INTERNATIONAL DATABASES ON PIPING FAILURES: DO THEY …

SE9800023

RSA-R-97-20(SKI Ref. No.: 14.2-940477)

INTERNATIONAL DATABASES ON PIPINGFAILURES: DO THEY EXIST - ARE THEY NEEDED?

Seminar on Piping Reliability

Sigtuna - SwedenSeptember 30 - October 1,1997

By

Bengt LydellRSA Technologies

San Diego, USA

INTERNATIONAL DATABASES ON PIPING FAILURES:DO THEY EXIST - ARE THEY NEEDED?1

Bengt O.Y. LydellRSA Technologies

Vista (San Diego), California 92083-6172U.S.A.

Abstract: The paper contends that no recognized PSA-oriented database on pipingfailures exists in the public domain. This fact is a reflection of the complex nature ofpiping reliability. Under the assumption that a need exists for database development,the paper gives an overview of fundamental issues associated with data on pipingfailures. The paper concludes by outlining a strategy for international cooperation todevelop a comprehensive piping reliability database.

1. INTRODUCTION

During 1994-1997, SKI funded a project to determine the viability of developing aPSA-oriented database on piping failures. Results from this project are documentedin SKI Report 97:262. In summary, a large body of operational data does exist forcommercial NPPs. That information is widely dispersed, and is typically of very lowpedigree. The effort to extract, interpret and classify these data is considerable.Furthermore, since the operational data 'reside' in a wide range of reporting systems,the determination of the accuracy and quality of data requires consultation of a largevolume of technical documents. As an example, for the qualification of a single pipefailure up to five different information sources were consulted.

The fundamental principle that databases on equipment failures must betailored to specific objectives is particularly relevant for pipe failure data collections.Data should be collected on an event basis. This entails detailed consideration of rootcause analysis. Since pipe failures are symptoms of several causal factors, the rootcause analysis principle assumes that a database design clearly distinguishes theunique reliability attributes and influence factors.

2. SKI's DATABASE ON PIPE FAILURES

The SKI database contains information on known pipe failures in nuclear power plantsworldwide. It covers the period 1970 to the present. In developing the database thescope of the work has emphasized piping failures in light water reactors (LWRs).Significant failures in heavy water reactors (e.g., CANDU plants) and graphite-

1 Seminar Topic: 'Data Collection' - Paper presented in the pm-session on September 30, 1997.2 SKI Report 97:26: Reliability of Piping System Components: Framework for Estimating FailureParameters from Operational Data. The report is to be published in November, 1997.

SEMINAR ON PIPING RELIABILITY 1Sigtuna (Sweden)September 30 - October 1,1997

moderated and channel-type boiling water reactors (LWGR plants) were selectivelyconsidered as well. Currently (September 1997), the database includes about 2,330failure reports; c.f. Table 1.

Table 1: The Database Content (September 1997).•

PLANT TWEC*)

BWRLWGRPHWR

PWR + WWER

Total:

NUMBER OFPLANTS

SURVEYED

711919164

274

RELATIVECOVERAGE

JFailure/PIant Type]

11.33.14.88.4

• • • •

FAILURE MODE

Crack<&)

1053855

171

Leak

6374174

1205

1957

Ruptured

621414113

203

Note: (a) Note, the material used in primary system piping differs among the plant types; e.g.,industrial grade vs. 'nuclear grade' stainless steel. Also, as an example, in WWER-1000, theprimary system piping material is ferritic steel with austenitic cladding as an anti-corrosionmeasure.(b) Significant events only, with generic implications,(c) Catastrophic loss of structural integrity and/or leak rate > 5 kg/s (80 gpm), no advancewarning.

The 'rupture category' in Table 1 includes catastrophic events, which occurredwithout advance warning, or failures resulting in major leakage in excess of 5 kg/s (80gpm). The failure reports were all classified according to leak rates. For the majorityof the reports, the leak rates were estimated based on event narratives.

In Table 1, the relative coverage is a measure of the scope of the surveys ofoperational data by this project. A low rate indicates that for the particular plant type,a systematic search for failure data was not within the current work scope. The readerof this report is advised not to apply any other interpretation of the population datapresented in Table 1. The project scope emphasized piping failures in BWRs andPWRs. The difference between BWRs and PWRs is explained by the generic IGSCCproblems, which affected BWR during the early to mid-1980's.

Except for the Swedish, U.S. and selected European plants, for which licenseeevent reports and special failure reports were used, the only primary reference usedwas the IAEA/NEA Incident Reporting System (IRS). By design, the IRS databaseonly includes nominated, or significant events as submitted by participatingorganizations. That is, an event report is submitted to IRS when the event isconsidered by a national coordinator to be of international interest. Only events ofsafety significance are reported. About 10% of all entries currently in SKI's databasewere extracted from the IRS database.

SKI's 'PSA-oriented' database content is compared with a recent, independentdata collection effort in Table 2. This independent effort was directed at U.S. plantexperience for the period 1961-1995. By contrast, the 'PSA-oriented' databaseincludes 1837 failure reports for U.S. plants for the period 1970-1997; i.e., about 80%


of the total number of records are for U.S. plants. Of the total database content, about5% of the records are for Swedish and Finnish plants. The categorization of eventreports into four pipe size classes in Table 2 follows the convention used in the SKIReport 96:203.

Table 2: Comparison of the Database Contents in SKI's PSA-Oriented Database andSKI Report 96:20V-2J.

