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Improvementoftheseismicsafetyofexistingnuclearpowerplantsbyanincreaseofthecomponentseismiccapacity:Acasestudy

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Nuclear Engineering and Design 238 (2008) 1410–1420

Improvement of the seismic safety of existing nuclear power plants byan increase of the component seismic capacity: A case study

Young-Sun Choun ∗, In-Kil Choi, Jeong-Moon SeoIntegrated Risk Assessment Center, Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu,

Daejeon 305-353, Republic of Korea

Received 1 September 2006; received in revised form 15 October 2007; accepted 31 October 2007

bstract

This case study produces the scenario earthquakes for an example nuclear power plant (NPP) site and suggests the effective seismic capacityf safety-related equipment and components which significantly contribute to a core damage to improve the seismic safety of an existing NPP bysing a probabilistic safety assessment. The response spectra for the scenario earthquakes show greater spectral accelerations than those for the

esign response spectrum in the frequency range higher than about 12 Hz. In order to improve the seismic safety of an example NPP, the effectsf the seismic capacity of safety-related equipment and components on the core damage frequency (CDF) are investigated, and their effectiveeismic capacities are determined. The results of the case study show that an increase of the seismic capacity of the equipment reduces the CDFonsiderably. The effective seismic capacities for the diesel generator, offsite power, condensate storage tank and battery rack are determined as

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.84, 0.35, 0.63 and 0.63 g, respectively.2007 Elsevier B.V. All rights reserved.

. Introduction

The safety-related structures, systems and componentsSSCs) in a nuclear power plant (NPP) which are designed to beafe for a design basis earthquake may be damaged or failed bytrong ground motions greater than a design basis earthquakes well as a particular earthquake of which the frequency con-ents are different from those of a design input motion. Due tohese uncertainties from a magnitude and frequency contents ofhe earthquake ground motions, it is necessary to maintain theeismic margin of the safety facilities high enough to ensure theeismic safety of a plant against probable strong-earthquakest a plant site. When new faults are found near a plant site, theesign basis earthquake should be determined by considering theew faults and then the seismic safety of the safety-related SSCshould be reevaluated under the revised design basis earthquake.

Recently, in Korea, Quaternary fault segments were found

ear NPP sites and a geological survey was performed to identifyhether they were active faults or not. When they are identified

s active faults, the seismic design spectra must be modified and

∗ Corresponding author. Tel.: +82 42 868 2036; fax: +82 42 868 8256.E-mail address: sunchun@kaeri.re.kr (Y.-S. Choun).

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029-5493/$ – see front matter © 2007 Elsevier B.V. All rights reserved.oi:10.1016/j.nucengdes.2007.10.008

he seismic safety of the NPPs near the faults must be reevalu-ted by using the modified design spectra. Then, if necessary,ome modification may be added to the vulnerable SSCs toncrease their seismic resistance capacities and to ensure theeismic safety of the plants. Actually, since a NPP consists ofumerous systems and components, the selection of the SSCsmportant to the seismic safety of a plant is not easy. Therefore,

ore efficient procedures are necessary to evaluate and improvehe seismic safety of a plant.

Probabilistic safety assessment (PSA) can be effectively usedor evaluating the seismic safety of a NPP. The use of a seismicSA cannot only provide the core damage frequency (CDF)ssociated with earthquakes, but also identify the dominant seis-ic risk contributors and the range of a peak ground acceleration

hat contributes significantly to a plant risk as summarized inhe IAEA Technical Document, IAEA-TECDOC-724 (1993),Probabilistic Safety Assessment for Seismic Events’. In thenited States, the NRC issued Generic Letter 88-20, ‘Individ-al Plant Examination for Severe Accident Vulnerabilities, 10FR 50.54(f)’ (USNRC, 1988) to conduct an individual plant

xamination (IPE) of a severe accident risk for internally ini-iated events. Subsequently, the NRC issued Supplement 4 toeneric Letter 88-20, ‘Individual Plant Examination of Exter-al Events (IPEEE) for Severe Accident Vulnerabilities, 10 CFR

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0.54(f)’ (USNRC, 1991) and NUREG-1407 (1991), ‘Procedu-al and Submittal Guidance for the Individual Plant Examinationf External Events (IPEEE) for Severe Accident Vulnerabilities:inal Report,’ to perform a complementary assessment to iden-

ify plant-specific vulnerabilities under severe accidents causedy external events. Finally, the NRC issued Supplement 5 toeneric Letter 88-20, ‘Individual Plant Examination of Exter-al Events for Severe Accident Vulnerabilities’ (USNRC, 1995)o provide guidance on modifications in the scope of a seismicPEEE for certain plants. One of the IPEEE objectives is toeduce the overall likelihood of a core damage and radioactiveaterial releases by modifying, where appropriate, hardware

nd procedures that would help prevent or mitigate severe acci-ents.

