ML13183A280

Entergy Nuclear Engineering, IP-RPT-11-00012, Rev. 0, Reassessment of Indian Point 3 Seismic Core Damage Frequency, Dated April 2011

ML13183A280
Person / Time
Site:	Indian Point
Issue date:	04/07/2011
From:	Drake R Entergy Nuclear Northeast
To:	Office of Nuclear Reactor Regulation
References
NL-13-084 IP-RPT-11-00012, Rev 0
Download: ML13183A280 (169)

Text

ENCLOSURE 1 TO NL-13-084 Entergy Nuclear Engineering Report No. IP-RPT-11-00012, "Reassessment of Indian Point 3 Seismic Core Damage Frequency", dated April 2011 Entergy Nuclear Operations, Inc.

Indian Point Unit 3 Docket No. 50-286

ENTERGY NUCLEAR Engineering White Paper Engineering Report No IP-R PT-1 1-00012 Rev. 0 EC No. 28797 April2011 Resesmn of Inia Pon Sesi Cor Daag Frqec AN

• ýEnteWg

Engineering Report No. IP-RPT- 11-00012 Rev 0 Entergy ENTERGY NUCLEAR Engineering White Paper Reassessment of Indian Point 3 Seismic Core Damage Frequency April 8,2011 EC No. 28797 Prepared by: Date: 4/8/I1 Responsible Lafgineer (Print Name/Sign)

Reviewed by:

Drago Nuta, 5 E Date: _17111 Reviewer (Print Name/Sign)

Reviewed by:

MIkA T ISE Date: 4/'/s Revmie er (Print NamehSign)

Reviewed by: Paul Baughman, PE - ARES Consultant Reiee (P* Nm/gn)

Date:

Reviewed by: Paul "f *L u n,*qonsultant

/4-,. ý r2*

f Reviewer (Print NanPSign)

Reviewed by: Date:

Date, Approved by: Date:

Joe Abisamra, PE Team Lead (Print Name/Sign)

Entrg ENG REPORT IP-RPT-I1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency TABLE OF CONTENTS 1.0 Executive Summary ..................................................................... 4 2 .0 P u rp o se .................................................................................... 4 3 .0 B ackg ro und ............................................................................. ... 4 4.0 Definitions and Acronyms ........................................................... 6 4 .1 D efi nitio ns .................................................... . ................. 6 4 .2 A cro nym s .................................................... .................. 7 5.0 General Discussion .................................................................... 9 5.1 Assessment Approach ................................ 9 5.2 NRC Review Results ......................................................... 10 6 .0 E va luatio ns ................................................................................. 10 6.1 NRC Methodology in Determining SCDF ............................... 10 6.2 Methodology that emulates NRC's results ............................. 12 6.3 USGS Hazard Curve for IPEC ............................................. 12 6.4 Identification of Margin ........................................................ 13 6.4.1 Identification of Low Capacity Components ................. 13 6.4.2 In-Structure Response Spectra ................................ 14 6.4.3 Component Fragility with (SBO) ................................ 15 6.4.4 Component Fragility without (SBO) ........................... 16 6.4.5 Fragility Calculation ............................................. 19 6.5 Comparison of IP3 HCLPF vs. SSE ....................................... 21 6.6 Concern of the Unit 1 Stack ................................................. 22 6.7 Reassessment of SCDF .................................................. 22 6.8 Seismic Design and the Ramapo Fault ................................... 24 7.0 C onclusions ....................................................... .................. ... 25 8.0 Recommendations ...................................................................... 25 8.1 Short term Recommendations ............................................ 25 8.2 Long term Recommendations ............................................. 25 9 .0 R efe re nces ................................................................................ 26 10 .0 A ttachm ents .................................................................................... 27 11.0 Team Members ........................................................................... 27 Sheet 3 of 27

-Entergy ENG REPORT IP-RPT- 11-000 12 Reassessment of Indian Point 3 Seismic Core Damage Frequency 1.0 EXECUTIVE

SUMMARY

Recent NRC Safety/Risk Assessment of US Nuclear Plant Seismic Core Damage Frequencies based on the 2008 US Geological Survey (USGS) Hazard Curves identified IP3 as the plant with the largest calculated Seismic Core Damage Frequency (SCDF). To address this issue Entergy assembled a Seismic Review Team.

Although the NRC estimated SCDF remained below an acceptable value, the review team reassessed key IP3 components and demonstrated a larger plant-level seismic capacity than that used in the NRC assessment. The NRC used values for plant capacity extracted from the IP3 IPEEE report submitted in 1997. This resulted in the NRC determining a very conservative SCDF estimate of 1.0E-04 per year, or 1 in 10,000 reactor-years. Using the improved plant capacities developed by the team a re-assessment of the SCDF estimate was performed. This resulted in a SCDF of 9.4E-06 per year, or 1 in 106,383 reactor-years, using the same USGS Hazard Curves.

With the use of the improved plant capacity and EPRI updated 2010 hazard curves; the SCDF estimate is further reduced to 7.1 E-06 per year (or 1 in 140,845 reactor-years).

2.0 PURPOSE The purpose of this Engineering Report is to provide an assessment relative to the seismic robustness of IP3. A Seismic Review Team was therefore formed with an established main goal to reassess the Seismic Core Damage Frequency (SCDF) for IP3 and demonstrate whether the actual IP3 SCDF value is in fact lower than the NRC reported estimate of 1.OE-04 per year. (SCDF if not explicitly shown is expressed in occurrences per year throughout this document).

3.0 BACKGROUND

During NRC's review of the Early Site Permit (ESP) applications for new nuclear plants, it appeared that the seismic hazard for operating plants in the Central and Eastern United States (CEUS) region may have increased for some sites. Based on the evaluations of the Individual Plant Examination of External Events (IPEEE) Program, the NRC had determined that seismic designs of operating plants in the CEUS provided an adequate level of protection. However, in light of the preliminary results from the review of ESP applications, the NRC also recognized that the probability of exceeding the Safe Shutdown Earthquake (SSE) at some of the currently operating sites in the CEUS may be higher than previously thought. Therefore, the NRC initiated Generic Sheet 4 of 27

SEntergy EN REPORT IP-RPT4 100012 Reassessment of Indian Point 3 Seismic Core Damage Frequency Issue GI-1 99 to assess the impact of increased estimates of seismic hazards on selected current nuclear power plants in the CEUS region that might be impacted by the updated seismic research, information, and models.

The NRC issued on 9/2/10 Safety/Risk Assessment (SRA) results for Generic Issue 199, "Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants." [9.3] The Safety/Risk Assessment consists of three separate analyses performed using respectively; the 1989 EPRI (Electric Power Research Institute), 1994 LLNL (Lawrence Livermore National Laboratory), and the updated 2008 USGS (U.S. Geological Survey) seismic Hazard Curves to calculate the SCDF for all US nuclear plants. The results of the GI-199 assessment raise the possibility that the probability of exceeding the design basis ground motion may have increased at some sites, but only by a relatively small amount. The NRC concluded that "no concern exists regarding adequate protection and that the current seismic design of operating reactors provides a safety margin to withstand potential earthquakes exceeding the original design basis." However, since the changes in seismic core-damage frequency estimated in the Safety/Risk Assessment Stage of GI-1 99 for numerous plants lie in the range of 1.OE-04 per year to 1.OE-5 per year, the issue will proceed to the Regulatory Assessment Stage of the Generic Issues Program. It should be noted that new seismic hazard curves are being developed by USGS. It is expected that these new seismic hazard curves will be less conservative than those presently available. Given that NRC already stated the implications are not significant, NRC potential industry actions may be delayed pending issuance of the new hazard curves.

The objective of the GI-1 99 Safety/Risk Assessment (SRA) was to perform a conservative, screening-level assessment to evaluate if further investigations of seismic safety for operating reactors in the CEUS are warranted consistent with NRC directives.

The results of the GI-199 SRA should not be interpreted as definitive estimates of plant-specific seismic risk. The nature of the information used (both seismic hazard data and plant-level fragility information) make these estimates useful only as a screening tool.

The NRC does not rank plants by seismic risk.

Analysis results showed IP3 as having the largest SCDF value of the U.S. nuclear plants (SCDF estimate of 1.OE-04 per year) using the 2008 updated USGS seismic Hazard Curves [9.13] for the IPEC site. Note that this is still within the range considered acceptable in the NRC's process for evaluating emergent issues.

The NRC used information from the IP3 IPEEE [9.1] submittal to derive the plant-level fragility used to calculate the SCDF. Conservatism in the information could have Sheet 5 of 27

SEntergy ENG REPORT IP-RPT-11-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency resulted in an overly conservative (high) SCDF estimate, over and above the conservatism inherent in the methodology used by USGS in developing the seismic hazard curves.

As a result of the Tohoku-Taiheiyou-Oki Earthquake in Japan and its effect on the Fukushima Daiichi Nuclear Station, increase attention has been directed to nuclear plants and in particular IP3's ability to withstand an earthquake without suffering core damage.

4.0 DEFINITIONS & ACRONYMS 4.1 Definitions

" Composite Uncertainty (9.) - The composite logarithmic standard deviation of the capacity about the best-estimate, or median value. It represents the variation due to both inherent randomness and uncertainty.

• Fragility - The conditional probability of failure as a function of a ground motion parameter (for example, peak ground acceleration).

• High Confidence of a Low Probability of Failure (HCLPF) - The ground motion parameter (for example, peak ground acceleration) at which there is at least 95% confidence of less than 5% probability of failure (for example, core damage).

• Median Capacity (Am) - The ground motion parameter (for example, peak ground acceleration) for which there is a 50% probability of failure (for example, core damage).

• Plant-Level Fragility - The probability of core damage as a function of the site earthquake-induced vibratory ground motion.

• Seismic Core Damage Frequency (SCDF) - The expected frequency of core damage caused by earthquake ground motion (expressed as occurrences per reactor-year) for a nuclear power plant at a specific location.

" Seismic Hazard Curve - The annual frequency at which the site earthquake-induced vibratory ground motion exceeds a given value of a ground motion parameter (for example, peak ground acceleration)

• Seismic Hazard Map - A seismic hazard map displays an earthquake ground motion parameter (for example, peak ground acceleration) for a specific probability of occurrence at many locations (for example, across the United Sheet 6 of 27

S-Ente&W ENG REPORT IP-RPTr- 1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency States). Seismic hazard maps are derived from seismic hazard curves calculated for a grid of sites across the United States.

Uniform Hazard Spectrum (UHS) - A plot of the maximum acceleration of a single-frequency system over a range of frequencies for a ground motion with a specific probability of occurrence.

4.2 Acronyms

• ATWS - Anticipated Transient Without Scram

• CAV - Cumulative Absolute Velocity SCCW - Component Cooling Water

• CDF - Core Damage Frequency

• CEUS - Central and Eastern United States

• EDG - Emergency Diesel Generator

• EPRI - Electric Power Research Institute

• ESP - Early Site Permit

• GI - Generic Issue

• GIP - Generic Implementation Procedure

• GMI - Ground Motion Incoherence

• GMRS - Ground Motion Response Spectrum

• GSI - Generic Safety Issue

• HCLPF - High Confidence of a Low Probability of Failure

• IP1 - Indian Point Unit 1

• IP2 - Indian Point Unit 2

• IP3 - Indian Point Unit 3

• IPEC - Indian Point Energy Center

• IPEEE - Individual Plant Examination of External Events

• ISRS - In-Structure Response Spectra

• LLNL - Lawrence Livermore National laboratory

• LOCA - Loss of Coolant Accident Sheet 7 of 27

ENG REPORT IP-RPT-1 1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency

• LOOP or LOSP - Loss of Off-Site Power

• NRC - US Nuclear Regulatory Commission

• PGA - Peak Ground Acceleration

• PRA - Probabilistic Risk Assessment

• PSD - Power Spectral Density

• RCP - Reactor Coolant Pump

• RHR - Residual Heat Removal

• SBO - Station Blackout

• SCDF - Seismic Core Damage Frequency

• SEWS - Screening Evaluation Worksheet

• SHIP - Seismic Hazard Integration Program

• SMA - Seismic Margins Assessment

• SPRA - Seismic Probabilistic Risk Assessment

• SRA - Safety/Risk Assessment

• SSE or DBE - Safe Shutdown Earthquake (SSE) which is equivalent to the IPEC terminology of Design Basis Earthquake (DBE)

• UHS - Uniform Hazard Spectra

• USGS - U.S. Geological Survey Sheet 8 of 27

'Enterg ENG REPORT IP-RPT- 1-000 12 Reassessment of Indian Point 3 Seismic Core Damage Frequency 5.0 GENERAL DISCUSSION 5.1 Assessment Approach The team effort consisted of completing the following:

" Understanding the current seismic risk to the station, in terms of SCDF.