Pipe Size

DN<2525<DN< 100100<DN<300

> DN300Unknown / Assumed Size(a)

Total:

SKI Report 97:26

[Number of Records]

958(41%)516(22%)441 (19%)167 (7%)

249(11%)

2331

SKI Report 96:20[Number of Records]

574 (38%)252(17%)155(10%)74 (5%)

456 (30%)

1511

Note: (a). Failure report contains no explicit information on diameter.

3. THE REPORTING OF PIPE FAILURES

The piping systems in nuclear power plants are designed to high standards, and majorfailures are rare events. The rare failures have a low frequency of occurrence (e.g.,less than, or much less than one failure per plant and year). Not only are the majorfailures rare events when viewed against a frequency-scale, they are also rare whenviewed against a passive component 'population-scale'. Each nuclear power plantcontains a very large volume of piping components (e.g., many thousands of welds,and several km of piping). Therefore, and for any given plant, the ratio of majorfailures by the total piping component population is a small number ( « 0.1). Mostpiping failure incidents are incipient or degraded failures with minor or no immediateimpact on plant operation and safety. The incipient or degraded failures have arelatively high frequency of occurrence; e.g., equal to or greater than one event perplant and year.

While the volume of technical information on operating experience withpiping systems is considerable, the quality of this information varies immensely.Some reports present detailed root cause analysis insights and result, while themajority of the reports contains cursory (and sometimes conflicting) information onthe causes and consequences. The determination of root cause involves interpretationof results from visual examinations and, sometimes, detailed metallurgical evaluationsof damaged or fractured piping components. In general, failure and reliabilityanalysis of incidents involving piping systems are complex and uncertain.

For the work documented in SKI Report 97:26, the main source of informationon piping failures was licensee event reports (LERs). The LERs are mainly prepared

3 Bush, S.H. et al, 1996. Piping Failures in Unites Stated Nuclear Power Plants: 1961-1995. SKIReport 96:20.


upon failure conditions, which place the plant operations outside the technicalspecifications. Rather than evaluations of the root causes, these reports concentrate onthe apparent causes of failure. Uniform regulatory reporting requirements do not yetexist, and no industry standards have been developed for the reporting anddissemination of information on piping failures. This lack of detailed reportingprotocols reflects the complex nature of piping reliability.

In the opinion of the authors of SKI Report 97:26, the lack of consistentreporting follows on not having a recognized model for analyzing piping reliability.Substantial interpretation of the available failure information is needed to determinethe where-why-how a particular piping system failed. The interpretation should reflectthe purpose of an analysis and the database design. It is not uncommon that thefailure reports include detailed narratives of the circumstances of a given event (e.g.,plant status and plant response). Reporting of the specifics of a piping failure (e.g.,exact description of fault location, mode of failure, type and diameter of the failedpiping component, trends and failure patterns) is beyond the scope of most LERsystems, however. Therefore, and accurate and consistent failure classification oftenrequires an 'interrogation' of several, independent information sources.

4. REPORTING PRACTICES

Typically, piping failures are reported as 'cracks/crack', 'indications', 'leaks' or'ruptures', corresponding to incipient, degraded and complete failure, respectively; c.f.Figure 1. In SKI Report 97:26, a 'rupture' was interpreted as a catastrophic loss ofmechanical integrity which occurs without advance warning. Ruptures potentiallyresult in large leak rates > 5 kg/s (80 gpm).

Piping System Incident(Crack or Series of Cracks inOne Heat Affected Zone or in

One Location of the Base-Metal)

T.S. = Technical SpecificationTO = Time at which 'degradation' beginsTW = Through-wall

"I TWC = Through-wall crackI

Incipient Failure(Crack Indication; < TW or

Full TWC Resulting in 'Seepage')

Degraded Failure(Detectable Leak; Within or In

Excess of T.S. Limitations)

Complete Failure(Severance, Break, Large

Leak » T.S. Limit)

(Potential 'latency'; NDE/ISI fails to detect an indication^Rupture

Leak Rate > 5 kg/s,No Advance Warning

TO + t TO + xt TO + yt

Figure 1: Piping Failure Mode Definitions - Except for Rupture, the DefinitionsAssume Leak-Before-Break (i.e., Non IGSCC-Susceptible Piping).

SEMINAR ON PIPING RELIABILITYSigtuna (Sweden)September 30 - October 1,1997

The classification of events and analysis of data should build on consistentapplication of clear definitions of failure. In the context of PSA, inadvertent orimproper classification of a piping failure event as rupture could result in significantover-estimation of the true rupture failure rate or failure probability. From the pointof reliability parameter estimation, there are several inherent limitations of LERs. Bydesign, LERs document the effects of failure on system and safety functions. They donot go into the details about the specific degradation or failure mechanisms,contributing causes, and required repair actions.

Functional and structural interpretations of the potential or actualconsequences of a given failure determine whether a formal, written report is preparedby a licensee for internal use or dispositioning with a regulatory agency. As anexample, the probable consequences of small cracks due to stress corrosion crackingin piping within the Reactor Coolant Pressure Boundary (RCPB) are crackpropagation in the through-wall direction and minor leakage of primary coolant.When small but detectable leaks occur, leakage monitoring systems detect the changeof leak rate, and a plant shutdown is required if allowable leak rate limits areexceeded. Such events are reportable according to technical specification reportingrequirements. These reporting requirements do not cover degradations or failures insteam piping or feedwater piping that are not part of the RCPB, however.Furthermore, the reporting of piping failures is a function of the approach toreplacement of degraded piping. The replacement of degraded piping prior todeveloping a gross leakage would normally not be a reportable event. With theexception for significant degradations and complete failures occurring within theRCPB, ad hoc reporting of piping failures is the norm rather than the exception.