This paper investigated the effect of the seismic capacity ofafety-related components on the seismic risk of a NPP and sug-ested their effective seismic capacities by using a seismic PSA.

case study produced the scenario earthquakes for a nuclearlant site and the corresponding response spectra by consideringhe potential active-fault effect, and then conducted a compari-on with the current design response spectra. For an evaluationf the seismic safety of the nuclear plant, this study selected theomponents important to the CDF from the results of a seismicSA, and then performed a seismic PSA by using different seis-ic capacities of the selected components. Finally, this study

uggested the effective seismic capacities of the selected com-onents for improving the seismic safety of the nuclear plant.

. Evaluation of the design response spectra

Most of the NPPs in Korea have been designed by usinghe design response spectra proposed in the USNRC Regulatoryuide 1.60 (1973) instead of the site-specific design response

pectra due to the lack of strong ground motion records.owever, recently, since many earthquake records have been

ccumulated and several new faults have been found near NPPites, it is necessary to evaluate whether the use of the NRCesign response spectra as their design bases could ensuren adequate level of a conservatism for the safety of theselants. This chapter proposes the scenario earthquakes and theiresponse spectra at one of the plant sites and compares themith the NRC design response spectra.

.1. Procedures for determining the scenario earthquakes

To define the scenario earthquakes at a site, in general,he deterministic and the probabilistic seismic hazard analysispproaches are used. The deterministic methodology estimateshe strong motion parameters for the maximum possible earth-uake assumed to occur at the closest area around a site, while therobabilistic methodology integrates the effect of all the earth-uakes expected to occur at different locations around a siteuring a specified life period. The seismic hazard analysis aims

t evaluating the annual probability of exceedance of variousarthquake sizes at a selected site. The seismic hazard of a NPPite is usually represented by a series of seismic hazard curves,hich shows a relationship between the annual probability of

and Design 238 (2008) 1410–1420 1411

xceedance and the ground motion intensity such as a peakround acceleration, velocity or spectral acceleration. Since arobabilistic seismic hazard analysis (PSHA) can determine thennual probability of exceedance for the intensity parametersf a ground motion, it is very useful to define the scenarioarthquakes (Ishikawa and Kameda, 1993).

The concept of the probability based scenario earthquakeriginates from McGuire (1995). The probability based scenarioarthquakes, represented by particular sets of an earthquakeource magnitude and the distance for a specified probabil-ty level, can be obtained from a de-aggregation of the PSHAesults. Currently, two typical methods are usually used for defin-ng the probability based scenario earthquakes at a NPP site.he first one was developed by USNRC. The scenario earth-uakes are called controlling earthquakes in the Regulatoryuide 1.60 (1973). The second one was developed by Japantomic Energy Research Institute (Hirose et al., 2002) base onstudy by Ishikawa and Kameda (1993). Takada et al. (1999)

dopted these two procedures to define the scenario earthquakesor an example site and showed that both methods generatedimilar results. However, the authors noted a methodologicalifference between them. The USNRC’s procedure cannot iden-ify an earthquake source location for the scenario earthquakes.his procedure uses coarser bins for an earthquake magnitudend distance, thus all the results including the earthquake sourceegions and faults are mixed up in the bins. On the other hand,shikawa’s procedure has the advantage that it produces sourceontribution factors that are an effective indicator for identifyinghich earthquake sources and/or which fault(s) are most influ-

ntial, and one or more scenario earthquakes can be determined.inally, the authors concluded that the USNRC’s procedureives a global view of the scenario earthquakes, while Ishikawa’srocedure provides a more precise view of the scenario earth-uakes.

In order to develop the scenario earthquakes for a KoreanPP site, this study used the USNRC Regulatory Guide 1.165rocedure because it was the original approach adopted for deter-ining the controlling earthquakes at the site. The procedure

s based on a de-aggregation of a probabilistic seismic hazardn terms of an earthquake magnitude and distance. A simpleescription of the procedure is as follows (USNRC Regulatoryuide 1.165, 1997):

Step 1: Perform a site-specific PSHA. The hazard assessment(mean, median, 85th percentile and 15th percentile) should beperformed for spectral accelerations at 1, 2.5, 5, 10 and 25 Hz,and the peak ground acceleration. A lower-bound magnitudeof 5.0 is recommended.Step 2: Using the reference probability, determine the groundmotion levels for the spectral accelerations at 1, 2.5, 5 and10 Hz from the total median hazard obtained in Step 1, andcalculate the average of the ground motion level for the 1 and2.5 Hz and the 5 and 10 Hz spectral acceleration pairs.

