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SEP 121983 T
MEMORANDUM FOR: Robert Jackson, Chief Geosciences Branch Division of Engineering FROM:
George Lear, Chief Structural and Geotechnical Engineering Branch Division of Engineering
SUBJECT:
LIMERICK GENERATING STATION - SEVERE ACCIDENT RISK ASSESSMENT (LGS - SARA)
Reference:
- 1) " Limerick Generating Station - Severe Accident Risk Assessment" - Philadelphia Electric Company Report No. 4161, April 1983
- 2) Memo for Reiter from Lear,
Subject:
" Severe Accident Risk Assessment (SARA) Limerick Generating Station", June 23, 1983
- 3) Letter to Schwencer Chief, NRC Division of Licensing, from Kemper, Philadelphia Electric Company,
Subject:
" Limerick Generating Station Units 1 and 2 Responses to NRC Ouestions on the SevereAccidentRiskAssessment(SARA)",
August 24, 1983
- 4) Letter to Schwencer, Chief, NRC Division of Licensin Company,g from Kemper, Philadelphia Electric
Subject:
" Limerick Generating Station Units 1 and 2 Response to NRC Questions on the Severe Accident Risk Assessment (SARA)",
August 29, 1983 5)
"A Preliminary Review of the L!merick Generating Station Severo Accident Risk Assessment", Oraft Report prepared by Brookhaven National Lab (SNL),
August 15, 1983 3I [(
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. Robert Jackson y gg j
We have completed our preliminary review and evaluation of the LGS -
SARA Report submitted by the applicant (Reference 1) and of the applicant's responses to NRC questions regarding that report (References 2, 3 and 4). In addition, we have also relied upon the review findings of Reference 5.
Our review and evaluation has been confined to the evaluation of geotechnical and structural engineering aspects of the analysis of accident sequences resulting from seismic events. The results of this review and evaluation are enclosed. The review of the geotechnical aspe:t was conducted by Dr. J. R. Pearring and the review of structural aspects was conducted by Dr. N. C. Chokshi of this Branch.
Significant conc)ssions of our preliminary review and evaluation include:
a) There are no open items associated with our review of the gaotechnical engineering aspects of the LG5 - SARA. Although the LG5 - SARA did not explicitly address geotechnical engineer < ng parameters impacting on core melt frequency, our evaluation of the LG5 - SARA indicates that the methodology used by the applicant for the most part adequately enveloped important geotechnical engineering aspects.
b) Due to uncertainties associated with geotechnical engineering related parameters, consideration should be given to verifying that relative building displacements, embedmont conditions, potential soil amplification, and potential slope instability will not significantly impact on the appitcant's estimated frequency of core melt at seismic levels greater than the $$E.
c) For structural engineering considerations, we concur, in general, with the reconmendations of Reference 5.
The methodology used in the LG5 - SARA is the state-of-the-art approach and similar to the one used in other PRAs. We emphasize that the methodology has not been validated and relies upon engineering judgment and, therefore, cannot be used for absolute risk prediction. Other conclusions are contained in Section 4.0 of the enclosed report, f
,/3NM George L4ek, Chief Structural and Geotechnical En Divi.gineering Branchsion of Engineering
Enclosures:
As stated I
cc:
J. Knight P. T. Kuo R. Jackson J. Pearring D. Jeng N. Chokshi L. Heller
a _,. -
7 e-Preliminary Review of Geotechnical Engineering Related Aspect of the Limerick Units 1 and 2 Severe Accident Risk Assessment I.
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~ The following summarizes the NRC staff's preliminary review of geotechnical engineering aspects of the Limerick Generating Station Units 1 and 2 Severe Accident Risk Assessment (LGS-SARA). Specific LGS-SARA elements reviewed include a) the applicant's consideration of geotechnical engineering re-lated potential failure mechanisms resulting from a seismic event, b) the applicant treatment of the geotechnical engineering aspects of the. seismic hazard analysis, and c) the applicant's treatment of geotechnical engineer-ing parameters affecting fragility analysis. The review was conducted in accordance with the applicable general guidance contained in NUREG-2300 Chapter 10, 11, and 12 (Ref. 1). This review also considered appropriate elements of the draft report of the preliminary review of the LGS-SARA performed by the NRC's consultant, the Brookhaven National Laboratory (Refer. 2).
II. Geotechnical Enaineerina Related Site Data 1.
General Site Description - Topography of the Limerick site area consists of gently rolling ridges dissected by the courses of the Schuylkill River and its tributaries. The main plant structures are on a broad ridge approximately 100 feet air.ie the river. The plant site is divided into three main subareas:
(1) the reactor / turbine area at grade elevation 217 feet as1, (2) the cooling tower area at grade elevation 257 to 265 feet asl, and (~s) the spray pond area with bottom of pond at elevation 241 feet as) and a normal still pond level of 251 feet as1. Ground water in the site area decreases from approximately 250 ft asi northeast of the spray pond area to an 09/02/83 1
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elevation of less than 120 feet asi southwest of the reactor / turbine area.
Detailed descriptions of the geotechnical engineering aspects of the
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general site area are presented in Limerick FSAR Sections 2.5.4 and 2.5.5 (Ref. 3).