" Identifying potential conservatisms within the IP3 IPEEE submittal information used by the NRC in the SCDF calculation.

" Exploring, and where appropriate, removing conservatisms in the IPEEE component fragility values. Plant-level fragility value is expected to increase significantly. It should be noted that original fragility calculations cannot be located, and the team computed fragility values for a sample of components that are high contributors to risk using original component design documents (See Attachment 7).

• Exploring, and where appropriate, reducing conservatism in the USGS seismic Hazard Curves. For example, the USGS seismic hazard calculations did not include Cumulative Absolute Velocity (CAV) filtering and ground motion incoherence, which tend to reduce the acceleration values.

• Developing a more realistic SCDF using the revised fragilities and hazard curves. It should be noted that the original PRA computer model Seismic Hazard Integration Program (SHIP) is not available to be modified and an alternate approach has been developed by the team - (See Attachment 6).

" Developing recommendations for long-term actions to improve the seismic risk estimates to more accurately reflect the robustness of the plant.

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ENG REPORT IP-RPT-1 1-00012 IMEntaWg Reassessment of Indian Point 3 Seismic Core Damage Frequency 5.2 NRC Review Results The results of the NRC analysis, described in Section 6.1, for IP2 and IP3 are summarized below, Table 5.1:

Table 5.1 NRC Results for SCDF SEISMIC CORE DAMAGE FREQUENCY INCREASE INCREASE PLANT WEAKEST LINK 2008 USGS 1989 EPRI 1994 LLNL USGS-EPRI USGS-LLNL IP2 3.3E-05 1.4E-05 2.6E-05 1.9E-05 TOE-06 1 in 30303 yrs 1 in 71429 yrs 1 in 38462 yrs IP3 1.OE-04 5.8E-05 5.8E-05 4.2E-05 4.2E-05 1 in 10000 yrs 1 in 17241 yrs 1 in 17241 yrs The IP2 and IP3 results are quite different and the IP2 results appear to be more consistent with similar plants. This indicates that there is excessive conservatism in the IP3 SCDF numbers. The NRC acknowledges [9.7] that "The results of the GI-199 safety risk assessment should not be interpreted as definitive estimates of plant-specific seismic risk because some analyses were very conservative making the calculated risk higher than in reality". It should also be noted that the two units were owned by two different companies [New York Power Authority (IP3) and Consolidated Edison (IP2)]

with separate contracting engineering firms performing the IPEEE study, which led to the differences in the analyses.

As a result of the GI-1 99 SRA, the NRC compiled a list of 27 plants requiring further review. Indian Point 3 and Indian Point 2 are among those selected plants. The criteria used for selection by NRC are described in Attachment 1.

6.0 EVALUATIONS 6.1 NRC Methodology in Determining SCDF The SCDF is calculated as:

sCDF = pa e fi da ft Where P(a) isthe plant-level fragility function and H(a) isthe seismic hazard curve.

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__Entarg ENG REPORT IP-RPT-II-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency The equation is evaluated numerically, as described in Attachment 5.

The fragility function used by the NRC was characterized by a median capacity, Am, and a composite uncertainty, Pc. The NRC derived these parameters from the IPEEE submittals for each plant. The submittal information varied from plant to plant; therefore, the NRC used different methods for different plants. The different methods are defined in Table C.1 of Appendix C [9.3]. The method used for each plant is listed in Table C.2 of Appendix C [9.3].

The method used for IP3 consisted of the following:

1. The NRC used 0.34g for median capacity, Am, the number given in Section 3.1.5.5 of the IP3 IPEEE submittal [9.1].

2. The NRC used 0.34 for composite uncertainty, Pc. This was calculated by assuming a HCLPF equal to the plant SSE, 0.15g, and calculating Pc = In(Am/HCLPF)/2.33. Note that Section 3.1.5.5 of the IP3 IPEEE listed the plant HCLPF as 0.13g; however, the NRC increased this to the 0.15g SSE (Refer to Section 6.5 for discussion).

For the seismic hazard the NRC used three sets of seismic hazard curves: USGS 2008, EPRI 1989 and LLNL 1994. Each set had individual hazard curves for PGA, 10 Hz, 5 Hz and 1 Hz. For a given set the NRC calculated the SCDF corresponding to each curve. The highest SCDF calculated, was used as the plant SCDF corresponding to the set of curves (the "weakest link approach"). The SCDFs for each plant for each set of curves are presented in Tables D.1, D.2 and D.3 of Appendix C [9.3].

The "weakest link" SCDF for IP3 for the USGS 2008 seismic hazard is listed in Table D.1 as 1.OE-04 per year [9.3]. The corresponding "weakest link" SCDFs for the EPRI 1989 hazard and LLNL 1994 hazard are 5.8E-05 and 5.8E-05 (coincidentally the same for both hazards). The NRC then used the differences between the SCDF for the USGS 2008 hazard and the SCDFs for the EPRI 1989 and LLNL 1994 hazards to select plants for further review.

The fragility information from the IPEEE submittals was in general characterized in terms of PGA. In order to calculate the SCDF for a 10 Hz, 5 Hz or 1 Hz seismic hazard curve, the fragility must be changed to be in terms of the spectral acceleration at that frequency. This was done by multiplying the median acceleration, Am, by a spectral ratio, m. The spectral ratios for all of the plants are given in Table C.2 of Appendix C of the GI-199 SRA [9.3].

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'-Entergy ENG REPORT IP-RPT-I 1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency 6.2 Methodology That Emulates NRC's Results A full understanding of the methodology used in the NRC report was achieved. The NRC Report results were duplicated, validating our understanding. The same methodology is used to develop the more realistic estimates of SCDF.

The methodology developed for calculating SCDF for a given hazard curve and fragility was derived and used in conjunction with the USGS Hazard Curves for which the acceleration interval was refined using a log-log interpolation between the raw acceleration data points for a 0.001 g interval. Excerpts from the SCDF analyses for the PGA, 10 Hz, 5 Hz, and 1Hz hazard curves showing the 0.001g discretization for each hazard curve are presented in Attachment 5.

The SCDF results obtained using the methodology described above, with the USGS Hazard Curve and IP3 fragility used by the NRC, match those reported in the NRC GI-199 Report. This same methodology was then used in the SCDF calculation estimates that more accurately reflect the robustness of the IP3 components as outlined in Attachment 6.

6.3 USGS Hazard Curves for IPEC In developing the SCDF estimates, the NRC report indicates that the IP3 Seismic PRA (SPRA) was used to establish plant-level fragility curve parameters. This plant level fragility was then convolved with seismic hazard curves derived by the United States Geological Survey (USGS) for the IPEC plant site longitude and latitude coordinates to obtain the estimates of the SCDFs based on the peak ground acceleration (PGA), 10 Hz, 5 Hz, and 1 Hz hazard curves. This is shown in Attachment 5. The curves were obtained from NRC [9.9].

Current NRC guidance for the estimation of design ground motions for new plants (Regulatory Guide 1.208 [9.10]) allows for a period and amplitude-dependent reduction in event frequency to represent the observation that some fraction of earthquake ground motions of a given amplitude fall below the threshold of damage for most engineered structures. One measure of this threshold is the cumulative absolute velocity (CAV) value. The present assessment has not included the CAV threshold effect [9.3].

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'*-Enter ENG REPORT IP-RPT-I 1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency 6.4 Identification of Margin 6.4.1 Identification of Low Capacity Components In the IP3 IPEEE report Section 3.1.5.5 [9.1], the mean core damage frequency (CDF) due to a seismic event is calculated to be 4.4E-05 per year. The dominant core damage sequence is a seismic-induced loss-of-offsite power (LOSP) and the subsequent loss of on-site AC power from all three emergency diesel generators. This sequence contributes, 43.5% of the total seismic CDF (1.92E-05 per year).

The second highest sequence contributing to the seismic CDF involves loss of secondary side cooling due to depletion of the condensate storage tank, and failure of RHR shutdown cooling due to the seismic event, which contributes 7.47E-06 per year (16.9% of the total seismic CDF).

The third highest sequence contributing to the seismic CDF involves a loss of component cooling water (CCW) due to a seismic event, which leads to an RCP seal LOCA and loss of decay heat removal. This sequence contributes 7.33E-06 per year (16.6% of the total CDF). This sequence includes success of the containment fan coolers, which was included in the sequence for use in the Level 2 analysis. An additional sequence, with a seismic CDF contribution of 2.84 x 10-6 per year, differs only in as much as it includes failure of the fan cooler units. Since these are essentially the same with regard to core damage, the overall contribution is the sum of these two sequences, or 1.02 x 10-5 per year (23.1% of the total seismic CDF).

With consideration of the above sequences, in-total six types of seismic-induced accidents dominate the seismic core damage frequency: station blackout (SBO), reactor coolant pump (RCP) seal loss-of-coolant accident, loss-of-offsite power (LOSP) transients, surrogate element, anticipated transient without scram (ATWS), and small break loss-of-coolant (LOCA) seismic accidents. Their mean contributions to the seismic core damage frequency for events up to seismic failure of the surrogate element (see Section 6.4.5 for additional discussion) are summarized below, Table 6.1:

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AEnerg ENG REPORT IP-RPT-11-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency Table 6.1 Events & Contribution to SCDF from IPEEE Seismic Accident Initiating Key Contributing Per year Mean Contribution Sequence Event(s) Components CDF to SCDF 1 LOSP/SBO SBO associated 1.92E-05 43.5 %

components LOSP/RHR- RHR pumps, heat 2 SD exchangers & CR 7.47E-06 16.9 %

supervisory panel 3 LOSP/CCW CCW surge tanks & heat 7.33E-06 16.6 %

exchangers 4 Seismic Failure of Surrogate 3.41 E-06 7.7 %

From a review of the contribution to SCDF, it is evident that the sequence associated with Station Blackout (SBO) resulted in significant risk contribution. As a result, components associated with the SBO were identified as requiring review to determine if conservatisms existed in the calculation of their fragility values. This was done by reviewing the IPEEE report Table 3.1.5.1 which lists the HCLPF84 of the components.

To broaden the review additional components were selected associated with other sequences which resulted in significant risk contribution to the SCDF, and with IPEEE calculated fragility levels below the surrogate element. In this approach a selection of the low capacity components were reviewed to ascertain if conservatisms exist, and if the findings, on a sample basis, are applicable to the remaining low capacity components with fragility levels below the surrogate element value.

Certain component groups identified in Table 3.1.5.1 with lower HCLPF84 values from the IPEEE assessment were excluded from review. HCLPF84 denotes a HCLPF calculated using SMA methodology; see Section 6.5 for further discussion. These low capacity components either do not initiate certain events, or were not significant in the seismic accident sequence with CDF contribution below the surrogate sequence.

