ML20248D112
| ML20248D112 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/31/1996 |
| From: | Kipp T, Shanlai Lu, Tiong L EQE INTERNATIONAL |
| To: | |
| Shared Package | |
| ML20248C969 | List: |
| References | |
| NUDOCS 9806020322 | |
| Download: ML20248D112 (140) | |
Text
{{#Wiki_filter:.-- - _ - - - _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ ______'---v---- - - - - , - - - - - - _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ l i i ATTACHMENT 6 DEVELOPMENT OF EQUIPMENT SEISMIC FRAGILITIES FOR OCNGS IPEEE IPIIERAI DOC U4 '98 i N M M 521 PDR ADOCK 05000219 4
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'O INT ERN A flON AL DEVELOPMENT OF EQUIPMENT SEISMIC FRAGILITIES FOR OCNGS IPEEE July,1996 Prepared by:
EOE International, Inc. L.W. Tiong S.Lu T.R. Kipp l Prepared for: 1 GPU NUCLEAR CORPORATION i Parsippany, New Jersey EQE Job Number: 42113.08 EOE INTERNATIONAL l l l
TABLE OF CONTENTS Pajte
- 1. I N T R O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.1 S c o pe o f Wo r k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 -
1.2 O bje c tiv e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.3 S u mm a ry o f Re sults . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3
- 2. M ETH O D O L O G Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1 O vervie w o f Methodology . . . . . . . . . . . .. . ... . . . . . . . . . . . . . . . .. . . . ... . . . . . . . . . . . . 2-1 2.1.1 Seismic Hazard Analysis . ......, ......... ....... . .. . ....... ... ..... . 2-1 2.1.2 Seismic Fragility Evaluation ........................................ 2-2 2.1.3 Analysis of Plant Systems and Accident Sequences ..... 2-3 2.1.4 Evaluation of Core Damage Frequency and Pu bli c R i s k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-4 2.2 Fragility M ethodology . . . . .. . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . 2-5 2.3 Fragility Cuto f f M e tho dolo gy . . ..... ... . . .... ... .. .. .. ... . . . . . . . ..... . .. . . . . . 25
- 3. SOURCES O F DESIGN CO NSERVATISMS ....................................... 3-1 3.1 Earthqu a ke I nput Motio n . . . .. . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '3-1 3.2 Damping............................................................................ 3-2 3.3 Calcul atio n o f Re spo nse . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-2 3.4 C a p a ci ti e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3
- 4. WALKDOWN............................................................................... 4-1
- 5. EVALUATION OF EQUIPMENT AND DISTRIBUTIVE SYSTEMS .......... 5-1 1
5.1 Screening of Flexible Equipment and Distribution S ys te m s By An alysi s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2 5.1.1 S tre ngth Facto r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5.1.2 Equipment Response Factor ....................................... 5-3 5,1.3 Structural Response Factor ........................................ 5-5 5.1.4 Fragility Description for Flexible Components De signed 6 y Analysis . . . . .. .. . . ... . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . .. . . . 55 f l
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f~ l 1 1 l i I I TABLE OF CONTENTS (CONTINUED) l Page , 5.2 Screening of Rigid Equipment ... ........ . .... .......................... ..... 5-6 5.3 Conservatism in input Motion and Response Analysis ........... 5-6 5.4 Functional Fragilities of Electrical Equipment .......................... 58 5.5 Summary of Fragility Results .... ..... ... . . . .................... .... ......... 5-9
- 6. REFERENCES................................................................................ 61 l
TABLES 31 Median Damping Recommended For Fragility Development ............... 3-5 5-1 I niti al Co mpo ne nt S cre e ning . . .. .. .. . . .. . . . .. ... . ..... . . .. .. . . . .. . .... . .. .. . .... . . ... .. 5 11 5-2 O C N G S Fra gility S umm ary . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-27 5-3 Electrical Panel Fragilities Based on Relay Function ........................... 5-39 FIGURES 2-1 Schematic Overview of a Seismic PS A ........................................... 29 2-2 Typical Seismic Hazard Curves For a Nuclear Power Plant Site .......... 2-10 2-3 Typical Family of Fragility Curves For a Component .......................... 2-11 24 Probability Distribution of Seismically-Induced Severe Core D a m age Fre q u e ncy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-12 3-1 Comparison of EPRI 10,000 Year Median UHS and A-46 Site-Specific Ground Response Spectrum vs. R.G.1.60 Spectral Shape ................ 3-6 3-2 Comparison of A-46 vs. 3xSSE Median Response Spectra, Reactor Build i n g Fo un d atio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-7 33 Typical Overtest at High Frequency ................................................ 3-8 P:92113-ossocreaToc.cocmv iv s _h
TABLE OF CONTENTS (CONTINUED) APPENDICES A Seismic Fragility Methodology B Comparison of 0.184g A-46 and 3XSSE Median Floor Response Spectra For Reactor Building C Comparison of 0.184g A-46 and 2XSSE Median Floor Response Spectra For Turbine Building D Comparison of 0.184g A 46 and 2XSSE Median Floor Response Spectra For Diesel Generator Building E Comparison of 0.184g A-46 and 2XSSE Median Floor Response Spectra For intake Structure I* P:\42113 o8\oCFRGToc. doc /irv V
- 1. INTRODUCTION In Generic Letter 88-20, Supplement 4 (Reference 1), the US Nuclear Regulatory -
Commission requested that the utilities for all operating nuclear power plants in the United States perform an evaluation of their nuclear power generating facilities to
- identify any vulnerabilities associated witn the occurrence of specified plant specific-external events and to assess the impact of these vulnerabilities on the potential for plant core damage or radioactive material release. This program, designated the Individual Plant Examination for External Events (IPEEE), is a corollary program to the Individual Plant Examination (IPE) which focuses on vulnerabilities associated with the occurrence of internal events. -
The general purpose of the IPEEE is similar to that for the IPE in that it is desired that each utility:
- 1) Develops an appreciation of the plant severe accident behavior.
- 2) Understands the most likely severe accident scenarios that could occur at the plant while at full power operating conditions.
- 3) Gains an understanding of the overalllikelihood of core damage and radioactive material release.
- 4) Reduces'the likelihood of core damage and radioactive material release, if necessary, by modifying hardware and/or procedures in a way that would prevent or mitigate severe accidents.
Based upon a review'of past Probabilistic Risk Assessments, the Generic Letter (Reference 1) identifies five specific external events that are to be evaluated. These o are: 1)- Se,ismic Events,
- 2) Internal Fires,-
3)' High Winds and Tornadoes,
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- 4) External Floods, and
- 5) Transportation and Nearby Facility Accidents.
In addition, any other plant specific event (e.g., volcanic activity) known to the utility is to be addressed. The Generic Letter (Reference 1) and NUREG-1407 (Reference I. which provide procedural and submittal guidance for implementation of the IPEEE, identify acceptable methodologies for the evaluation of the various IPEEE initiating events. For the seismic evaluation of Oyster Creek Nuclear Generating Station (OCNGS), the Seismic Probabilistic Risk Assessment (PRA) approach was selected. Each of the seismic evaluation methodologies requires the development of a Safe Shutdown Equipment List (SSEL) which is to include structures and equipment necessary to achieve four specified safe shutdown objectives at the Review Level Earthquake (RLE). These are:
- 1) Achieve a safe shutdown condition (e.g., hot or cold standby) and rnaintain that condition for a period of 72 hours.
- 2) Mitigate the effects of a Loss Of Offsite Power (LOOP). ]
- 3) Mitigate the effects of a Small Break Loss Of Coolant Accident (SBLOCA). )
- 4) Ensure containment performance functions and the ability to isolate containment.
In addition, each of the seismic evaluation methodologies requires that a documented walkdown of the items included on the SSEL be performed for the purpose of identifying equipment / system seismic weak links in either the component I load path or anchorage, potential seismic f ailure of non-safety items resulting in i 1 l f alling and proximity interactions, potential flooding or fluid spray interactions and l potential seismic fire interactions. For a PRA, seismic fragilities are developed for 1 those items which, based upon the walkdown and reviews of the design basis and equipment qualification, appear to be the weak links. These fragilities are then utilized, along with the seismic hazard and plant function models to compute the PA42113 oBioCFRGCH1.DoCfrv 1-2 , tw = L
probability of core damage. This report describes the development of equipment fragilities for OCNGS. 1.1 SCOPE OF WORK EQE's scope of work consists of performing a walkdown of OCNGS, screening components and structures based upon their relative seismic capacity, and calculating selected fragilities. Included in the scope of work are components on the SSEL that perform safety functions, components that are potential seismic spatial interactions, and components which are potential seismic / fire interactions. Development of the SSEL and the utilization of the fragility descriptions in quantifying seismic risk is outside the scope of this report. 1.2 OBJECTIVE The objective of this effort is to identify seismically vulnerable plant equipment which could potentially affect the safe shutdown of the plant and to develop corresponding equipment fragility descriptions so that potential seismic risk can be quantified for OCNGS. 1.3
SUMMARY
OF RESULTS The Safe Shutdown Earthquake for OCNGS is 0.184g peak ground acceleration. It is noted that the seismic IPEEE work is performed in parallel with an A-46 study. The findings of the A-46 program, coupled with lower amplifications in the median IPEEE ground input, indicate that most equipment have significant seismic margins ; and may be screened for the IPEEE. Some exceptions to this generci rule were } noted during the walkdowns and subsequent review of A-46 documentation. Calculations were performed to estimate median seismic capscity for these items. ) Several potential spatial systems interaction problems were identified, including possible cabinet impacts with adjacent structural elements which could affect relay I performance, pounding of adjacent distribution panels that are not bolted together, 4 housekeeping items such as unrestrained toolboxes in vicinity of sensitive equipment, and unrestrained fluorescent lights in panels. It is expected that these seismic interaction issues will be resolved under the OCNGS A-46 program. No i i PA42113-o8\oCFROCH1. doc /irv 1-3 g L __ . . _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . . _ _ . _
spray or flood issues were noted. The seismic / fire interaction walkdown findings are documented in a separatueport. New probabilistic soil-structure interaction analyses were performed based on the 10,000 year median uniform hazard spectral shape defined in EPRI document
-(Reference 5). The floor spectra generated from the probabilistic response analysis show significant margin when compared to the floor respont.e spectra used for A-46 evaluation. A comparison of the A-46 and the probabilistic spectra, coupled with a review of A 46 packages for margins to code allowables and plant walkdown observations, enable most equipment to be screened out. The exceptions to this are certain marginal equipment items identified irs the A 46 program.. Where modifications to the equipment has been conceptually designed or installed, the fragility calculation is based on the upgraded configuration as shown in Table 5 2.
The overall impression gained from the walkdowns and a review of A-46 SEWS packages is that the plant is adequately constructed for this low seismicity site. The A-46 packages demonstrate a sufficient degree of conservatism in almost all cases to satisfy the screening criteria. Where this degree of conservatism does not exist, detailed estimates of median capacities are developed for the seismic PRA. The screening threshold fragility descriptions are: Am ER SU HCLPF' Equipment structural integrity 1.0g 0.40 0.32 0.30g This threshold fragility, when convolved with the mean seismic hazard curves l L results in an annual probability of failure of the individual component of
- approximately 8.8E-6. A specific fragility has been completed for equipment not satisfying this threshold fragility. The results of the equipment evaluation are
- provided in Section 5.