Step 3: Perform a complete probabilistic seismic hazard anal-ysis for each of the magnitude–distance bins.Step 4: From the de-aggregated results of Step 3, the medianannual probability of exceeding the ground motion levels of

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Step 2 (spectral accelerations at 1, 2.5, 5 and 10 Hz) are deter-mined for each magnitude–distance bin. Using the medianannual probability, the fractional contribution of each mag-nitude and distance bin to the total hazard for the average of 1and 2.5 Hz and 5 and 10 Hz, are computed, respectively.Step 5: Review the magnitude–distance distribution for theaverage of 1 and 2.5 Hz to determine whether the contributionto the hazard for distances of 100 km or greater is substan-tial. If the contribution to the hazard for distances of 100 km orgreater exceeds 5%, additional calculations are needed to deter-mine the controlling earthquakes using the magnitude–distancedistribution for distances greater than 100 km.Step 6: Calculate the mean magnitude and distance of thecontrolling earthquake associated with the ground motionsdetermined in Step 2 for the average of 5 and 10 Hz.Step 7: If the contribution to the hazard calculated in Step 5fordistances of 100 km or greater exceeds 5% for the average of 1and 2.5 Hz, calculate the mean magnitude and distance of thecontrolling earthquakes associated with the ground motionsdetermined in Step 2 for the average of 1 and 2.5 Hz.Step 8: Determine the SSE response spectrum.

.2. Probabilistic seismic hazard analysis

The PSHA was performed for a selected NPP site. The teampproach developed by EPRI (Electric Power Research Insti-ute) was adopted for the hazard analysis. Three seismicityxpert teams and one attenuation team were composed to obtainhe PSHA input parameters. At least one non-seismologist wasncluded in each seismicity team. However, in the attenuationeam, only one expert recommended several different attenuationquations with a weighting factor (Seo et al., 1999).

A questionnaire was compiled and distributed to the team.he contents of the questionnaire are as follows:

Seismicity(1) Matrix of the physical characteristics.(2) Assessment of the tectonic features according to the

matrix of the physical characteristics.(3) Seismic sources (source zone) and their inter-dependency.(4) Maximum magnitude of each zone.(5) Seismic parameters of each zone.

(6) Backup data (or interpretation) on the given figures.Attenuation (strong ground motion)(1) Equations and their weights.(2) Background.

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able 1round motion attenuation models

round motion measure Model Descriptio

eak ground acceleration Baag et al. (1998) South KorToro et al. (1997) Central anXu et al. (1984) North Chi

pectral acceleration Toro et al. (1997) Central anBaag et al. (1998) South KorAtkinson and Boore (1995) Eastern No

Fig. 1. Example seismic source map used for the PSHA.

Fig. 1 shows one of the seismic source maps which wassed for the evaluation of the seismicity by the expert team, andable 1 lists the six attenuation equations for the peak groundcceleration and the spectral acceleration recommended by thexpert.

Using the input data proposed by the expert teams, the PSHAas performed for the example NPP site. Fig. 2 shows the seis-ic hazard curves for the site.

.3. Scenario earthquakes for a Korean NPP site

The scenario earthquakes, which are specified in terms ofhe magnitude and the distance from the site under considera-

ion, can be obtained by a de-aggregation of the PSHA resultsccording to the USNRC Regulatory Guide 1.165 procedureummarized in Section 2.1. In this case study, the seismic hazardas de-aggregated to determine the dominant magnitudes and

n Minimum distance (km) Weight

ea 0 0.5d Eastern North America 0 0.3na 0 0.2

d Eastern North America 0 0.5ea 0 0.3rth America 0 0.2

Y.-S. Choun et al. / Nuclear Engineering and Design 238 (2008) 1410–1420 1413

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Fig. 3. Contribution factors for 1 Hz.

Fig. 2. Seismic hazard curves for the example plant site.

istances at 1, 5 and 10 Hz at the median 1.0E−5 annual proba-ility of exceedance level because the ground motion attenuationquations proposed by the experts did not include an equationor 2.5 Hz. The fractional contribution of the magnitude andistance bin to the total hazard for 1 Hz was used for the devel-pment of a low frequency scenario earthquake. Because theontribution of the distance bins greater than 100 km containedess than 5% of the total hazard for the 1 Hz, additional calcu-ations to consider the effects of distant and larger events wereot needed.

