping; The American National StandardsInstitute.

25. ASME Boiler and Pressure Vessel Code Section XI, Division 1, CodeCase N-513, Evaluation Criteria for Temporary Acceptance of Flawsin Class 3 Piping; The American Society of Mechanical Engineers.

26. ASME Boiler and Pressure Vessel Code Section XI, Division 1, CodeCase N-640, Alternative Reference Fracture Toughness forDevelopment of P-T Curves for ASME; The American Society ofMechanical Engineers.

27. NRC Regulatory Issue Summary 2005-20, Revision 1, U.S. NuclearRegulatory Commission, Washington D.C. (ML0373440103).

28. Standard Technical Specifications Babcock and Wilcox Plants,NUREG-1430, Vol. 1, Rev. 3.0, U.S. Nuclear Regulatory Commission,Washington D.C. June 2004.

29. Standard Technical Specifications Westinghouse Plants, NUREG-1431,Vol. 1, Rev. 3.0, U.S. Nuclear Regulatory Commission, WashingtonD.C. June 2004.

30. Standard Technical Specifications Combustion Engineering Plants,NUREG-1432, Vol. 1, Rev. 3.0, U.S. Nuclear Regulatory Commission,Washington D.C. June 2004.

31. Standard Technical Specifications General Electric Plants, BWR/4,NUREG-1433, Vol. 1, Rev. 3.0, U.S. Nuclear Regulatory Commission,Washington D.C. June 2004.

32. Standard Technical Specifications General Electric Plants, BWR/6,NUREG-1434, Vol. 1, Rev. 3.0, U.S. Nuclear Regulatory Commission,Washington D.C. June 2004.

33. Cooper, W.E., 1992. The initial Scope and intent of the Section IIIDesign Procedures, Paper presented at PVRC Workshop on CyclicLife and Environmental Effects in Nuclear Applications, January 1992.

34. Gosselin, S.R., F.A. Simonen, P.G. Heasler, and S.R. Doctor, 2007.Fatigue Crack Flaw Tolerance in nuclear Power Plant PipingABasis for Improvements to ASME Code Section XI AppendixL, NUREG/CR-6934, U.S. Nuclear Regulatory Commission,Washington, D.C.

35. Simonen, F.A. and S.R. Gosselin, 1999. Life Prediction and Monitoringof Nuclear power Plant Components for Service Related Degradation,J. Pressure Vessel Technology, Vol. 123, pp. 58-64, February, 2001.

36. Lydell, B.O.Y., 2007. PIPExp-2007 High Level Summary ofDatabase Content as of July 31, 2007, SPI-R-2007-01.07, Sigma-Phase Inc., Vail, AZ, August 6, 2007.

37. Lydell, B., Huerta, A. and Gott, K., 2007, Progress with theInternational Pipe Failure Data Exchange Project, PVP2007-26278,Proc. 2007 ASME Pressure Vessel and Piping Division Conference,July 2226, 2007, San Antonio (TX).

38. Lydell, B., Huerta, A. and Gott, K., 2008, Characteristics of Damage& Degradation Mechanisms in Nuclear power Plant Systems,PVP2007-61914, Proc. 2008 ASME Pressure Vessel and PipingDivision Conference, July 2731, 2007, Chicago (IL).

39. Olson, D.E., B.J. Voll, and H.T. Tang, 1994, Behavior and FailureMode of Standard Components Beyond their Design Condition:Recommendations for Support Evaluation and ReconciliationCriteria, PVP-Vol. 285, Codes and Standards for QualityEngineering, American Society of Mechanical Engineers, N.Y.


Functionality Definition

Documents

components sscs

operable operability

sscs capability

typical sscs

nonsafety related sscs

longterm operability

operability determinations

related support functions