The bedrock in the site area consists of interbedded red sandstones, siltstones, and shales indurated to a depth of several thousand' feet, which are moderately to closely jointed. Bedrock in the immediate plant area dips 8 to 20 degrees to the north. The soils at the site consist of red sandy and clayey silts with rock fragments derived from weathering of the underlying bedrock. Soil thickness ranges from 0 to 40 feet, averaging 10 to 15 feet. For the most part, soils below a depth of 10 feet consist of highly weathered and fractured rock with intermixed silts and clays.
The main seismic Category I plant structures including the reactor enclosure, control structure,-diesel generator enclosure, spray pond pamp house, spray pond spray network, turbine enclosure and radwaste enclosure are founded on unweathered bedrock. Seismic Category I facilities not founded completely on bedrock-are founded totally or in part on natural soil or manmade fill and include the diesel fuel oil storage tanks, buried cooling water piping, a pipe valve pit, electrical ducts and the northwestern portion of the spray pond.
2.
Properties of Site Subsurface Materials The Limerick site investigation program, which was accomplished -
between 1969 and 1977, was accomplished in a deterministic manner.
.The applicant has reported in the PSAR (Ref. 4) and FSAR (Ref. 3) that site exploratory investigations included 380 borings, 17 test pits, and 10 seismic refraction traverse lines, totalling 5180 linear feet.
In addition, a surface shear wave velocity survey, a seismic uphole survey, inhole permeability testing, plate bearing testing, micromotion measurement testing, and in situ bedrock stress testing e
were accomplished in the site vicinity.
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l As sampling theory was not used in the development of the' site l
investigation program and 'as the number of'gamples of rock and soil materials tested would be considered small in a statistical sense, the NRC staff concludes that the application of probability theory to determine statistically significant estimates of means, variance, and probability density functions for pertinent soil and rock prop-arties at the Limerick site is n'either justified nor appropriate for use in the LGS-SARA analyses. Therefore, it is the conclusion of the NRC staff that the use of deterministically estimated repre-sentative rock and soil material properties values in the LGS-SARA is acceptable.
III. Failure Mechanisms In the LGS-SARA, the applicant addressed earthquake-induced acceleration as the potential failure mechanism capable of producing structural and component failures at the Limerick site. Other potential failure mech-i anisms, including subsidence and acceleration-induced liquefaction and settlements were not explicitly considered.
1.
Subsidence - The NRC staff review of site data contained in the Limerick PSAR (Ref. 4) and FSAR (Ref. 3) presented no evidence of zones of solutioning, caverns, or highly weathered areas in the foundation bedrock or soils which would allow significant subsidence under any proposed seismic loading. The NRC staff therefore concurs with the applicant's exclusion of subsidence as.a probable failure mechanism requiring consideration in the LGS-SARA.
2.
Liquefaction - The probability of failure of structures, systems, and components under seismic loading conditions due to liquefaction of foundation and backfill soil was not explicitly addressed in the applicant's report. Based on the site data presented by the appli-cant in the FSAR,-the NRC staff independently analyzed the liquefac-tion potential of the natural residual soils and backfill materials at the Limerick site and finds the following:
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The natural residual soils at the site evidence an average SPT a.
resistance of 46 blows per foot, exhitpit cohesive characteris-
. tics (as evidenced by an average plasticity index of 8), and fa general can be considered to have a low potential for saturation 1
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due to the relatively severe water table gradient at the site (Refs. 4 and 3).
The NRC staff considers that such soils hav2 a negligible potential for' liquefaction.
b.
The applicant reported in the FSAR that all fills associated with seismic Category I structures and piping were classified as either mass concrete fill, cementitious backffil, select granular backfill, or Type 1 random fill. The mass concrete fill and the cementitious backfill were batched to attain a 28-day compressive strength of 2000 psi and 80 psi, respectively.
Such materials are not capable of liquefaction. The select granular backffll material. consists of 3/4-inch maximum size aggregate with less than 10% by weight passing a No. 200 sieve and was compacted to 95% AASHTO-T-180 maximum dry density. The Type I random fill consists of 8-inch maximum size broken rock 7
graded course to fine and was compacted to 90% AACHTO T-180 maximum dry density (Refs. 4 and 3).
Due to the relatively dense nature'of the select backfill and Type I random fill material, the NRC staff considers these backfill materials would not be susceptible to liquefaction under the seismic loading conditions postulated for the Limerick site.
The NRC staff therefore concludes that the potential for liquefac-tion of the Limerick site soil and backfill material due to postu-lated seismic loadings may be neglected without significantly influ-encing the overall LGS-SARA results.
3.
Settlements - The applicant has not explicitly analyzed rock and soil settlement and differential settlement as potential failure mechanisms in the LGS-SARA.
In the Limerick FSAR the applicant has
- deterministically estimated a maximum settlement of the reactor build-ing, the heaviest structure at the site, to be on the order of one I
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i quarter of an inch or less due to pseudo-elastic compression of the
-rock occurring upon application of loading',nith construction (Ref. 3).
The NRC staff has independently verified these findings ~using the 1
procedures in Reference 5 and has further considered the potential for total settlement under possible seismic loadings.
