Seismic reevaluation and upgrading of nuclear power facilities outside the US using US developed...

Nuclear Engineering and Design 181 (1998) 115–129

Seismic reevaluation and upgrading of nuclear power facilitiesoutside the US using US developed methodologies

R. Campbell a,*, S. Short a, M. Ravindra a, G. Hardy a, J. Johnson b

a EQE International Incorporated, 18101 Von Karman 400, Ir6ine, CA 92715, USAb EQE International Incorporated, San Francisco, California, USA

Abstract

Several seismic licensing and safety issues have emerged over the past fifteen years for commercial US NuclearPower Plants and US Government research reactors, production reactors and process facilities. The methodologiesfor the resolution of these issues have been developed in numerous government and utility sponsored researchprograms. The resolution criteria have included conservative deterministic design criteria, deterministic seismicmargins assessments criteria (SMA) and seismic probabilistic risk assessment criteria (SPRA). The criteria for SMAsand SPRAs have been based realistically on considering the inelastic energy absorption capability of ductilestructures, equipment and piping and have incorporated the use of earthquake and testing experience to evaluate theoperability of complex mechanical and electrical equipment. Most of the applications to date have been confined tothe US, however there have been several applications to Asian, Western and Eastern Europe reactors. This papersummarizes the major issues addressed, the development of reevaluation criteria and selected applications to non USreactors including VVER reactors of Soviet origin. © 1998 Elsevier Science S.A. All rights reserved.

1. Introduction

Earlier nuclear power plants (NPPs) con-structed throughout the world did not have therigorous seismic design basis that is applied tocurrent plants. In many cases, the perceived seis-mic hazard was less than is currently postulated.As a result, there has been an evolution of activi-ties to reevaluate and upgrade these older plantsto increase their seismic resistance. In most cases,the focus has not been guided by risk assessments

and the evaluations and upgradings have oftenbeen disproportionate. The emphasis has been onpiping and structures whereas other essential andoften more vulnerable equipment have receivedmuch less attention. There are still operatingplants with poorly anchored or unanchored essen-tial equipment which have an excessive number ofpipe supports. The biggest issue in most of theolder US plants was ‘operability’ of equipmentduring and after a seismic event. Operability isnot usually amenable to analysis and as a result,other methods must be applied to demonstratefunction. The US Department of Energy reactorsand process facilities have not been subjected to

* Corresponding author. Tel.: +1 714 8333303; fax: +1714 8333392; e-mail: [email protected]

0029-5493/98/$19.00 © 1998 Elsevier Science S.A. All rights reserved.

PII S0029-5493(97)00339-7

R. Campbell et al. / Nuclear Engineering and Design 181 (1998) 115–129116

the regulatory practice for commercial NPPs. Inresponse to DOE order 5840.28, they must nowbe evaluated for natural phenomena hazards(NPH) to similar criteria developed for commer-cial NPPS. For the older reactors in Easter Eu-rope and the CIS, the current status is quitevaried. Many seismic reevaluation and upgradeprograms are just beginning and could benefitfrom the US experience over the past 15 years.

This paper is an overview of the methodologiesthat have been applied to resolve outstandingseismic issues in the US. This paper also discussescase studies where similar seismic evaluationmethodologies have been applied in selectedplants outside the US.

2. Seismic evaluation methodologies in the US

The evaluation of seismic vulnerabilities in ear-lier operating plants in the US began in the late1970s, during the Systematic Evaluation Program(SEP). Some of the initial activities were con-ducted by a Senior Review Panel funded by theUSNRC, and consisted of the analytical evalua-tion of selected structures, walkdowns and samplecalculations for equipment. Subsequent activitiesby the utilities and their contractors were in re-sponse to the findings of the USNRC consultantsand focused on piping analysis and evaluation ofselected structures and equipment. Operability ofequipment was not verified during the SEP pro-gram. Unfortunately, these utility programs didnot always focus on priority issues, often becauseof non-risk based perceptions of governing vul-nerabilities and excess conservatism contained inregulatory requirements at that time.

In general, the SEP program allowed for moreliberal acceptance criteria for structures by allow-ing the response to go beyond the elastic limit.Newmark and Hall (1978) developed criteria forevaluation of structures and equipment which in-cluded the use of inelastic response spectra. Forstructural systems which undergo inelastic defor-mation, the effective dynamic response could bedefined by a linear elastic response analysis usinga reduced response spectrum to define the inputmotion. The reduction in spectral acceleration

was based upon the allowable inelastic deforma-tion (ductility) and the frequency of the structuralsystem and was based upon exhaustive analyticalstudies of structures subjected to real earthquakerecords and on observed behavior of structuresduring strong motion earthquakes.

The evaluation of structures in many cases uti-lized the inelastic spectra concept, however, thiswas not carried over to piping systems. Pipingevaluations were very conservatively conductedusing classic linear elastic response spectrum anal-ysis, low damping and conservatively defined in-put motion. Subsequently, some of theseconservatisms have been reduced in efforts todevelop more rational criteria for the resolutionof other seismic issues and for new design.