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- 2. METHODOLOGY 2.1 OVERVIEW OF METHODOLOGY The key elements of a seismic PRA are:
- 1. Seism /c hazard ene/ys/s - estimation of the frequency of various levels of seismic ground motion (acceleration) occurring at the site.
- 2. Se/sm/c frag /#ty eve /uet/on - estimation of the conditional probabilities of structural or equipment failure for given levels of ground acceleration.
- 3. ~ Systems / accident sequence analysis - modeling of the various combinations of structural and equipment failures that could initiate and propagate a seismic core damage accident sequence.
- 4. Evaluat/on of core damage frequency (CDF) andpublic risk -
assembly of the results of the seismic hazard, fragility and systems analyses to estimate the frequencies of core damage and plant damage states; assessment of the impact of seismic events on the containment integrity; and integration of these results with the core damage analysis to obtain estimates of seismic risk in terms of effects on public health. The process is shown schematically in Figure 2-1. Following is a brief description of the four steps utilized in the PRA process. 2.1.1 Seismic Harard Analvsis Seismic hazard is usually expressed in terms of the frequency distribution of the _ peak val ue of a ground motion parameter (e.g., peak ground acceleration) during a specified time interval. The different steps of this analysis are as follows: C Pn42113 o8\OCFRGCH2.DoClirv 2-1 gg
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- 1. Identification of the sources of earthquakes, such as faults and seismo-tectonic provinces.
- 2. Evaluation of the earthquake history of the region to assess the -
frequen . n of occurrence of earthquakes of different magni'.. a or epicentral intensities, i
- 3. Development of attenuation relationships to estimate the intensity of earthquake-induced ground motion (e.g., peak 3 ground acceleration) at the site.
4 Integration of the above information to estimate the frequency of exceedance for selected ground motion parameters. The hazard estimate depends on uncertain estimates of attenuation, upper bound magnitudes, and the geometry of the postulated sources. Such uncertainties are included in the hazard analysis by assigning probabilities to alternative hypotheses regarding these parameters. A probability distribution for the frequency of occurrence is thereby developed. The annual frequencies for exceeding specified values of the ground motion parameter are displayed as a family of curves with different probabilities (Figure 2 2). A Bayesian estimate of the frequency of exceedance at any peak ground acceleration is obtained as the weighted sum of the frequencies of exceedance at this acceleration given by the different hazard curves. The weighting factor is the probability assigned to each hazard curve.
- 2.1.2 Seismic Fraaility Evaluation The methodology for evaluating seismic fragilities of structures and. equipment is documented in Reference 3. Seismic fragility of a structure or equipment item is defined as the conditional probability of its failure at a given value of the seismic input or response parameter (e.g., ground acceleration, stress, moment, or spectral acceleration). Seismic fragilities are needed in a PRA to estimate the conditional probabilities of occurrence of initiating events (i.e., loss of emergency AC power, loss of forced circulation cooling systems) and the conditional failure probabilities of different mitigating systems (e.g., auxiliary cooling system).
P:\42113-o8\oCFRGCH2.DoClirv 2-2
_The objective of a fragility evaluation is to estimate the ground acceleration capacity
~
of a given component. This capacity is defined as the peak ground acceleration value at which the seismic response of a given component located at a specified - point in the structure exceeds the component's resistance capacity, resulting in its failure. The ground acceleration capacity of the component is estimated using information on plant design bases, responses calculated at the design analysis stage, as-built dimensions, and material properties. Because there are many variables in the estimation of this ground acceleration capacity, component fragility is described by a family of fragility curves; a probability value is assigned to each curve to reflect the uncertainty in the fragility estimation (Figure 2 3). 2.1.3 Analvsis of Plant Svstems and Accident Seouances Frequencies of severe core damage and radioactive release to the environment are calculated by combining plant logic with component fragilities and seismic hazard estimates. Event and fault trees are constructed to identify the accident sequences that may lead to severe core damage'and radioactive release. The plant systems and sequence analyses used in seismic PRAs are based on the PRA Procedures Guide (Reference 4) and can generally be summarized as follows:
- 1. Fault trees are constructed which reflect (a) failures of key system components or structures that could initiate an accident sequence, and (b) failures of key system components or structures that would be called on to mitigate the accident sequence.
- 2. The fragility of each such component (initiators and mitigators) is estimated.
- 3. Fault trees are used to develop Boolean expressions for severe core damage that lead to each distinct plant damage state sequence.
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- 4. Considering possible severe core damage sequences and containment mitigation systems (e.g., fan coolers, containment l
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-- - 9
sprays, and containment), Boolean expressions are developed for each release category. As an example, the Boolean expression for severe core damage in a prior Probabilistic Safety Study is: MS = 4 + 8 + 10 + 14 + 17 + 21 + (12 + 22 + 26)
- 9 (2-1)
The numbers represent components for which seismic fragilities have been developed. The symbols " +" and "*" indicate "OR" and "AND" operations, respectively. Plant level fragility cur'ves are obtained by aggregating the fragilities of Individual components according to Equation 2-1, using either Monte Carlo simulation or numerical integration. The plant level fragility is defined as the conditional probability of severe core damage as a function of the peak ground acceleration at the site. The uncertainty in plant level fragility is displayed by developing a family of fragility curves; the weight (probability) assigned to each curve is derived from the fragility curves of components appearing in the specific plant damage state accident sequence. 2.1.4 Evaluation of Core Damaae Freauenev and Public Risk Plant level fragilities are convolved with the seismic hazard curves to obtain a set of doublets for the plant damage state frequency, { < pij , f i j > } (2-2) where f ij is the seismically-induced plant damage state frequency and pij is the discrete probability of this frequency. pij = qipj (2-3) fj. i f(a)dHj i da (2-4) l Here, Hj represents the jth hazard curve, if the ith plant damage fragility curve; qi i s the probability associated with the ith fragility curve and pj is the probability associated with the jth hazard curve. j
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The above equations state that the convolution between the seismic hazard and plant level fragility is carried out by selecting hazard curve j and fragility curve i; the f l probability assigned to the plant damage frequency resulting from the convolution is j the product of the probabilities pj and gj assigned to these two curves. The l convolution operation given by Equation 2-4 consists of multiplying the occurrence l frequency of an earthquake peak ground acceleration between a and a + da (obtained as the derivative of Hj with respect to a) with the conditional probability of l the plant damage state, and integrating such products over the entire range of peak ground accelerations O to oc. In this manner, a probabilistic distribution on the frequency of a plant damage state can be obtained. Severe core damage occurs if any one of the plant damage states occurs. By probabilistically combining the plant darnage states, the plant level fragility curves for severe core damage are obtained. Integration of the family of fragility curves over the family of seismic hazard curves yields the probability distribution function of the occurrence frequency of severe core damage (Figure 2-4). By extending this procedure, probability distribution functions of the occurrence of different release categories are obtained. Public risks in terms of early fatalities, long term adverse health effects, and property damage are evaluated by developing a site-consequence model and inputting the release frequencies calculated above. This analysis would produce seismic risk curves showing frequencies of exceedance at different levels of consequences. 2.2 FRAGILITY METHODOLOGY The scope of work performed by EOE and documented in this report pertains to the development of seismic fragilities. The general methodology forfragility development is contained in Appendix A. Further details are provided in Reference 3. 2.3 FRAGILITY CUTOFF METHODOLOGY lt is not practical to calculate fragilities for all components which are included in the risk model. Most components and distribution systems are inherently rugged and Pn42113 o8\oCFRGCH2.ooClirv 2-5
can be screened,out on the oasis that their seismic induced failure rate is low in comparison to the items which will ultimately dominate seismic risk. It is desirable to establish a fragility target above which components exceeding this target may be screened.out without significantly affecting the overall results. In developing the target, three variables must be considered; seismic hazard, uncertainty in the median fragility and frequency of failure (potential core damage) relative to that for other events. A fourth variable, consequence of failure is important, but for purposes of establishing a fragility cutoff it is assumed that all failures have equal consequence. Parametric studies were conducted using the hazard and candidate fragility curves as input variables and examining the resulting failure frequency. The candidate fragility descriptions were convolved with the seismic hazard using the EQE computer code EQESRA (Reference 13) to compute seismic failure rates. The example cases were then studied to determine an acceptable cutoff fragility level. Seismic Harard: The EPRI and LLNL studies on seismic hazard are documented in References 5 and 6. The peak ground acceleration (pga) vs. frequency of occurrence is provided up to 1.Og in the LLNL study. Also, the predicted seismic hazard is higher in the LLNL study. NUREG-1407 states that the seismic hazard must be carried out to 1.5g unless sensitivity studies can show that a lower cutoff is justified. In the fragility cutoff study, the pga hazard was extrapolated to 1.5g. It was found that the extension of the hazard did not make a significant difference in the computed mean CDF. Uncertainty in the Median Fragilitv: The uncertainty range for fragilities varies with the failure mode. For ductile modes of failure, such as for structures or piping, the margin to failure relative to code allowables is larger than for brittle or functional failure modes, however, the uncertainty is also larger. Thus, a dual criteria must be implemented to establish a minimum value of both the median capacity and the HCLPF. HCLPF is an acronym for High-Confidence-of-Low-Probability of-Failure, it is defined mathematically as approximately 95% confidence of less than about 5% probability of failure. If the fragility curve is as described in Section 2.1 by the median, Am, the PA42113-o8\0CFRGCH2. doc /irv 26 gg
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[ randomness, SR, and uncertainty, SU, where the Ss are logarithmic standard
- deviations, the HCLPF may be computed from:
HCLPF = A,,, exp (-1.65) (SR+EI U For ductile failure modes of flexible systems, such as for structures, the ratio of median to HCLPF is typically about three. For brittle failure modes of rigid equipment or functional failure modes such as relay chatter, the ratio of median to HCLPF tends to be closer to two. Thus, for the same seismic failure rate, the . flexible, ductile items may have both a higher median and a lower HCLPF than for a non-ductile failure mode. The final cutoff 'value for fragility was targeted to a rnedian value, wherein the HCLPF value is implied. Failura Rate Relative to Other Event Failure Rataa: Core damage from internal events usually governs the plant risk. Internal event core damage frequencies typically are on the order of 10E-5. For OCNGS, the largest internal event CDF was found to be 8.5E-5, resulting from failure of Balance-of-Plant (BOP) systems which include the feedwater, condensate, main condenser, TBCCW, circulating water, service water, and instrument air systems (Reference 7). Typically, seismic failures that could contribute more than about 10% of the internal events value are not
' screened out. Thus, the target for screening was set at 8.5E-6 or less for seismic failure rate for a 1.5g cutoff.
Applying the logic described above, several EQESRA analyses were performed and the following approximate fragilities were determined to be the threshold for screening out components: Characteristics of itemL TAmi 2 HCLPFf Mean CDF Flexible, ductile structures or equipment 0.80g 0.24g 8.7E-6 Brittle or functional modes of failure 0.68g 0.32g 8.8E-6 e
These two cases result in seismic failure rates of approximately 8.8E-6 for a 1.5g cutoff. However, it is desirable that the HCLPF be at least equal to 0.3Og in conformance with the seismic margin guideline. For flexible systems, the median capacity screen needs to be raised to 1.Og to meet the 0.3Og minimum HCPLF capacity. For convenience, the 1.Og median capacity was used for both flexible and brittle systems. This is overly conservative for brittle systems, but removes the need to distinguish between brittle and flexible failure modes. Also, the 1 Og median capacity value with Bn of 0.4 and Bu of 0.32 is recommended for use as a surrogate fragility for components or seismic accident sequences where all components in a system are screened out. In this manner the failure rate of the screened out components is included in the seismic risk analysis. i l l l. P:\42113-08\oCFRGCH2.DoCfrv 2-8 , cc:s - l u _ _ _ _ _ . _ _ . ._______________-________0
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- Frequency P:\42113 08\OCFROCH2.DOCArv 2-12
. 3. SOURCES OF DESIGN CONSERVATISM For an IPEEE, evaluations are performed for an earthquake ground motion which is typically greater than the design basis for the plant. However, in accordance with the documented program methodologies, the evaluations are to utilize " realistic" values for the various seismic response and equipment capacity (i.e., median-centered;. In contrast, the original design basis is conservative. Therefore, even though a larger ground input is used for the IPEEE, the origina! design can often be shoy!n to have significant reserves to meet the demands of the earthquake prescribed for the IPEEE if median-centered (realistic) parameters ar :onsidered.