These observations would be not be of concern to PSA practitioners, were itnot for the fact that piping failures are rare events. The believable reliabilityestimation based on the operational data requires full consideration of the entire bodyof operating experience, and a consistent interpretation of the diverse failureinformation. There needs to be assurance about the completeness and relevance of theoperational data to be considered in piping reliability analysis.

A range of different reporting criteria is in current use. These criteria mostlyfollow structural reliability considerations and RCPB leak rate criteria as defined bythe technical specifications for plant operation, and applicable piping codes andstandards. The piping codes define minimum requirements for design, materials,fabrication, installation, test and inspection. The standards contain design andconstruction rules and requirements for individual piping components such as elbows,tees, flanges and other in-line items. Compliance to Code is mandated by regulationsimposed by regulatory agencies. The codes and standards encompass consideration ofmetallurgical degradation mechanisms.

The purpose of NDE is to determine the suitability for continued use of agiven piping system after a predetermined in-service time. Depending on the extentof degradation, the findings of NDE could result in formal or informal reporting toregulatory agencies. Some examples of typical NDE-based reporting criteria aresummarized in Table 3. While there are regional differences among the criteria, most


of them are adaptations of the ASME BPVC Section XI and the applicable AmericanNational Standards Institute (ANSI) standards. In Sweden, SKIFS 1994:1 documentsregulatory requirements for the mechanical integrity of piping system components.

Table 3: Examples of NDE-Based Reporting Criteria.

ISI Acceptance Standards & Reporting Criteria - Some Examples

• Formal dispositioning with regulatory agency for pipe wall thickness < 50% ofnominal wall thickness (NWT).

• Increased inspection frequency for wall thickness < 75% NWT; discretionaryreporting may be acceptable.

• Using radiography, any elongated indication with a length greater than 1/3 T forT (= thickness of weld being examined) from 6 mm to 57 mm inclusive isunacceptable.

In addition to the structural reliability considerations, functional requirements(e.g., acceptable leak rates) also determine the reporting of piping failures. Thedefinition of failure criteria based on leak rates is difficult and must, as a minimum,acknowledge the design criteria as defined in Final Safety Analysis Reports; e.g., leakdetection capability and reliability, and make-up capacity of engineered safetysystems. The majority of documents surveyed during the database development anddata collection did not include explicit leak rate or leak duration information.

A large portion of reported incipient and degraded failures within the RCPBare detected by inservice inspection (ISI) during annual refueling and maintenanceoutages. Relaxations in the plant technical specifications (TS) and reportingrequirements during outages result in discretionary reporting of the ISI-findings,however. This means that while formal licensee event reports (LERs) would not befiled based on the NDE/IS findings, other means of reporting could be prepared aspart of summaries of the performance of outage activities (i.e., outage inspectionreports). If a 'significant' ISI-finding by one licensee is believed to have potentialgeneric, industry-wide implications, then that finding would be reported and result informal dispositioning. Not only would the 'discovering' licensee provide a report, butalso the other licensees, which are affected by the original ISI results. The NDE-basedreporting criteria are interpreted and implemented on a case-by-case basis, and a lackof functional considerations could impose restrictions on the dissemination of reportswithin and outside an organization. Examples of reporting practices include:

Significant incipient or degraded failures discovered during refueling orextended maintenance outages normally are reported to regulatory agencies.

Some degraded failures during routine power operation are reported;especially those with assumed generic implications.

Most degraded failures within the RCPB are reported, especially where thereis an external leakage which is detected by the leak detection system(s). Thereporting is almost guaranteed whenever the plant-specific TS define leak ratecriteria with limiting conditions for operation (LCO).


There are many exceptions to the above practices, however. As an example, toeffect repairs, a RCPB leak could result in a planned shutdown of the unit.While progressing with the manual shutdown, an equipment failure occurswhich is unrelated to the leak but possibly triggered by the change of plantstatus and causes an automatic reactor trip, say, from 50% power. In this casea LER may be filed for the equipment failure which caused the trip directly,but none filed for the piping failure. Therefore, a search for failure data onpiping often must include more than one information source.

Complete failures (e.g., ruptures) which result in manual or automatic reactortrip are reported most of the time, especially if they occur within the RCPB.Discretionary reporting applies to failures outside the RCPB.

There is no all-encompassing definition of piping failure modes. Differentinterpretations based on functional and/or structural interpretations lead toinconsistent reporting of failures, and complicates data analysis. Insights from thedata collection effort in this project seem to imply that ruptures and major leaks arereported at all time, while the incipient and degraded failures (e.g., leaks near or wellbelow the TS limitations) at best are reported on an ad hoc basis.

In simple terms, a rupture is a major loss of mechanical integrity withoutadvance warning. Using a functional definition, a rupture is a piping failure whichcauses a loss of coolant (or process medium) inventory in excess of the make-upcapability of an engineered safety system (or non-safety-related make-up system). Thedifferent interpretations of failure potentially influence the formal reporting of eventsinvolving piping degradations.

The reliability of reactor pressure vessels and primary system piping is animportant topic for nuclear safety R&D as well as plant operations. The earliestnuclear safety debates kept addressing this complex reliability issue; sometimes in anhighly unbalanced way. With this debate followed a 'sensitized' awareness about thepotential implications of including too detailed accounts of the evaluations of resultsfrom NDE/ISI in the licensee event reports. Non-stringent use of technical termscould be misinterpreted. The historical developments within the nuclear safety haveinfluenced the way piping failures are documented and reported today.