6.4.2 In-Structure Response Spectra To determine component seismic response within a structure, floor response spectra are typically generated. For IP3, floor response spectra based on design ground input spectra were generated as part of the original plant design [9.5]. Within the IP3 IPEEE assessment to develop component capacity, in-structure response spectra (ISRS) were Sheet 14 of 27

EnteENG REPORT IP-RPT-11-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency estimated utilizing a power spectral density (PSD) direct generation technique. Based on a review of the resulting spectra and acceleration levels utilized in individual component assessments from this process, the review team concluded that the approach utilized to perform this estimate resulted in significant amplified response at some structure elevations. The review team assessed that the ISRS direct generation approach resulted in instances of over-estimation of in-structure floor amplification. This resulted in underestimating component HCLPFs for equipment evaluated in those structures.

For this evaluation, a spectral amplification method was used for the structures and floor elevations where the low capacity components selected for assessment are located.

The scaling approach utilized is associated with a method as defined in EPRI NP-6041

[9.2]. The design basis floor response spectra were then amplified by the developed amplification factor to obtain ISRS associated with the LLNL Hazard Curves. The spectra plots for selected structures and elevations within those structures were developed and are included in Attachment 3.

6.4.3 Review of Fragility for Components associated with Station Blackout (SBO)

Sequence The weak link components associated with the SBO sequence are identified in Table 6.2. These components were re-assessed to determine if increased HCLPF values could be derived, increasing the values previously calculated and reported in Table 3.1.5.1 of [9.1]. The original fragility calculations from the IPEEE assessment were not available. However, capacities were generally developed using the Screening Evaluation Worksheets (SEWS) developed as a part of the A-46 Generic Implementation Procedure (GIP) [9.6].

In reviewing the SEWS for the specific components associated with the SBO sequence, it was apparent that conservatisms were utilized within these calculations to demonstrate acceptance. Such conservatisms included using bounding acceleration values, minimizing effective anchorage to simplify the assessments, and grouping components and then assessing the "weakest' configuration to bound all other assessments. Such conservatisms are appropriate to demonstrate acceptance to meet plant design requirements and/or meet the intent of the A-46 program. However, these methods result in under predicting the true capacity of a component, resulting in an artificially low, conservative, HCLPF value.

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S~Entergy REPORT IP-RPT-I 1-00012

• ENG Reassessment of Indian Point 3 Seismic Core Damage Frequency Individual assessments of the eight component groups associated with the SBO sequence are contained within Attachment 7. The approach utilized in assessing fragilities for these components was to screen the likely failure mode, and using simplistic methods, consistent with EPRI NP-6041, derive the HCLPF value. Input ISRS used to derive seismic acceleration values was determined based on the amplification approach, as outlined within Attachment 3. The results from this review are summarized within Table 6.2. From the review of all component groups within the SBO sequence, HCLPF for these low capacity components was demonstrated to be above the surrogate element value.

6.4.4 Review of Fragility for Components associated with Non-SBO Sequences The weak link components associated with high CDF contribution sequences below the surrogate sequence are identified in Table 6.2. Based on the results from review of components associated with the SBO sequence, a sample selection of the remaining component groups was identified for review. These components were reviewed to determine if increased HCLPF values was likely, based on information within the SEWS, plant drawings, and experience of the review team.

Observations on the prior assessments performed within the SEWS for these sample components are the same as per the components reviewed associated with the SBO sequence. Review notes associated with the sample components selected for review are contained within Attachment 7 and/or summarized in Table 6.2. In addition, the Unit 1 stack was reviewed as documented in Attachment 4. The Unit 1 stack was dynamically analyzed for the 0.15g PGA design basis earthquake and found to be adequate. The median capacity for the stack is documented in the IPEEE assessment at 0.72g PGA, which is approximately equal to the surrogate elements median capacity.

Further discussion on the Unit 1 stack is contained in Section 6.6. The results from the sample review of components associated with non-SBO sequence are summarized within Table 6.2.

From the explicit review of all component groups associated with the SBO sequence, and a sample of the remaining components below the surrogate element, the review team concluded that the seismic capacity of all components associated with sequence events with significant contribution to the CDF could be raised to the surrogate element level.

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IP-RPT-1I-00012 e'Entergy ENG REPORT Reassessment of Indian Point 3 Seismic Core Damage Frequency Table 6.2 Seismic Comnonent Fraailitv Re-assessment IPEEE Recalculated Component fragility (g) fragility (g) Notes Description HCLPF 84 HCLPF84 1 Battery Chargers 31 0.24 0.66

& 32 Motor Control 2 Centers 36A, 36B, 0.24 0.55 37 Station Service These are part of Switchgear 31, 3 Transformers 2, 3, 5, 0.31 0.90 32.

6 4 Switchgear 31, 32 0.31 0.90 Service Water 5 System Pumps 31, 0.31 1.12 32, 33, 34, 35, 36 6 Battery Banks 31, 0.41 0.76 32, 34 7 EDG 31, 32 & 33 0.41 0.93 Control Panels EDG Current 8 Transformers 31, 32, 0.46 0.58 33 9 Supervisory Panel 0.24 0.56 Central Control Room Racks 11 Control Room Flight 0.31 0.68 Panel Current analysis shows that capacity is controlled by bending Component Cooling stress of tank support 10WF33.

12 Water (CCW) Surge 0.19 0.50 The current SF is 1.9. If margin Tanks 31, 32 capacity is used, the SF will increase to about 3.0. Therefore it is judged that there is adequate margin to reach a HCLPF of 0.50.

Current analysis shows that capacity is controlled by the upper support cast-in-place anchors. 2%

damping floor response spectra 13 CCW Heat 0.24 0.50 was used in analysis. The current Exchangers 31, 32 SF is 1.7. If margin capacity is used, the SF will increase to about 3.0. Therefore it is judged that there is adequate margin to reach a HCLPF of 0.50.

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ENG REPORT IP-RPT- 11-00012

~Etergy Reassessment of Indian Point 3 Seismic Core Damage Frequency

-r T ~ T - . - . - r Recalculated

-I. 4 fragiility (gi)

- 4 fraqility (g)

- .'-. I.

These are in-line components to piping systems. The heat exchangers are seismically supported at their base and with seismic tie-rods at the top of the heat exchanger. Based on the configuration of the heat exchanger, the qualification of the RHR Heat 14 0.23 0.50 attached piping (including the heat Exchangers 31, 32 exchanger) to conservative /%

damping floor response spectra), it is concluded by the review team that these components have sufficient structural capability and inherent margin in the design qualification to reach a HCLPF of

-~ 4 0.50.

These pumps are in-line components to piping systems.

The pumps are supported on steel slide plates and are located within a basement level of the PAB, subjected to ground response.

Based on the qualification of the attached piping (including the 151 RHR Pumps 31, 32 0.29 0.50 pump) to conservative 1 2%

damping floor response spectra, it is concluded by the review team that these components have sufficient structural capability and inherent margin in the design qualification to reach a HCLPF of 0.50.

Current analysis shows that capacity is controlled by drilled- in anchor bolts at the base. 2%

damping floor response spectra Motor Control was used in analysis The current 16Centers 36, 36E N/A 0.50 SF for the anchor bolts is 1.6. If margin capacity is used, the SF will increase to about 2.7.

Therefore, it is judged that there is adequate margin to reach a HCLPF of 0.50.

Sheet 18 of 27

Entergy ENG REPORT IP-RPT-11-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency 6.4.5 Fragility Calculations The 0P3 IPEEE SPRA analysis used fragilities for components modeled in the SHIP code. Two types of fragilities were used:

• Fragilities for components with relatively low seismic capacity compared to the rest of the plant, and

• A "surrogate" fragility for all of the components with relatively high seismic capacity.

A specific structure or component was designed as relatively low or high capacity based on the USI A-46 plant walkdown and evaluation. The criterion was whether or not the HCLPF was below (relatively low capacity) or above (relatively high capacity) the screening level of 0.5g.

The fragilities for components with relatively low capacity were determined as follows:

• The HCLPF 84 was calculated based on the critical failure mode. These components, their HCLPFs and the critical failure mode are listed in Table 3.1.4.3 of the IP3 IPEEE submittal [9.1]. The HCLPF84 denotes a HCLPF calculated according to the SMA procedures in EPRI NP-6041. (See Section 6.5 for further discussion).

• The HCLPF 84 was converted to a HCLPF 5o by dividing by 1.3, a factor that accounts for the peak-to-valley uncertainty in the seismic hazard curve.

• A 13c, of 0.46 was assigned. This is a generally agreed-upon generic Pc to be used for fragilities of specific components that are based on HCLPF calculations rather than full fragility analysis.

• A median capacity, Am, was calculated using the equation Am = HCLPF 50

• e2"33pc These component fragilities in terms of Am and Pc were input into the SHIP model in the places where the components appeared in each sequence.

The surrogate fragility was calculated as follows:

• A generic HCLPF8 of 0.50g was assigned as a lower-bound HCLPF for any screened-out (relatively high capacity) component. Obviously, the actual HCLPFs for the components are higher than this.

• The generic HCLPF84 was divided by 1.3 to obtain the HCLPF5 o.

• A generic Pc of 0.3 was assigned. This is taken from EPRI TR-103959 [9.11].

• A median capacity, Am, of 0.75g was calculated from the HCLPF 50 and Pc.

Sheet 19 of 27

'Ente ENG REPORT IP-RPT-I1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency A surrogate component with this fragility was modeled into each sequence to represent all of the screened-out components.

The SHIP model was run with these fragilities and the LLNL PGA seismic Hazard Curve. The SCDF was calculated by the SHIP code, along with the plant-level median capacity and HCLPF. A P3c could be inferred from the median capacity and HCLPF.

The reported HCLPF is a HCLPF 5o. The reported median capacity in the IPEEE submittal was 0.34g, and the HCLPF was 0.13g.

The NRC desired to calculate the SCDF for a different seismic hazard (e.g., the 2008 USGS hazard). The NRC did not have the ability to rerun the SHIP model with the different seismic hazard. Therefore, it derived a plant-level seismic fragility and convolved the fragility with the hazard curve. The NRC derived the fragility parameters, Am and Oc, from the IPEEE information.

The NRC used the reported median capacity and HCLPF to compute Pc, although it changed the HCLPF to 0.15g. The Pc they calculated was 0.34. For verification the seismic review team convolved this fragility with the USGS seismic Hazard Curves and matched the SCDFs reported in the GI-199 SRA.

The seismic review team identified conservatisms in the fragilities reported in the IPEEE submittal. The most significant conservatisms were:

• The HCLPF84 for the lower capacity components seemed too low, based on the judgment of the seismic review team members. Therefore, the HCLPFs for these components were recalculated, based on assessment and sampling.

• The division of the HCLPF8 by 1.3 to account for peak-to-valley variability in the ground spectral accelerations was too conservative because this variability is accounted for in the USGS hazard methodology. Therefore, the Am values for all components, including the surrogate, should be increased by 1.3 when convolving with the USGS Hazard Curves.

• The use of Pc = 0.3 in the surrogate fragility is too conservative when the plant-level fragility is equal to the surrogate fragility. A value between 0.4 and 0.5 is more appropriate, with 0.4 being conservative. The NRC used 0.4 for a generic Pc in the plant-level fragility of other plants. Therefore, the Am of the surrogate fragility should be increased by a factor of 1.3, and a Pc of 0.4 used for the case where the capacities of all components are shown to be greater than the surrogate capacity (HCLPF = 0.5g). The 1.3 factor is e 2 33x°4 /e 2 33 x0".

These factors are multiplicative. If the HCLPF84 of each lower capacity component is calculated and found to be higher than 0.5g, then all components in the plant have a HCLPF8 4 greater than 0.5g and the plant-level HCLPF84 is greater than 0.5g. The Sheet 20 of 27

Entergy ENG REPORT IP-RPT-11-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency surrogate fragility, Am = 0.75g and Pc = 0.3, could be used as the plant-level fragility.