The contribution of the magnitude–distance bins for 1 Hznd the average of the 5 and 10 Hz are shown in Figs. 3 and 4,espectively. The characteristics of the scenario earthquakeshich were determined based on their contributions are shown inable 2. The magnitudes and distances of the two scenario earth-uakes are very similar. It may be due to the small contributionf the distant earthquakes for the 1 Hz scenario earthquake.

.4. Response spectra for the scenario earthquakes

The spectral shape for the scenario earthquakes were devel-ped by using the attenuation relationships derived by Toro etl. (1997), those by Baag et al. (1998) and those by Atkinsonnd Boore (1995) as shown in Table 1. Fig. 5 shows the spec-ral shapes for the scenario earthquakes normalized to 0.2 gPA (zero period acceleration) together with the design based

esponse spectrum proposed in the USNRC Regulatory Guide.60 (1973) and the site-specific response spectrum to compareheir spectral shapes.

able 2robability based scenario earthquakes

requency (Hz) Scenario earthquake Remarks

Magnitude Distance (km)

M6.4 9.0 Scenario I–10 M6.2 13.0 Scenario II

Fig. 4. Contribution factors for an average of 5 and 10 Hz.

1414 Y.-S. Choun et al. / Nuclear Engineering

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Mean ground response spectra obtained from 270 earthquakeecords with magnitudes of 3–5 which occurred in the Koreaneninsula are shown in Fig. 6. It is found that the spectral shapesor the scenario earthquakes shown in Fig. 5 are very similar tohe mean response spectrum developed from the real earthquakeata. This shows that the three attenuation relationships used inhis paper reflect the ground motion attenuation characteristicsnd the site soil conditions relatively well.

Spectral accelerations for the response spectra of the scenarioarthquakes are smaller than those for the NRC design responsepectrum in the frequency range lower than about 12 Hz, whilereater in the frequency range higher than about 12 Hz. Thiseans that even though the SSCs were designed with a sufficient

eismic margin for the NRC design response spectrum, the SSCs

ith a natural frequency higher than 12 Hz become unsafe under

he scenario earthquakes. Conversely, the SSCs with a naturalrequency lower than 12 Hz become more conservative underhe scenario earthquakes than under the design based spectrum.

pK

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Fig. 6. Mean response spectra for the earthquake records for the K

and Design 238 (2008) 1410–1420

imilarly, spectral accelerations for the response spectra of thecenario earthquakes are smaller than those for the site-specificesponse spectrum in the frequency range lower than about0 Hz, while greater in the frequency range higher than about0 Hz. Since there is a considerable amount of equipment andomponents with a high natural frequency greater than 10 Hz inPPs, their seismic safety may be threatened under the scenario

arthquakes. Considering this circumstance, therefore, a seismiceevaluation for the safety-related equipment and components athe example NPP should be conducted and, if necessary, some

odification should be made to the vulnerable equipment andomponents to increase their seismic resistance capacities ando ensure the seismic safety of the plant.

. Contribution of the safety-related equipment andomponents to a core damage

Since many of the safety-related equipment and componentsave a high natural frequency, a considerable difference betweenhe response spectra of the scenario earthquakes and the designesponse spectra in the high frequency range may threaten theireismic safety. To cope with this problem appropriately, first ofll, it is important to elucidate the effect of the failure of eachafety-related equipment or component on the CDF of a plantnder a seismic event.

.1. Evaluation methodology for a CDF

This study used the computer program EQESRA (1995)or evaluating the CDF of a NPP. EQESRA was developedo evaluate the probability distribution of a system failure fre-uency from information about component fragilities (seismic oron-seismic failures). The program performs component com-inations in accordance with the Boolean expressions, and thenonvolves the system fragility with the seismic hazard to yieldprobability distribution for a failure frequency. The EQESRA

rogram uses the methodology proposed by Kaplan (1981) andaplan and Lin (1987).The methodology for evaluating the seismic fragility of

structure or equipment, which is defined as a conditional

orean Peninsula. (a) NS component and (b) EW component.