In the early 1980s, there were a number ofunresolved seismic safety issues in the US. Manyof these issues were consolidated into two majorprograms. The first of these programs is thedemonstration of the operability of safety-relatedequipment during and after the design basis earth-quake. This activity is limited in scope to onlyaddress those issues which have not previouslybeen resolved for the design basis earthquake. Thesecond major issue is the evaluation of the plantresponse to seismic events beyond the designbasis.

The first issue was the Unresolved Safety Issue(USI) A-46, dealing with operability of safe shut-down equipment in 72 of the earlier US NPPs.The scope was, however, expanded to includelong-term decay heat removal equipment (USIA-45), selected seismic design basis issues (USIA-40) and seismic spatial systems interactions(USI A-17). Some additional passive items havealso been included (cable raceways, tanks andheat exchangers). Structures and piping were notincluded in this program since they had beenaddressed in other programs. The second issuewas the Individual Plant Examination of ExternalEvents (IPEEE). In this program, all structures,piping and equipment essential for safe shutdownmust be evaluated for seismic events greater thanthe design basis. This program also includes seis-mic spatial systems interactions, and to some ex-tent, Generic issue (GI)-57 which includes theconsequences of inadvertent activation of fire sup-pression systems during a seismic event.

R. Campbell et al. / Nuclear Engineering and Design 181 (1998) 115–129 117

The US Department of Energy has numeroustest and production reactors and process facilitieslocated on government reservations which havenot been subject to the US nuclear regulatoryprocess for power reactors. DOE order 5840.28requires that these facilities be reevaluated fornatural phenomena hazards and brought up tosafety standards commensurate with the publicrisk involved.

2.1. Resolution for USI A-46

A Generic Implementation Procedure (GIP)(Seismic Qualification Utility Group, 1991), hasbeen prepared over a several-year period toprovide criteria and methods to resolve most ofthe outstanding seismic issues related to the de-sign basis earthquake. The GIP is based heavilyupon the use of earthquake and testing experiencein lieu of analysis and testing of components. Alarge database of earthquake and testing experi-ence has been reviewed by a SSRAP, 1991, andthe USNRC and rules have been formulated todemonstrate survivability and operability of sev-eral generic classes of equipment. A testing data-base has also been collected and reviewed toestablish operability limits for equipment and re-lays (Merz, 1991a,b).

A final Safety Evaluation Report (SER) on theGIP has been issued and the affected US utilitieshave initiated programs to apply this procedure totheir NPP’s. More recently established plantswhose equipment has been seismically qualified toIEEE 344–1975 or later are exempt from thisissue.

The steps involved in the applications of theGIP to resolution of USI A-46 are:

Development of a safe shutdown equipment listDevelopment of a seismic demand (floorspectra)Equipment walkdown and screeningRelay evaluationOutlier resolutionReporting

The safe shutdown equipment list states thatequipment must function to safely shut down thereactor after a design basis earthquake. A singleshutdown path is defined, but redundancy must

be maintained for decay heat removal functions.Accident mitigation equipment is not required.

The seismic demand is that specified for thedesign basis (safe shutdown) earthquake. Manyplants chose to develop new spectra using moremodern and less conservative methods than origi-nally used. In some cases, the NPPs have electedto change their licensing basis by using a USNRCRegulatory Guide 1.60 spectral shape to definethe ground motion rather than the spectrum orig-inally used. By using the Regulatory Guide Spec-tral Shape, more recent and liberal regulatorycriteria may be used for the analysis of the struc-tural and equipment response.

The equipment walkdown and screening andrelay evaluation procedures are based upon seis-mic and testing experience. With regard to an-chorage evaluation, exhaustive studies have beenconducted to develop inspection and strengthcriteria for concrete expansion anchors.

2.2. IPEEE

Criteria for IPEEE have also been developedsimultaneously with the GIP, but are applicableto seismic levels beyond the plant design basis.The USNRC issued the Generic Letter, (USNRC,1991a) and NUREG 1407 (USNRC, 1991b) forIPEEE. There are three methodologies which maybe used: Seismic Probabilistic Risk Assessment(USNRC, 1983) NRC Seismic Margins Method(Budnitz et al., 1985, Prassinos et al., 1986) EPRI,1988.

For all of the methods, the goal is to determinethe seismic shaking level at which there is ahigh-confidence-of-low-probability-of-failure(HCLPF). This HCLPF is mathematically definedwith a confidence of 95% and 5% probability offailure.

2.3. Seismic PRA

In the PRA method, fragility curves for essen-tial equipment, piping and structure are defined asa conditional probability of failure versus a seis-mic input parameter (either peak ground accelera-tion or spectral acceleration within a definedfrequency range). A seismic hazard is defined as a


Fig. 1. Risk assessment methodology for seismic events.

frequency of occurrence versus a seismic inputparameter (peak ground acceleration or spectralacceleration). The plant systems are modeled asevent trees and fault trees from which Booleanequations are derived. Using the Boolean equa-tions, the seismic hazard and the component fra-gility curves, the frequencies of core damage andrelease from containment can be derived. Fig. 1shows the seismic PRA process from the modelingand input parameters up through the analysis ofthe consequences of an accident. As a by-productof the risk modeling, the plant level HCLPF canbe computed from the Boolean equations and thefragility curves. This computation defines thedominant accident sequences that lead to coredamage and release and the HCLPF for each.