. Tyoically, the first task in a seismic IPEEE is to quantify the available margins by comparing the originallicensing basis floor spectra against median responses generated for the IPEEE. In the case of OCNGS, it was decided to use the A-46 qualificat;o.1 data as a point of reference for the IPEEE program since the SEWS packages contain the most current as-built information. The A-46 program is also conservatively biased in the prescription of ground input, response parameters, and capacity evaluation. The following paragraphs compare the A-46 and IPEEE requirements.
3.1 EARTHQUAKE INPUT MOTION . l The DBE for OCNGS has a peak ground acceleration of 0.184g. The A-46 evaluations were performed to a site-specific 84% Non-exceedance Probability (NEP) horizontal grourid motion spectral shape anchored to 0.184g pga and a corresponding site-specific vertical ground motion spectral shape. Meanwhile, the IPEEE ground input is the 10,000-year median uniform hazard spectrum (UHS) for OCNGS specified in Reference 5 with the vertical spectrum taken as 2/3 of the horizontal spectrum. Figure 3-1 compares the RG.1.60, A-46, and IPEEE ground input anchored to the same PGA. It is observed that the site-specific spectral shape has significantly higher amplification over the UHS shape, especially at frequencies below 10 Hz. This difference propagates through the OCNGS l- structures in varying degrees, depending on the fundamental frequencies of the soil-l structure systems.
. PA42iisossocFRoCH3. doc /irv 3-1 t .
W_____-__-__-__________-_--________. - _ - _ _
l l 3.2 DAMPlNG For both A-46 and IPEEE purposes, the median damping appropriate for the response level is used to estimate median seismic demand. If the failure mode is functional without inelastic deformation, lower damping values are used than if the ) failure mode is purely structural. Table 3-1 tabulates damping levels recommended in Reference 3 for calculation of median fragilities. There is little difference i between damping values recommended for the A-46 and IPEEE programs. The recommended damping ratio for most structure / equipment is about 5% for the two programs. 3.3 CALCULATION OF RESPONSE As stated earlier, the A-46 evaluation was used as the reference point in equipment screening for the seismic IPEEE. In the generation of floor spectra for the A-46 I program, SSI response analyses we're performed using input motion based on a site-specific 84% NEP spectral shape. Conservatively biased structural damping was used. Uncertainty in the soil properties was accounted for by shifting the best estimate soil shear modulus to a lower and upper bound, and enveloping the resulting floor spectra. The envelope of the various soil cases was then peak broadened. The above steps are consistent with the current regulatory guidance in the USNRC Standard Review Plans (Reference 8). For the IPEEE evaluation, probabilistic structural response was conducted to develop loads and floor response spectra for seismic loading. The methodology for probabilistic structural response is documented in Reference 9 and the results are presented in Reference 10. Probabilistic analyses were conducted for two levels of earthquake,0.368g and 0.552g (i.e. 2xSSE and 3xSSE levels). At each increasing level, the soil properties and demping and structural damping were adjusted so that I their expected range is enveloped. The resulting IPEEE median floor spectra were found to have lower amplifications than the conservatively biased A-46 spectra. A typical comparison is shown in Figure 3-2, where it can be seen that the IPEEE 3xSSE median peak response at the Reactor Building foundation is about the sams as the peak of the conservative A-46 spectra. Since the A-46 evaluations conservatively used a static coefficient which P:\42113-o8\oCFRGCH3.DoCrrv 3-2 gg e_t = L ;
is based on the peak of the A-46 response spectra, this implies that there is a factor of conservatism of needy 3.0 on response for components located in the Reactor Building. For the other structures, the factor of conservatism is generally about 2.0. Overplots of the IPEEE 3xSSE median and the A-46 response spectra l were made for all locations of the Reactor Building where potentially sensitive equipment was located. These comparative plots are included as Appendix B, together with Table B-1, which summarizes critical ratios for each response direction. Similarly, Appendices C, D and E contain overplots of the 3xSSE median and the A-46 conservative response spectra for the Diesel Generator Building, Turbine Building, and intake Structure. The summary tables in these appendices present the critical response ratio in terms of the comparison of the 2xSSE median and A-46 conservative response spectra since 2xSSE is used as the reference input for these structures. The spectra comparison is a significant aspect of the equipment screening process which is described in snapter 5. 3.4 CAPACITIES The design of OCNGS followed codes and stendards that were in effect at the time the plant was designed. The general philosophy is to design structures to nominally remain elastic at the OBE level. Equipment and piping may exceed elastic limits during an SSE, but demonstration of functionality is a requirement for active equipment and piping. Material properties used in design are code specified values that are set at about the 95% confidence level. Typically, the median material strength has a factor of 1.1 to 1.25 over the code specified value, depending on the steel materials. Equipment qualified by test theoretically has the least margin. The tested equipment is subjected to test response spectra (TRS) that must envelop te required response spectra (RRS) defined by the amplified response spect.. at the equipment mounting location. There is no required safety factor on t!w test input and no knowledge obtained from the tests as to the espacity beyond the test level. Thus, the conservatism in the response spectra generation is the only definable margin although significant margin may exist beyond the achieved test level.
. In some cases, particularly for relays, results of fragility tests are used. The fragility level is defined as the threshold of malfunction or merely the ruan s oe\oCm0CH3. doc /ir, 3-3 g
a e L__i______._________.___________x . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ __ ._. 1
manufacturer's highest achioved test level. The margin above a fragility level established by test is typically the conservatism in the response spectra generation. The not result of the design and qualification process is that there are inconsistent margins above the qualification level. Ductile structural failure modes tend to have the larger margins, whereas non-ductile failure modes or equipment function demonstrated by testing tend to have the smaller' demonstrated margin. In developing fragilities, the knowledge of the conservatism in the design process is utilized to determine the median capacity. Median material properties are utilized rather than code properties. For ductile structural failures, the concept of ductility is utilized to define the margin beyond the elastic limits of the structure or equipment. The median capacity is well beyond the elastic limit and applicable ductility factors are derived to reflect this. For ductile modes of feilurs, the ductility factor of safety is discussed in Appendix A. For non-ductile structural failure modes, the margin is defined as the median ultimate strength divided by the applied load or stress. Non-ductile failure modes tend to have less margin than ductile modes if the load is at code limits. However, in most cases, the design applied load is significantly less than the code allowable, resulting in considerable margin in equipment capacity. Reference 3 provides guidance for estimating margins for tested components. This guidance results from a consensus of several experts in equipment qualification testing. A single proof test does not demonstrate a high-confidence-of-low-
. probability-of failure (HCLPF) which is an industry measure of acceptance for demonstrating adequate seismic margin. Thus, without a significant margin in the RRS or a significant overtest, equipment qualified by test would be computed to have low margin above the test level. The industry practice is to significantly overtest at frequencies beyond the frequency of the peak of the RRS. An example is shown in Figure 3 3. Since OCNGS is a soil site, the peaks of the RRS occur at low frequencies and most equipment is expected to be significantly overtested.
The overtest provides additional margins that may be considered in the screening process. .: l 1 mamessocmacHz.oocti,, 3-4 gg 9 _ - _ _ . _ _ _ _ . _ _ - _ - . - - - _ _ _ _ _ _ _ . _ - - . _ _ _ _ . . ~ _ _ . . _ _ _ _ . _ - _ _ . _ _ _ , . . _ - . _ . ._ _ -. -
I Table 3-1 MEDIAN DAMPING RECOMMENDED FOR FRAGILITY DEVELOPMENT (Reference 3)
, STRUCTURE OR COMPONENT , - MEDIAN DAMMNG.
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- 4. WALKDOWN The walkdown activity is emphasized in NUREG-1407 (Reference 2) as being a particularly important ingredient for equipment screening and the identification of l concerns in the IPEEE seismic evaluation. The Oyster Creek IPEEE and A-46 walkdown efforts were coordinated and conducted in accordance with the
! recommendations of References 11 and 17. The documentation of all walkdown activities was accomplished using Screening Evaluation Work Sheet (SEWS) forms similar to those in Reference 17. The IPEEE walkdown was performed from October 3 through October 19,1994, near the end of the A-46 walkdown effort, to ensure completeness' of the documentation for evaluation of IPEEE issues and the beyond-design-basis seismic excitations. The walkdown team was comprised of the following experienced personnel: NAME AFFILIATION AREA OF EXPERTISE Thomas R. Kipp EOE International, Inc. Seismic Capability Ken Canavan GPU Nuclear Systems Analysis Swadesh Ramdeen GPU Nuclear Seismic Capability Mr. Canavan is intimately familiar with the operation of the Oyster Creek safety systems and is responsible for the development of the seismic PRA risk model. Mr. Ramdeen is trained as a seismic capability engineer, participated in the A-46 l walkdown, and is intimately familiar with the findings and observations from the A-46 walkdown evaluation. Mr. Kipp is trained as a seismic capability engineer, has been intimately involved in the development and application of the seismic fragility methodology, and is responsible for the development of Oyster Creek seismic 1 fragility descriptions for SSEL equipment. The scope of the walkdown included all components on the A 46 and IPEEE combined SSEL. Accessible components w'ere examined internally and all components were examined externally. ' During the walkdown inside containment, PA42113-o8\0CFRGCH4. ooc /irv 4-1 {g]
= ---._m._ _ _ . __.. _ _ _ . _ _ . _ _ _ . _
mechanical penetrations were examined to determine if any concerns existed associated with containment performance resulting from potential differential motions between the Reactor Building and the Drywell. No concerns were noted. Piping and distribution systems inside containrnant appeared to be robust and no concerns were noted. Due to the relatively low seismicity at Oyster Creek, and the correspondingly low in-structure response, many of the equipment items were screened out on the basis of Table 2-4 of Reference 11 and thus the walkdown focused on anchorage conditions. A number of outlier conditions were noted. These include: l
- 1. Electrical panels which were not anchored to the floor ,
- 2. Electrical panels containing relays were not bolted together
- 3. Outdoor MCC with fully rusted anchorage load path.
- 4. Block walls near containment spray pumps
- 5. Heavy electrical panel mounted by boiting through sheet metal of panel
- 6. Relay panel with very flexible interior panel supporting relays q
- 7. Unanchored skid . mounted diesel generators
- 8. Numerous badly corroded supports for air handling units
- 9. Switchgear with anchorage that failed A-46 anchorage check
- 10. Failed holt on instrument rack wall strut j i
- 11. MCCs located near wall with impact probable i
- 12. Unanchored startup and SBO transformers For these cases, either calculations were prepared to demonstrate acceptable j performance for the A-46 evaluation or modification requests have been issued to correct the outlier condition. Fragilities for these items are based on the as-modified i
PA42113-o8\oCFRGCH4. doc /iry 4-2 ms 4
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condition. Other vulnerabilities were identified which, in the mind of the seismic
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capability engineers, potentially limit the seismic capacity. These include:
- 1. Block walls strengthened by unistrut
- 2. Non-metal battery racks with minimal battery support
- 3. Rotary inverter with unsymmetrical anchorage and vibration isolators .