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Table 3Occurrence frequency and CDF for the seismic-initiating events

Initiating event Occurrence frequency CDF (ry−1)

Loss of essential power 3.68E−06 3.68E−06Loss of secondary heat removal 1.16E−06 1.16E−06Loss of component cooling

water/essential chilled water2.48E−06 5.25E−08

Small LOCA 3.82E−08 3.82E−08Loss of offsite power 1.12E−04 1.20E−06G

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robability of its failure at a given value of a peak ground acceler-tion, is described by Kennedy and Ravindra (1984). The groundcceleration capacity is modeled as

= AmεRεU (1)

n which Am is the median ground acceleration capacity, εR andU are the random variables with unit medians representing annherent randomness about the median and the uncertainty of a

edian value, respectively. In this model, it is assumed that bothR and εU are lognormally distributed with logarithmic standardeviations, βR and βU, respectively.

With perfect knowledge, the conditional probability of fail-re, f0, for a given peak ground acceleration level, a, is giveny:

0 = φ

[ln(a/Am)

βR

](2)

here φ[·] is the standard Gaussian cumulative distribution func-ion.

When the modeling uncertainty βU is included, the fragilityan be represented by a subjective probability density functiont each acceleration value as given by

′ = φ

[ln(a/Am) + βUφ−1(Q)

βR

](3)

here Q = P[f < f′|a] is the subjective probability (confidence)hat the conditional probability of failure, f, is less than f′ for

peak ground acceleration a and φ−1[·] is the inverse of thetandard Gaussian cumulative distribution function.

The accident-sequence definition and system modelingnclude the identification of accident initiating events, seis-

ic component failures, random failures, human errors andependent failure mechanisms that could cause these accidentequences to occur. Event and fault trees are constructed toescribe the accident sequences in the form of Boolean expres-

3

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able 4quipment and components selected through the fragility and screening analysis

quipment/component Natural frequency(Hz)

Am (g) βR

ffsite power – 0.30 0.22iesel generator >33 1.13 0.36ssential chilled water compression tank >33 1.00 0.35attery charger 11.5 1.03 0.28

1.54 0.33ondensate storage tank 9.95 0.91 0.21ssential chilled water chiller 8 1.08 0.28egulating transformer 9.9 1.30 0.33ssential service water pump 33.98 1.20 0.29omponent cooling water surge tank 17.6 2.00 0.41.16 kV switchgear 6 1.33 0.33nverter 13.8 1.37 0.33attery rack 23 1.46 0.3380 V load center 5.5 1.50 0.32witches – 2.33 0.41nstrumentation tube (primary system) – 1.50 0.3025 V DC control center 8 1.58 0.33VAC ducting and supports – 2.06 0.32ssential chilled water pump 37.2 1.85 0.36

eneral transient 2.79E−03 8.73E−07

otal 6.96E−06

ions. The family of plant level fragility curves is evaluatedy combining the component fragility curves according to theoolean expression for an accident sequence.

Each of the n plant level fragility curves for an accidentequence are convolved with each of the m seismic hazard curvesor the site; the convolution is expressed by the

∞

0

−dH(a)

daS(a) da (4)

here −dH(a)/da is the frequency with which earthquakes occurn the size range da about a and S(a) is the conditional probabilityf accident sequence.

Assuming that pi (i = 1, 2, . . ., n) and hk (k = 1, 2, . . ., m) arehe probabilities associated with n plant level fragility curvesnd m seismic hazard curves, respectively, n × m convolutionsre performed, resulting in n × m frequencies of a failure, withssociated probabilities pihk. This assumes that the uncertaintyn a seismic hazard is independent of an uncertainty in fragility.

.2. Seismic-initiated events and core damage

In the event of a strong earthquake, the failure of a safetyystem in a NPP brings the plant to a safe shutdown condition.

βU HCLPF(g)

Failure mode Related initiating event

0.20 0.15 Functional failure Loss of offsite power0.30 0.38 Concrete coning Loss of essential power0.20 0.40 Anchorage Loss of essential chilled water0.28 0.41 Functional failure Loss of essential power0.33 0.52 Structural failure Loss of essential power0.27 0.41 Sliding Loss of secondary heat removal0.27 0.44 Structural failure Loss of essential chilled water0.30 0.46 Functional failure Loss of essential power0.28 0.47 Anchorage Loss of component cooling water0.47 0.47 Concrete coning Loss of component cooling water0.29 0.48 Functional failure Loss of essential power0.30 0.49 Functional failure Loss of essential power0.31 0.51 Structural failure Loss of essential power0.29 0.54 Functional failure Loss of essential power0.45 0.55 Functional failure Loss of essential power0.30 0.56 Piping break Small LOCA0.29 0.57 Structural failure Loss of essential power0.41 0.62 Functional failure Loss of essential power0.27 0.65 Heavy duty bolt Loss of essential chilled water

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ince the shutdown is caused through several paths of the sys-em, the possible paths that a plant would follow are identified.hese paths involve a seismic-initiated event that causes a shut-own, and a success or failure designation for the plant systemsffecting the course of the events. Typically, the minimum set ofnitiating events includes both loss of coolant accidents (LOCA)nd transient events. In addition, the site-specific failure events,hich act as initiating events, may be added to a minimum set.For the Yonggwang Nuclear Units 5 and 6 in Korea, the

eismic-initiated events include the following six events, byonsidering the results of a seismic fragility analysis for theirtructures and equipment, and a result of the evaluation of aelay chatter and its effect analysis (KEPCO, 2001). The seismic-nduced medium and large LOCAs were excluded through thereliminary evaluation for the initiating events.