If a PRA is elected to satisfy the IPEEE, it isonly required that core damage frequency (Level 1PRA) plus an evaluation of containment perfor-

mance be performed. The computation of releasefrequency (Level 2 PRA) is not required, however,many utilities have elected to go to this extent. Itis further stipulated that only a point estimate ofcore damage frequency is required. This involvesthe use of a mean seismic hazard prediction and asingle mean fragility curve (Fig. 2). The use of thefull uncertainty spread of the seismic hazard andfragility curves was not required. However, manyutilities elected to carry out this uncertaintyanalysis.

2.4. NRC Seismic Margins Method

The NRC seismic margins method was devel-oped by USNRC contractors and is a truncationof PRA. The plant systems are modeled andseismic fragility curves are developed, just as in aPRA, and the plant level HCLPF is computed.


Fig. 2. Component fragility curves.

However, only the most important safety func-tions are considered. The frequency of core dam-age and release are not determined. In applyingthe NRC margins method, seismic capacityscreening is conducted to eliminate many compo-nents from fragility computations. This capacityscreening is based primarily on the results of pastseismic PRAs and on the successful performanceof certain classes of equipment in previous strongmotion earthquakes.

The NRC seismic margins method involves thefollowing steps (Prassinos et al., 1986):

Selection of the review level earthquakeDevelopment of systems modelsInitial component ruggedness screeningPlant walkdownDevelopment of component and structural/fragilitiesSystem analysisDetermination of the plant level HCLPF

This procedure is virtually identical to the PRAprocedure with the exception that the systemsanalysis step does not involve the use of a seismichazard in the computation of core damage fre-quency. The systems models and fragility curvesare used to determine the dominant accident se-quences and the plant level HCLPF.

2.5. EPRI seismic margins method:

A deterministic seismic margins method wasdeveloped by Electric Power Research Institute(EPRI) contractors and is very similar to themethodology contained in the GIP for the resolu-tion of USI A-46. This similarity was deliberatedto minimize the required activity necessary toresolve both USI A-46 and IPEEE. In thismethod, safe shutdown paths are defined andcomponents and structures in the safe shutdownpaths are deterministically evaluated to calculatecomponent HCLPFs. The weakest component in


a shutdown path then defines the plant levelHCLPF for that path.

The steps in the EPRI seismic margins evalua-tion methodology are:

Selection of the review level earthquakeSelection of the assessment teamPreparatory work prior to the walkdownSuccess path selectionSeismic capability walkdown and screeningSeismic evaluation of unscreened componentsDocumentations

In this case, the success path selection must in-clude a primary success path and an alternatesuccess path utilizing different equipment to thegreatest extent possible. One of the paths mustalso have the capability to mitigate a small loss ofcoolant accident (SLOCA). This process is virtu-ally identical to the A-46 process with the excep-tion that the alternate success path and theSLOCA mitigation are additional requirements.Also, since the review level earthquake is specifiedas being beyond the design basis, all structuresand equipment, including piping that are impor-tant to the success paths must be included.

2.6. Selection of method

One of the above three methods is being ap-plied to all US operating plants. The choice ofmethod was determined by the review level earth-quake specified for the plant, the utility desire tocombine USI A-46 and IPEEE resolutions andthe utility preference for methodology.

The plants have been placed into three reviewlevel earthquake (RLE) bins. Most plants are tobe evaluated for a 0.3 g RLE and have elected todo an EPRI seismic margin methodology evalua-tion, although some have elected to do PRA.Only a few are opting for the NRC marginsmethodology with the goal of expanding the mar-gins evaluation to a PRA at some future date.There are a few plants which are placed in the 0.5g RLE level and all have elected seismic PRA fortheir IPEEE. Two California NPPs have RLEsexceeding 0.5 g and were required to conduct aPRA.

Even though the steps to perform the evalua-tion are summarized somewhat differently in the

governing documents, all of the methods requiresimilar procedures, as does the resolution of USIA-46. The NRC has emphasized the integration ofthe A-46 and IPEEE programs for plants whichmust carry out both. Fig. 3 compares the A-46resolution process to the EPRI seismic marginsprocess. Fig. 4 compares the A-46 resolution pro-cess to the seismic PRA process. The NRC seis-mic margins process follows the steps in Fig. 4 tothe point of seismic risk quantification. At thatpoint, the margins process involves the computa-tion of the plant level HCLPF using the systemsmodels and fragility curves. As can be seen, theactual steps and scope of work are very similar.

Numerous seismic PRAs have been conductedin the US prior to the IPEEE requirements. ThesePRAs require enhancements, principally the per-formance of a detailed walkdown, the addition ofequipment associated with containment perfor-mance and the use of more recent estimates ofseismic hazard.

Pilot studies that have been conducted using themargins methodologies include: the NRC methodfor Main Yankee (Ravindra et al., 1987), theEPRI margins method for Catawba (Campbell etal., 1989), and a combined EPRI margins andA-46 methodology for Plant Hatch (SouthernCompany Services, 1991).