- 4. Emergency Condenser anchorage spacing violations
- 5. Accumulators for vacuum valves restrained by friction straps
- 6. Battery charger with anchor bolts threaded into standoff channel (no nuts)
Numerous problems were noted for non-safety related equipment located in the Turbine Building. As a result, it was decided to remove these systems from the SSEL and to consider them to be inoperable following a seismic event. The complete IPEEE SSEL is included in Chapter 5, together with the derived fragility description or the screening disposition for each component. 1
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- 5. EVALUATION OF EQUIPMENT AND DISTRIBUTION SYSTEMS Most commercial equipment and distribution systems are inherently rugged if they are adequately anchored. When this same equipment is qualified for seismic loading in a nuclear power plant environment, the qualification requirements and conservatism in the specification of seismic demand and the calculation of equipment response combine to produce very large margins. In Reference 11, screening guidelines are provided that allow most equipment and distribution systems to be screened for earthquakes up to 1.2g peak spectra acceleration (ground), subject to a walkdown of the equipment to search for vulnerabilities. This 1.2g peak spectral value is compared to the peak spectral acceleration of the 84th percentile ground motion spectrum defined at 5% damping. If the median UHS of Figure 3-1 is compared to the 1.2g screening level,'the peak ground acceleration is about 0.75g. Assuming a coefficient of variation (COV) of 0.2 on the median UHS, the ratio of the 84th percentile UHS to the median is given by exp(0.2) = 1.22.
Therefore, when equipment is screened per the 1.2g caveats in the screening tables in Reference 11,it implies a HCLPF capacity of about 0.75g/1.22 = 0.6'g. This HCLPF value is significantly higher than the risk-based screening level of 0.3g established in Section 2.3, and therefore indicates that most OCNGS equipment can be screened out. The screening criteria in Reference 11 are based upon collection and review of a large suite of earthquake experience data which is summarized in Reference 12. The criteria also take into account the experience from many seismic PRAs in which excess design margins have been quantified. Given the low amplification of the UHS for OCNGS, meeting the walkdown screening caveats in Reference 11 provides a high degree of confid'ence that the component is inherently rugged and has low probability of failure in an earthquake that could potentially occur at OCNGS. i in addition to the above general screening based on 1.2g peak spectral acceleration, the findings of the A-46 evaluation program are also used in conjunction with the walkdown to screen out generic classes of components and distnbution systems. 4 This is a two-step process:
- First, overall factors of conservatism in the input motion, response analysis, and material strength are developed. Based on the risk studies described in Section 2.3, equipment with l
P:\42113 o8\oCFRGCH5. doc /irv 5-1
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median capacity in excess of 1.0g does not contribute to the risk and may be screened out. To achieve a median capacity of 1.Og, the ovarall factor of safety in the various response and strength factors described in Appendix A needs to be at least 5.43 (1.0g/0.184g). In the following subsections, generic factors of conservatism for flexible and rigid equipment items are developed. d
. Next, if the overall factor of safety is less than the requisite 5.43, " worst case" components are selected for median capacity analysis. This analysis uses either the original design basis document, or the A-46 SEWS package as a starting point. Also, other components that are suspected to have vulnerability based on walkdown observations or past experierce are subjected to similar analysis.
5.1 SCREENING OF FLEXIBLE EQUIPMENT AND DISTRIBUTION SYSTEMS BY ANALYSIS Flexible equipment and distribution systems usually have auctile failure modes, but for purposes of screening it is assumed that the failure mode is non-ductile such as for failure of anchor bolts, welds or buckling instability. Fillet welds are rhown in Reference 11 to have a much larger margin than implied by the design codes so fillet welds are not the basis for the screening computation. The following assumptions are made for the screening calculation:
- Probable frequancy range is 5-10 Hz (flexible equipment)
- Code margins are those inherent in the ASME code where the allowable stress may be as high as 70% of the specified ultimate strength, resulting in a nominal safety factor' of 1.43.
The safety factors of expansion anchors and welds are ! greater, so the ASME criteria governs the screening calculation. The methodology follows that described in Appendix A. l P:\42113MoCFROCH5.DoClirv 52
I l 5.1.1 Strength Factor The Strength Fa6 tor, Fs, is given by: Fu - N Fs = l l A-46 l Fu is the ultimate strength (stress or load), N is the load or stress under normal conditions and A-46 !s the seismic load or stress due to the A-46 ground input. For l floor mounted equipment, the gravity weight effects is beneficial to the equipment anchorage capacity. Thus, conservatively assume the gravity weight effect to be zero. The median strength is about 1.2 times the code value; the code value is set at the 95% confidence level, which is a -1.65 BU value. It is assumed that the average stress induced by the A 46 ground input is 70% of the code allowable for non-ductile failure mode such as anchor bolt failure. The code allowable is as high as 70% of the code ultimate strength, therefore the A-46 load / stress is (.7)(.7) = 0.49 of the eltimate code capacity. The strength factor is then: 1.2 - 0 Fs = = 2.45 0.49 Note that for ductile failure modes, the Fsmay be lower than 2.45 as the average stress may be held closer to the code allowable. But the overall capacity factor which includes the inelastic energy absorption factor will be higher than that for non-dr Ae failure modes. , 5.1.2 Equipment Response Factor The equipment response factor consists of the product of the individual factors for the variables of Qualification Method, Damping, Modeling, Mode Combination and Earthquake Component Combination. Qualification Method: In the A-46 evaluations of non-distributive flexible systems, a conservative static coefficient method was generally used for lateral load analysis. The static coefficient is based on the peak of the A-46 response spectra. The factor of conservatism is therefore a function of the equipment fundamental P:\42113 o8\oCFRGCH5. ooc /irv 5-3
=
T' _ _ .___________.__..___..___._._______._________________w
frequency, and is given by the ratio of the A-46 spectral peak to the median probabilistic spectral value at the frequency of interest. This ratio is always greater thEn the ratio of the A-46 peak to the peak of the probabilistic median spectra l given in the last column of Tables B-1, C-1, D-1, and E-1. For the purpose of a generic screening discussion, it is conservatively assumed that there is no additional' margin over that given by the A-46 peak to median peak. As shown in sub-section 5.3, the Fou is about 3.0 for the RB, and about 2.0 for the TB, EDG, and IS. For distribution systems such as piping and cable trays, the original design analyses were performed using the plant's Housnor spectra at design damping values much lower than 5%. It is reasonable to assume that the seismic demand imposed by these earlier vintage floor spectra would exceed the median seismic demand of these systems when high ductility of the piping systems and high damping value of the cable trays are considerod. Hence, for screening purposes, the Fou discussed in the foregoing may also be applied to distribution systems. Damping: in the A-46 evaluations, equipment damping is assumed to be 5% in most cases. This damping ratio coincides with that specified for the IPEEE, thus no further factor of conservatism is available for this parameter. Modeling: Modeling error can arise from frequency error and mode shape difference between the model and the actual response. The model would normally be median centered so the modeling factor would be unity. Mode Combination: As mentioned earlier, the static coefficient method is used in the A-46 evaluation of non-distributive equipment. The use of static coefficient implies single mode response. This is reasonably true for discrete equipment items, but may be conservative for distribution systems. The Fue is taken to be 1.0. 4 Earthouake Component Combination: Earthquake components were combined by the SRSS or 100-40-40 rule for both the A-46 and IPEEE evaluations. This is considered a median approach. The Fece is then 1.0. Equipment Response Factor Results: Combining the response factors as the product of the individual factors, the equipment response factor is:
, Fm = 3.0 for the RB and 2.0 for the other structures P:\42113-o8\oCFRoCH5. Doc /irv 5-4
5.1.3 Structural Response Factor
. Median response spectra were developed by probabilistic methods using a Latin Hypercube simulation process in which all important variables associated with structural response are included. The median results were used to derive the strength and response factors so the structural response factor is unity.
5.1.4 Fragility Description for Flexible Components Designed by Analysis The median peak ground acceleration capacity is the product of the strength, equipment response, and structural response factors times the reference 0.184g peak ground acceleration. A = (F.) (Fac) (Fas) (0.184) A = 2.45(3.0)(1.0)(0.184)
= 1.35g for equipment within the RB.
A = 2.45 (2.0) (1.0) (0.184)
= 0.90g for equipment within the TB, DGB, and IS.
This generic derivation shows that appropriately designed and installed equipment within the RB meets the screening threshold of 1.0g median peak ground acceleration capacity established in Section 2.3. For equipment within the other structures, additional capacity above the 0.99 median is assessed by examining worst cases of specific equipment classes. Table 5-1 lists components for which the SEWS package were reviewed and in many cases, median capacities estimated. The above generic screening of components and distribution systems is coupled with a walkdown verification to ensure that there are no configurations 'which appear vulnerable. In general, if the walkdown does not uncover unusual as-built conditions, or potential interaction problems, subsystems such as piping, cable trays and valves as well as anchorage of passive instrument racks or electrical distribution cabinets are screened out. 1 Valves are rigid but the piping systems in which they are mounted are flexible and the piping response dictates the demand for the valves. Valvos usually have a large design margin above the specified demand so the above derivation is also considered applicable to valves with the exception of those identified during the l1 1 P:\42113 o8\oCFRGCH5. doc /iry 5-5 r-
walkdowns that appeared to be outside of the seismic experience database. The MSlVs are examples of valves that were initially identified as outliers in the A-46 program because the operator weights were outside the bounds of the experience database. - Other than the operator weight, there were no other concerns, in this case, analysis was used to demonstrate adequacy for A-46. 5.2 SCREENING OF RIGlD EQUIPMENT in general, rigid equipment has very high capacity due to the low seismic demand defined as the zero period acceleration (ZPA) and the rugged features that lead to its rigidity. Failure of rigid equipment with heavy weight is almost always due to
- failure of anchorage. For the most part, compact rigid equipment such as small pumps and tanks, chillers, small air handlers, small instrumentation and sensors, switches, etc. were screened out on the basis of ttie walkdown. Rigid equipment that appeared to be vulnerable on the basis of weak anchorage or heavy mass, were selected for median capacity derivation. This selected list includes the unanchored DG units and startup transformers, isolation condensers and liquid poison tank.