Considering pseudo-linear-elastic response and input seismic ~ accelerations up to 4 times the SSE, the NRC-staff has deterministically estimated an upper bound total rock deformation of less than 0.5 inches. The NRC i-staff therefore considers that there is a negligible low likelihood of rock settlement becoming a failure mechanism that would require i
assessment. Consideration should however be given to verifying that potential failure of safety-related piping and of small lines attached to safety-related piping near the junction of the containment building
-and the reactor enclosure due to impact or relative displacement of the' buildings will not contribute to the frequency of core melt (Ref. 2).
The NRC staff has also analyzed the potential for settlement of residual soils and backfill materials due to a seismic event. Using the procedures of References 6 and 7, the staff concludes that the maximum upper-bound settlement of soils and backfill materials support-ing seismic Category structures systems on components would be ex-pected to be less than 1.0 inch for seismic loadings up to 4 times the design SSE. The NRC staff considers that there is very little likelihood that soil' settlement or differential settlements due to
' seismic events of this magnitude could become a significant failure mechanism for. structural systems and components founded on soils.
The NRC staff therefore concurs with the applicant's exclusion of settlements and differential settlements as significant potential j
failure mechanisms requiring detailed analyses in the LGS-SARA.
-IV.
Seismic Hazard Analysis 4
The procedures used by the applicant in the development of the Limerick seismic hazard model do not explicitly consider the specific Limerick site 09/02/83 5
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soil and. rock engineering properties.
In his analysis the applicant used an attenuation. function to estimate peak ground.dccelerations that was developed from an analysis _of existing recorded strong motion data. The function waJ cerived from a regression of peak ground acceleration against magnitude and distance and assumes a regionally constant anelastic atten-uation factor (Q).
Local site characteristics relating to the geometry and engineering characteristics of the near surface soil and rock materials associated with the strong motion records were not separately accounted for in the regression. Uncertainties in the peak ground accelerations 4
predicted through the use of this function attributable to local site conditions are therefore combined with those associated with source and propagation path effects.
In applying the developed attenuation function to the Limerick site the applicant assumed a lognormal distribution of teceleration about the mean value with a standard deviation of 0.6 selected as typical of the scatter associated with strong motion data sets from a specific geological region. This standard deviation value corresponds to a factor of 1.8 times the median value.
There are no geotechnical engineering related local site features known to the NRC staff which would preclude considering this site to be within the geologic regional ~ average for which the applicant-developed attenua-tion equation is intended to apply. The NRC staff recognizes that the physical processes affecting local site response are not well understood.
The NRC staff is also aware that large uncertainties due to source and path parameters as well as local site-specific geotechnical engineering related parameters are already reflected in the variance in the data used to estimate peak ground acceleration. The staff therefore considers that a vigorous analysis of the influence of local soil and rock property param-t-
eters on the attenuation of acceleration at the Limerick site is not warranted nor apprapriate in keeping with the general level of the state i
of the art in predicting ground motion in the eastern U.S.
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2 V.
Fracility Analysis a.
1.
Structures, Systems, and' Components Founded on Rock In the LGS-SARA the seismic fragility of structures, systems, and components founded on rock were described in terms of the median ground acceleration capacity asiociated with seismic-induced failure and_of the logarithmic st.% rd deviation of this median value. As an aid to computation, the applicant : sed an interrediate variable called t'.e " median factor of safe ~ty."
It was defined as the ratio of the estimated median ground acceleration capacity causing failure to tha Safe Shutdown Earthquake (SSE) acceleration used in the design analysis. Thus, rather than directly estimating the seismic fragil-
'ity of structures and components founded on rock, the applicant estimated median factors of safety and logarithmic standard deviations against failure based upon the deterministically accomplished design response analysis.
In the detenninistic design 'of the Limerick Plant, two-dimensional. lumped-mass models were developed for the major seismic Category structures founded on rock. Separate models were t'
developed for the north-south, east-west, and vertical responses E
analyses for each structure.
Because of the relatively high stiff-ness of the rock the applicant treated the. foundation rock as a fixed p
boundary for the analysis of all structures excepting the primary I
containment and the reactor enclosure and control structure.
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floor response spectra developed for these two structures for equip-ment analysis purposes were based upon models considering the elastic deformation of the supporting medium. The shear modulus, shear wave L
velocity, and the density of the supporting rock used in the analysis were 1.2 x 10s psi, 6000 ft/sec., and 150 lbs/fta respectively.
(On page 4-22 of Appendix B of the LGS-SARA the reported bedrock modulus of elasticity of 7.3 x 108 psi is in error. Revision 19 to the LGS FSAR-corrected that value to 3.0 x 10s psi to reflect the value actu-ally used in design (Ref. 3)).
Embedment conditions were neglected in the design analysis. To account
.~t variations in the structural responses owing to uncertainties in thu i.>terial properties and to a9proximations associated with the mooeling techniques used in the 09/02/83 7
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design analysis, the computed floor response spectra were smoothed
'and the spectra were broadened on either side of the peak value by 15% of the frequency at which the peak occurred. Additional soil structure interaction analysis were performed to access the sensitiv-ity of the design models to variation in rock modulus. Model analysis demonstrated that for a variation in rock modulus of i 50 percent,.
variations in structural frequericies did not exceed 10 percent for predominant. modes.