2.7. Department of energy criteria

DOE 5480.28 requires that the US Departmentof Energy facilities be designed or evaluatedagainst performance goals consistent with theirsafety mission and cost importance. In order tocomply with this order, the US Department ofEnergy has developed criteria for the evaluationof DOE reactors and process facilities for naturalphenomena hazards (DOE, 1994). The criteria arestructured with respect to the performance goalsand risk inherent in the process. There are fourcategories of facilities. The performance goal foreach category is based upon a frequency of occur-rence of the event and the probability of failure,given the event. The most critical of the facilitieshas a seismic hazard defined for a very low fre-quency of occurrence, similar to that for definingthe safe shutdown earthquake (SSE) for power


Fig. 3. Seismic IPEEE integrated seismic margin assessment and A-46 evaluation.


Fig. 4. Seismic IPEEE integrated seismic PRA or NRC margins and A-46 evaluation.


Table 1Performance goals for each usage category

Specified hazardPerformance goal descriptionUsage category Performance goal annual probabil-ity of exceedanceprobability

General use (normal buildings) 2×10−3 1×10−3 of the onset of majorMaintain occupant safetystructural damage to the extentthat occupants are endangered

Important or low hazard (es- Occupant safety, continued operation 1×10−3 5×10−4 of facility damage to thesential buildngs) extent that the facility cannot per-with minimal interruption

form its functionOccupant safety, continued function, 5×10−4 1×10−4 of facility damage to theModerate hazard (nuclear lab-

extent that the facility cannot per-hazard confinementoratories)form its function (confinement)

Occupant safety, continued function, 1×10−5 of facility damage to theHigh hazard (reactors) 1×10−4

very high confidence of hazard confi- extent that the facility cannot per-form its function (confinement)nement

reactors. The use categories and performancegoals are shown in Table 1.

There is an ongoing effort to update and ex-pand the DOE criteria. In particular, the DOEhas undertaken a program to develop a completeevaluation criteria parallel to the GIP for theresolution of USI A-46. This procedure will bebased principally upon earthquake and testingexperience, with supplemental analysis for an-chorage and strength of supports. Several DOElaboratories are currently in the process of evalu-ating their major structures and equipment. Sometest and production reactors have completedPRAs.

3. Applications outside of the US

Some past and ongoing projects outside the UShave utilized the methods described above as fullor partial resolution of seismic issues. SeveralPRAs which include external events have beenperformed for plants in Switzerland, Taiwan, Ko-rea, Japan, Finland, the Czech Republic, andBulgaria.

None of these PRAs have been conducted spe-cifically to address seismic issues; external eventshave been a logical extension of the PRAs ini-tiated to study internal event vulnerabilities.Other selected applications of the above describedmethodologies have been applied in Switzerland,

Finland, Sweden, Belgium, France, Bulgaria,Hungary and Slovakia. Finland and Sweden havea low seismic hazard and as a result, the emphasison seismic events is somewhat limited. At theTihange plant in Belgium and the Beznau plant inSwitzerland, the seismic design basis is similar tothat of an eastern US site. At the Kozloduy site inBulgaria, Paks in Hungary and Bohunice in Slo-vakia, the seismicity has recently been redefinedand the results in ground motion input levels aresignificantly greater than the previously predictedlevel. In Bulgaria, several earthquakes have oc-curred at the site. The largest earthquake was in1977, resulting in approximately 0.1 g peakground acceleration at the site, resulting in somestructural damage. At this site, the currently spe-cified earthquake for seismic reevaluation is 0.2 g.At Paks, it is 0.25 g and 0.3 g at Bohunice.

3.1. Sweden

The older nuclear power plants in Sweden werenot specifically designed to withstand earthquakesdue to Sweden’s relatively low seismicity. In thepast several years, a number of studies have beenconducted to assess the seismic hazard at theSwedish nuclear power plant sites and to estimatethe seismic margins of older plants.

The first study was to estimate the seismicmargin of the mitigation systems at OskarshamnUnits 1 and 2. The mitigation concept developed


by OKG to address the unlikely event of a severeaccident at Oskarshamn consists of the filtervented containment and an independent contain-ment spray system. These systems were designedto meet seismic standards currently used in Swe-den. Some of the components of these systemsinterface with the existing systems in OskarshamnUnits 1 and 2 which were not designed to currentseismic criteria. A pilot study (Landelius et al.,1989) was performed to assess the seismic marginsof these interfacing systems and to verify that theywould perform successfully in a major earth-quake. In the first phase of the study, the seismicmargin of the containment spray system was eval-uated using the earthquake experience database,plant walkdown and seismic fragility develop-ment. It was concluded that the system had aHCLPF seismic capacity of 0.23 g peak groundacceleration anchored to the USNRC RegulatoryGuide 1.60 ground response spectrum, providedthat the seismic adequacy of anchorage of theheat exchangers’ framing was further verified.