Also, since the spectral acceleration is accounted for in the USGS Hazard Curves, this does not need to be accounted for in the fragility. Thus the plant-level fragility based on the surrogate component becomes Am = 1.3 x 0.75g = 0.98g and Pc = 0.3. Further, since the plant-level Pc should be at least 0.4 rather than 0.3, and the surrogate fragility is based on a HCLPF of 0.5g, the plant-level fragility becomes Am = 0.98 x 1.3 = 1.27g and Pc = 0.4.

This plant-level fragility was used to recalculate the SCDF reflecting the removal of conservatism.

6.5 Comparison of IP3 HCLPF vs. SSE IP3 was designed for a SSE of 0.1 5g PGA. The plant HCLPF should be higher than this. The HCLPF reported in the IP3 IPEEE report is only 0.13g. This apparent contradiction is explained below.

The purpose of the HCLPF is to provide a measure of the seismic margin available in the structural integrity of structures and equipment to resist earthquakes over and above the design basis earthquakes. The HCLPF is defined as the ground acceleration at which there is at least 95% confidence of less than 5% probability of failure, or a conditional probability of failure of 1% (0.01). As noted in NUREG-1407 [9.12] the plant HCLPF determined in a seismic PRA (SPRA) is different from the result that would be obtained in a seismic margin assessment (SMA). It is different in three respects.

First, in the SPRA, random failures and operator actions are included in the quantification of the plant logic model. The contribution of these basic events to the plant fragility is greatest at low ground motions where the conditional probabilities of failure are small (in the lower tail of the fragility curve). As a result, the conditional probability of failure of 0.01 is reached at a ground motion that is lower than it would otherwise be if only seismic failures were considered.

Second, in the SMA methodology, the plant level HCLPF is quantified using a different approach. In the SPRA the lower tail of component fragility curves are combined (as defined by the plant logic model) at ground motion levels below the component HCLPF levels. This is not done in an SMA.

The plant fragility calculated in a SPRA is therefore more appropriately called the total fragility. The seismic fragility can be determined by removing the contribution from random failures and operator actions. The IP3 IPEEE reported the total fragility and did not calculate the seismic fragility. Thus the reported HCLPF, which is associated with the total fragility, is lower than the seismic HCLPF. In comparison, the IPEEE submittal for another eastern coast plant reported both the total HCLPF and the seismic HCLPF.

The seismic HCLPF was 30% higher than the total HCLPF.

Sheet 21 of 27

eEntrg ENG REPORT IP-RPT-I 1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency Third, in a SPRA the HCLPF is computed directly from the plant-level fragility. It is based on the median spectral shape of the ground motion, and uncertainty is explicitly accounted for. The SPRA-calculated HCLPF is denoted HCLPF 50 to reflect this. The HCLPF calculated in a SMA is calculated directly from the plant design capacities, and uncertainty is conservatively accounted for implicitly by using conservative failure capacity and median plus one standard deviation ground motion spectral shape (84%

non-exceedance probability). This HCLPF is denoted HCLPF84. The difference between HCLPF 50 and HCLPF84 can be calculated from the uncertainty in the spectral shape associated with the seismic hazard curve. In the IP3 SPRA, the HCLPF84 is 30%

higher than the HCLPF5s.

The appropriate HCLPF for comparison to the plant SSE is the HCLPF84 calculated using the SMA methodology (the CDFM method). This would be roughly 30% higher than the HCLPF 50 calculated in the IP3 IPEEE. Further, the HCLPF should be calculated from the seismic fragility rather than the total fragility. This would be expected to be about 30% higher than what IP3 reported. Therefore, the IP3 HCLPF suitable for comparison to the IP3 SSE would be about 70% higher than the HCLPF reported in the IPEEE submittal, or about 0.22g. Considering the conservatisms in the calculation of component capabilities used in the IP3 IPEEE SPRA model, this demonstrates substantial margin above the design basis.

6.6 Review of the Unit I Stack The Unit 1 Superheater Stack and supporting steel work has been analyzed for seismic, tornado and vortex shedding wind load as outlined in Attachment 4. The concern associated with the Unit 1 stack as it affects IP3, is for a seismic failure of the stack to impact the condensate storage tank. This structure was evaluated in earlier PRA studies [9.8] as documented in the IP3 IPEEE report [9.1]. The report documents a median capacity of the Unit 1 stack of 0.73g PGA. This value is approximately equal to the surrogate elements median capacity of 0.75g PGA. The Unit 1 stack is not considered a large contributor to SCDF for IP3. A discussion on the qualification and condition of the stack is included in Attachment 4.

6.7 Reassessment of SCDF The IP3 SCDF was calculated in the IP3 IPEEE SPRA using the systems model as input to the SHIP model. The model incorporated seismic fragilities for components into event trees and fault trees, producing sequences representing combinations of system failures that could lead to core damage following the seismic event. The effect on plant SCDF of changing the fragility of a given component (structure, system or equipment) would normally be determined by changing the fragility parameters in the model and re-running it using the SHIP software. However, it is not possible, at this time, to change the inputs to the model. Therefore, the change in plant SCDF was estimated using the process described in Attachments 5 through 9.

Sheet 22 of 27

• Entergy ENG REPORT IP-RPT-1 1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency The evaluation process began by developing a method that could be shown to be equivalent to the NRC approach by benchmarking it against the USGS results, and then used to develop SCDFs using the more robust fragilities and refined input parameters developed in this effort. These additional runs allowed the review team to assess the potential impact on the Indian Point 3 seismic core damage frequency (SCDF) of increasing the calculated seismic capacity of components and structures, whose failure contributes significantly to the dominant seismic core damage sequences. Since the seismic level associated with the "surrogate element" component fragility encompassed essentially all of the IP3 seismic risk contribution, this seismic level (0.75g) was chosen as the objective for this task.

The dominant sequence in the IP3 seismic analysis was seismically induced station blackout events (i.e. loss of all offsite and onsite AC power). The initial effort therefore concentrated on components whose failure would lead to loss of the emergency onsite diesel generators. Since the seismic events of concern would render the offsite power sources and the Appendix R/SBO EDG unavailable due to their low seismic ruggedness, loss of the EDGs following such a "beyond design basis" seismic event would lead to an unrecoverable loss of all power (Station Blackout - SBO). The reassessment concluded that all significant component contributors to the SBO sequence have seismic capacities at least equivalent to the surrogate component.

Given this success, the effort was expanded to components that were contributors to the remaining significant non-SBO sequences. The reassessment concluded that the significant component contributors to the non-SBO sequence also have seismic capacities at least equivalent to the surrogate component.

Setting the median capacity of those components equivalent to the surrogate element, the equivalent method developed in this effort was run with both the updated USGS

[9.9, 9.13] and updated EPRI Hazard Curves [9.14], for a number of varied input parameter assumptions, as described in Attachment 6.

Seismic core damage frequency analyses using USGS Hazard Curves used by NRC for 1 Hz, 5 Hz, 10 Hz, and PGA discretized to acceleration intervals of 0.001 g were performed using fragilities adjusted for identified conservatisms. The analyses consist of convolving the USGS Hazard Curves for different frequencies with the various fragilities.

Areas of identified conservatisms included the double-counting of the peak to valley variability in both the hazard curves and the calculated fragilities (i.e., a the hazard accelerations were increased by a 1.3 factor and the median capacities were also reduced by a 1.3 factor - Section 6.4.5), and the underestimated fragilities for a series of components, most of which were, as discussed above, associated with the SBO sequence (Section 6.4.3, 6.4.4). Reassessment of the fragilities for the high contributors to risk discussed in Section 6.1, showed their median capacity to be equal to, or higher than, the median capacity selected for the robust components that were screened out and considered as a surrogate element in the SCDF calculations.

Sheet 23 of 27

SEntergy ENG REPORT lP-RPT-I1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency As shown in Attachment 6, the most accurate representation of seismic risk is believed to be represented by Analysis Series 6. Using the value calculated for the limiting frequency, the SCDF using the USGS Hazard Curve for the 10Hz frequency case was estimated at 9.4E-06 per year.

Attachment 6 also presents results obtained using the EPRI Updated Hazard Curves and the same fragilities considered in conjunction with the USGS Hazard Curves. The EPRI Hazard Curves were prepared using different Ground Motion experts and use a Cumulative Absolute Velocity (CAV) filter to remove high frequency wave with limited energy. The highest SCDF estimate obtained using the EPRI updated Hazard Curves is 7.1 E-06.

6.8 Seismic Design and the Ramapo Fault Micro seismic activity recorded by the seismic monitoring network in the Indian Point region is evidence of minor crustal adjustments due to regional stresses. However, neither the readings from the seismic monitoring networks nor the bore-hole experiments in the fault regions show the evidence of any contemporary (geologic) movement along faults exposed at the surface. Many researchers have determined that the Ramapo faults appear to be from compaction and original deposition of material during geologic formation rather than tectonic in origin. The seismic activity in the area is due to rebounding from previous glacial pressure. The conclusions on seismic activity that were established at the time of initial licensing are still valid today. The Ramapo Fault and the Ramapo Fault Zone do not represent a seismic hazard to the Indian Point plant. The Ramapo Fault is not a Tectonic plate that is capable of generating large earthquakes and shows no signs of movement. The USGS has not identified any Quaternary faults (younger than 1.6 million years) in this region. This means there has not been any significant earthquake movement since that time. Further discussion is outlined in Attachment 2.

It is therefore concluded that the seismic design criteria for structural analysis at the Indian Point site is satisfactory and that the plant is set on solid bedrock. The current analyses envelope any effect due to the Ramapo fault. No public hazard can be expected from the plant due to an earthquake in the region that would exceed its design margin.

Sheet 24 of 27

'*Enterg ENG REPORT IP-RPT-1 1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency

7.0 CONCLUSION

S It was concluded that the IP3 plant is significantly more robust than shown in .the IP3 IPEEE study and the NRC GI-199 SRA indicated.

The site median capacity (Am) equal to 0.34g and a composite uncertainty (91c) equal to 0.34 used by the NRC resulted in a very conservative estimate of SCDF of 1.OE-04 per year or 1 in 10,000 reactor-years.

The improved values for the median capacity (Am = 1.27g), and composite uncertainty (lBc = 0.4) calculated by the review team resulted in a SCDF estimate of 9.4E-06 per year (1 in 106,383 reactor-years) associated with the current USGS Hazard Curves which represents a significant reduction from the GI-199 reported SCDF estimate (1.OE-04 per year or 1 in 10,000 reactor-years) and is a much more realistic estimate of the seismic risk.

It should also be noted that using the improved fragility (Am = 1.27g, and f~c = 0.4 ) along with the 2010 EPRI updated seismic Hazard Curves instead of the USGS curves further reduces the SCDF value to 7.1 E-06 per year (or 1 in 140,845 reactor-years) 8.0 RECOMMENDATIONS 8.1 Short Term Recommendations:

- Consider developing more formal calculations per EN-DC-126 to enhance the supporting documentation associated with this Engineering Report.

- Write an official letter to Stevenson and Associates requesting them to locate original calculations associated with the September 1997 IPEEE report for IP3.

- Write official letters to NUS Corporation and EQE International requesting them to locate original calculations associated with the December 1995 IPEEE report for IP2.

- Review the IP2 IPEEE report to understand the current seismic risk to the station, in-terms of SCDF.

- Identify any potential conservatism within the existing IPEEE submittal for IP2. Investigate removing, where appropriate, conservatisms in the IPEEE fragility values.

- If beneficial and warranted, repeat this IP3 effort for IP2.