Since piping reliability and reporting of failures are so difficult, is there a wayof determining the coverage and completeness of failure reports? A philosophyadopted in this project is the notion that piping failures of varying severity haveoccurred at each operating plant worldwide. Failure reports qualified for entry into thedatabase came from the plants subjected to a detailed survey of its operating history.Plants not yet included in the database are the those for which operational data wereunavailable, or for which reports on occurred failures had not yet been dispositioned.In developing the SLAP database the emphasis of the detailed surveys of operationaldata was on Swedish and U.S. plants. Based on the plant population and operationaldata for these two countries, the annual frequency of a piping degradation is on theorder of 0.5 event per year, which should be compared with the following publishedestimates:


According to Rodabaugh (1985)4, a "reasonable pipe failure rate" is about 1event per year and plant;

Recent German information on degradations and failures in reactor andfeedwater-condensate piping systems indicates a failure rate of about 0.2 eventper year; c.f. Reck and Bieniussa (1995)5.

5. A PLAN FOR FURTHER WORK

Many operating nuclear power plants are undergoing renovation and modernization aspart of plant life extension projects. In some cases, the renovation activities aredirected at improving the primary system piping reliability by incorporating detailedconsiderations of the current state-of-knowledge about degradation and failuremechanisms and structural reliability. Increasingly, PSA applications are performed(or are being considered) to evaluate the effects the modified primary system pipingdesigns have on plant risk. Also, PSA applications are performed to support thedefinition of enhanced strategies for in-service (ISI) objectives or targets. With theelevated expectations on the PSA technology follows an urgent need for acomprehensive database on pipe failures.

The SKI project has demonstrated the viability of database development. It isa highly resource intensive activity. Further work is required to improve the databasecoverage and completeness. // is strongly recommended that future efforts to improvethe database should be pursued within the international cooperative nuclear safetyR&D. Specifically, the recommendations are:

(1) Taxonomy on pipe failure modes. To improve the quality of the reporting, aneffort should be initiated to develop detailed reporting guidelines.

(2) Database coverage and completeness. The current event-based, relationaldatabase should be further enhanced by surveying additional sources ofexperience data. A philosophy for dissemination of experience data toqualified PSA practitioners should be developed.

4 Rodabaugh, E.C., 1985. Comments on the Leak-Before-Break Concept for Nuclear Power PlantPiping Systems, ORNL/Sub/82-22252/3 (NUREG/CR-4305), Oak Ridge National Laboratory, OakRidge (TN),pp 10-12.

5 Reck, H. and K.W. Bienussa, 1995. "Auswertung von Betriebserfahrungen Teil 1: Schaden an DWR-und SWR- Rohrleitungen der J- und K-System," GRS Fachseminar Ermittlung der Haufigkeiten vonLeeks und Briihchen in druckfuhrenden Systemen fur probabilistische Sicherheitsanalysen, Koln, 18-20September, 1995.


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SKI's Seminar onPiping reliability

Stocholm,Sept.30-Octl.1997

International data bases onpiping failures

Bengt LydellBojan Tomic

SKI Seminar on Piping Reliability (1997) 2 1

Background

SLAP is not the first project which attempt to collect the data from operationalexperience internationally and try to estimate the piping reliability using operationalexperience data. From the times before WASH 1400 (1975) there were attempts tocollect the information relevant for piping reliability from nuclear plants as well as fromnon-nuclear installations.This presentation summarize the main data collection projects which were conducedinternationally and present their scope and extent, discuss some of the methods suedand present the final results reached. The summary covers work performed in the period1964-1996.

The reader shall recognize the fact that no recognized, validated data collection of pipefailure events exist. While the information on piping failures may be collected atnuclear plants (or indeed other industries), no data collection exercise managed to get acomprehensive population data and a comprehensive failure count to enable formalstatistical analysis. The data collection exercises described below used various sourcesof data, and different assumptions to generate the reliability parameters.

1. Data Collections Prior to WASH-1400

Preceding WASH-1400 by about ten years, General Electric (GE), under contract withthe U.S. Atomic Energy Commission, performed the 'Reactor Primary Coolant SystemRupture Study' (GEAP-4574)"1. This study surveyed available experience with steamplant piping and provided frequencies for the failure modes 'leaks' and 'severance'taking into consideration the impact of ultrasonic testing (UT) on overall reliability;Table 1.

Table 1: Early Pipe Failure Rate Estimates.

FAILURE MODE

"General Failure" - LeakageSeverance

Severe Service Failure

Leakage - Without UTSeverance - Without UT

Leakage - With UTSeverance - With UT

LeakageSeverance

FAILURE RATE[Events/Plant Year]

4.4E-021.9E-034.4E-04

2.6E-014.0E-031.3E-011.5E-03

6.8E-011.5E-02

SOURCE

GEAP-4527 (1964); Conventional utilityindustry steam piping. About 9000plant-years of experience.

GEAP-10207-23 (1970); Steam pipingin conventional power plant and NPP.

ORNL-TM-3425 (1970); Oak RidgeNational Laboratory. Review of NPPexperience (75 reactor-years) withinterpretations and additional analysis byHolt[2i. High failure rates attributed tohuman error/ design error/constructionerror resulting in severe loadingconditions. According to the report byOak Ridge, about 57% of all LERsattributed to human error.