The Swedish nuclear industry and the regula-tory agency funded an investigation—ProjectSeismic Safety—to develop a characterization ofseismic ground motions for the probabilistic anal-ysis of nuclear facilities in Sweden. The study(Engelbrektson, 1989) has produced uniform haz-ard ground motion spectra for hard rock and softsoil sites at annual frequencies exceeding of 10−5,10−6 and 10−7. The utilities began studying waysof using these ground motion spectra in seismicevaluation of existing plants. A pilot study wasconducted to evaluate the probabilistic responseand capacity of the reactor/containment buildingat Oskarshamn 2 and Barsebeck 1 and 2 (Asfuraand Baltus, 1991). The objective was to demon-strate that probabilistic response analysis wouldlead to a more realistic response prediction whichcould be used in seismic evaluation of existingplants. OKG, the proprietor of Oskarshamn, ulti-mately joined SQUG (Seismic Qualification Util-ity Group) and selected the EPRI seismic marginmethod developed in the US to conduct a fullseismic evaluation of Oskarshamn II. Unit 1 isexpected to follow, along with other NPPs inSweden.

3.2. Finland

Imatran Voima Oy has performed a probabilis-tic risk assessment of the Soviet-designed LoviisaNuclear Power Plant. This PRA explicitly treatedthe seismic event. The Loviisa plant was notdesigned for any specific seismic criteria. There-fore, the seismic fragility evaluation had to rely onseismic walkdown, selected calculations, and theuse of earthquake experience data in the develop-ment of seismic capacities of structures and equip-ment. An initial scoping study (Ravindra et al.,1989) identified certain components needing fur-ther fragility evaluation. Imatran Voima Oy per-formed the probabilistic response analysis todevelop realistic floor spectra (Varpasuo and Put-tonen, 1991). Using these spectra and based onthe plant information and walkdown findings, theseismic fragilities have been developed for selectedcomponents (Ravindra et al., 1991). The results ofthe seismic risk analysis shows a very low (4×10-

7/year) frequency of seismic induced core damage(Varpasuo et al., 1993).

Teollisnuden Voima Oy (TVO) is performing aseismic probabilistic risk assessment of theSwedish designed BWR plant at Olkiluoto—twounits of 710 MW each. The objective is to verifyseismic adequacy of the plants in support of reli-censing. This ongoing project was initiated in1996. The seismic PRA procedure follows thosedeveloped in the US. Also, in conjunction withrelicensing of the plant, other studies for theresponse to BWR hydrodynamic loads are beingconducted. The safety relief valve discharge loadsmust be combined with seismic loads in the seis-mic PRA.

3.3. Switzerland

Seismic PRAs have been conducted for Beznau,Gosgen and Muelburg in Switzerland, with anongoing study at Liebstadt. The Beznau PRA wasthe first in Switzerland and was used to specifydesign requirements for a dedicated safe shut-down facility which has been added to each of thetwo PWRs. All of the new equipment in the safeshutdown facilities has been seismically qualifiedby currently specified US standards (ASME,


IEEE etc.) The piping and equipment in thecontainment and the steam and feedwater pip-ing outside the containment ahead of their re-spective isolation valves were required to berequalified. In this requalification program, avariety of methods were utilized (Sahgal etal., 1990). Large bore piping has been evalu-ated to current ASME standards using dy-namic analysis. Small bore piping has beenevaluated using chart type screening methodsbased on ASME code stress allowables, se-lected dynamic analysis with increased allow-ables and to some extent, using seismicexperience based criteria. All equipment(valves, heat exchangers, tanks, piping penetra-tions) and systems interaction issues have beenresolved using deterministic seismic marginsmethods which rely heavily on seismic experi-ence based screening and selected calculations.This program has worked well to apply prac-tical, yet technically justifiable methods forseismic requalification of existing piping andcomponents. In this program, the applicabilityof the seismic experience based screening crite-ria to European equipment had to be demon-strated. Only minor modifications have beenrequired to demonstrate the ability of existingequipment to withstand the safe shutdownearthquake.

3.4. Belgium

The Belgian Utility, Electrabel, was an earlymember of the SQUG and has applied the GIPto the Tihange 1, 2 and 3 nuclear power plants.The Belgian work began before the finalizationof the GIP. As reported in Lafaille et al., 1990,the issues to be resolved with the Belgian au-thorities could mostly be addressed by using theGIP methodology, with some minor changes inprocedures and some additional study for equip-ment that could not be demonstrated to be rep-resented by the earthquake experience databasewhich forms the basis for GIP screening rules.This program appears to have been very suc-cessful and was the first full application of theGIP in Europe.

3.5. Taiwan

The three NPPs in Taiwan have conductedPRAs including seismic events. The three plantsare of US origin and the PRAs were conducted byUS contractors with the participation of localengineers. In the oldest NPP in Taiwan, seismicexperience was utilized to resolve seismic qualifi-cation issues. However, in this case, the plant wasa turn key plant which utilized almost all USmanufactured equipment, thus there was little is-sue regarding the applicability of seismicexperience.