Loss of essential power (LEP).Loss of secondary heat removal (LHR).Loss of component cooling water/essential chilled water(LOCCW).Small loss of coolant accidents (SLOCA).Loss of offsite power (LOOP).Seismic induced general transient (GTRN).

In computing the frequency of the initiating events, a hier-rchy between them must be established. The order of thisierarchy is defined such that, if one initiating event occurs, theccurrence of other initiating events further down the hierarchyas no significance in terms of a plant’s response. The seismicvent trees should be taken directly from those developed forhe internal events analysis, with modifications to include anyeismically induced failures.

The occurrence frequencies for the initiating events are cal-ulated as

Seismic CDF for Younggwang Units 5 and 6 is calculated as.96E−06 per reactor year (ry) as shown in Table 3. The loss ofssential power is a governing initiating event in calculating theotal CDF. The seismic CDF due to the loss of an essential powerccupies more than half of the total value. The loss of an essentialower, loss of a secondary heat removal, and a small LOCAirectly induce a core damage, whereas the loss of a componentooling water/essential chilled water, loss of an offsite powernd a general transient are coupled to the secondary event trees.

.3. Safety-related equipment and components important

or a seismic risk

Component seismic fragilities are obtained from a data basef generic fragility functions for seismically induced failures

eFt

and Design 238 (2008) 1410–1420

(5)

r developed on a plant-specific basis for components not fit-ing the generic component descriptions. Fragility functions forhe generic categories are developed based on a combinationf experimental data, design analysis reports and an extensivexpert opinion survey. A generic fragility for any particularomponent can be estimated by selecting a set of site-specificragilities for that component.

For Younggwang Nuclear Units 5 and 6, after the seismicazard and seismic fragility analysis, a screening analysis waserformed to identify the structures, components and equipmentmportant for the seismic risk of the plants, and to minimize theumber of calculations for the significant contributors to theeismic CDF. The result of the seismic fragility analysis for theite-specific spectrum shown in Fig. 5 included 7 civil structuresnd 67 items for the equipment and components whose fail-re during seismic events might conceivably affect the safetyhutdown function of the plants. The screening criterion wasased on HCLPF (high confidence of a low probability of fail-re) value, which has a 95% confidence of not exceeding a 5%robability of producing a failure and indicates the seismic resis-ance of the equipment in terms of a gravitational accelerationalculated by

CLPF(g) = Am(g) × exp[−1.65(βR + βU)] (6)

here Am is the median seismic capacity and βR and βU arehe lognormal standard deviation for the randomness and uncer-ainty, respectively.

For the screening analysis, the HCLPF value of 0.65 g waselected as a screening criterion under the assumption that theffects of seismic-induced failures or the combined effects ofheir failures with a loss of offsite power directly lead to coreamage. This corresponds to a ground acceleration with a meanrequency higher than 1.0E−5 per year. Thus, this criterion

liminates all the equipment and components that can contributeo more than 5.0E−7 per year to the seismic CDF with a con-dence level of 95%. As a result of the screening analysis,ighteen equipment and component items, which have HCLPFalues lower than 0.65 g, were finally selected from the resultsf the site-specific fragility analysis. Table 4 lists the selectedquipment and component items, together with their naturalrequencies, median fragilities, uncertainty parameters, HCLPFalues, failure modes and related initiating events. The offsiteower, switches, instrumentation tube and HVAC ducting andupports listed in Table 4 used the generic values.

The comparison between the response spectra of the scenarioarthquakes and the site-specific response spectrum, shown inig. 5, reveals that there may be a change in the fragility value of

he equipment and components with a natural frequency between

Y.-S. Choun et al. / Nuclear Engineering and Design 238 (2008) 1410–1420 1417

ent a

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Fig. 7. Contribution of the equipm

0 and 50 Hz. This paper excludes a detailed evaluation of theffect of the scenario earthquakes on the fragility of the SSCs.