In the LGS-SARA seismic fragility analysis the applicant did not explicitly consider geotechnical engineering parameters beycnd those in the deterministic design analysis. The applicant treated the uncertainty introduced into the calculated design response of struc-tures -due to variability in geotechnical engineering related parameters only by including it as an element of all randomness and uncertainty associated with soil-structure interaction effects. No uncertainty was assigned to the ground response spectrum factor used in the analysis due to variation in foundation material properties.
Lacking quantitative evidence from site specific sensitivity analyses data to estimate the variability in the median factors of safety for structural capacity due to geotechnical eng'ineering related parameters required to define the fragility curves for the plant structures, the applicant used subjective engineering judgment.
In the applicant's judgment the major plant seismic Category 1 structures are considered to be founded on competent rock and the design of the structures was con-ducted using assumptions and methods of analysis that result in small variation in frequency and response when_significantly large variations
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in the flexibility of rock and of energy dissipation in rock by.
radiation damping are considered. The applicant therefore concluded i
that the design results would have a median factor of safety of 1 based upon soil structural interaction considerations. Using'similar reasoning and considering the nature of the model used in the determi-nistic design, the applicant assigned a relatively small logarithmic standard deviation of 0.05 to the uncertainties in the median factor of safety for the overall structural acceleration capacity due to'all soil structure interaction effects.
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f The NRC staff, considers that although the deterministically derived design structural response to the SSE cannot be accepted as an "abso-lute best estimate" of a pedian value.for structural acceleration capacity when considering geotechnical engineering parameters at the Limerick site, it is an acceptable' estimate co'nsidering the level of effort and the analytical model used.
In the applicant's methodology the individual uncertainties associated with each factor bearing on the variance of the mean structural capacity are summed to obtain an overall logarithmic standard deviation using the " square root of the sum of the squares" process.
In this process, because of the number of factors considered and the relative size of the uncertainties for each factor, the addition of reasonable amounts of uncertainties to a few factors would result in' only a very small increase in the overall summation. The NRC staff therefore concludes that although the appli-cant has not incorporated the total effect of variation due to geo-technical engineering related factors into the total overall struc-tural acceleration capacity'by the procedures used, the inclusion of a reasonable additional uncertainty value for geotechnfal-parameters would only produce a small effect which would not be significant in the final product of the analysis (Ref. 2).
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The NRC staff also considers that neglecting embedment considerati,ons in the basic design tends to bias the "best estimate" of the mediac.
structural capacity to the conservative side especially at higher acceleration levels. This conservatism is acceptable to the NRC staff. However, because embedment conditions were neglected in the original deterministic design the effect of soil pressure on buried walls expressed in terms of variance of the mean factor of safety for capacity, were not explicitly addressed-in the LGS-SARA. Considera-tion should be given to evaluating the impact of embedment on the fragility of affected seismic Category I walls and any supported systems or equipment under greater than SSE loadings (Ref. 2).
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Structures, Systems, and Components not Founded on Rock Th.e applicant did not address fragility of structures, systems, and
-. components which are partially or totally founded on soils in the LGS-SARA. -Seismic Category I facilities not found completely on competent bedrock include the diesel oil tanks, underground piping, a piping value pit and electrical ducts. These structures, systems, and components have been evaluated by the applicant deterministically 1
for an SSE of 0.15g in conjunction with the Litaerick Safety Evaluation. Soil response studies were performed by the applicant using the' computer program " Shake" to estimate ground motion induced by a safe shutdown earthquake in the backffil material surrounding -
and. supporting buried seismic Category I piping system. Earthquake motion was specified at the level of the top of rock and resulting peak accelerations were computed at the level of the pipe. The sensitivity of output to variation in soil shear modulus was also considered. The applicants reported results indicate an approximate.
2 fold amplification of input acceleration results when input accelerations are equal to the SSE.
Since the response of soil to seismic input motion is nonlinearly strain dependent consideration should be given to verifying that soil supported safety-related piping, and other soil supported structures and components are not stressed to the point that they would signifi-l cantly contribute to the frequency of core melt when considered to
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be exposed to seismic events greater than the SSE.
i 3.
Spray Pond The, fragility of the seismic Category I spray pond which provides the
.ultis: ate heat sink for cooling water was not addressed by the applic-ant in the LGS-SARA.
Based upon a review of information presented by the applicant in References 3, 4, 8, and 9, the NRC staff has evalu-ated the stability of the spray pond slopes. The slopes of the ulti-
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mate heat sink spray pond were excavated partly in soil and partly in l
rock. The applicant has deterministically designed the spray pond l
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l slopes using protection riprap stone materials with stone size, layer thickness, and slope geometry governed by t,he anticiated wave condi-
'tions exp$cted during the Probable Maximum Flood (PMF).
In addition..
the applicant has deterministically analyzed the stability of the spray pond slopes to demonstrate the stability of the soil and rock'
. slopes under the design basis conditions of an SSE of 0.15g.
The NRC s,taff concluded that the scope of the applicant's field and labora-tory efforts was adequate to define the bedrock and foundation condi-tions at the spray pond site and to establish appropriate determini-stic design basis strength parameters of the slope materials. The NRC staff also found the rock and soil slopes acceptably stable under a design basis SSE of 0.15g.