SKI Seminar on Piping R e l i a b i l i t y (1997) 12

2 Pipe Failure Data in WASH-1400

At the time of publication of the RSS in 1975 only about 150 reactor years of U.S.operating experience had been accumulated, and limited experience data were availablefor estimating pipe break frequencies. As part of the WASH-1400 effort a limitedevaluation of nuclear pipe reliability, based on actual failures in nuclear systems relatedto the operating period of nuclear systems. The database included 11 'significant'events. The emphasis was on the derivation of order-of-magnitude LOCA frequenciesfor input to event tree analysis (Table 2) and pipe failure rates for input to system faulttrees (Table 3). WASH-1400 examined several different sources to obtain failure ratesfor small-diameter and large-diameter pipe. The reason for using several data sourceswas the interest in pipe ruptures (complete pipe severances) resulting in reactor coolantloss, and none had occurred in the 150 U.S. NPP operating years considered by thestudy. Therefore, other pipe failure data sources were sought for extrapolating pipefailure rates for use in the RSS.

Table 2: LOCA Frequencies in WASH-1400.

LOCA CLASS

Small

Medium

Large

INITIATING EVENT FREQUENCY[I/Year]

Median

l.OE-3

3.0E-4

l.OE-4

Range (90%)

l.OE-4- 1.0E-2

3.0E-5 - 3.OE-3

1.0E-5- l.OE-3

Table 3: Pipe Failure Rates in WASH-1400.

PIPE SIZE[DN, mm]

<75

>75

FAILURE RATE, RUPTURE [1/hr.m]

Median

l.OE-9

l.OE-10

Range (90%)

3.3E-1O-3.OE-8

3.3E-1O-3.OE-9

Several different means of extrapolating the data were devised because the data weregiven in different forms. Details such as leak rates, pipe diameter, cause of failure,system in which the failure occurred, and other pertinent information were not supplied.As a result, weighting factors based on 'average plant characteristics' were used to

relate total plant piping to LOCA-sensitive piping and to large- and small-diameterpiping and complete severance to large pipe. LOCA-sensitive piping was defined as:

SKI Seminar on Piping Reliability (1997) 13

LOCA-sensitive piping; 10% of total piping in the reported data base.LOCA-sensitive small piping (< DN 100); 4.7% of total piping in the reporteddata base, 10% of small piping.LOCA-sensitive large piping (> DN 100); 5.3% of total piping in the reporteddata base, 10% of large piping.

3 Pipe Failure Data by PNL (1976)

After the publication of WASH-1400 in 1975, Battelle Pacific Northwest Laboratories(PNL) performed an assessment of piping reliability based on available U.S. LWRoperating experience and non-nuclear operating experience13'41; Table 4. Differencesbetween WASH-1400 and PNL results were due differences in the interpretation oflimited failure data. The study by PNL addressed the role of periodic inspection, andaddressed failures due to intergranular stress-corrosion cracking (IGSCC). Among theconclusions by PNL were:

• The failure probabilities for larger sizes of nuclear piping were considered to bein the range of 1.0E-4 to 1.0E-6 per reactor year (exclusive of IGSCC).

• Smaller pipe sizes, of lesser safety significance, have much higher failure rates.

• In BWRs, IGSCC can cause failure rates much higher than 1 .OE-4 (e.g., 1 .OE-2)in piping DN 100 to DN 250.

• Catastrophic failures would appear more likely from operator error or designand construction errors (water hammer, improper handling of dynamic loads,undetected fabrication defects) rather than conventional flaw initiation andgrowth by fatigue.

Table 4: Pipe Failure Rates in PNL Study (DN > 100).

Note:study, afeet and

FAILURE RATE [1/hr.m]

BWR

4.9E-9

The failure rates are in terms of failures per mBWR contains 94,500 m of piping, and a PWR280,000 feet, respectively.

of pipingcontains

PWR

5.3E-9

. According to the84,000 m of piping

PNL-; 317,000


4 Pipe Failure Data by AECL (1981)

The Atomic Energy of Canada Limited (AECL) performed a study[5] of U.S. LWRpiping operating experience for the period 1959 through 1978, representing 409 reactor-years of experience. The study was initiated in support of the analysis of theconsequences of pipe rupture in the Primary Heat Transport System (PHTS) forCANDU power stations. Another objective was to establish whether the additionaloperating experience that had accumulated since publication of WASH-1400 warrantednew pipe failure rates to be used in PSA applications.

The pipe failure events were classified according to: (i) severance, (ii) leak, and (ii)defect. Of the total of 840 failure events considered by the study, 87 pipe failures wereinterpreted to be severances (8 events in small-diameter primary system piping). Table5 summarizes failure rate estimates for primary system pipe severances. Statisticalanalysis was limited to estimation of confidence limits for failure rates using the Chi-square distribution. Because of uncertainties in the pipe failure event data base andassumptions in interpretation of the data, the order-of-magnitude failure rate estimatesby WASH-1400 were viewed by the AECL study as representative of 'true' failurerates.

Table 5: Pipe Failure Rates in AECL Study (1981).

PIPE SIZE

DN<25

25 < DN < 150

DN>150

FAILURE RATE, RUPTUREUPPER LIMIT AT 95% CONFIDENCE [1/hr.plant]

4.4E-6

8.3E-7

8.3E-7

5. Pipe Failure Data by Thomas (1981)

In 1981 H.M. Thomas of Rolls Royce & Associates Ltd. published a modeling systemfor interpretation of pipe failure data (usually referred to as the 'Thomas model'), andfor 'adjusting' generic industry data to plant-specific data'61. Among reliabilityinfluence factors acknowledged in updating generic data were: design learning curve,pipe diameter, plant age, fracture toughness, pipe length, number of cycles, parentmaterial versus weld material, fatigue stress, crack dimensions, and wall thickness. Thepaper by Thomas included no experience data, however. Thomas made the followingstatement on the subject of pipe length:

"... It is known that a typical [nuclear power] plant contains about 16,500 feetof pipe less than 4 inch diameter and about 18,500 feet of pipe greater than 4inch diameter, making a total of 35,000 feet..."