3.6. Korea

Two US-supplied NPPs in Korea have com-pleted PRAs, including seismic events. The resultswere comparable to the results obtained for plantsin the US. A third plant of Canadian origin isundergoing a PRA which includes seismic events.In this case, the site hazard is quite high relativeto the standard 0.2 g pga design basis and it isanticipated that the seismic risk may be higherthan normal, but within the spread of past stud-ies. Much of the equipment in this plant is ofKorean origin. However, it is built under licenseto Canadian or Japanese companies and is similarto seismic experience database equipment.

3.7. Japan

PRAs including seismic events have been con-ducted for a Japanese PWR and for the MonjuFast Breeder Reactor (FBR). Results of the PWRwere comparable to results obtained for USplants. For the FBR, the site hazard was quitehigh relative to the design basis (0.46 g S2 earth-quake). However, due to the passive design, thecalculated core damage frequency was within therange of US LWRs. If there is no loss of primarycoolant, the passive safety features can insure corecooling without electrical power or a water sourceas an ultimate heat sink (Campbell et al., 1991).

In the FBR PRA, where specific qualificationdata were not available, the earthquake experi-ence database was used in a few instances todevelop fragilities. In cases where there was a lot


of data, especially at high acceleration sites, atechnique known as survival analysis was used todevelop the probability of failure of specificclasses of equipment at increasing accelerationlevels. In conducting this study, it was determinedin a walkdown that the Japanese manufacturedequipment was just as rugged as database equip-ment. Due to the frequency and severity of earth-quakes in Japan, most manufactured productshave good seismic resistance by way of appropri-ate anchorage and attention to the load path.

3.8. Slo6enia

The KRSKO NPP in Slovenia is a Westing-house PWR and was designed for a 0.3 g peakground acceleration. This site is quite seismicallyactive and high acceleration, low energy, earth-quakes frequently occur. These close in, low en-ergy earthquakes are, however, not damaging toengineered structures and equipment. A short du-ration earthquake of 0.46 pga occurred in 1989with no damage to the plant. Recent seismichazard studies show that a 50 percentile, 10000-year return period earthquake producing low fre-quency damaging vibratory motion isapproximately 0.4 g, which exceeds the designbasis. This is similar to many situations in USplants where beyond design basis earthquakesmust be addressed in IPEEE. The utility elected toconduct IPE and IPEEE of the plant using the USPRA methodology. This work was carried out byWestern European and US contractors. Resultswere comparable to the US plant results (Vermautand Monette, 1995).

3.9. Spain

A consortium of Spanish utilities have beenlong-standing SQUG members and are now be-ginning to implement their A-46 program. Acountry-wide hazard study has been conducted todefine the seismicity at each NPP site and recom-mendations have been made as to the methodol-ogy to be used when performing IPEEE.Vandellos II, a modern Westinghouse PWR, isthe leading plant for IPEEE studies.

3.10. Bulgaria

Kozloduy units 1–4 are Soviet designed VVER440 model 230 PWRs. In an initial IAEA mission,(Monette et al., 1991), a short walkdown wasconducted and fragilities were calculated for themost seismically vulnerable components identifiedin the walkdown. HCLPF’s were back calculatedfrom the fragility curves. The principal reason forselecting the fragility method was it’s ability totreat uncertainties regarding incomplete informa-tion as to the seismic input, structural responseand equipment construction. The deterministicmethod requires that these parameters be definedin accordance with stated rules or to be estimatedconservatively.

Subsequent to this initial IAEA study, twofollow-on programs were simultaneously initiated.IAEA defined terms of reference for a WANOsponsored program to design priority seismic up-grades for Kozloduy 1 and 2. The scope of workfor the terms of reference was developed basedupon the prior IAEA mission and risk prioritiesderived from results of a toplevel risk assessmentof Kozloduy 1–4 (BEQE, 1992). The programwas defined in four phases and the first twophases have been completed. They included theevaluation and upgrade design for equipment an-chorage, the diesel generator building, and theservice water pump house. In addition, the mainbuilding, which consists of the reactor confi-nement, the auxiliary building and turbine hall,have been analyzed and in-structure spectra havebeen developed. Phases 3 and 4 would includefurther evaluation and design of upgrades for themain building, evaluation and upgrade of theprimary circuit, and a walkdown and experiencebased evaluation of piping and cable raceways.Some of this work has been accomplished andmany equipment items have been upgraded. How-ever, major structural upgrades have not beenconducted.

A very similar program funded directly by theplant was carried out for Kozloduy units 3 and 4by local Bulgarian engineers with assistance fromtheir US counterparts. To date, most equipmentanchorage and masonry wall upgrades have beencompleted. Structural upgrades have not been ini-


tiated to date. Upgrading of the unit 3 and 4pumphouse is planned for the near future. Theupgrade program must be compatible with out-ages and electricity demand, thus the design ofstructural upgrades must minimize outage time.

A PRA including seismic events was conductedfor Kozloduy 5 and 6, which are 1000 MWVVERs. The internal event core damage fre-quency was higher than most Western PWRs. Theseismic portion of the PRA was not completeddue to software limitations. There were a fewseismic vulnerabilities observed in the study, mostof which have been easily fixed.