.The applicant has not presented an analysis of the effects of seismic loading greater than 1 times the design SSE on the stability of the spray pond slopes and of the water holding capability of the spray
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pond'after such seismic events. However, it is the NRC staff judgement that considering the configuration of the spray pond, the topography of the site, and the geometry and strength of the rock and soil slopes, the impact would be small for acceleration levels up to 2 times the SSE. Uncertainties associated with the strain dependent non-linear response of the soil slopes of the spray pond founded above near h
surface bedrock, preclude fragility judgements when greater input
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accelerations are considered. Consideration should therefore be given to accurately defining the impact of the hypothesized exposure of the spray pond soil and rock slopes to seismic events 2 to 4 times the I
design SSE on the overall core melt frequency.
VI.
Conclusion The NRC staff review of the LGS-SARA indicates that the report of the applicant did not explicitly address geotechnical' engineering parameters
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impacting upon core melt frequency. The staff's evaluation of the LGS-SARA, however, indicates that the methodology used by the applicant in the seismic hazard and seismic fragility analysis for the most part adequately envelops geotechnical engineering parameters considering the 09/02/83 11 LIMERICK SEV ACC RISK ASSMT l
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state of -the art of the methodology and the large uncertainties associ-ated with the overall analysis.
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The NRC staff review of the LGS-SARA also found that failure mechanisms relating to geotechnical engineering parameters other than acceleration, i.e., subsidence, liquefaction, and settlements were not explicitly addressed in the LGS-SARA. Using the' site data presented by the. applicant
'in the PSAR and FSAR, the NRC staff analyzed the potential for occurrence of these failure mechanisms in the soil and rock areas of the site. The results of this analysis indicate that there is a negligibly low potential for structures, systems, and components failures due to possible effects of these mechanisms. The NRC staff therefore concurs with their exclusion from consideration in the LGS-SARA.
Based upon the NRC staff and consultants review of the information present-ed in the LGS-SARA the following items related to geotechnical engineering aspects of the review are presented _.for consideration (a) -Although settlement and differential settlements of structures are not considered to be viable failure mechanisms for the Limerick site requiring comprehensive treatment in the LGS-SARA, consideration should be given to verifying that potential failure of safety-related piping and of the small attached lines located near the junction of the containment.hsilding and the reactor enclosure caused by impact or relative displacement of the buildings will not contribute to the frequency of core melt.
l-(b) Because embedment condition were neglected in the original design, the l
effect of soil pressure on buried walls expressed in terms of variance of the median factor of safety for structural capacity was not
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explicitly addressed in the LGS-SARA. Consideration should be given to evaluating the impact of embedment on the fragility of affected walls and supporting systems.
(c) Because of soil amplification considerations related to structures founded upon soil, consideration should be given to evaluating the 09/02/83 12 LIMERICK SEV ACC RISK ASSMT l
fragility of soil supported safety-related piping and other soll supported structures at seismic levels greater than the SSE.
(d) Due to geotechnical engineering related uncertaintics associated with the capability 'of the spray pond slopes to withstand seismic loadings greater than about 2 times the SSE, consideration should be given to accurately defining the fragility of the spray pond at SSE levels greater than 2 times the SSE.
VII Reference 1.
NUREG-2300, "PRA Procedures Guide, Volume 1 and 2," U.S. Nuclear Regulatory Commission, Washington, D.C.
2.
" Draft-Preliminary Review of the Limerick Generating Station Severe Accident Risk Assessment," Brookhaven National Laboratory, August 15, 1983.
3.
Limerick Generating Station Units 1 and 2 FSAR Volumes 2 and'3.
4.
Limerick Generating Station Units 1 and 2 PSAR Volume 1.
5.
Bowles, J. E. " Foundation Analysis and Design," McGraw-Hill Book Company NY, NY 1979.
6.
Lee, Kenneth L.,
and Albaisa, Aurelio
" Earthquake Induced Settlements in Saturated Sands," Journal of the Geotechnical Engineering Division, ASCE, Vol 100, April 1974.
7.
Silver, Marshall L., and Seed,- H. Bolton, " Volume Changes in Sands During Cyclic Loading," Journal of the Soil Mechanics and Foundations Division, ASCE Vol 97, Sept. 1971.
8.
Limerick Generating Station Units 1 and 2, Philadelphia Electric Company -
Yard Work Spray Pond Drawings, Bechtel Drawings Nos. C-1103; C-1104; and C1105, Revision 12, 9/17/82.
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9.
Geotechnical Engineers Inc., " Report of Soil Testing - Limerick Nuclear
. Station Spray Pond, Winchester, Mass.," September 1974.
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e Preliminary Review of Structural Fragility Aspects of the Limerick Generating Stations Severe Accident Risk Assessment 1.0 Scope The review comments and evaluation presented here are based on the preliminary review of those portions.of.Section 3.0 and Appendix B of the Limerick Generating Station Units 1 and 2 Severe Accident Risk; Aisessment-(LGS-SARA) which are related to structural response
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and structural fragility formulation for a seismic event. The major aims of this preliminary, review are:
(a) to identify, where possible, sources of conservatism and nonconservatism; (b) form general impressions regarding the adequacy of the approach used of the findings of the LGS-SARA; (c) to identify key contributing structural components, if any; (d) compare the LGS-SARA with recent probabilistic risk assessments (PRAs) for other plants, if applicable; and (e) to gain insights regarding probable seismic-capacity of the plant beyond SSE.