8.2 Long Term Recommendations:

- Consider modification for the removal of the Unit 1 stack, in order to reduce seismic risk for Units 2 and 3.

- Continue monitoring GI-199 issue progress and implementation.

- Respond to the associated NRC Generic Letter when issued. If required by the GL, reconfirm the seismic robustness of the plant via a new analysis.

Sheet 25 of 27

OEntmWy ENG REPORT IP-RPT-I 1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency

9.0 REFERENCES

9.1 UNIT 3 IPEEE report IP3-RPT-UNSPEC-02182, September 1997.

9.2 EPRI Nuclear Plant Seismic Margin R-1, Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1), NP-6041-SL, August 1991 9.3 NRC, "Safety/Risk Assessment Results for Generic Issue (GI) 199, Implications of Updated Probabilistic Hazard Estimates in Central and Eastern United States on Existing Plants," dated September 2, 2010 (ADAMS Accession Number ML100270582) 9.4 NRC Information Notice 2011-05: Tohoku-Taiheiyou-Oki Earthquake Effects on Japanese Nuclear Power Plants 9.5 Westinghouse Electric Co. "Seismic Response Curves for Class I Buildings for Indian Point Generating Station Unit No. 3" 9.6 Entergy IP3 GIP SEWS 9.7 NRC frequently asked questions related to the March 11, 2011 Japanese Earthquake and Tsunami 9.8 SMA, "Conditional Probability of Seismic Induced failures for Structures and Components for Indian Point Generating Station Units 2 and 3", SMA 12901.01, Oct 1980.

9.9 Email from Martin Stutzke (USNRC) to Dan Nuta (Entergy), "Indian Point Seismic Hazard Curves used in GI-199", dated 3/29/2011.

9.10 NRC Regulatory Guide 1.208, "A Performance Based Approach to Define the Site Specific Earthquake Ground Motion", March 2007.

9.11 EPRI TR-103959, "Methodology for Developing Seismic Fragilities", 1994.

9.12 NRC NUREG-1407, "Procedural & Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities".

9.13 US Geological Survey:

http://earthquake.usgs.gov/hazards/products/conterminous/20 08/update 201001/curves/

9.14 EPRI, "Updated Seismic Hazard Results for the Arkansas, Fitzpatrick, Grand Gulf, Indian Point, Pilgrim, River Bend, Vermont Yankee, and Waterford Nuclear Sites." August 2010 (Not published, Proprietary).

Sheet 26 of 27

ENG REPORT IP-RPT-1 1-00012

'z-Entergy Reassessment of Indian Point 3 Seismic Core Damage Frequency 10.0 ATTACHMENTS 10.1 Selection Criteria for 27 Plants that Require Further Investigation (3 pages) 10.2 Indian Point Site Seismic Review and effect of the Ramapo Fault (12 pages) 10.3 IP3 In-structure Response Spectra LLNL & USGS Uniform Hazard Spectra (43 pages) 10.4 Review of Concerns for the Unit 1 Stack (4 pages) 10.5 Develop Methodology that emulates NRC's use of IPEEE Parameters to Calculate Seismic Core Damage Frequency (SCDF) Values (22 pages) 10.6 Revised SCDF for IP3 using USGS & EPRI Hazard Curves (2 pages) 10.7 Revised Calculations for Identified Low Capacity Components (33 pages) 10.8 Calculate the CDF for SBO for surrogate level fragility using LLNL (9 pages) 10.9 Plan for Adjusting SCDF without Re-Running SHIP Model (3 pages) 11.0 REVIEW TEAM MEMBERS Joe Abisamra, PE - Chief Engineer (ECH) - Team Lead Richard S. Drake, PE - Engineering Supervisor (IPEC)

Dragos (Dan) Nuta, PE - Senior Engineer (IPEC)

Doug Gaynor - Senior Lead Engineer (IPEC)

Ahmet Unsal, PE - Senior Staff Engineer (WPO)

Clem Yeh - Engineering Supervisor (WPO)

John Bretti - Senior Lead Engineer (WPO)

John Favara - Senior Lead Engineer (WPO)

Paul Baughman, PE - Senior Consultant (ARES Corporation)

Paul Bruck- Principal (Lucius Pitkin, Inc.)

Sheet 27 of 27

Entergy ENG REPORT IP-RPT-II-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency Attachment 1 Selection Criteria for 27 Plants that Require Further Investigation

Indian Point Energy Center Position Paper 201 1-ISRT-01 Selection Criteria for the 27 Plants that Require Further Investigation NRC's August 2010 List

• This is NRC's explanation of how they compiled the list that was shared at the NRUG meeting at Amelia Island in early August and again during the presentation to the ACRS in Nov 2010.

Please note that NRC says they were not trying to identify specific plants but, instead, were trying to determine iffurther evaluation by the Generic Issue Program was warranted. Because there are plants in the "continue" region (with a change in seismic CDF greater than 10E-5),

their conclusion is that they should continue to evaluate the Generic Issue.

Two sets of seismic hazard curves were developed to support the Individual Plant Examination of External Events (IPEEE) that was requested by Generic Letter 88-20, Supplement 4:

1. Lawrence Livermore National Laboratory (LLNL), as reported in NUREG-1488

2. Electric Power Research Institute Seismic Owners Group (EPRI/SOG), as reported in EPRI NP-6395.

Note that there are LLNL seismic hazard curves all for Central and Eastern United States (CEUS) plants, but that some plants did not participate in the EPRI/SOG. The observation that there are two sets of credible, IPEEE-era seismic hazard curves implies that there are two ways to compute the change in seismic core-damage frequency (delta-SCDF). Specifically, you can compute the change with respect to the LLNL curves and/or with respect to the EPRI/SOG curves.

For plants that had both sets of curves, we computed both delta-SCDFs and developed the so-called "delta-delta plot" (Figure 1) to decide which plants warranted further investigation. If one of the delta-SCDFs was greater than or equal to 1E-5/y and the other was positive, then we binned the plant into the "continue" region. This process identified 24 plants as listed below:

1. Crystal River 3

2. Dresden 2

3. Dresden 3

4. Farley 1

5. Farley 2

6. Indian Point 2

7. Indian Point 3

8. Limerick 1

9. Limerick 2

10. North Anna 1

11. North Anna 2

12. Oconee 1

13. Oconee 2

14. Oconee 3

15. Peach Bottom 2

16. Peach Bottom 3

17. Perry 1

18. River Bend 1 Page 1 of 3

Indian Point Energy Center I Position Paper 2011-ISRT-01 Selection Criteria for the 27 Plants that Require Further Investigation

19. Seabrook 1

20. Sequoyah 1

21. Sequoyah 2

22. Summer

23. Watts Bar 1

24. Wolf Creek 1 For plants that did not participate in the EPRI/SOG, we were limited to considering only the delta-SCDF with respect to the LLNL seismic hazard curves. If this delta-SCDF was greater than or equal to 1E-5/y, then we binned the plant into the "continue" zone. This process identified 3 plants as listed below:

1. Duane Arnold

2. Saint Lucie 1

3. Saint Lucie 2 Hence, there are a total of 27 plants (24 + 3) that fall into the "continue" zone.

Reference:

NRC staff presentation to ACRS Siting Subcommittee, November 30, 2010.

Page 2 of 3

Indian Point Energy Center IPosition Paper 201 1-ISRT-01 Selection Criteria for the 27 Plants that Require Further Investigation I

Figure 1 5.E-05 4.E-05 3.E-05 2.E-05 T

S1.E-05 zml 0.E+050

-j

-j O.EO_00 Evud

-2.15-05

-3.E-05

.-4.E-05

-5.E-05

-1.E-05 0.E+00 .E-05 2.1-05 3.E-05 4.E-05 5.E-05 6.-05 7E-05 Change in SCDF (EISOG Baseline)

Page 3 of 3

'Entergy ENG REPORT [P-RPT-11-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency Attachment 2 Seismic Review and effect of the Ramapo Fault

SI Paper IPosition 201 1-ISRT-02 Indian Point Energy Center Indian Point Site Seismic Review and affect of the Ramapo Fault I

Indian Point Site Seismic Review and effect of the Ramapo Fault Prepared by: Richard S. Drake PE Reviewed by: Dragos Nuta, PE and Paul Bruck 1.0 Purpose The purpose of this paper is to summarize the geological and Seismological region around the Indian Point Energy Center. This will include the localized effects of the Ramapo Fault and other "Intra-plate Earthquakes".

2.0 Seismic Design Summary 2.1 Seismology Seismic activity in the Indian Point area is rare and no damage has resulted from any historical basis. As stated in IP3 FSAR Section 2.8, the site is "practically non-seismic" and is "as safe as any area at present known." Not withstanding such assurance, the safety related equipment, components and structures of the plant were designed to withstand an earthquake of the highest intensity which can reasonably be predicted from geologic and seismic evidence developed for the site.

The design of the Containment and other seismic Class I structures was based on a "response spectrum" approach in the analysis of the dynamic loads imparted by earthquake. The seismic design took into account the acceleration response spectrum curves developed by G. Housner.

Seismic accelerations were computed as outlined in the AEC TID-7024, "Nuclear Reactors and Earthquake", for large-magnitude earthquakes at moderate distances from the epicenter. As such, the curves are made up of the combined normalized response spectrum determined from components of four strong-motion ground accelerations: El Centro, California, December 30, 1934; El Centro, California, May 18, 1940; Olympia, Washington, April 13, 1949 and Taft, California, July 21, 1952.

Since no strong motion records were available for the Eastern United States, the method used appeared to be the most conservative and rational considering the amount of earthquake data available at that time. In addition, this method was consistent with the procedure being carried out on the majority of the nuclear plants under construction at that time in the United States.

As indicated in IP2 and IP3 FSARs Chapter 16, peak ground accelerations used in the seismic design are:

Page 1 of 12

Indian Point Energy Center Position Paper 201 1-ISRT-02 Indian Point Site Seismic Review and affect of the Ramapo Fault Operational Basis Earthquake accelerations are 0.1 g horizontally and 0.05 g vertically and for the Design Basis Earthquake are 0.15 g horizontally and 0.10 g vertically.

2.2 Background and Seismic Design Bases Geographic areas of the continental United States have been subdivided into regions of known or assigned seismic probability or risk and this has served as a useful basis for generating code provisions for earthquake - resistant structures. The Seismic Risk Map adopted by the International Conference of Building Officials for inclusion in the 1970 edition of Uniform Building Code, divided the United States into four (4) major zones of seismic risk or probability. The Indian Point Site is located in Zone I of this map with intensities limited to V and VI on the Modified Mercalli Intensity Scale of 1931 and only slight earthquake activity can be expected.

However, the Indian Point 3 facility was actually built per requirements of Zone 2 of the Uniform Building Code i.e., corresponding to an intensity of VII of the Modified Mercalli Scale. The range of expected horizontal acceleration of ground motion for earthquakes of this intensity is70-150 cm/sec2 near the epicenter or a maximum peak ground acceleration of about 0.15 g, Reference 4.0.1. At a distance of 100 miles from the epicenter, the acceleration drops to 50%. The nearest event larger than intensity VII occurred near Cape Ann, Massachusetts, a distance of more than 200 miles from the site, in 1755. This event was classified as intensity VIII on the Modified Mercalli Scale. It was believed, therefore, that the plant's structural design, allowing for safe shutdown in the event of an earthquake of intensity VII on the Modified Mercalli Scale was adequate.

The definition of a Modified Mercalli Intensity Scale level VII earthquake is: People have difficulty standing. Drivers feel their cars shaking. Some furniture breaks. Loose bricks fall from buildings.

Damage is slight to moderate in well-built buildings; and considerable damage in poorly built buildings. See Seismic scale attachment to this paper.