3.11. Hungary

There are four WWER 440 model 213s at Paksin Hungary. Several ongoing programs are ad-dressing different aspects of the seismic issues. Aunified criteria has evolved for applications atPaks which was the merging of criteria being usedby several contractors for structural evaluationsand easy-fixes of equipment and masonry walls.The criteria are a combination of the SQUG GIP,Seismic Margins Methodology and DOE standardfor structures. The easy-fix project for anchorageof equipment and stabilization of masonry wallshave been completed. The phase 1 evaluation ofpiping and mechanical equipment is completed,however, the actual backfit is not commencinguntil there is further refinement in the siteseismicity.

The main building complex at Paks consists ofthe reactor building and turbine hall, intercon-nected by gallery buildings with each building ona common, flexible foundation mat. The struc-tures are a combination of reinforced concreteand structural steel frames with concrete infillpanels. These structures were not designed forearthquakes, hence are vulnerable to current esti-mates of the seismic hazard. Seismic evaluationsof the existing structures were conducted based oncurrent seismic hazard estimates and upgrade con-cepts were developed for identified deficiencies.The criteria being considered for the design ofupgrades is that defined in the DOE Standard1020–94 with suitable restrictions placed on theductility of existing precast structures. These stud-

ies and fixes were conducted for a 0.35 g pga,earthquake. Recently, the hazard was finalized at0.25 g. As a result, the remaining work to accom-plish the more difficult fixes can proceed withmore realistic loading.

Earthquake experience has been used on a lim-ited basis to quality a new diesel generator to beinstalled at PAKS. This was acceptable to theauthorities on the basis that similar projects hadbeen conducted in the US for the qualification ofDiesel Generator Systems.

3.12. Slo6akia

There are four WWER 440s at Bohunice, twomodels 230s and two models 213s. These plantswere not originally designed to resist earthquakes.However, more recent seismic hazard assessmentsreveal that the hazard could be higher than that inBulgana and similar to Paks in Hungary. Majorstructural backfits for the two model 230s havebeen conducted by Czech and Slovak engineers. AWestern European contractor has recently beenselected to do a complete modernization of theolder model 230s. The seismic portion of thiswork will utilize US developed experienced-basedmethodology for evaluating existing equipment.The contractor joined SQUG to have access to allof the US technology and seismic experience data-base for use in such projects.

4. Conclusions

Well-defined criteria for evaluation of outstand-ing seismic issues in the US have been developedand are rapidly being applied to existing powerreactors, test reactors and nuclear facilities. Somelimited applications of these methodologies havebeen made for European plants. In Europe, thereis a wide diversification of regulating authoritiesand seismic hazard at plant sites. As a result, it isunlikely that US requirements for IPEEE, USIA-46 and DOE will be applied across-the-boardfor all plants in all countries. There is merit inselecting practical aspects of these methodologiesand methods for application to specific issues.Some limited applications have been presented todemonstrate the applicability.


However, Screening criteria associated withthese methods must be used with caution. Someequipment in European reactors is not adequatelyrepresented in the experience database to confi-dently apply the US screening criteria. While mostequipment is generically rugged, there are someunique constructions that have been observed,where the screening criteria are clearly not appli-cable without additional justification. In a SQUGtraining course held in Brussels, a trial plantwalkdown was conducted on a Belgian fossilpower plant. In this brief training walkdown,several instances were found where the intent ofthe screening criteria was not satisfied. In particu-lar, the 6.3 kv switchgear contained ceramic parts.The GIP criteria limit voltage is 4.1 kv due to thefact that ceramics are often used in higher voltageswitchgear. In Eastern European plants, 6.3 kv iscommon and a similar design with ceramic partshas been observed in Russian supplied switchgear.

The seismic evaluation procedures as character-ized by the GIP and earthquake experience data,in general, are becoming standardized by theIEEE and the ASME for electrical and mechani-cal equipment, respectively. It is anticipated thatultimately such procedures will become industrystandards worldwide.

References

Asfura, A.P. and Baltus, R., 1991. Pilot Study of Reactor/Containment Building: Oskarshamn 2 and Barsebeck 1and 2, Probabilistic Response and Capacity, Report pre-pared for Sydkraft and OKG Aktiebolag by EQE Engi-neering, Westinghouse Energy Systems International.

BEQE, 1992. Top Level Risk Study For Kozloduy Units 1 to4, Prepared for the Committee on the Use of AtomicEnergy for Peaceful Purposes by BEQE Ltd., April.

Budnitz, R.J., Amico, P.J., Cornell, C.A., 1985. An Approachto the Quantification of Seismic Margins in Nuclear PowerPlants, Lawrence Livermore National Laboratory,NUREG/CR-4334.

Campbell, R.D., Tiong, L., Tong, W., Kipp, T., Nafday, A.,Yamaguchi, A., 1991. Seismic Fragility Methodology forEvoluation of Liquid Metal Reactors, Structural Mechan-ics in Reactor Technology, Paper Ml l(H)/1.

Campbell, R.D., Henley, B.F., Buttemer, D.R., 1989. SeismicMargin Assessment of the Catawba Nuclear Station, EPRINP-6359.