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In this review, the findings of the draft report prepared by the Brookhaven National Laboratory (BNL) (Reference 1) are relied upon heavily. The earlier review report prepared by Sandia Nationa'l Laboratory (Reference 2) for the -Indian Point Probabilistic Safety Study (Reference 3) is also relied upon extensively in this review.
Additional information obtained from the licensee and its representatives in a meeting of August 5,1983 is also reflected in this review findings.
2.0 Methodolocy of LGS-SARA A brief description of the methodology used for developing structural fragilities in the LGS-SARA is given in this section.
Structural fragility data are presented in the form of fragility curves which plot the fraction of expected failures versus effective peak ground acceleration.
In Fig.1, an example of a fragility curve _is shown.
In the LGS-SARA, generally, Seismic Class I Structures are considered to fail functionally when inelastic deformations of the structure under seismic load are estimated to be sufficient to potentially interfere with the operability of safety related equipment attached to the structure.
Thus, the conditional probabilities of failure for a given free field ground acceleration for Class I structures are for operability limits and should not necessarily correspond to structural collapse.
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In order to obtain fragility curves, the approach adopted in assigning capacities (failure fraction as function of effective peak ground acceleration) for the structures was to first determine the median factor of' safety against failure and its statistical variability under the safe shutdown earthquake (SSE). Then the median effective ground acceleration causing failure was estimated by multiplying the SS,E acceleration level by this factor.
The overall safety factor was determined by evaluating the safety factors for a number of parameters, which fell into two categories:
structural capacity and structural response. Parameters influencing the factor of safety on structural capacity include the strength of the structure compared to the design stress level and the inelastic energy absorption capacity'(ductility) of a structure to carry load beyond yield.
In the LGS-SARA, an additional parameter, earthquake duration factor, is also included in computing the median factor of safety on structural capacity. The parameters in structural response for_ a given ground acceleration are made up of many factors. The most significant of these include: (1) ground motion and the associate ground response spectra for given peak free field ground acceleration, (2) energy dissipation (damping), (3) structural modeling, (4) method of analysis, (5) combination of dynamic response modes, and (6) combination of earthquake components. The derivation of each factor of safety considered variability.
In each case, a median safety factor was assigned along with a variability. When combining the median safety factors W
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e of contributing parameters, their variabilities were also combined to define the overall safety factor. From this'overall safety factor, the median effective or sustained peak ground acceleration associated with failure was determined as explained earlier.
The entire fragility curve for any structure can be expressed in terms of the best estimate of the median ground acceleration capacity and two random variables, one representing the inherent randomness of the event (B ) and the other corresponding to R
uncertainty associated with predicting response to an event (B ).
g B by definition is irreducible. For example, it is not possible, R
at least in the foreseeable future, to predict the exact time-history of an earthquake event at a given site, assuming that the occurrence of the event can be predicted. B, in a sense, g
represents a measure of our lack of knowledge for. example, the mathematical modeling of a structure to predict the responses to a seismic ev4nt. As our knowledge advances, this uncertainty can be reduced.
In Fig.1 an example of log-normally distributed fragility curve is shown. The solid curve is, effectively, the median fragility curve incorporating inherent randomness uncertainty, B. The left and R
right dashed curves represent certain percentile curves to reflect the uncertainty (B )-in the median curve.
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Note that seismically induced failure data are generally unavailable for structures. Therefore, each factor of safety and its variability and hence the final fragility curyes are developed primarily from analysis and engineering judgment supported by limited test data. Table 1 lists the key structural component with assoc-iated capacity data in terms.of median acceleration capacity and asso'ciated log-nopnal standard deviations.
.e The earthquake duration factor used in LGS-SARA has not been used, explicitly, in other PRAs. This factor, according to LGS-SARA, reflects the additional capacity due to the shoiter duration with correspondingly lower energy content and fewer stirong motion cycles present in the Limerick median expected earthquake as compared to the earthquake which would generate the number of cycles used in the determination of the median factor of safety related to the ductility of the structure.
It should be noted that the methodology used in the LGS-SARA, as in other PRAs, does not include an explicit consideration of design
[
and construction errors and, hence, may be biased (Reference 1).
3.0 Evaluation of Findings In this section, review comments are presented on both general methodology and the key structural components listed in Table 1.
I l
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f, The following seismic Category I structures were evaluated in the LGS-SARA.
Primary Containment Structures.
Reactor Enclosure and Control Structures Spray Pond Pump Structure Diesel Generator Enclosure Spray Pond It was judged in LGS-SARA-that the failure of non-Category I buildings would not affect the seismic capacities of the Category I structures and, hence, fragility evaluation were not conducted for the non-Category I structures as part of this evaluation.
It is our understanding that the selection of the critical structural components was based on the identification by NUS of system and components important to safety and a plant walk-down performed by SMA. As indicated in Reference 1, since the plant is still under construction, a systematic review of the potential for secondary components failing, falling, and impacting primary components was not undertaken. Therefore, we concur with the recomendation in Reference 1 (p. 4-7) to conduct a systematic review of this aspect after the construction of the plant is completed.
.i 3.1 ' Comments on General Methodology The methodology used in the LGS-SARA is very similar to the methodology used 16 other PRAs (e.g. Reference 3), and as such is a
. state-of.-the-art approach. However, the methodology is based on simple probabilistic models and hence, in our opinion, contains large uncertainty due to. methodology.itself. Specific comments on the methodology are a,s'.follows.