5.3 CONSERVATISM IN INPUT MOTION AND RESPONSE ANALYSIS The conservatism in the input motion and response analysis were estimated by-comparing the A-46 and median probabilistic floor response spectra for the buildings housing essential equipment. Each building of interest at OCNGS is discussed in the following paragraphs. The last column of Table B 1 in Appendix B surnmarizes the spectra comparison for the Reactor Building (RB). It is observed that for the two horizontal and the vertical directions, the ratio of the A-46 peak to the 3xSSE median spectral peak is about equal to or greater than 1.0. Thus, a factor of conservatism of at least 3 is obtained for components that are susceptible to seismic inertial effects. This factor of conservatism increases with elevation from about 3.0 and reaches 5.2 (1.73x3) at El. 95'-3" of the RB which is the highest elevation for essential equipment. Above El. 37'-3" of the Drywell, the factor of conservatism is less than 3.0 for the vertical spectra. However, the equipment of interest at these elevations comprise
~
valves (isolation, SRV, and check valves), and RPV water level transmitters which are inherently rugged and are easily screened cut (sca Table 5-2). PA42113-o8\oCFRGCH5.DoClirv 5-6 m
For the Turbine Building (TB), the ratio of the horizontal A-46 peak to the 2xSSE median spectral peak is close to or greater than 1.0 (see Table C-1 of Appendix C). Thus a factor of conservatism of about 2.0 for the horizontal components is obtained. This factor drops below 2.0 at Elevations 63'-9" and 74'. However, Table 5-2 shows that at these locations, the equipment consists of switches, transformers, dampers, chilled water unit, power supply, and a/c unit including fan. With the exception of the a/c unit, these equipment items are rigid and are generically screened out on the basis of low seismic demand in the rigid range. The A-46 SEWS packages for the a/c unit was reviewed and the median capacity for this item was determined to be above the 1.0g cut-off (see Table 5-1). For the Diesel Generator Building (EDG), the horizontal and vertical A-46 peaks are close to the 2xSSE median spectral peaks (see Table D-1 in Appendix D). This indicates that a factor of conservatism of about two is available in the response analysis. The essential equipment housed within this structure consists of fuel oil transfer pumps, local shutdown panel, switchgear bus, fuel oil storage tank, and the unanchored DG units. The local shutdown panel is well anchored when compared to similar panels in other locations and is screened out. The switchgear bus may also be screened out provided interaction concerns are addressed. It is expected that the seismic interaction issues will be resolved under the OCNGS A-46 program. The diesel fuel oil storage tanks, and fuel oil transfer pumps, are rigid and are well supported. The only items not screened out are the unanchored DG units which may slide during ground shaking. The median ground acceleration capacity of the DGs based on incipient sliding is 0.56g. However, failure of the DG does not occur until the DG set slides enough to rupture the connected fuelline. Based on the configurations of the connected fuel line, a lateral displacement of one. inch was estimated to rupture the fuel line. Nemark's rigid body sliding approach (Reference
- 18) which accounts for the cyclic nature of earthquake motion was used to calculate the peak ground acceleration at which the DG set will slide one inch.
l Based on this approach, the median ground acceleration capacity was calculated at 0.81 g. L The only items of equipment of interest in the intake Structure are the emergency service water pumps. Fragility calculations performed for these components yielded median capacities of 1.18g. The load center and other control panels which are not located in the Intake Structure were screened. A comparison of the A-46 P:\42113 o8\oCFRGCH6. doc /irv 5-7 i
floor spectra to 2xSSE median floor response spectra is shown in Table E-1 of Appendix E. , 5.4 FUNCTIONAL FRAGILITIES OF ELECTRICAL EQUIPMENT As part of the A46 evaluation of OCNGS equipment, a review of the electrical systems was performed and a list of " essential" relays was developed. Based upon type, location, seismic input, and demonstrated GERS capacity for " essential" relays within a given panel, the functional fragility of the panel was developed. Reference 3 provides a methodology for developing fragilities for components qualified by test. The median capacity is expressed as: Au = C - Fo Fas PG A RRSC where: TRSc = Clipped test response spectra RRSc = Clipped required response spectra Fo = Device capacity factor Fas is the structural response factor and PGA is the peak ground acceleration for development of the RRS TRSc = TRS (Cr)(Ci) where: Cr is unity for broad banded multiaxis testing
)
Ciis a capacity increase factor and is recommended to be 1.1 with a pu of 4 0.05. (Reference 3) RRSc = RRS (Cc)(Dn) where:
. P:\42113-oe\ocFRGcH5.DoClirv G8
== \
. lj
RRS is the required response spectrum Cc is the RRS clipping factor calculated per Reference 3 ' ' Da is a demand reduction factor if the RRS is calculated by deterministic means. Probabilistic methods were used for the development of median floor spectra so Da is unity Fo is recommended in Reference 3 to be 1.07 and 1.45 for components whose fragility description is based on relay GERS and non-relay GERS, respectively. The variabilities on Fo for components based on relay GERS are pn = 0.09 and pu = 0.18, while those for components based on non-relay GERS are pn = 0.11 and pu = 0.23. Fns is the structural response factor which is unity for probabilistic spectra. The pcis computed as the naturallogarithm of the ratio of the 84th percentile response to the 50th percentile response at the frequency of the equipment. The pc varies with the location of the equipment and is evaluated accordingly. PGA is the peak ground acceleration of the reference earthquake. Reference 3 also provides the randomnes.s and uncertainty variability for the various equation parameters. Based on the computation of the median capacity and variabilities for each " essential" relay within a panel, the functional fragility of the panel is equal to that of the essential relay with the lowest fragility. 5.5
SUMMARY
OF FRAGILITY RESULTS
~ Table 5-1 lists the results of the screening and specific fragilities determined for the equipment on the SSEL. Several notes are included that provide the basis for screening or for the tabulated fragility. Most cornponents are inherently rugged and are screened out on the basis of satisfactory A-46 SEWS resolution. Examples of equipment types that are considered inherently rugged are rigid switches, pressure and level transmitters, temperature sensors, and valves. For other less rugged equipment classes such as MCC's, electric panels, switchgear, fans, pumps transformers, tanks, heat exchangers, and generators, a sample of the most vulnerable configurations documented in the A-46 SEWS packages were reviewed PA42113 o8\oCFRGCH5. doc /irv 5-9
to develop fragility estimates. These are tabulated in Table 5-2. In many cases, potential spatial interaction problems have been noted and are expected to be ~ resolved as part of the A-46 program. The condensate storage tank is lightly anchored and represents one of the weak components, as shown in Table 5-2. It should be noted that the estimated fragilities shown in Table 5-2 represent non-recoverable structural failure modes, and do not consider the effect of relay functional f ailures. Table 5-3 lists the fragilities of electrical panels based upon chatter of " essential relays" for the lowest capacity relays housed within the important electrical panels. Reference 16 documents the evaluation of " essential relays" for which relay chatter is unacceptable. Relays which may be reset by operators, should chatter occur, are also included in the " essential relays" list (Table 5-3) in order to maintain accountability of the number of operator action which may be required. GERS were used for the relays which have no qualification test data. For those untested relays with no GERS data, as identified in Table 5-3 with an asterisk, a conservative peak spectral acceleration capacity of 3g was assumed for the seismic fragility evaluation. This 3g capacity is judged to be reasonable so long as the relay normal operating status is not de-energized /normally closed. However, if this normal operating status cannot be confirmed by OCNGS or if the 3g spectral capacity is found to penalize the computed plant seismic risk, relay testing to justify a higher chatter capacity is required. I i i P:\42113-o8\oCFRGCH5. doc /rv 5 10
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- 6. REFERENCES
- 1. Nuclear Regulatory Commission, "IndividualMont Examination of Extemal Events (IPEEE) For Severe Accident Vulnerabilities - 10CFR 50.54(f),"
Generic Letter No. 88-20, Supplement 4, June 28,1991.
- 2. Nuclear Regulatory Commission, " Procedural and Submittel Guidance for the Individual Mont Examination of Extemal Events (IPEEE) for Severe Accident .
Vulnerabilities," NUREG-1407, April,1991.
- 3. EPRI, " Methodology for Developing Seismic Fragilities," EPRi-TR-103059,
, April 1994.
- 4. USNRC "PRA Procedures Guide," NUREG/CR-2300, January 1983.
- 5. EPRI "Probabilistic Seismic Hazard Evaluations at Nuclear Mont Sites in the Centraland Eastem United States: Resolution of the Charleston Earthquake issue," EPRI NP-6395 D, April 1989.
- 6. NUREG-1488, " Revised Livermore Hazard Estimates for 69 Nuclear Pdwer Mant Sites East of the Rocky Mountains," USNRC, October 1993.
- 7. Personal Communication from Ken Campbell to Tom Kipp dated January 9, 1995.
- 8. USNRC, " Standard Review Mans," NUREG-0800,1989.
~
9.- Johnson, J.J., G.L. Goudreau, S.E. Bumpus, O.R. Maslenikov, "SSMRP Phase i Final Report - SMACS Seismic Methodology Analysis Chain with Statistics (Project Vill," NUREGICR 2015 Vol. 9, UCRL-53012, Vol. 9,
. Lawrence Livermore National Laboratory,1981.
- 10. EQE Engineering Consultants, Report No. 50124-R-003, "Probabilistic Seismic Response Analyses of the Oyster Creek Nuclear Generating Station
- in Support of the IPEEEProgram," July 1994.
P:\42113 o8\oCFRoCH6. doc /irv . 6-1 e
- 11. EPRI "A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 11," Electric Power Research Instituta, EPRI NP-6041-SL, Revision 1, August 1991.
- 12. Senior Seismic Review and Advisory Panel (SSRAP), "Use of Seismic and Testing Experience Data to Show Ruggedness of Equipment in Nuclear Power Plants," Rev. 4, February 1991.
- 13. EOE International, *EQESRA, Computer Code for Calculating Seismic Risk,"
Doc. No. 52049.01-R1, Rev. O, May,1990.
- 14. Newmark, N.M., " Inelastic Design of Nuclear Reactor Structures andits Implication on Design of Critical Equipment," SMiRT Paper 411, Structural Mechanic in Reactor Technology, San Francisco, CA,1977.
- 15. Riddell, R. and N.M. Newmark, " Statistical Analysis of the Response of Nonlinear Systems Subjected to Earthquakes," Department of Civil Engineering Report UILV-2016, University of Illinois, Urbana, IL, August 1979.
- 16. EOE Engineering Consultants, Calculation 42113-C-020," Oyster Creek IPEEE: Electrical Panel Fragility Based on Relay Function," October 1995.
- 17. SOUG " Generic Implementation Procedures (GIP) for Seismic Verification of Nuclear Plant Equipment," Revision 2, February 1992.
. 18. Nemark, N.M., " Effects of Earthquakes on Dams and Embankments,"
Geotechnique, Volumn XV, No. 2,1965. I t 1 PA4211348\oCFPtGCH6. doc /irv 6-2 ______s- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
APPENDIX A f SEISMIC FRAGILITY METHODOLOGY - I i l { f
/
a.
A. SEISMIC FRAGILITY METHODOLOGY The seismic fragility of a structure or equipment is defined as the conditional probability of its failure at a given value of peak ground acceleration. The methodology for evaluating seismic fragilities of structures and equipment is documented in the PRA Procedures Guide (Reference 4) and is more specifically described for application to NPPs in Reference 3. This general methodology has been applied in over 30 seismic Probabilistic Risk Assessments of nuclear power plants. The objective of a tragility emuation is to estimate the capacity of a given-component relazive to a ground acceleration parameter such as peak ground acceleration or spectral acceleration. The seismic hazard for OCNGS is defined by peak ground acceleration (pga) hence all fragility estimates are referanced to pga. The ground acceleration capacities of the components are estimated using
~information on plant design basis and responses calculated at the design-analysis stage. The ground acceleration capacity is a random variable which can be described completely by its probability distribution. However, there is uncertainty in the estimation of the parameters of this distribution, the exact shape of this distribution, and in the appropriate failure model for the component. For any postulated failure model and set of parameter values and shape of the probability distribution, a fragility curve depicting the conditional probability of failure as a function of peak ground acceleration can be obtained. Hence, for different models and parameter assumptions, one could obtain different fragility curves. A satisfactory way to consider these uncertainties is to represent the component fragility by means of a family of fragility curves obtained as above. A subjective probability value is assigned to each curve to reflect the analyst's degree of belief in the model that yielded the particular fragility curve. When represented in this fashion, the fragility curves need not appear to be smooth S-shaped curves, approximately parallel to each other; they could intersect each other and they may not even be nondecreasing functions of peak ground acceleration. The only requirement is that fragility being a probability should be between 0 and 1 (see l Figure A-1). .