DOE Standard 1020-94, 1994. Natural Phenomena HazardDesign and Evaluation Criteria for Department of EnergyFacilities, April 1994.

Electric Power Research Institute (EPRI), 1988. A Methodol-ogy for Assessment of Nuclear Power Plant Seismic Mar-gin, NP-6041.

Engelbrektson, A., 1989. Characterization of Seismic GroundMotions for Probabilistic Safety Analysis of Nuclear Facil-ities in Sweden, vol. K1. Trans. of the 10th Int. SMiRTConf., pp. 37–42.

Lafaille, J.P., Aelbrecht, D., Lepiece, M., Detruoux, P., 1990.Experience of Seismic Walkdowns of Belgian Plants. Proc.3rd Symp. on Current Issues Related to Nuclear PowerPlant Structures, Equipment and Piping, North CarolinaState University.

Landelius, M., Ravindra, M.K., Hardy, G.S., Hashimoto,P.S., 1989. Seismic Margin Assessment of Mitigation Sys-tems in Oskarshamn. 10th Int. Conf. on Structural Me-chanics in Reactor Technology, Anaheim, California.

Merz, K.L., 1991a, Generic Seismic Ruggedness of PowerPlant Equipment, Prepared by ANCO Engineers for theElectric Power Research Institute, EPRI NP-5223.

Merz, K.L., l991b, Seismic Ruggedness of Relays, EPRI NP-7147, Prepared by ANCO Engineers for the Electric PowerResearch Institute.

Monette, P., Baltus, R., Yanev, P., Campbell, R., 1991. Seis-mic Assessment of Kozloduy VVER 440, Model 230 Nu-clear Power Plant, Structural Mechanics in ReactorTechnology, Paper SD 006/5.

Newmark, N.M. and Hall, W.J., 1978. Development of Crite-ria for Seismic Review of Selected Nuclear Power Plants,NUREG/CR-0098.

Prassinos, P.G., Ravindra, M.K., Savay, J.D., 1986. Recom-mendations to the Nuclear Regulatory Commission onTrial Guidelines for Seismic Margin Reviews of NuclearPower Plants, Lawrence Livermore National Laboratory,NUREG/CR-4482.

Ravindra, M.K., Hardy, G.S., Hashimoto, P.S., Griffin, M.J.,1987. Seismic Margin Review of the Main Yankee AtomicPower Station, NUREG/CR-4426, vol. 3, Prepared byEQE Inc. for Lawrence Livermore National Laboratory.

Ravindra, M.K., Tong, W.H., Tiong, L.W., Monette, P., 1991.Seismic Fragilities of Selected Components in Loviisa Nu-clear Power Plant, Prepared for Imatran Voima Oy byEQE Engineering, Inc. and Westinghouse Energy SystemsInternational.

Ravindra, M.K., Hardy, G.S., Hashimoto, P.S., 1989. ScopingStudy on Seismic Fragilities for Seismic Risk Analysis ofLoviisa Nuclear Power Plant, report prepared for ImatranVoima Oy by EQE Engineering, December.

Sahgal, S., Culot, M., Campbell, R., Monette, P., 1990. Appli-cation of Experience Based Methodology to the SeismicQualification of Beznau Nuclear Power Plant. 3rd Symp.on Current Issues Related to Nuclear Power Plant Struc-tures, Equipment and Piping, North Carolina StateUniversity.


Seismic Qualification Utility Group (SQUG) 1991. GenericImplementation Procedure (GIP) for Seismic Verificationof Nuclear Plant Equipment, Rev. 2.

Senior Seismic Review and Advisory Panel (SSRAP), 1991.Use of Seismic Experience and Test Data to Show Rugged-ness of Equipment in Nuclear Power Plants, Rev. 4.

Southern Company Services, 1991. Seismic Margin Assess-ment of the Edwin I. Hatch Nuclear Plant, Unit 1, EPRINP-7217, Prepared by Southern Company Services forElectric Power Research Institute.

USNRC, 1983. PRA Procedures Guide, NUREG/CR 2300.USNRC, 1991a. Generic Letter 88–20, Supplement 4, Individ-

ual Plant Examination for External Events (IPEEE) forSever Accident Vulnerabilities-10 CFR 50.54(f).

USNRC, 1991b. Procedural and Submittal Guidance for theIndividual Plant Examination of External Events (IPEEE)for Severe Accident Vulnerabilities, NUREG-407.

Varpasuo, P. and Puttonen, J., 1991. Development of Proba-bilistic Floor Spectra for Loviisa Nuclear Power Plant.11th Int. Conf. on Structural Mechanics in Reactor Tech-nology, Tokyo, Japan.

Varpasuo, P., Puttonen, J., Ravindra, M.K., 1993. SeismicProbabilistic Safety Analysis of Loviisa NPP, Unit 1, Proc.Structural Mechanics in Reactor Technology, 12, PaperMK05/3, August.

Vermaut, M.K. and Monette, P., 1995. Methodology andResults of the Seismic Probabilistic Safety Assessment ofKRSKO Nuclear Power plant, SMiRT 13. Post Confer-ence Seminar 16, Iguazu, Argentina, August.

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