SKI Seminar on Piping Reliability (1997)

Thomas references WASH-1400, Appendix IE. There is discrepancy between WASH-1400 and the Thomas paper, however. Let us speculate how the information on pipelength was developed. Some insights can be gleaned by assuming that Thomas arrivedat a number of 350,000 feet being the total length of piping in a typical nuclear powerplant. By multiplying this length by 4.7% and 5.3%, respectively, we would (consistentwith WASH-1400) get the total lenght of small-diameter, LOCA-sensitive piping andlarge-diameter, LOCA sensitive piping, respectively; i.e., together about 35,000 feet ofpipe. It is feasible that Thomas was influenced by the paper of Spencer Bush publishedin 1976P1 in which a typical BWR is stated as having 315,000 feet of (LOCA-insensitive) piping. Under the set of assumptions there would be consistency betweenThomas and Bush; i.e., 315,000 + 35,000 = 350,000 feet).

6. Pipe Failure Data by Ris0 (1982)

Within the framework of the SAK-1 (Probabilistic Risk Assessment and Licensing)project sponsored by the Nordic Liaison Committee for Atomic Energy (NKA), Ris0performed the 'Pipe Failure Study'[7>8]. Derived failure rates were based on Swedishand Finnish nuclear plant operating experience for the period 1975-1981, correspondingto 43 reactor-years. A total of 73 events were recorded in Swedish plants for the studyperiod, of which 12 events represented 'breaks and rupture' in small-diameter piping.Repair times were not specified in 41 of the 73 incident reports. Based on the availableinformation, the mean repair time was about 15 hours with an observed maximum ofabout 150 hours. A summary of the derived pipe failure rates is given in Table 6.Unfortunately, the raw data assembled by the study were not been retained for futureuse, and no re-validation of the information has been possible.

Table 6: Pipe Failure Rates in Ris0 Study.

PIPE RUPTURE SIZE

Small

Medium

Large

FAILURE RATE, RUPTURE(90% Range) [1/hr]

Water Pipe

6.63E-5 - 1.17E-4

8.68E-6 - 3.15E-5

9.13E-7 - 1.26E-5

Steam Pipe

6.96E-6 - 2.79E-5

9.13E-7 - 1.26E-5

< 6.16E-6

7 Pipe Failure Rates by AECL (1984)

As a continuation of the study by AECL in 1981 (Section 4), an assessment of thepiping system component reliability in CANDU plants was published in 1984[9].Failure event data from Pickering-A and Bruce-A for the period 1971-1981 wasanalyzed using an approach similar to AECL (1981) study. A total of 158 failureevents were recorded for the study period. Of these, 6 events were pipe severances intotal plant. Only one primary system severance was reported.


Table 7: Pipe Failure Rates in AECL Study (1981).

PIPE SIZE

DN<25

25 < DN < 150

DN>150

FAILURE RATE, RUPTUREUPPER LIMIT AT 95% CONFIDENCE [1/hr.plant]

1.2E-5

6.4E-6

6.4E-6

8 Pipe Failure Data by EG&G Idaho, Inc. (1987)

Objective of the EG&G-study[101 was to update the failure rate estimates of WASH-1400 by utilizing the accumulated U.S. nuclear operating experience available as ofDecember 1984. About 800 reactor years of operation were considered. Derived LOCAfrequencies and pipe failure rates are shown in Tables 8 and 9, respectively. Relative toWASH-1400 an additional 650 reactor years were accounted for to improve theuncertainties of the pipe failure rates. Whereas RSS accounted for a total of eleven (11)significant pipe failures, the EG&G-study identified twenty (20) 'significant' pipefailure events for input to failure rate estimation.

Table 8: LOCA Frequencies in EGG-2421.

LOCA CLASS

Leak rate > 3 kg/s

INITIATING EVENT FREQUENCY [I/Year]

Median

3.0E-4

Range (90%)

0 - 3.8E-3

Table 9: Pipe Failure Rates of Non-LOCA-Sensitive and LOCA-Sensitive Piping.

PIPE RUPTURE SIZE[mm]

BWR12-50

50- 150> 150

PWR12-50

50- 150> 150

FAILURE RATE [1/hr]

5th

3.0E-71.3E-77.3E-7

8.0E-83.2E-71.9E-7

Median

1.1E-67.3E-71.8E-6

4.7-79.5E-77.1E-7

95th

2.8E-62.3E-63.8E-6

1.5E-62.2E-61.8E-6


For LOCA-sensitive piping, failure was defined as a leak rate of at least 3 kg/s forPWRs and 30 kg/s for BWRs. These rates are the normal reactor coolant makeupsystem capacity for each plant type. For non-LOCA-sensitive systems, several factorswere considered in the definition of failure. One factor considered was whether onecould determine the leak rate necessary to disable a system from performing itsintended function. Since the leak rate value is system and location dependent, the datawere instead placed in two discrete categories (> 0.06 kg/s and > 1 kg/s). These leakrate categories were selected because the few actual known leak rates reported occurredroughly in the range 0.06 - 1 kg/s.