.4. Contribution of the safety-related equipment andomponents to a core damage

Fig. 7 shows the contribution of a failure of the equipmentnd component item to a seismic core damage in Youngg-ang Units 5 and 6. It is found that the high contributors toseismic core damage of the plant are the diesel generator

29.8%), offsite power (18.3%) and condensate storage tank17.7%).

. Determination of an effective seismic capacity

As shown in Fig. 5, the spectral accelerations for the site-pecific response spectrum are smaller than those for the

ig. 8. CDF with an increase of the seismic capacity of the selected equipmentnd components.

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esponse spectra of the scenario earthquakes in the frequencyange between 10 and 50 Hz. In a severe case, the scenario earth-uakes may threaten the seismic safety of the equipment andomponents with a high natural frequency greater than 10 Hz.hus, in order to improve the seismic safety of the plant under

his circumstance, a seismic reevaluation for the safety-relatedquipment and components should be conducted and, if nec-ssary, some modification should be made to the vulnerablequipment and components to increase their seismic resistanceapacities.

.1. Effect of the seismic capacity of the equipment andomponents on a seismic core damage

To investigate the effect of the seismic capacity of thequipment or components on the frequency of a seismic coreamage of the plant, a seismic PSA is performed by using dif-

erent seismic capacities of the four selected equipment andomponents—diesel generator, offsite power, condensate stor-ge tank and battery rack. Although the contribution of theatter rack to a core damage is not high, it is also considered

able 5ecrease of the CDF with an increase of the equipment seismic capacity

quipment Increase ratioof seismiccapacity (%)

CDF (ry−1) CDF decreaseratio (%)

iesel generator 25 5.83E−06 16.250 5.41E−06 22.3

ffsite power 25 6.36E−06 8.650 5.93E−06 14.8

ondensate storage tank 25 6.02E−06 13.550 5.86E−06 15.8

attery rack 25 6.78E−06 2.650 6.74E−06 3.2

ll 25 3.77E−06 45.850 2.47E−06 64.575 2.30E−06 67.0

1418 Y.-S. Choun et al. / Nuclear Engineering

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ig. 9. CDF ratios with an increase of the seismic capacity of the selectedquipment and components.

n the investigation because it has a natural frequency of 23 Hznd its fragility may be considerably affected by the scenarioarthquakes.

Fig. 8 shows the relations between the cumulative mean fre-uency of the failure and the peak ground acceleration for theelected equipment items with different increased ratios of 25,0 and 75%, and Table 5 summarizes the CDF and their ratios tohe original value 6.96E−06 ry−1 shown in Table 3. It is foundrom Fig. 8 and Table 5 that the failure of the diesel generatoras a greater influence on the CDF than the other equipment.n other words, increasing the seismic capacity of the dieselenerator can improve the seismic safety of the plant consider-bly. As shown in Table 5, when the diesel generator has a 25

nd 50% increased seismic capacity, the CDF will decrease by6.2 and 22.3%, respectively. This indicates that an increase ofore than 25% in the seismic capacity of the diesel generator

an improve the seismic safety of the plant by more than 16%.

ded

able 6ariation of the CDF for different HCLPF values of the equipment and components

edian seismic capacity (g) Diesel generator Offsite power

HCLPF (g) CDF (ry−1) HCLPF (g) CDF (ry−

.05 0.02 2.79E−03 0.03 3.58E−0

.1 0.03 1.67E−03 0.05 2.38E−0

.3 0.10 1.49E−04 0.15 6.98E−0

.5 0.17 3.94E−05 0.25 6.04E−0

.7 0.24 1.65E−05 0.35 5.86E−0

.0 0.34 8.13E−06 0.50 5.82E−0

.2 0.40 6.59E−06 0.60 5.82E−0

.4 0.47 5.95E−06 0.70 5.82E−0

.6 0.54 5.65E−06 0.80 5.82E−0

.8 0.61 5.52E−06 0.90 5.82E−0

.0 0.67 5.45E−06 1.00 5.82E−0

.5 0.84 5.39E−06 1.25 5.82E−0

.0 1.01 5.38E−06 1.50 5.82E−0

and Design 238 (2008) 1410–1420

n addition, when the condensate storage tank has an increasedeismic capacity, the CDF will decrease by more than 10%. Theffect of the seismic capacity of the battery rack on the CDFs not considerable. When all of the selected equipment itemsave a 25, 50 and 75% increased seismic capacity, the CDF willecrease by 45.8, 64.5 and 67%, respectively.