Several well known experts at the time, such as the Reverend Joseph Lynch, S. J., while Director of the Fordham University Seismic Observatory stated ".... that the probability of a serious shock occurring in this area for the next several hundred years is practically nil. The area therefore would certainly seem to be as safe as any area at present known." Captain Elliott B. Roberts, while Chief of the Geophysics Division of the Department of Commerce, substantially agreed with the conclusions of Rev. Lynch. Rev. Lynch also stated that the "estimated maximum ground acceleration of 0.03 g is reasonably conservative for the area." This has been established as the basis for design of the plant. Rev. Lynch stated further that the "safety factor for a horizontal stress of 0.1 g is therefore.... more than adequate". The IP2 and IP3 plants are designed to have no loss of function of systems important to safety for earthquakes having a horizontal acceleration of 0.1 g and a vertical acceleration of 0.05 g acting simultaneously at zero period,.

2.3 Public Concerns and Resolutions during Initial Plant Licensing Subsequent to the plant's construction, public concerns were raised on the following issues:

• A series of N-NE trending faults pass through the area surrounding the site - collectively known as the Ramapo Fault System. The concern raised was whether the Ramapo Fault, which is located approximately 4 miles northwest of the plant, is "capable" of causing an earthquake at the site. (Reference 4.1.1)

Page 2 of 12

Indian Point Energy Center IPosition Paper 2011-ISRT-02 Indian Point Site Seismic Review and affect of the Ramapo Fault

• Because of the lack of historical records of earthquakes in the plant area (see attached table) -

older than the Cape Ann earthquake, concerns were raised, by reference 4.1.1, whether the safe shutdown earthquake (SSE) for the plant's design should be greater than intensity VII on the Modified Mercalli Scale.

" As stated earlier, the plant was designed for 0.15 g maximum peak ground acceleration for safe shutdown. Concern was that, if the SSE ground acceleration be raised from intensity VII to VIII on the Modified Mercalli Scale, would the 0.15 g peak ground acceleration still be adequate?

" An extended micro-monitoring system for measuring magnitude, accurately determining the location and even focal mechanism behavior of small magnitude earthquake near the plant and Ramapo Fault Zone was required by condition 2.C.4(c) of Amendment No. 2 to the Operating License of the plant. Concern was raised whether this extended micro-seismic-measuring instrumentation was considered as a licensing requirement.

An 18 month proceeding was held on the above concerns before a U.S. Nuclear Regulatory Commission (NRC) Atomic Safety and Licensing Board. Also, there have been numerous extensive studies on these concerns. The findings by the board and these studies may be summarized as follows:

• In testimony before the board, Charles F. Richter, who developed the Richter Scale, stated that the earthquakes in the Ramapo region are "of minor magnitude and relatively trivial".

Radiometric age determination of un-deformed minerals that have grown within fault zones was studied by Ratcliffe and fault related deformation of Pleistocene deposits and surface features by Dames & Moore (Ref 4.1.3) and Ratcliffe((Reference 4.1.5). Both prove that the faults in the Indian Point area have not moved in at least the last 2 million years. The Ramapo Fault, therefore, is considered to be old, inactive and not a "capable" fault under Appendix A to 10 CFR 100.

• The unanimous ruling by the board was - "In accordance with Appendix A to 10 CFR 100, neither the Cape Ann earthquake nor any other historic event requires the assumption of a safe shutdown earthquake for the Indian Point Site of greater than a Modified Mercalli intensity of VII". Hearings were also held before the Advisory Committee on Reactor Safeguards. Despite some controversy over appropriate tectonic divisions, the Committee, scientists in the TERA Corp.(references 4.1.8 and 9) and Dames & Moore(reference 4.1.4) concluded that an event of intensity VII on the Modified Mercalli Scale is adequate as the design earthquake for the Indian Point Site.

• Consistent with the above ruling, the board also ruled - "The ground acceleration value used for the design of Indian Point Units 2 and 3 should remain at 0.15 g".

• Amendment 2 to the Technical Specifications stated in Section 2(c)(4)(c) that an extended micro seismic instrumentation network must be operated for at least two years following complete installation of all stations. The Atomic Safety and Licensing Appeal Board repealed this decision in hearings held on October 12, 1977 and the NRC issued Technical Specification Amendment 9 to reflect this. However, a instrumentation network was operated from 1975-1990 by the lamont-Doherty laboratory and a final report was published.

Page 3 of 12

Indian Point Energy Center Position Paper 2011-1SRT-02 Indian Point Site Seismic Review and affect of the Ramapo Fault 2.4 Recent analysis of North-East United States Earthquakes References 4.0.4, 5, 9 and 10 describe the type of earthquakes that occur in the interior of the North American Plate. These types of Quakes are called "Intra-plate Earthquakes". Theses types of quakes are common but do not follow the tectonic model for earthquakes and are typically of lower magnitude. Many seismologists try to use tectonic plate theory for older non-active faults.

There are many hypotheses for the cause of the quakes in intra-plate zones but "there is no obvious relationship between earthquakes and geologically mapped faults in most intra-plate areas."

At the present time, a commonly accepted explanation for the cause of earthquakes in the Northeast, specifically including the Ramapo fault zone, is that "ancient zones of weakness" are being reactivated in the present-day stress field. In this model, preexisting faults and/or other geological features formed during ancient geological episodes persist in the intra-plate crust, and, by way of analogy with plate boundary seismicity, earthquakes occur when the present-day stress is released along these zones of weakness. The theory of the re-activated Ramapo fault is specifically disputed below. Some seismologists explain that these small earthquakes are caused by the ground rebounding from the stress of the glaciers millions of years ago that pressed the area down. References 4.0.11 and 4.0.12 suggest that the primary deformation within eastern North America results from de-glaciations or postglacial rebound. The New Jersey Highlands are at the southernmost limit of the area covered by the Laurentide ice sheet. The Laurentide ice sheet expanded outward from accumulation centers in northern Quebec and Labrador. This ice sheet has grown and melted away about 10 times within the past 2 million years. The Hudson valley and the Highlands were glaciated by the Laurentide at least three times within this period. Using this model as a guide, much of the research on northeastern United States earthquakes has involved attempts to identify preexisting faults and other geological features that might be reactivated by the present-day stress field. While this concept of reactivation of old zones of weakness is commonly assumed to be valid, in reality the identification of individual active geologic features has proven to be quite difficult. Unlike the situation for many plate boundary earthquakes,it is not at all clear whether faults mapped at the earth'ssurface in the Northeast are the same faults along which the earthquakesare occurring. Contrary to reference 4.1.1, other researchers have shown that the earthquakes do not specifically fit or are caused by the Ramapo fault. (References 4.0.3, 4, 5, 9, 10, 11, 12, 14, and 16).

Because Intra-plate earthquakes are not the result of Tectonic plate movements that slide along each other there is no energy or large movement that is built up which would cause the strong motion earthquakes.

The USGS database, reference 4.0.6, contains no known Quaternary faults (younger than 1.6 million years) in this region because geologists have not found any faults that young that are at the Earth's surface. The Quaternary began 2,588,000 years ago, at a time when rock strata show extensive evidence of widespread expansion of ice sheets over the northern continents.

Earthquakes occur by sudden movement on faults deep in the Earth's crust. Earthquakes larger than about magnitude 6, Richter scale, can rupture from those depths upward to the surface. This database summarizes surficial geologic evidence of faults that have ruptured during the Page 4 of 12

Indian Point Energy Center Position Paper I 201 1-ISRT-02 Indian Point Site Seismic Review and affect of the Ramapo Fault Quaternary. The most common kinds of surficial evidence are (1) Quaternary geologic strata that are offset across faults, and (2) Quaternary liquefaction features formed when strong earthquake shaking caused water-saturated sand, a few meters below the surface, to liquefy and erupt upward onto the surface. Many faults older than Quaternary are known in this region, but young movement on them has not been proved. In some areas, particularly east of the Rocky Mountains, the evidence of young offsets and liquefaction might have been destroyed by erosion or glaciations, buried under younger deposits, or are ambiguous. This proves that the Ramapo Fault has not moved for the past 2 million years.

2.5 Ramapo fault and resolution of third party questions Reference 4.0.14 states, The Ramapo fault system is composed of southeast dipping normal faults based on research by Ratcliffe, in the 1970s and 1980s. The faults form the northwest boundary of the Mesozoic Newark Basin in Putnam, Rockland, and Westchester Counties of southeastern New York State and eight counties across northern New Jersey. In northern New Jersey most of the Mesozoic normal slip of the fault system is concentrated on a single strand, the Ramapo fault. It was suggested that the fault system in general and the Ramapo fault in particular are seismically re-activated as suggested by Aggarwal and Sykes in 1978, reference 4.1.1. However as summarized by Crone and Wheeler (2000, Reference 4.0.16), the more recent studies have not replicated most of the results of Aggarwal and Sykes (1978). Structural and petrofabric analyses of fault rock, which was cored from the Ramapo and other basin-border faults of the fault system at six localities, showed that the most recent slip was extensional at each locality (Burton and Ratcliffe, 1985; Ratcliffe, 1980, 1982a; Ratcliffe et al., 1990; References 4.1.5 and 4.0.17 thru 19).

The last widespread extensional episode was Mesozoic, and modern-day extensional slip would be inconsistent with the existing, east-northeast-trending, contractional stress field (Zoback and Zoback, 1991). Seismic-reflection profiles showed that the Ramapo fault and other strands of the fault system dip more shallowly than inferred by Aggarwal and Sykes (1978), with the result that most earthquake hypocenters were in the footwalls, not on the faults (Ratcliffe and Costain, 1985).

Stone and Ratcliffe (1984) and Ratcliffe et al. (1990) trenched the up-dip projection of the Ramapo fault at two localities. Neither trench revealed evidence of quaternary tectonic faulting. The faults appear to be compaction, syndepositional faults rather than tectonic in origin.

Crone and Wheeler (2000, reference 4.016 & 20) evaluated additional suggestions of Quaternary tectonic activity that were based on pollen data, sea-level curves from tidal marshes, soft-sediment deformation observed in cored sediments from a glacial lake, and geomorphic observations. None of these suggestions provide evidence of sudden seismic slip as opposed to slow aseismic creep.

Most are inconsistent with the orientation of the existing compressional stress field and the absence of significant post-Mesozoic slip. No available arguments or evidence can preclude the possibility of occasional small earthquakes on the Ramapo fault or other strands of the fault system, or of rarer large earthquakes whose geologic record has not been recognized.

Nonetheless, there is no clear evidence of quaternary period tectonic faulting on the fault system aside from the small earthquakes scattered within and outside the Ramapo fault system (Kafka et al., 1985; reference 4.0.5).

The small to moderate earthquakes that make up most of the four-century-long Eastern US historical seismicity record rarely produces surface deformation or evidence of strong ground Page 5 of 12

Indian Point Energy Center Position Paper 201 1-ISRT-02 Indian Point Site Seismic Review and affect of the Ramapo Fault motion that is currently recognizable in the geologic record. A pre-Quaternary fault probably has an annual probability of slip that is too small to significantly affect these hazard computations (Wheeler, 2002, reference 4.0.21). Without large movements a large earthquake would not be expected to be generated.

The USGS Report 00-260 and NRC letter dated December 15, 2004 refuted the Riverkeeper/Sykes theory of the Ramapo Fault being an active fault which could cause a large earthquake. The USGS report, reference 4.0.16 identified the Ramapo Fault as a "Class C" fault.

Class C faults by definition: Geologic evidence is insufficient to demonstrate (1) the existence of a tectonic fault or (2) Quaternary slip or deformation associated with the feature. Faults and features assigned to Class C do not have demonstrated Quaternary activity and are not considered to be potential earthquake sources. The report also stated that reactivation of the Ramapo or other border faults in the present day compressional stress field would be inconsistent with the results of the core analyses. NRC response, reference 4.0.3, identified the Ramapo fault system as a key fault system in the area and is very unlikely to generate any earthquakes larger than historical earthquakes based on paleoseismic and geologic investigations. The original design of Indian Point and later analysis has taken these earthquakes into consideration. The USGS Hazard maps listed in References 4.0.6 and 4.0.13 show the area around Indian point as low seismic areas.