P:\42113 08\oC-APA.coC\lRV A-1 _ 1 - L_ _ _.-_---- __
At any acceleration value, the component fragility (i.e., conditional probability of failure) varies from 0 to 1; this variation is represented by a subjective probability . distribution. On this distribution we can find a fragility value (say,0.01) that corresponds to the cumulative subjective probability of 5% We have 5% cumulative subjective probability (confidence) that the fragility is less than 0.01. Similarly, we can find a fragility value for which we have a confidence of 95%. Note that these statements can be made without reference to any probability mode,l. Using this procedure, the median and high (95%) and low (5%) confidence fragility curves can be drawn. On the high confidence curve, we can locate the fragility value of 5% the acceleration corresponding to this fragility on the high confidence curve is the so called "high-confidence-of-low-probability-of-failure" (HCLPF) capacity of'the component. By characterizing the component fragility through a family of fragility curves, the analyst has expressed all his knowledge about the seismic capacity of the component along with the uncertainties. Given I the same information, two analysts with similar experience and expertise would produce approximately the same fragility curves. Development of the family of fragility curves using different failure models and parameters for a large number of components in a seismic PRA is impractical if it is done as described above. Hence, a simple model for the fragility was propored as described in the above cited references. In the following section this fragility model is described. A 1 Fragility Model l The entire fragility family for an element corresponding to a particular failure mode can be expressed in terms of the best estimate of the median ground acceleration capacity, Am, and two random variables. Thus, the ground ace,eleration capacity, A, is given by: A'= AmE R EU,' (A-1) in which ER and su are random variables with unit medians, representing, respectively, the inherent randomness about the median and the uncertainty in the median value, in this model, we assume that both ER and EU are ir.Caormally
' distributed with logarithmic standard deviations, RS and SU , respectively. The 'I formulation for fragility given by Eq. A 1 and the assumption of lognormal distribution allow easy development of the family of fragility curves which +-- y 2.r 22 i o
- appropriately represent fragility uncertainty. For the quantification of fault trees in the plant system and accident sequence analyses, the uncertainty in fragility needs to be expressed in a range of conditional failure probabilities for a given ground acceleration. This is achieved as explained below.
With perfect knowledge (i.e., only accounting for the random variability, BR), the conditional probability of failure, f a, for a given peak ground acceleration level, a, is given by: r a'~ law f=@ o (A-2) OR where @(.) is the standard Gaussian cumulative distribution function. The relationship between foand a is the median fragility curve plotted in Figure A-2 for a component with a median ground acceleration capacity Am = 0.90g and GR= 0.30. For the median conditional probability of failure range of 5% to 95% the ground acceleration capacity would range from 0.55g to 1.48g. When the modeling uncertainty G U is included, the fragility becomes a random variable (urartain). At each acceleration value, the fragility f can be represented by a subjective probability density function. The subjective probability, Q (also known as " confidence") not exceeding a fragility f' is related to f' by: a in + U e-1(O) f' = $ - - (A-3) OR where: O = P[f < f' l al i.e., the subjective probability (confidence) that l the conditional probability of failure, f, is less than f' for a peak l l ground acceleration a.
@-l (.) = the inverse cf the standard Gaussian cumu?ative distribution function.
P:\42113 o8\oC.APA. doc \lRv A3 (
1 1 For example, the conditional probability of failure f' at acceleration 0.4g that has a 95% nonexceedance subjective probability (confidence) is obtained from Eq. A-3 as 0.22. The 5% to 95% probability (confidence) interval on the failure at 0.4 9 is 0 to 0.22. Subsequent computations are made easier by discretizing the random variable probability of failure f into different intervals and deriving probability q; for each interval (Figure A-3). Note that the sum of gj associated with all the intervals is unity. The process develops a family of fragility curves, each with a'1 associated probability gj. The median ground acceleration capacity Am, and its variability estimates GR and GU are evaluated by taking into account the safety margins inherent in capacity predictions, response analysis, and equipment qualification, as explained below. A.2 Failure Modes The first step in generating fragility' curves such as those in Figure A-2 is to develop a clear definition of what constitutes failure for each of the critical elements in the plant. This definition of failure must be agreeable to both the structural analyst generating the fragility curves and the systems analyst who must judge the consequences of component failure. Seteral modes of failure (each with a different consequence) may have to be considered and fragility curves may have to be generated for each of these modes. For example, a motor-actuated valve may fail j in any of the following ways: l
- 1. Failure of power or controls to the valve (typically related to j the seismic capacity of such items as cable trays, control panels, and emergency power). Since these failure modes are not related to the rpecific item of equipment (i.e., motor actuated valve) and are common to all active equipment, such faiiure modes are most easily handled as failures of separate systems linked in a series to the equipment.
l
- 2. Failure of the motor, ,
l 1 l l 3. Binding of the valve due to distortion and, thus, failure to operate.
- 4. Rupture of the pressure boundary.
P:\42113-o8\oC-APA. doc \lRV A-4
I i l i it may be possible to identify the failure mode most likely to be caused by the seismic event by observations during the walkdown or by reviewing the equipment design and considering only that mode. Otherwise, fragility curves are developed based on the premise that the component could fail in any one of all potential failure modes. Identification of the credible modes of failure is largely based on the analyst's experience and judgment. Review of plant design criteria, calculated stress levels in relation to the allowable limits, qualification test results, seismic fragility evaluation studies done on other plants, and reported failures (in past earthquakes, in licensee event reports and fragility tests) are useful in this task. Structuus are considered to have failed functionally when they cannot perform their designated functions, in general, structures have failed functionally when inelastic deformations under seismic load are estimated to be sufficient to potentially interfere with the operability of safety-related equipment attached to the structure, or fractured sufficiently so that equipment attachments fail. These failure modes represent a conservative lower bound of seismic capacity since a larger margin of safety against collapse exists for nuclear structures. Also, a structural failure has been generally assumed to result in a common cause failure of multiple safety systems, if these are housed in the same structure. For example, the service water pumps may be assumed to fail when the enclosure pump house roof collapses. Structures which are susceptible to sliding are considered to have failed when sufficient sliding deformation has occurred to fail buried or interconnecting piping or electrical duct banks. For piping, failure of the support system and fracture of the pressure boundary are credible failure modes. Failure modes of equipment examined may include structural failure modes (e.g., bending, buckling of supports, anchor bolt pullout, etc.), functionai failures (binding of valve, excessive deflection in rotating equipment), and breaker trip or relay chatter. Consideration should also be given to the potential for soil failure modes (e.g., liquefaction, toe bearing pressure failure, base slab uplift, and slope failures). For buried equipment (i.e., piping and tanks), failure due to lateral soil pressures may be an important mods. Seismically induced failures of structures or equipment under .
\
P342113 06\oC APA. doc \lRv A5
impact of another structure or equipment (e.g., a crane) may also be a consideration. seismically induced failures of dams, if present, resulting in either flooding or loss-of-cooling-source, should also be investigated.
~
A.3 Estimation of Fragility Parameters in estimating fragility parameters, it is convenient to work in terms of an intermediate random variable called the factor of safety. The factor of safety, F, on peak ground acceleration capacity above a reference level earthquaka specified for design,i.e., the safe shutdown earthquake level specified for design, ASSE, is defined as follows: A =FASSE p = Actual seismic capacity of element Actual response due to SSE
= Actual capacity ___
Calculated capacity X Calculated capacity Design response due to SSE X Design response due to SSE Actual response due to SSE F is further simplified as: p = Actual capacity
~
. Design response due to SSE X Design response due to SSE Actuai response due to SSE F =FFC RS (A-4)
Note that F can also bo defined with reference to a different earthquake such as the operating basis earthquake (OBE) level. The median factor of safety Fm, can be directly related to the median ground acceleration capacity, Am, as:
- l e
L
Fm * (A-5) ASSE The logarithmic standard deviations of F, representing inherent randomness and - uncertainty, are identical to those for the ground acceleration capacity A. For structures, the factor of safety can be modeled as the product of three random variables: F = FsF pFRS , (A-6)
' The strength factor, FS, represents the ratio of ultimate strength (or strength at loss-of-function) to the stress calculated for ASSE. In calculating the value of FS , the .
nonseismic portion of the total load acting on the structure is subtracted from the strength as follows: S-P3 FS'= (A-7) PT-PN' where S is the strength of the structural element for the specific failure mode, PNs i the normal operating load (i.e., dead load, operating temperature load, etc.) and PT is the totalload on the structure (i.e., sum of the seismic load for ASSE and the normal operating load). For higher earthquake levels, other transients (e.g., SRV discharge) may_ have a high probability of occurring simultaneously with the earthquake. The definition of P N in such cases should be extended to include the loads from these transients. The inelastic energy absorption factor (ductility factor), Fy , accounts for the fact that an earthquake represents a limited energy source and many structures or
, equipment items are ' capable of absorbing substantial amounts of energy beyond
. yield without loss-of function. A suggested method to determine the deamplification effect resulting from inelastic energy dissipation involves the use of ductility modified response spectra (Newmark, Reference 14). The deamplification factor is primarily a function of the ductility ratio p defined as the ratio of maximum j .
displacement to displacement at yield. Mcre recent analyses (Riddell and . p . Newmark, Reference 15) have shown the deamplification factor to also be a function of system damping. One might estimate a median value of for low-rise PM2113 o8\oC-APA. doc \mV A-7 1 .J ,
. - ~ . . ~ . . . , , . -
concrete shear walls (typical of auxiliary building walls) of 4.0. The corresponding median F value would be 2.45 at 7% damping. The variabilities in the inelastic energy absorption factor, Fp, are both estimated for this case as BR = 0.21 and GU
= 0.21, taking into account the uncertainty in the predicted relationship between p , and system damping.