9 Pipe Failure Data by GRS (1987)

In support of the Phase B of the German Risk Study, GRS sponsored R&D on pipingreliability1"1. This R&D was sponsored in recognition of the significant limitations ofthe available pipe reliability estimation approaches, and the significant limitations in theapproaches to LOCA frequency estimation practiced in PSA projects. GRS elected toapply two general analysis approaches: (i) statistical evaluation of operatingexperience, and (ii) probabilistic fracture mechanics studies. The former approach wasapplied to small-diameter piping for which failure experience existed, while the latterapproach supported analysis of piping for which some experimental data existedtogether with insights from the German NDE experience. Table 10 summarizes pipefailure probabilities by GRS. Reliability influencing factors were recognized in thework. According to GRS:

• The worldwide operating experience with LWRs is of limited use as a datasource. Observed failure mechanisms are partly design dependent. Problemswith pooling of data.

• The available operating experience with German NPPs showed only a smallnumber of leakage events. Therefore the statistical uncertainty bands wereconsiderable.

Rather than using equivalent leakage/rupture sizes, pipe failure data were estimated forthree categories of piping: (i) < DN 25, (ii) > DN 25 - < DN 250, and (iii) > DN 250.Statistical analysis of operating experience was used for < DN 25, while probabilisticfracture mechanics studies were used for the large nominal diameters; > DN 250. Forthe range > DN 25 to < DN 250 insights from operating experience was applied in aqualitative sense together with experimental data and LBB-reasoning.


Table 10: Failure Probabilities ofPWR Piping Inside Containment.

PIPE FAILURE CLASS[Break Size]

DN25DN50DN80

DN 100DN 150

DN 250(a)

> DN 300(a)

Break (DEGB) - > DN 250(b)

Leakage - > DN 250<b)

FAILURE PROBABILITY[Mean]

1.7E-O31.7E-045.7E-059.6E-061.4E-05

<1.0E-07< 1.0E-07

< 1.2E-10< 2.0E-07

Notes: (a). Evaluated using probabilistic fracture mechanics. Stated value interpreted asupper bound,(b). FromNUREG/CR-3660-VI[12]. Stated values are the upper bounds. The DEGB isinduced by fatigue crack growth. The leakage is assumed to result from a through-wall crack.

10 Pipe Failure Data by EG&G Idaho, Inc. (1991)

Building on earlier work (Section 8) EG&G Idaho, Inc., under contract with the U.S.Department of Energy, developed an updated study on leakage and rupture events forpiping and piping components such as valves, flanges, fittings'131. This new data sourcewas developed to support internal flooding risk analyses; Table 11. Licensee EventReports (LERs) contained in Nuclear Power Experience (NPE) were searched forleakage and rupture events. Extracted failure reports covered the period 1960-1990.Some of the qualitative insights from the data analysis were:

• There appeared to be no significant difference in external leakage or rupturefrequencies between piping with diameters < DN 75 and larger piping.

• There appeared to be no significant difference between PWR and BWRcomponent external leakage and rupture frequencies.

• It was possible to distinguish between external rupture frequencies forcomponents in primary coolant systems and external rupture frequencies forcomponents in other systems.

• External rupture frequencies were found to generally be factors 25 or 100 timeslower than external leakage frequencies and are dependent on the type ofcomponent and whether the component is in the primary coolant systems.

Based on derived leakage frequencies a rupture frequency was estimated by firstcalculating the conditional rupture probability given failure. For piping the externalrupture probability given that an external leakage or rupture has occurred was given as0.04 for non-PCS piping and 0.01 for PCS piping. Table 12 summarizes the pipefailure rates.


Table 11: Piping Component Failure Probabilities in EG&G-Study (1991).

PIPING SYSTEMCOMPONENT

Non-RCS (b):Piping (including elbows)Valve, pump, heat exchanger, tankFlange

RCS<C):Piping (including elbows)Valve, pump, flange, heat exchanger, tank

Notes: (a). Conditional (given an external leakage or rupturederiving the probabilities, the ratio of external ruptureevents was determined. Leakage rate > 3 kg/s.(b). Non primary system components.(c). Primary system components.

RUPTURE PROBABILITY1"1

[Mean]

3.3E-O35.2E-021.0E-02

8.OE-O39.0E-03

event) mean rupture probability. Inevents to external leakage and rupture

Table

Note:

12: Pipe Failure Rates in EG&G Study (1991).

PIPE FAILURE MODE

Leakage (PCS & Non-PCS)(1)

Rupture (PCS)

Rupture (Non-PCS)

MEAN FAILURE RATE[1/hr.m]

1.0E-08

l.OE-10

4.0E-10

(1). Leakage defined as < 3 1/s / Rupture defined as > 3 1/s or complete severance.

11. Pipe Failue Data by EPRI (1990-1993)

Originally undertaken for Northeast Utilities Service Company"41, and later co-sponsored by EPRI, Jamali1151 developed a methodology and data base for pipe failurerate estimates. A first report documenting results was published in 1990. To allow forwider access, EPRI later published enhanced and updated versions of this report in1992 and 1993, respectively116'171. The EPRI-studies were undertaken to provide a U.S.nuclear plant pipe failure data base reflecting the additional experience generated sinceWASH-1400 was published. The principal sources of pipe failure information wereLicensee Event Reports (LERs), Nuclear Power Experience (NPE) , and the NuclearPlant Reliability Data System (NPRD) operated by the Institute of Nuclear PowerOperations (INPO). Table 13 summarizes pipe failure rates presented in the interim(1992) EPRI-study. A final report published in 1993 included an updated list of piperupture events; 105 events in 1993 versus 41 events in 1992. EPRI adopted theWASH-1400 definition of pipe section; i.e., a segment of piping between majordiscontinuities such as valves, pumps, reducers, etc. Pipe section counts are providedfor typical U.S. BWRs and PWRs, and these counts are consistent with WASH-1400.


Table 13: Pipe Failure Rates in EPRI-Study (1992).