~~
~ The multiplicative model (i.e. the median of the overall (a) factor of safety is a product of the median factors of safety for each variable) proposed in the section 2 of Appendix B of the LGS-SARA requires the mathematical condition that each median factor of safety be an independent variable.
The. licensee in a meeting and in a subsequent letter (Reference 4) indicated that the use of the above model is not intended to imply that each of these variables are totally independent. The estimated influence of dependency was.
considered in developing the factors of safety and log normal standard deviations for each variable. The applicant further stated that the overall median factor of safety and associated variabilities are checked for reasonableness for each structure and mode of failure.
Considering the state-of-the-art, the methodology is reasonable when the above fact is taken into consideration and the evaluation is performed by an experienced engineer.
.3 However, it must be noted that the above methodology has not been verified by either analytical investigation or adequate test data and, therefore, contains a great deal of uncertainty.
~
(b) The explicit use of the factor of safety associated with the expected duration of the earthquake is unique to LGS-SARA.
In other published PRAs, the duration effect is accounted for by considering it to be incorporated in the concept of effective peak acceleration in the hazard estimation. 'In the LGS-SARA, the duration factor is considered independently and in addition to the use of effective peak acceleration.
Reference 1 contains a detailed discussion of this factor including its effects on the risk results. The licensee provided additional information (Reference 4) to indicate that when the three factors of safety of (effective peak acceleration, ductility, and duration are considered simultaneously, the combined median factor of safety and uncertainty values are reasonable as compared to other PRAs.
This information is currently being reviewed by the staff.
(c) As discussed earlier, the design and construction errors are not accounted for explicitly in the fragility development.
It is recognized that this is the limitation of the current state-of-the-art.
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(d)' As discussed in Reference 1 and reviews of other PRAs (e.g.,
Reference,3), additional studies and research efforts are required to justify the use of ductility factor for single degree of freedom models to represent multidegree'of freedom structures. We concur with Reference 1 that higher
... uncertainty value should be assigned to this factor.
('e) 'We' concur with the Reference 1, that uncertainty in some of the' parameters has.been understated (particularly, modeling uncertainties). The median capacity value may be on the high side in some cases.
(f) As in other published PRAs, LGS-SARA does not contain r.
1 sensitivity analysis to indicate the robustness of the assumptions. However, in a August 5 meeting the licensee provided some discussion-on the results of a recent sensitivity study which examined, for example, the effects of the assumption of different distribution (other than log-normal). We have not reviewed the results of this study. Reference 1 has included.some sensitivity studies for some assumptions. It appears that effects of the assumption used in the developing fragilities LGS-SARA on seismic risk are minor.
1
, f, (g)' It was not clear to us whether or not dynamic lateral earth pressures were considered in the structural fragility evaluation.
In a response to (Reference 4) the staff inquiry, the licensee stated the following:
"The Limerick structures are generally embedded in rock with lean concrete backfill. The rock and concrete backfill are separated from the structures by several inches of rigid
~
insulation so that essentially no lateral loads can be transmitted.to the structure from the rock. The seismic models, which were developed for the design analysis and which were used for the capacity evaluations, reflect this separation. No lateral loads are transferred from structure to rock or vice versa except at the base slab, and all shears and moments developed.in the structure are transferred down to the base slab rather than being taken out at higher elevations.
Any local soil loads on the walls were judged to be small in comparison to the out-of-plane capacities of the walls, and therefore no reduction in the seismic capacities of the Limerick structures was judged appropriate. There is no evidence that dynamic soil pressures have ever failed basement walls unless gross soil failures have occurred. Such a failure is considered incredible at the Limerick site."
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It is recommended that the above judgment should be verified by a specific analysis of the embeddad portion of a reactor enclosure wall which takes into account the effects of soil pressures, where applicable.'
(b) We concur with the Reference-1 that the implications of impact between the containment building and the reactor enclosure should be a'ddressed for the following concerns:
I a.
Failure of-safety-related electrical and control equipment located in the reactor enclosure.
b.
Failure of safety-related piping which crosses between the two buildings due to relative displacements.
In addition, it should be verified that no safety-related components will be damaged by spilled concrete caused by impact of the two structures.
Finally, it should be verified that failure of small lines attached to the safety-related piping near the junction of the two structures and anchored to the reactor enclosure will not r.
contribute to the frequency of core melt.
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' Based on the above discussion of the general, methodology, it is apparent that current state-of-the-art for seismic risk evaluation
- ~~
precludes the determination of " absolute" seismic risk and it only provides a relative measure of risk between different sites and plants provided the methodologies and assumptions are consistent in each risk evaluation.-
3.2 Coments on Critical Structures / Components Listed in Table 1 Table 1 lists the critical structural components or components which are affected by the estimation of structural response parameters. We have not performed a detailed review of the
- calculations for each of these critical components. However, we have reviewed the findings reported in the Reference 1 and the discussions in LGS-SARA.
(a) ~ Condensate Storage Tank (S.,)
We concur with the findings in the Reference 1 that the fragility _ parameters for the condensate storage tank appear to be reasonable.