F, The structure response factor, FRS, is based on recognition that in the design analyses, structural response was computed using specific (often conservative) deterministic response parameters for the structure. Because many of these parameters are random (often with wide variability) the actual recponse may differ substantially from the calculated response for a given peak ground acceleration. The structure response factor, FRS, is modeled as a product of factors influencing the response variability: FRS =FSA FGMI FS FM FMC F EC Fggi (A-8) where: FSA = spectral shape factor representing variability in grcund motion and associated ground response spectra. FGMI = ground motion incoherence factor which accounts for the fact that a traveling seismic wave does not excite a large foundation uniformly. i FS = damping factor representing variability in response due to difference between actual damping and Osign damping. FM = modeling factor accounting for uncertainty in response due to modeling assumptions. FMC = mode combination factor accounting for variability in response due to the method used in combining dynamic ! modes of response. l l 1 P:\42113-08\oC-APA. doc \lRv A-B
l l FEC. = earthquake component combination factor accounting for variability in response due to the method used in combining earthquake components. FSSI = factor to account for effect of soil-structure interaction including the reduction of input motion with depth below the surface. The median and logarithmic standard deviations of F are expressed as:
= F F Fm_ Sm #m FSA m FGMi m F6 m Mm F F (A-9)
MCm ECm Fssl m and~ r (A-10) Sp' = (Bs' + Sp ' + BsA* + GGMI' + + BSSi )1/2 The logarithmic standard deviation Bp is further divided into random variability, GR, and uncertainty, BU. To obtain the median ground acceleration capacity Am the median factor of safety, Fm,is multiplied by the referen::e earthquake peak ground acceleration. I For OCNGS, probabilistic structural response was conducted and probabilistic loads and spectra were derived. The methodology for development of probabilistic response is contained in Reference 9. Reierence 10 contains the results, when - probabilistic response analyses are conducted', the above described factors and then l variability are' inclusive and are not developed separately. 1 For equipment and other components, the factor of safety is composed of a capacity factor, FC a structure response factor, FRS: and an equipment response (relative to :.he structure) factor, FRE. Thus, F .= F C FRE FRS (A-11) l The capacity factor F C for the equipment is the ratio of the acceleration level at
' which the equipment ceases to perform its intended function to the seismic design level. . This acceleration level could correspond to a breaker tripping in a switchgear, PA42 tis-osioc APA.ooCVRV A9 . _ . ,
i
excessive deflection of the control rod drive tubes, or failure of an equipment support. The capacity factor for the equipment may be calculated as the product of FS and Fp . The strength factor, Fs,is calculated using Eq. (A-7). The strength, S, of equipment is a function of the failure mode. Equipment failures can be classified into three categories:
- 1. Elastic functional failures
- 2. Brittle failures
- 3. Ductile failures Elastic functional failures involve the loss of intended function while the component is stressed below its yield point. Examples of this type of failure include the following:
. Elastic buckling in tank walls and component supports
. Excessive blade deflection in fans The load level at which functional failure occurs is considered the strength of the I component.
1 Brittle failure modes are those which have little or no system inelastic energy absorption capability. Examples include the following:
. Anchor bolt failures J I
. Component support weld failures
. Shear pin failures Each of these failure modes has the ability to absorb some inelastic energy on the component level, but the plastic zone is very localized and the system ductility for an anchor bolt or a support weld is very small. The strength of the component failing in a brittle mode is therefore calculated using the ultimate strength of the material.
Ductile f ailure modes are those in which the structural system can absorb a
.significant amount of energy through inelastic deformation. Examples include the j following:
P:142113-oMoC APA.DoCilRv A 10
- Pressure boundary failure of piping or vessel nozzles e Structural failure of cable trays and ducting
- Failure of component support members (plastic bending, plastic buckling)
The strength of the component failing in a ductile mode is calculated using the yield strength of the material for tensile loading. For flexural loading, the strength is defined as the limit load or load to develop a hinge mechanism. The inelastic energy absorption factor, Fp, for a piece of equipment is a function of the ductility ratio,p. The median value of Fp is considered close to 1.0 for brittle and functional failure mode's. For ductile failure modes of equipment that respond in the amplified acceleration region of the design spectrum, the ductility is calculated in a manner sirnitar to structures (Reference 15). The equipment response factor FRE, is the ratio of equipment response calculated in the design to the realistic equipment response; both responses being calculated for design floor spectra. F REi s the factor of safety inherent in the computation of equipment response. It depends upon the response characteristics of the equipment and is influenced by some of the variables listed under Eq. (A-8). These variables differ according to the seismic qualification procedure. For equipment qualified by dynamic analysis, the important variables that influence response and variability are as follows:
* . Qualification method (QM)-
- Spectral shape (SA) -including the effects of peak broadening and smoothing, and artificial time history generation e Modeling (affects mode shape and frequency results) (M)
- Damping (5) e Combination of modal responses (for response spectrum method) (MC) 7
- Combination of earthquake components (EC)
;. =
a (
For rigid equipment qualified by static analysis, the variables, except the qualification method, and combination of earthquake components are not significant. The equipment response factor is the ratio of the specified static coefficient divided by the zero period acceleration of the floor level where the equipment is mounted if the equipm'et is flexible and was designed via the static coeffic.ent method, the dynamic characteristics of the equipment must be considered. This requires estimating the fundamental frequency and damping, if the equipment responds predominantly in one mode. The equipment response factor-is the ratio of the static coefficient to the spectral acceleration at the equipment fundamental frequency. Where testing is conducted for seismic qualification, the response factor must take , into account the following:
. - Qualification method (QM) e Spectral shape (SA) e Boundary conditions in the test versus installation (BC) e Damping (6) e Spectral test method (sine beat, sine sweep, complex waveform, etc.) (STM) e Multi-directional effects (MDE)
The overall equipment response factor is the product of these factors of safety corresponding to each of the variables identified above. The median and logarithmic standard deviations for randomness and uncertainty are estimated following Eqs. (A-9) and (A-10). The structural response factor, FRS, is based on the response characteristics of the structure at the location of the component (equipment) support. The variables pertinent to the structural response analyses used to generate fl'oor spectra for
. equipment design are the only variables of interest to equipment fragility. Time-history analyses using the same structural models used to conduct structural response analysis for structural design are typically used to generate floor spectra.
The applicable variables are as follows:
.
- Spectral shape
= Ground motion incoherence
. Damping
- Modeling
- Soil-structure interaction including reduction with depth of seismic input Fcr equipment with a seismic capacity level that has been reached while the structure is still within the elastic range, the structural response factors should be calculated using damping values corresponding to less than yield conditions (e.g.,
about 5% median damping for reinforced concrete). The combination of earthquake components is not included in the structural response since the variable is to be addressed for specific equipment orientation in the treatment of equipment response. Median Fmand variability BR and By estimates are made for each of the parameters affecting capacity and response factors of safety. These median and variability estimates are then combined using the properties of lognormal distribution in accordance with Equations (A-9) and (A-10) to obtain the overall median factor of safety Fmand variability BR and GU estimates required to define the fragility curves for the structure or equipment. For each variable affecting the factor of safety, the random (GR) and uncertainty (GU) variabilities must be separately estimated. The differentiation is somewhat judgmental, but it can be based on general guidelines. Essentially, BR represents variability due to the randomness of the earthquake characteristics for the same acceleration and to the structural response parameters which relate to these characteristics. The l dispersion represented by G U is due to factors such as the following:
- Our lack of understanding of structural material properties such as strength, inelastic energy absorption, and damping.
l
- Errors in calculated response due to use of approximate modeling of the structure and inaccuracies in mass and stiffness representations.
PM2113-08;oC APA. doc \lRV A-13
e Usage of engineering judgment in lieu of complete plant-specific data on fragility levels of equipment capacities, and responses. Probabilistic floor response spectra were developed for OCNGS (Reference 10) and all important variables were included in the spectra development. Spectra were developed for different ground motion input levels to capture the variation of damping in the structures and soil, b 4 I t
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C o O O Peak Ground Acceleration Figure A-1: Fragility Curves
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--i._____________.______._______
1.0 e 95% . B Nonexceed. -
% P r o ba bili t y-
- u. 0. 0 Tc onfidence Median /
o Curve Fragility E Curve
- 0.6 2
r; D' o . E 0.4 - l
% l -
b %' E / Nonexceed, 5 Probability 0 .2 -
'9 (Confidence) 5 Curve U g L ! I i 0.4 0,8 1.2 1.6 2.0 SSE Peak Ground A6celera tion o l
Figure A 2: Median,5% Non-Exceedance, and 95% Non-Exceedance Fragility Curver for a Component
^ .:
PA42113mioc-APA. Doc \my ' A 16 C-- _. I
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O O i i i 0 0.4 0.8 1.2 1.6 2.0 Peak Ground Acceleration, g Figure A-3: Family of Fragility Curves for a Component f P:\42113-08\0C APA. DOC \lRV A-17
E I 6 APPENDIX B COMPARISON OF 0.184G A-46 AND 3XSSE MEDIAN FLOOR RESPONSE SPECTRA FOR
, REACTOR BUILDING l
i l . I l l
3 3 4 2 2 4 2 2 4 3 3 5 4 4 5 4 4 5 3 3 6 2 2 4 3 2 4 3 3 4 3 3 4 3 3 4 - b hm0 ? 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 e eK 34
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f-l l l-l l ATTACllMENT 7 INFORMATION NOTICE 93-53 EVALUATION IPEEERAI.IXX' 04/10/98
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1 l Nuclear memorandum OC LICENSING ACTION ITEM NO. 93090.03 August 5, 1994 subject. 0' ate: IN 93-53, 51: EFFECT OF HURRICANE ANDREW ON TURKEY POINT NUCLEAR GENERATING STATION AND LESSONS LEARNED
* ' 8
- Morr.is Corp. Center R. Earadaran - Engineer, Risk Analysis 1 5430-94-0016 B. Demerchant - Engineer, Licensing OC NRC IN 93-53 requests that all holders of operating licenses for nuclear power reactors review the information provided for the effects of Hurricane Andrew 6n the Turkey Point Nuclear Generating Station and l'essons learned for a. applicability to their facilities. The concerns are mostly applicable to the pre-1975 nients that do not meet Regulatory Guide 1.76, " Design Basis Tornado for Nuclear Power Plants," and NUREG 0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants."
Oyster Creek was evaluated against above guidelines by NRC and the findings were reported in the " Integrated Plant Safety Assessment Report (IPSAR)." All the concerns raised as part of that effort have been resolved. I The following are excerpts of the evaluation performed for Oyster Creek in this area as part of the Individual Plant Examination of External Events (IPEEE), which provides more details. It is deemed that this issue has beeri fully investigated for Oyster Creek and all the relevant concerns have been reviewed.
"5.1.3 Tornado Wind Fragility,of Structures 10 CFR Part 50 (GDC 2), as implemented by SRP Sections 3.3.1 and 3.3.2 and Regulatory Guides 1.76 and 1.117, requires that new plants be designed to withstand the effects of natural phenomena such as wind and tornados.
The effects of tornados were not considered.in the original design of the Oyster Creek' structural systems. In Integrated Plant Safety Assessment Report (IPSAR), Sections 4.3.1 through 4.3.9, the staff identified some structures and components important to safety that did not meet current licensing criteria, which require that they be adequate to resist tornado winds of 250 miles per hour and a differential pressure of 1.5 pounds per square
. inch (Reference 9). The following were identified in the IPSAR as not meeting the prescribed criteria (Reference 9):
N 0648 (06-86)
B. Demerchant-August 5, 1994 l 5430-94-0016 Page 2
- 1) reactor building steel structure abos the operating floor
- 2) ventilation stack
- 3) effects of failure of non-qualified structures
- 4) components not enclosed in qualified structures 5 exterior masonry walls -
6 roof decks (7 intake structure, oil tanks, and diesel generator building (8) load combinations (9) soil and foundation capacities However, in IPSAR Sections 4.3.4, 4.3.5, and 4.3.9, the staff concluded that further evaluation of items (4), (5), and (9) was not warranted (Reference 9). GPUN responded to the remaining issues in submittals dated February 2 and October 25, 1983, and February 2, March 13, and June 4, 1984. On the basis of a letter from the staff to GPUN dated March 8, 1986 (Reference 8), which provided an evaluation of the responses, in IPSAR Supplement 1, Sections 2.3.2, 2.3.3, and 2.3.4, the staff reported that items (2), (3), and (6) were resolved. Item (7)has been resolved on the following basis: The licensee proposed that safe shutdown could be achieved and maintained in case of loss of intake structure, oil tanks, and diesel generator building with makeup water provided to the isolation condenser from the torus i by the main core spray pumps. The staff reviewed this proposal and concluded that although the flow path itself is acceptable,
. the core spray pumps rely on the emergency diesel generators for motive power.