PIPE SIZE - INNER DIAMETER (ID)[mm]

12 < ID < 50

50 < ID < 75

75 < ID < 150

ID > 150

FAILURE RATE [1/hr.section]

EPRI

6.0E-10

3.0E-10

3.0E-10

7.0E-10

WASH-1400

3.6E-9

3.6E-9

3.6E-10

3.6E-10

12 Pipe Failure Data by Bush & Chockie (1996)

In 1995-96, Bush & Chocke[18] surveyed U.S. operating experience with pipingsystems. The work was sponsored by SKI (Dept. RH) and focused on counts of failureevents (leaks, severances and ruptures). Table 14 compares the data by Bush &Chockie and the SLAP database.

Table 14: Comparison of the Database Contents in 'SLAP' and SKI Report 96:20t'51.

Pipe Size

DN<2525<DN<100100 < DN < 300

>DN300Unknown / Assumed Size'"'

Total:

SLAP Version 7.5{Number of Records]

958(41%)516(22%)441 (19%)167 (7%)249(11%)

2331

SKI Report 96:20[Number of Records]

574 (38%)252(17%)155(10%)74 (5%)

456 (30%)

1511

Note: (a). Failure report contains no explicit information on diameter.

13. Summary

Several attempts have been made to collect data on pipe failures in NPPs. Mostly, thescope of these data collections has been limited to counts of significant failures.Meaningful parameter estimation requires detailed event analysis to identify pertinentreliability attributes and influence factors, however. A data collection 'infrastructure'such as the SLAP project has been demonstrated as a promising alternative.


14. References

[1]. Gibbons, W.S. and B.D. Hackney, 1964. Survey of Piping Failures for theReactor Primary Coolant Pipe Rupture Study, GEAP-4574, Atomic Power EquipmentDepartment, General Electric Company, San Jose (CA).

[2]. Holt, A.B., 1974. "The Probability of Catastrophic Failure of Reactor PrimarySystem Components," Nuclear Engineering and Design, 28:239-251.

[3]. Bush, S.H., 1976. "Reliability of Piping in Light-Water Reactors," NuclearSafety, 17:568-579.

[4]. Bush, S.H., 1985. "Statistics of Pressure Vessel and Piping Failures," inSundararajan, C. (Editor): Pressure Vessel and Piping Technology 1985. A Decade ofProgress. The American Society of Mechanical Engineers, New York (NY), pp 875-893.

[5]. Janzen, P., 1981. A Study of Piping Failures in U.S. Nuclear Power Reactors,AECL-Misc-204, Atomic Energy of Canada Limited, Special Projects Division, ChalkRiver Nuclear Laboratories, Chalk River (Canada).

[6]. Thomas, H.M., 1981. "Pipe and Vessel Failure Probability," ReliabilityEngineering, 2:83-124.

[7]. Petersen, K.E., 1982. "Pipe Failure Study," Probabilistic Risk Analysis andLicensing, NKA/SAK-1-D(82)9 (Ris0-M-2363), Proceedings of Seminar 2, Helsing0r(Denmark), March 29-31, pp 129-149.

[8]. Petersen, K.E., 1983. "Analysis of Pipe Failures in Swedish Nuclear Plants,"Proceedings of the 4th EuReDatA Conference, Venice (Italy), March 23-25.

[9]. Janzen, P., 1984. Piping Performance in Canadian CANDU NGS, AECL-Misc-252, Atomic Energy of Canada Limited, Special Projects Division, Chalk River NuclearLaboratories, Chalk River (Canada).

[10]. Wright, R.E., J.A. Steverson and W.F. Zuroff, 1987. Pipe Break FrequencyEstimation for Nuclear Power Plants, EGG-2421 (NUREG/CR-4407), Idaho NationalEngineering Laboratory, Inc., Idaho Falls (ID).

[11]. Beliczey, S. and H. Schulz, 1987. "The Probability of Leakage in PipingSystems of Pressurized Water Reactors on the Basis of Fracture Mechanics andOperating Experience," Nuclear Engineering and Design, 102:431-438.

[12]. Holman, G.S. and C.K. Chou, 1985. Probability of Pipe Failure in the ReactorCoolant Loops of Westinghouse PWR Plant. Volume 1: Summary Report, UCJJD-19988(NUREG/CR-3660-VI), Lawrence Livermore National Laboratory, Livermore (CA).


[13]. Eide, S.A. et al, 1991. Component External Leakage and Rupture FrequencyEstimates, EGG-SSRE-9639 (DE92 012357), INEL, Idaho Falls (ID).

[14]. Northeast Utilities was lead participant responsible for construction andoperation of Millstone-1 (GE-BWR), Millstone-2 (ABB-CE-PER), and Millstone-3(WE-PWR). The three Millstone units are located in Connecticut (USA) on the LongIsland Sound.

[15]. Jamali, K., 1990. A Study of Pipe Failures in U.S. Commercial Nuclear PowerPlants, Halliburton NUS Corporation, Gaithersburg (MD).

[16]. Jamali, K., 1992. Pipe Failures in U.S. Commercial Nuclear Power Plants,EPRI TR-100380 (Interim Report), Electric Power Research Institute, Palo Alto (CA).

[17]. Jamali, K. and J.-P. Sursock, 1993. Pipe Failures in U.S. Commercial NuclearPower Plants, EPRI TR-100380, Electric Power Research Institute, Palo Alto (CA).

[18]. Bush, S.H. et al, 1996. Pipe Failures in United States Nuclear Power Plants:1961-1995, SKI Report 96:20, Swedish Nuclear Power Inspectorate, Stockholm.


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