2.6 Other Issues The Japanese earthquake in March of 2011 caused a huge tsunami which resulted in the heaviest damage to the coastal area. At this time it appears that the tsunami was instrumental in causing the damage to the Fukushima Daiichi nuclear power plans. A "Subduction zone" earthquake is required for such a significant tsunami to be generated, as occurred recently in Japan. No such subduction zones exist off the east coast of the United States.

Subduction zone earthquakes occur when one of the many tectonic plates that make up Earth's outer shell descends, or "subducts," under an adjacent plate. This kind of boundary between plates is called a "subduction zone." When the plates move suddenly in an area where they are usually stuck, an earthquake happens. The sudden uplift or drop of the sea floor displaces large volumes of water towards the shore, which then rises as a standing wave as the shallower water is encountered near the shore line.

Page 6 of 12

Indian Point Energy Center Position Paper 201 1-ISRT-02 Indian Point Site Seismic Review and affect of the Ramapo Fault 3.0 Conclusions Micro seismic activity recorded by the seismic monitoring network in the Indian Point region is evidence of minor crustal adjustments due to regional stresses. However, neither the readings from the network nor the bore-hole experiment (reference 4.1.11) at Kent Cliffs, NY show the evidence of any contemporary (geologic) movement along faults exposed at the surface as was suggested by Aggarwal and Sykes (reference 4.1.1). On the contrary, the last movement of the region was during the Mesozoic Period (The Mesozoic Era is an interval of geological time from about 250 million years ago to about 65 million years ago) and was in a direction normal to the proposed direction. The conclusions on seismic activity that were established at the time of initial licensing are still valid today. The Ramapo Fault and the Ramapo Fault Zone do not represent a seismic hazard to the Indian Point plant. The Ramapo Fault is not a Tectonic plate that is capable of generating large earthquakes and shows no signs of movement.

It is therefore concluded that the seismic design criteria for structural analysis at the Indian Point site is satisfactory and that the plant is set on solid bedrock. The current analyses envelope any effect due to the Ramapo fault. No public hazard can be expected from the plant due to a probable earthquake in the region that would exceed its design margin.

Page 7 of 12

Indian Point Energy Center Position Paper 201 1-ISRT-02 Indian Point Site Seismic Review and affect of the Ramapo Fault

4. 0

References:

1. Unit 2 and Unit 3 FSAR, Final Safety Analysis Reports.

2. Unit 2 and Unit 3 TS, Technical Specifications.

3. NRC Letter to Riverkeeper dated Dec 15, 2004 and attachment.

4. Alan L. Kafka, Weston Observatory, Boston College, Seismological Research Letters, Volume 71, Number 3, May/June 2000. "PUBLIC MISCONCEPTIONS ABOUT FAULTS AND EARTHQUAKES IN THE EASTERN UNITED STATES: IS IT OUR OWN FAULT?

5. ALAN L. KAFKA, ELLYN A. SCHLESINGER-MILLER and NOEL L. BARSTOW, Weston Observatory, Boston College, "Earthquake activity in the greater New York City area:

Magnitudes, seismicity, and geologic structures", Bulletin of the Seismological Society of America; October 1985; v. 75; no. 5; p. 1285-1300

6. US Geological Survey database and website, http://earthquake.usas.aov

7. Lamont-Doherty Final Report on the Indian Point Seismic Network: Effective Monitoring of Seismicity and Ground Motions. NYPA doc 0001A, dated 7-17-95

8. Observations and tectonic setting of historic and Instrumentally Located Earthquakes in the greater NYC-Philadelphia Area, Sykes, Armbruster, Won-Young Kim, and Seeber, Bulletin of the Seismological Society of America, Vol 98. August 2008

9. Alan L. Kafka, Weston Observatory, Boston College, "Why does the Earth Quake in New England"

10. Henry N. Berry, Maine Geological Survey, "Earthquakes in Maine, Seismic Activity."

11. Stephane Mazzotti and John Townend, "State of stress in central and eastern North American seismic zones", Geological Survey of Canada, Natural Resources Canada, 9860 West Saanich Rd., Sidney, British Columbia V8L 4B2, Canada;. Pages 76-83.

12.Wolin, E, and Stein, S., "Passive Margin Earthquakes as Indicators of Intraplate Deformation", Northwestern University, Evanston, II.

13. Documentation for the 2008 Update of the United States National Seismic Hazard Maps, USGS Open-File Report 2008-1128.

14. Russell L. Wheeler, "Quaternary tectonic faulting in the Eastern United States", US Geological Survey, accepted 26 October 2005.

15. Scott D. Stanford, "Glacial aquifers of the New Jersey Highlands", 23rd Annual Meeting of the Geological Association of New Jersey, Ramapo College of New Jersey, New Jersey Geological Survey

16. Anthony J. Crone and Russell L. Wheeler, "Data for Quaternary faults, liquefaction features, and possible tectonic features in the Central and Eastern United States, east of the Rocky Mountain front", USGS Open-File Report 00-260

17. Burton, W.C., and Ratcliffe, N.M., 1985, "Compressional structures associated with right-oblique normal faulting of Triassic strata of the Newark basin near Flemington, New Jersey":

Geological Society of America Abstracts w i t h Programs,v. 17,p. 9.

18. Ratcliffe, N.M., Burton, W.C., 1985., "Fault reactivation models for the origin of the Newark basin and studies related to U.S. eastern seismicity.", U.S. Geological Survey Circular 946, 36-45.

19. Ratcliffe, N.M., 1984, Analysis of faulted glacial pavement in southeastern New York:, U.S.

Geological Survey Professional Paper 1375, p. 48 & 49 Page 8 of 12

Indian Point Energy Center Position Paper 201 1-ISRT-02 Indian Point Site Seismic Review and affect of the Ramapo Fault 20.Wheeler, R.L. and Crone, A.J., 2001, Known and suggested Quaternary faulting in the Midcontinent United States:, Schweig, E.S., ed, Earthquake hazard evaluation in the central United States, Engineering Geology, v. 61, no. 1-3, p. 51-78.

4.1 Historical

References:

1. Aggarwal, Y. P. and Sykes, L. R., 1978, "Earthquakes, Faults and Nuclear Power Plants in Southern New York and Northern New Jersey", "Science," V. 200.

2. Ratcliffe, N. M. 1976, "Final Report on Major Fault Systems in the Vicinity of Tomkins Cove

- Buchanan, New York": Report for Consolidated Edison Company of New York, Inc.

3. Dames & Moore, 1977, "Geotechnical Investigation of Ramapo Fault System in the Region of the Indian Point Generating Station."

4. Dames & Moore, 1980, "Seismic Ground Motion Hazard at Indian Point Nuclear Power Plant Site" - A report prepared for Pickard, Lowe & Garrick, Inc.

5. Ratcliffe, N. M. 1981, "Brittle Faults (Ramapo Fault) and Phyllonitic Ductile Shear Zones in the Basement Rocks of the Ramapo Seismic Zone, New York and New Jersey, and their Relationship to Current Seismicity" published in: Manspeizer, W., editor, Field Studies of New Jersey Geology and Guide to Field Trips, 52nd Annual Meeting of the New York State Geological Association, Newark, New Jersey, Rutgers University

6. United States Nuclear Regulatory Commission/Atomic Safety and Licensing Appeal Board, Farrar, M. C. - Chairman, Buck, J. H. and Quarles L. R. - Members at a Hearing cited as 6 NRC 547 (1977), ALAB -436.

7. Gutenberg, B. and Ritchter C. F., "Earthquake Magnitude, Intensity, Energy and Acceleration", BSSA, 32,(3) July 1942.

8. TERA Corporation, "Seismic Hazard Analysis - A Methodology for the Eastern United States", NUREG/CR-1 582, 2, 1980.

9. TERA Corporation, "Seismic Hazard Analysis Solicitation of Expert Opinion", NUREG/CR-1582,3,1980.

10. Woodward-Clyde; "Scientific Results of Seismic Monitoring Network near the Indian Point Nuclear Generating Facilities", Final Report (10/27/92); R&D Project 92284

11. Woodward-Clyde, 1986: "Kent Cliffs Bore-hole Research Project: A Determination of the Magnitude and Orientation of Tectonic Stress in Southeastern New York"; Research Report EP 84-27, Empire State Electric Energy Research Corporation, New York, New York.

12.TID-7024, "Nuclear Reactors and Earthquakes," August 1963

13. G. W. Housner, "Design of Nuclear Power Reactors Against Earth-quakes," Proceedings of the Second World Conference on Earthquake Engineering, Volume I, Japan, 1960, Pages 133, 134 and 137.

Page 9 of 12

Indian Point Energy Center I Position Paper 201 1-ISRT-02 Indian Point Site Seismic Review and affect of the Ramapo Fault I Attachment LARGEST EARTHQUAKES NEAR NEW YORK CITY Max.

DATE TIME LAT. LONG. MAGNITUDE Intensity Richter (ML) yr/mo/da hh:mm:sec (*N) (°W) LOCATION est. (MM) Remark Greater 1737 Dec N.Y. City 19 3:45 40.8 74 area- 5.2 VII Threw down chimneys 1783 Nov N. Central 30 3:50 41 74.5 N.J.* 4.9 VI Threw down chimneys Greater 1845 Oct N.Y. City 26 23:15 41.22 73.67 area* 3.8 VI Greater 1847 Sep N.Y. City 29 0:00:00 40.5 74 area* 4.5 V Probably Offshore Greater Many people in the NY 1848 Sep N.Y. City City area felt the 09 41.11 73.85 area* 4.4 V earthquake Near Nyack 1874 Dec and Tarry-11 3:25 41.05 73.85 town, N.Y. 3.4 VI 1878 Oct 4 7:30 41.5 74 V Greater 1884 Aug N.Y. City Threw down chimneys -

10 19:07 40.45 73.9 area 5.2 VII felt from Virginia to Maine; 1885 Jan Hudson 04 11:06 41.15 73.85 Valley 3.4- VI 1895 Sep N. Central Location determined by 01 11:09 40.55 74.3 N.J. 4.3 VI fire and aftershock Near Very high intensity in 1927 Jun Asbury Asbury Park, NJ -

01 12:23 40.3 74 Park, N.J. 3.9 VI-VII perhaps shallow event Western 1937 Jul Long Is., One or few earthquakes 19 3:51 40.6 73.76 N.Y. 3.5 IV beneath Long Island 1938 Aug Central 23 5:04:53 40.1 74.5 N.J. 3.8 VI 1951 Sep Rockland 03 21:26:24 41.25 74 Co., N.Y. 3.6 V 1952 Oct 8 21:40 41.7 74 V 1957 Mar Central 23 19:02 40.6 74.8 N.J. 3.5 Page 10 of 12

Indian Point Energy Center Position Paper 201 1-ISRT-02 I Indian Point Site Seismic Review and affect of the Ramapo Fault I 1964 Nov 17 17:08 41.2 4 73.7 4-

• V 1967 Nov 22 22:10 41.2 73.8 V 1976 Mar 11 21:07 41 74.4 V 1976 Apr 13 15:39 40.8 74 VI 1979 Jan 30 16:30 40.32 74.26 VI Felt by some people in 1979 Mar Central Manhattan[it is called 10 4:49:39 40.72 74.5 N.J. 3.2 V Chesequake earthquake]

1979 Dec 30 2:15:12 41.14 73.69 V 1984 Apr 23 39.92 76.36 4.4 IV Many people in the NY 1985 Oct Ardsley, City area felt this 19 10:07 40.98 73.83 N.Y. 4 IV earthquake 1991 Jun 17 42.63 74.68 4.1 IV 1994 Jan 16 40.33 76.04 4.6 V Manhattan, Felt in Upper East Side of 2001 Jan New York Manhattan, Long Island 17 12:34:22 40.78 73.95 City 2.4 IV City and Queens, NYC Manhattan, Felt in Upper West Side 2001 Oct New York of Manhattan, Astoria and 27 1:42:21 40.79 73.97 City 2.6 IV Queens, NYC Au Sable 2002 Apr Forks, New 20 York 5.1 V -VI 2002 Dec Redford, 25 New York 3.3 III 2002 May Plattsburgh 24 Aftershock 3.6 III

• Location very poorly determined; may be uncertain by 50 miles.