In its safety evaluation dated February 26, 1990, the staff q concluded that the walls of the diesel generator vaults and the oil tank compartment are capable of withstanding the loads generated by a tornado having a wind speed of 168 miles per hour and are acceptable. However, the staff required that GPUN provide adequate protection to the outside fuel supply line against the potential missile strike (Reference 9). In letters dated April 16 and July 27, 1990, GPUN committed to install a safety-grade check valve and a safety-grade gate valve . in the supply line inside the emergency diesel generator fuel tank room. The installation of these valves is intended to prevent the fuel oil supply from backflowing out of the 15,000-gallon diesel l l N
? B. Dcierchant August 5, 1994 , 5430-94-0016 Page 3 I i l generator fuel storage tank (day tank) in the event of a rupture of the fuel supply line outside the fuel storage tank room. The staff reviewed the licensee's proposed changes to the diesel generator supply line and the proposed modification to protect the day tank fJel supply to the diesel generators and found them acceptable in its safety evaluation dated November 28, 1990. However,. the staff's acceptance of the proposed design is predicated on its finalization and implementation (Referen:e 9). In IPSAR Supplement 1, Sections 2.3.7 and 2.3.8, the staff identified the following two additional items related to tornado missile damage on the basis of the letter af March 8,1986: (10) control room and (11)' architectural components, respectively. GPUN addressed items (1), (8), (10), and (11) in letters dated November 15, 1990 and October 3, 1991. In. the letters, GPUN described a planned upgrading of the upper reactor building structure by adding cross bracos to the roof fr ming and provided justifications for other items of concern. NRC issued letters dated July 29, 1992 and December 7, 1992 (References 10 and 11) in which the staff concluded.that the upper steel-framed portion of the Reactor Building will be able to withstand the tornado-wind loading generated by a total wind speed (i.e., rotational plus transnational) of up to 306 KM/h (190 . mi/h), when the planned modifications to the steel . framing are completed (GPUN has committed to complete the modification during Cycle 14, 1993 - 1994 time frame, Reference 12). NRC also considered all other tornado missile related ssues resolved by above letters on the basis of having a system consisting of the isolation condenser (IC), torus, and a core spray pump available for both shutting down the plant and maintaining it in a shutdown mode. More discussion of this shutdown path is available in Section 5.1.6 titled: " Tornado Missile Fragility of Structures." , In summary, all the structures and components important to safety are capable of withstanding at least a tornado of a wind speed of 1 168 miles per hour after all the planned modifications are I I completed. k i 1 1 e wh___.__._. ___.__-_._
.B. Demerchant August 5, 1994' 5430-94-0016 Page 4 This is the wind speed that the staff has concluded that the most
-vulnerable structure, the walls of the diesel generator vaults and the oil tank compartment,:can withstand and the other structures can take much higher wind speeds."
and "5.1.6 Tornado Missile Fragility of structures 4 10CFR (GDC2), as implemented by Regulatory Guide 1.117, . prescribes structures, systems, and components that ,should be designed to withstand the effects of a tornado, including tornado missiles, without loss of. capability to perform their safety functions. Regulatory Guide 1.117 requires that structures, systems, and - components that should be protected from the effects of a design-basis tornado are.-(1) those necessary to ensure the integrity of the reactor coolant pressure boundary, (2) those necessary to ensure the capability to shut down the reactor and maintain it in a safe shutdown conditions (including both hot standby and cold shutdown), and (3) those whose failure could lead to radioactive releases resulting in calculated offsite exposures greater than 25%~of the guideline exposures of 10 CFR 100 using appropriately conservative analytical methods and assumptions. The physical separation of redundant or alternate structures or components required for the safe shutdown of the plant is not considered acceptable by itself for providing protection against the effects of tornados, including tornado-generated missiles, because of the large number and random direction of potential missiles that could result from a tornado as well as the need to consider the single-failure criterion (Reference 15).
- The following structures and components were found vulnerable to tornado missiles at Oyster Creek according to NRC staff IPSAR (Reference 15):
l (1) emergency diesel generators and fuel oil day tank I (2) mechanical equipment access area
- l. (3) ' control room, reactor building, and turbine building heating, ventilating, and air conditioning systems condensate storage tank torus water storage tank.
service water and emergency service water pumps ( j, l: . l L1-
m ,
. B. Deserchant August 5 1994-i '5430-94-0016:
Page 5; L l 1 Item-(3) was considered resolved by the_ staff on the fo11' owing
-basis (Reference 15):
Since the-intake for the control room HVAC system is located in the reactor building wall and is. not protected from tornado missiles, GPUN has committed to install a remote shutdown
, capability as part of his Appendix R modification; therefore,.the plant would be able to shut down in'the event the control room is lost.
The staff has found this acceptable because GPUN has committed that the location of the remote shutdown capability systems and all supporting systems would be protected from the effects of t:renado missiles. As far as the Reactor Building and Turbine Building HVAC is concerned, GPUN has concluded that the plant can be safely shutdown with a loss of HVAC. The demonstration of that ability has been evaluated in conjunction with SECY 82-207, "The Final Rule for Environmental Qualification of Safety Related Electrical Equipment for Nuclear Power Plants." Items 2, 4, 5 and 6 were found acceptable by the staff (Reference 9 and 11) based on the following available alternate shutdown path (Reference 16):
"GPUN indicateil in its August 14, 1987 submittal that a detailed I field walkdown identified an existing system interconnection between co,'e spray and the isolation condenser using the suppression chamber (torus) water as the protected water ' supply.
This water supply and the interconnection are located in the Reactor Building, below 119' -3" elevation, where they are protected from tornado missiles or potential . external flooding. During a tornado event at Oyster Creek, decay heat from the ' reactor will be removed by the isolation condenser where the boiloff from the shell side will be released to the atmosphere. The licensee proposed to supply the necessary makeup, due to the boiloff, by using the system interconnection and the torus as water supply. Credit for torus water is given during a tornado missile event for emergency fill, because a design basis accident, plant transient, or failure of core spray system isolation valves
. are not postulated to occur coincidentally with a tornado missile event._. The fill flow will be pumped by the Core Spray System I main pump from the torus to the isolatiot !.ondenser through the 9
iii___E_._._____________i____i_________...__z___.____
4 -
- 8; Demerchant' August 5,1994' 5430-94-0016 ~
1 Page' 6
' fill line. The licensee calculated that by core adding spray systp(differential 10,000 ft volume between maximum and minimum torus water level) of torus water to the isolation condenser shall along with the amount of water initially in the isolation condenser shell, the system is capable af removing decay heat for i about 13 hours which provides sufficient time to ensure an emergency reactor shutdown. Also, the licensee indicated that the inventory water in the isolation condenser could remove decay heat for the initial 1.5 hours or until the fill path is established.
This emergency- fill path involves manual operation of the following valves: valves V-11-110, V-11-111, and V-11-346 change position- from close- to- open; valve V-11-15 changes from open to close. The licensee indicated in their August 12, 1988 submittal that these valves are readily accessible for manual operation during either a tornado event or an external flooding event. The operator would take these actions as soon as the procedure for the event is initiated. Because the inventory water in the isolation condenser could remove decay heat for 1.5 hours, the staff concledes that the operator would have sufficient time to manually L operate these valves for the emergency fill path. , The licensee indicated that the Core Spray Pump NPSH is approximately 39' available at - the minimum torus water level during the transient and that the required NPSH is 12' at 2600 gps. Therefore, the required NPSH for' the Core Spray Pump is about 1/3 of the NPSH available. This ensures the pump sufficient NPSH if torus water level drops to the end of the downcomer. The licensee performed an analysis which determined that the system flow rate:using the system interconnection will provide sufficient decay heat removal for a safe reactor shutdown. The licensee also performed a radioactivity dose calculation for the small amount of radioactive cesium in the torus water. In a tornado missile event, the radioactivity releas~e would be within permissible levels of radiation to an unrestricted area according to 10 CFR 20.105. The calculated dose is only a fraction of the
' 10 CFR 20.105 limit.
' GPUN indicated that the revision schedule for the' Iso-Condenser- -
- Diagnostic and Restoration Actions, procedure 2000-0PS-3024.18 is
.0ctober 30, 1988. This procedure is intended to provide specific
. guidance. to the operators to identify the needed actions to initiate emergency fill path. . The training of the operators to
- follow this procedure and the. implementation of this procedure 4
u__ _
B. Demerchant August 5, 1994- . 5430-94-0016 Page 7 f will be completed within a few days after NRC's approval of this I proposal." Based on above, the staff concluded:
" Based on our review, the staff concludes ' that the proposed emergency fill path is protected from tornado missiles or external '
flooding and that the reactor. emergency safe shutdown pathway
' during a tornado missile event is ensured for the following reasons:
- a. The core spray pump using the system interconnection will- have a system flow rate to provide ' sufficient >
heat removal, and the system is . capable of removing decay heat for 13 hours.
- b. The manual operated valves in the emergency fill flow path are. readily accessible during either- a tornado missile event or an external flooding event. .
- c. The core spray pump has sufficient NPSH to operate should the torus water level drop to downcomer submergence.
- d. The small amount of radioactive cesium in the torus water released to the environment through the isolation condenser shell side boiloff is only a '
fraction of the 10 CFR 20.105 limit. ,
- e. The licensee's revision schedule for the Iso-Condenser-Diagnostic and Restoration Action procedure is October 30, 1988.
Therefore, the staff concludes that the proposed emergency fill ~ path is acceptable for a tornado missile event or an external flooding event with the loss of all external water sources. The viability'of this water supply is dependent on diesel generators for motive power. Protection for the diesel generators including l missile protection is being r.eviewed under a related issue (see Safety Evaluation dated February 26,.1990)." l As discussed in Section 5.1.3, the staff has accepted the changes proposed by GPUN to protect the outside fuel supply line to the
~
. diesel generators.against potential missile strike (Reference 9)
.p, d
A B. Demerchant' August 5, 1904 4 , !- 5430-94-0016 l - Page 8 -
- and the issue with the diesel generators protection is considered resolved. .
One' issue that surfaced during - the review of. the proposed alternate shutdown path was protection of the controls for this system in the control room, wh:ch are located adjacent to the north wall. . The staff, in evaluating the susceptibility of the control room (CR) to damage by tornados, found that the north wall of the CR would be damaged if struck directly by a utility pole of the type and size described in SRP Section 3.5.1.4, " Missiles Generated by Natural Phenomena," during a tornado strike with a wind velocity of 168 MPH. Such a strike could cause spalling of the inner portions of the wall which could, in turr., damage the controls for the proposed shutdown system; there is no potential for missile penetration (Reference 10). However, the staff concluded that (Reference 10): "the utility pole would have to hit the wall almost perpendicularly and in the specific locale of the contro' in order to preclude use of the shutdown system. GPUN has a established procedure for shutting down using normal shutdown methods in the event of a tornado watch or tornado warning, thereby reaching shutdown before a tornado could strike the plant. GPUN also has an established procedure for shutting the plant down, employing the' system described above (IC, torus and core spray pump) as nedded. A In view of the foregoing, the staff considers that GPUN complies with the intent of the criterion of SRP Section 2.2.3 and also, therefore, with the intent of SRP Section 3.5.1.4, " Missiles Generated by Natural Phenomena," with regard to safe shutdown in
. - the event of tornado missiles."
" e 4
F Q aradaran Extension 7549 RB/jh
- cc: _
L. C. Lanese - Manager, Risk Analysis / Human Engrg. RA File # 4.93090.03 4-4
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