(b) Reactor Internals (S3), CRD Guide Tube (SS), Reactor Pressure Vessel (56)
In the fragility estimation for these three components, a value of 10% damping was assigned to concrete portions of the support structure (also see Reference 1).
i 4
In Reference 4, the licensee quoted results of a recent study (Reference 5) which indicates median dampings at various stress levels.. Based on these damping values, the licensee performed an analysis to indicate that composite damping value to be between 9 and 10 percent for the RPV support system at
.the. median. RPV fragility leyel., P.rovided the values in the Reference are acceptable, the issue of the damping value in
- concr te structure'can be considered resolved.
~
We concur with, Reference.,1 that, as for other components, the modeling uncertainties are underestimated. According to Reference 1, the effect of doubling the uncertainty for modeling would have a small affect in the frequency.of core melt.
(c) Reactor Enclosure and Control Structures (54)
In Reference 1,. it is estimated that the median capacity for this component should be 0.90g as opposed to 1.05g as indicated in LGS-SARA (p. 4-25 of Appendix B).
It is further estimated that the mean frequency core melt would increase, approximately, by 20 percent because of the lower median L
capacity.
l (d) SLC Test Tank (58)
We concur with the recommendations in Reference 1 that the component specific analysis is needed to verify the parameters used in the fragility development.
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(e) SLC Tank (S10)-
As suggested in Reference 1, the possible failure of the SLC tank due to tearing of the base plate flange near anchor bolts should be checked to verify that it is not the weak'est capacity.
4.0 Conclusions (a) Sources of Conservatism and Unconservatism Following is a partial and preliminary list of possible sources of conservatism and unconservatism.
(1) Conservatism structs.al fragility formulation does not have a lower-bound cut-off value. It is believed that belcw certain acceleration value, an engineered stnJcture or components will not fail.
~
The use of low ductility value (2 to 2.5) for flextural made of failure of shear walls compared to other PRAs (4 to 4.5).
J It appears that median values are generally
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conservatively estimated.
(ii) Uncenservatism OmissTon of explicit consideration of design and construction error. Note that, in some cases, this
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4 cmission may 1,ead to conservative results.
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' Inclusion of. duration factor in conjunction with the
. effective peak acceleration.- This is, in part, compensated by the use of low ductility values.
The modeling uncertainties both due to probabilistic
-- - model 'to determine fragility and original. design model, are,' generally, underestimated. THe effects of-increase in these uncertainties are discussed in the Reference 1.
(b) It is concluded that the methodology used in the LGS-SARA is a state-of-the-art approach and this approach, although considered reasonable has not been validated and contains a great deal of. uncertainty in itself. The estimated median values in LGS-SARA appear reasonable or conservative (except for the case of the reactor enclosure building as discussed 3'
in Section 3.0) while the uncertainties are underestimated in some cases.
is (c)._ In Reference 1, it is indicatea that further analyses are required to determine whether the mean frequency of core melt is-dominated by contributions from structural failures or electrical component failures. Therefore, the issue of significantly contributing structural components will be 4
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It should be noted that any significant imprevement in risk is not anticipated from any structural fixes.
(d) The methodology used in the LGS-SARA is essentially ider.tical to the one used in the recent PRAs, such as, Zicn Probabilistic Safety Study (ZPSS) and Indian Point Probabilistic Safety Study (IPPSS). Following are the major differences between the methodology used in LGS-SARA and ZPSS and IPPSS.
The LGS-SARA includes an explicit factor of safety for earthquake duration (see (a) above).
The LGS-SARA includes random failure (non-seismic) of components.
In the ZPSS and IPSS, structural components were found to be dominant contributor, while in the LGS-SARA electrical components have been found to be dominant contributors (see(c)above).
(e) It appears that plant structures and structural components can withstand the earthquake levels well beyond the SSE level. Based on the median acceleration values and associated
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e variability listed in Table 1,in general, no significant probabilities of failure can be identified for these-components in the range of the SSE level.
References 1.0 A Preliminary Review of the Lime. rick Generating Station Severe Accident Risk Assessment" (Draft), Brookhaven National
~
.~ I Labordtory; August' 15, 1983.
2.0 Review and Evaluation of..the Indian Point Probabilistic Safety Study", NUREG/CR-2934, December, 1982.
3.0- Indian Point Probabilistic Safety Study, Power Authority of the State of-New York, Consolidated' Edison Company of New York, Inc., Spring 1982.
~
4.0 Letter'from Philadelphia Electric Company'(PECO) to A. Schwencer of HRC dated August 29, 1983.
5.0 Stevenson, J.D., " Structural Damping Values as a Function of Dynamic Stress," Nuclear Engineering and Design, Volume 60(2) pp. 211-237, September 1980.
4 4*
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- Table 1.
Significant Structural Fragility Components Median Ground Failure Cause Acceleration O
8 No.
Component or Mode Capacity R
R S
Condensate storage tank Tank-wall rupture 0.24 0.23 0.31 2
S Reactor internals Loss of shroud support 0.67 0.28 0.32 3
E Reactor enclosure and Shear-wall collapse 1.05 0.31 0.25 4
control structure S
CRD guide tube Excess bending 1.37 0.28 0.35 5
S Reactor pressure vessel Loss of upper support 1.25 0.28 0.22 6
bracket S
SLC-test. tank Loss of support 0.71 0.27 0.37 8
S SLC tank Wall buckle 1.33 0.27 0.19 10
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