Fig. 9 plots the ratios of the CDF for the equipment andomponents with an increased seismic capacity to that withn original capacity according to the peak ground accelera-ion. It is found that the ratios of the CDF are considerablynfluenced by the value of the peak ground acceleration, upo 1.0 g. The effect of the seismic capacity of the equipmentn the CDF is remarkable in the PGA range of 0.2–0.6 g. Ifhe seismic capacities of all the selected equipment items aremproved, the CDF may be decreased by about 5 and 30% at 0.2nd 0.3g, respectively. At 0.4 g, increasing the seismic capac-ty of the offsite power will be more effective, and, under 0.6 g,ncreasing both the seismic capacities of the offsite power andhe diesel generator will be more effective. In the case of theffsite power, at 0.4 g, an increase of its seismic capacity of5 and 50% leads to a reduction of 33 and 45% in the CDF,espectively.

.2. Effective seismic capacity of the selected equipmentnd components

Fig. 10 shows the relations between the HCLPF of the equip-ent and the CDF according to the median value of the seismic

apacity and Table 6 summarizes the CDF for the differentCLPF values of the selected equipment and components. Fromig. 10 and Table 6, the effective HCLPF values for the dieselenerator, offsite power, condensate storage tank and batteryack are determined as 0.84, 0.35, 0.63 and 0.63 g, respec-ively. For a larger HCLPF value than the effective value, even

ecrease in the CDF. In the case that all the four selectedquipment items have an increased seismic capacity, the CDFecreases to 2.30E−06 ry−1 from 6.96E−06 ry−1.

Condensate storage tank Battery rack All

1) HCLPF (g) CDF (ry−1) HCLPF (g) CDF (ry−1) CDF (ry−1)

5 0.02 2.89E−03 0.02 2.82E−03 2.91E−035 0.05 1.68E−03 0.03 1.66E−03 2.39E−036 0.14 1.14E−04 0.10 1.39E−04 2.33E−046 0.23 2.78E−05 0.17 3.65E−05 6.14E−056 0.32 1.10E−05 0.24 1.58E−05 2.34E−056 0.45 6.44E−06 0.35 8.67E−06 8.10E−066 0.54 5.99E−06 0.42 7.50E−06 5.00E−066 0.63 5.88E−06 0.49 7.06E−06 3.76E−066 0.72 5.86E−06 0.56 7.02E−06 3.01E−066 0.82 5.86E−06 0.63 6.79E−06 2.67E−066 0.91 5.85E−06 0.70 6.76E−06 2.48E−066 1.13 5.85E−06 0.87 6.73E−06 2.33E−066 1.36 5.85E−06 1.04 6.72E−06 2.30E−06

Y.-S. Choun et al. / Nuclear Engineering and Design 238 (2008) 1410–1420 1419

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ig. 10. Relationship between the HCLPF values of the equipment and compond) battery rack and (e) all.

. Conclusions

This paper evaluated the effects of the seismic capacity ofafety-related equipment on the CDF of an existing NPP whichhould be reevaluated for the modified seismic design spectrand suggests the effective seismic capacities of the selectedafety-related equipment and components by using a probabilis-ic safety assessment.

For the NPP considered in this case study, the failures of theiesel generator, offsite power, condensate storage tank and bat-ery rack contribute remarkably to the CDF. When the dieselenerator or condensate storage tank has an increased seismic

apacity, the CDF will be decreased considerably, while in thease of the battery rack the CDF does not decrease signifi-antly. Increasing the seismic capacities of the diesel generatory more than 25% can improve the seismic safety of the plant

cpaa

nd the CDF. (a) Diesel generator, (b) offsite power, (c) condensate storage tank,

y more than 16%. In the case of increasing the seismic capac-ties of the equipment which reveals a high contribution to aore damage, the CDF may be decreased by more than 50%.he effective HCLPF values for the diesel generator, offsiteower, condensate storage tank and battery rack were deter-ined as 0.84, 0.35, 0.63 and 0.63 g, respectively. In the case that

ll four selected equipment and components have an increasedeismic capacity, the CDF decreases to 2.30E−06 ry−1 from.96E−06 ry−1.

The seismic safety of existing NPPs can be improved easilyy an increase of the seismic capacity of the dominant equip-ent or components. Since the relationship between the seismic

haracteristics of the facilities and the seismic risk of a nuclearlant can be obtained by using a probabilistic safety assessment,n improvement of the seismic safety of existing NPPs can bechieved effectively.

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420 Y.-S. Choun et al. / Nuclear Enginee

cknowledgement

This study was supported by the Ministry of Science andechnology, Korean government, through its National Nuclearechnology Program.

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