Note: Data taken from References 4.0.1 along with USGS and LCSN Databases.

Comment - The General area around the epicenter of the Oct. 17, 1989 Loma Prieta (World Series) was characterized as a VIII with other area in the San Francisco Bay area as a VII or VI using the MODIFIED MERCALLI SCALE.

Page II of 12

Indian Point Energy Center I Position Paper 201 1-ISRT-02 I

Indian Point Site Seismic Review and affect of the Ramapo Fault I Relationship of the Mercalli Scale to the Richter scale Modified Richter Maximum Description of effects Mercalli Acceleration Intensity (g)

I 0-2.5 .001 Not felt.

II 2.5-3.1 .002 Felt by persons at rest, on upper floors, or favorably placed.

III 3.1 -3.7 .003 to .007 Felt indoors. Hanging objects swing. Vibration like passing of light trucks. Duration estimated. May not be recognized as an earthquake.

IV 3.7-4.3 .007 to .015 Hanging objects swing. Vibration like passing of heavy trucks; or sensation of a jolt like a heavy ball striking the walls. Standing cars rock. Windows, dishes, door rattle. Glasses clink. Crockery dashes. In the upper range of IV, wooden walls and frame creak.

V 4.3-4.8 .015 to .03 Felt outdoors; direction estimated. Sleepers awakened. Liquids disturbed, some spilled. Small unstable objects displaced or upset. Doors swing, close, open.

Shutters, pictures move. Pendulum clocks stop, start, change rate.

VI 4.8-5.5 .03 to .09 Felt by all. Many frightened and run outdoors. Persons walk unsteadily. Windows, dishes, glassware broken. Knickknacks, books, etc., off shelves, pictures off walls.

Furniture moved or overturned. Weak plaster and masonry D cracked. Small bells ring (church, school). Trees, bushes shaken visibly, or heard to rustle.

VII 5.5-6.1 .07 to .22 Difficult to stand. Noticed by drivers. Hanging objects quiver. Furniture broken.

Damage to masonry D, including cracks. Weak chimneys broken at roof line. Fall of plaster, loose bricks, stones, tiles, cornices, also unbraced parapets and architectural ornaments. Some cracks in masonry C. Waves on ponds, water turbid with mud. Small slides and caving in along sand or gravel banks. Large bells ring.

Concrete irrigation ditches damaged.

VIII 6.1-6.7 .15 to .3 Steering of cars affected. Damage to masonry C; partial collapse. Some damage to masonry B; none to masonry A. Fall of'stucco and some masonry walls. Twisting, fall of chimneys, factory stacks, monuments, towers, elevated tanks. Frame houses moved on foundations if not bolted down; loose panel walls thrown out. Decayed piling broken off. Branches broken from trees. Changes in flow or temperature of springs and wells. Cracks in wet ground and on steep slopes.

IX 6.7 - 7.3 .3 to .7 General panic. Masonry D destroyed; masonry C heavily damage, sometimes with complete collapse; masonry B seriously damaged. General damage to foundations. Frame structures, if not bolted, shifted off foundations. Frames racked. Serious damage to reservoirs. Underground pipes broken. Conspicuous cracks in ground. In alluviated areas, sand and mud ejected, earthquake fountains, sand craters.

X 7.3-7.9 .45 to 1.5 Most masonry and frame structures destroyed with their foundations. Some well-built wooden structures and bridges destroyed. Serious damage to dams, dikes, embankments. Large landslides. Water thrown on banks of canals, rivers, lakes, etc. Sand and mud shifted horizontally on beaches and flat land. Rails bent slightly.

Xl 7.9-8.5 .5 to 3 Rails bent greatly. Underground pipelines completely out of service.

XII 8.5-9.0 A .5 to 7 Damage nearly total. Large rock masses displaced. Lines of sight and level distorted. Objects thrown into the air.

Notes:

Masonry A. Good workmanship, mortar and design: reinforced especially laterally and bound together using steel, concrete etc. Designed to resist lateral forces.

Masonry B. Good workmanship and mortar. Reinforced but not designed in detail; to resist horizontal forces.

Masonry C. Ordinary workmanship and mortar. No extreme weaknesses like failing to tie in at comers but neither reinforced nor designed to resist horizontal forces.

Masonry D. Weak materials such as adobe; poor mortar; low standards of workmanship; weak horizontally.

(Modified From Elementary Seismology by C.F. Richter)

Page 12 of 12

• 'Entergy ENG REPORT IP-RPT-11-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency Attachment 3 Indian Point 3 In-structure Response Spectra for LLNL &

USGS Uniform Hazard Spectra

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011 -SRT-03I Uniform Hazard Spectra Indian Point 3 In-structure Response Spectra For LLNL & USGS Uniform Hazard Spectra Prepared By: Paul Bruck Reviewed By: Paul Baughman 1.0 Problem

Description:

Perform an estimate of the In-Structure Response Spectra (ISRS) to be utilized in estimating component fragilities. The ISRS will be estimated using the Individual Plant Examination of External Events (IPEEE) [1] Review Level Earthquake (RLE) as defined by the set of earthquake curves from the NRC-sponsored Lawrence Livermore National Laboratories (LLNL) revised hazard estimates documented in NUREG-1488 [2]. Input to the assessment is a Uniform Hazard Spectrum (UHS), defined with a 0.23 g Peak Ground Acceleration (PGA) 10,000 year mean peak value associated with a 1.OE-04 frequency of exceedance per year, (LLNL 10,000 UHS). Additionally, a review of the recent United States Geological Survey (USGS) hazard estimates [3] was performed relative to the LLNL UHS. A Uniform Hazard Spectrum for the Indian Point site was obtained from USGS [3], defined with a 0.3g PGA, and 1.0E-04 frequency of exceedance per year. The shear wave velocity and site longitude and latitude are required to extract specific hazard curves from USGS.

From the LLNL UHS, the dominant mode scaling method outlined within EPRI NP-6041

[4], was used to scale ISRS from design basis ISRS for specific Indian Point Unit 3 (IP3) structures. The structures selected were associated with equipment and components identified in the IPEEE report for IP3 [5], which exhibited low or "weak link" HCLPF capacities.

2.0 Background Information The Indian Point structures are founded on rock, which is typical of the area and consists primarily of fractured, seamy limestone and dolomite. Rock excavations for the entire plant were carried out until firm rock was uncovered and fill concrete was added to bring the excavated areas to appropriate foundation level. The rock at the Indian Point site has a shear wave velocity in excess of 9,300 ft/second [6].

The Indian Point plant was designed in the mid 1960's and early 1970's in accordance with the codes, standards and criteria in effect at that time. The plant was designed to withstand a design earthquake presently designated as Operating Basis Earthquake Page 1 of 43

Indian Point Energy Center-Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 201 1-ISRT-03 Uniform Hazard Spectra (OBE) and a maximum potential earthquake, later referred to as the Design Basis Earthquake (DBE). The DBE level earthquake is commonly referred to as Safe Shutdown Earthquake (SSE). All Indian Point Structures, classified as Seismic Class I, per the UFSAR [6], are designed to resist both the OBE and DBE. The horizontal PGA for the OBE is defined as 0.1 g, and the DBE is defined as 0.15 g. Vertical response is considered as two-thirds horizontal response. Associated structural damping, as defined within the UFSAR for 1P3, considered 2% for the OBE and 5% for the DBE.

For the Indian Point structures, founded on rock, the unmodified shape but scaled amplitude versions of response spectra defined by G. W. Houser were utilized [6]. The "Housner" spectra are idealized response spectrum curves based on the average of recorded ground motions of larger US earthquakes [7]. Horizontal "design" spectra for the Indian Point site are shown in Figures 1 and 2.

IP3 Class 1 structures have ISRS generated as documented in [8]. The spectral plots from [8] have been digitized and are outlined within [9]. Earthquake time-histories, enveloping the design response spectra (as defined in Figure 1 and 2) were used as inputs into dynamic models of each structure to develop the design ISRS in [8].

3.0 Discussion of Hazard Curves For the IPEEE, the LLNL UHS was utilized, [2]. Per NUREG-1407 and as used in the assessment [5], the median spectral shape for the 10,000 year return period was utilized for the evaluation of components and structures. The UHS, normalized to the mean peak ground accelerations for the return periods 5,000 and 10,000 years, together with the design basis earthquake, are depicted in Figure 3. The resulting control points for the LLNL UHS, based on an interpolation of the ZPA for the median peak acceleration data, are listed in Table 1.0.

For GI-199, NRC utilized the latest USGS hazard curves [3]. For the Indian Point site the seismic hazard curve is shown in Figure 4, and data is listed in Table 2. From this data, the 10,000 year return USGS UHS is listed in Table 3.

4.0 Method of Evaluation The method of response spectra scaling as outlined in EPRI NP 6041 is as outlined in Figure 5. This approach utilizes the dominant mode frequency of the structure to define control acceleration values from the design ground response spectrum (GRS) and the applicable UHS. The ratio of these spectral accelerations defines a scaling factor. This Page 2 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011 -ISRT-03 Uniform Hazard Spectra scaling factor is then used to scale the design ISRS (from [8]), to develop ISRS applicable to the UHS.

In calculating fragilities, seismic demand may be reduced to account for ground motion incoherence (GMI). GMI functions have been developed for use in soil-structure interaction (SSI) analyses. These can be applied to both soil and rock sites, and they result in significant reductions in seismic demand especially for rock sites. SSI analysis is complex and would require too much time to use for this effort. However, EPRI NP-6041 [4] and ASCE 4 [11] contain conservative scale factors that can be used in lieu of a soil-structure interaction analysis - see Figure 5. They are a function of frequency and foundation size.

For a 150 foot plan dimension, the factors are 1.0 for frequencies less than or equal to 5 Hz, 0.9 for 10 Hz and 0.8 for 25 Hz and above. For other foundation sizes, a linear interpolation or extrapolation of the change in reduction (i.e., 1 minus the reduction factor) proportional to the plan dimension. Between frequencies log-log interpolation is used.

The assessment approach is as follows:

1. Design GRS spectral acceleration at structure fundamental mode = a

2. LLNL UHS spectral acceleration at structure fundamental mode = b

3. Incoherence factor at structure frequency, applicable to foundation size = IF

4. Hazard scale factor = Amp = (b/a) x IF

5. The resulting LLNL UHS floor response spectra are defined as:

LLNL UHS FRS = Design FRS x Amp 5.0 Structures Based on the weak link fragilities, components were selected for fragility evaluation.

These components were identified and located to applicable structures and floor elevations. These structures and floor elevations were then assessed to estimate ISRS for the LLNL UHS, utilizing the scaling method outlined above.

Page 3 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 201 1-ISRT-03 Uniform Hazard Spectra The following structures at selected floor elevations were evaluated:

• Control and Diesel Generator Building

• Primary Auxiliary Building

" Intake Structure

" Inner Containment Structure ISRS for the LLNL UHS are summarized in Attachments to this document.

6.0 References

1. Nuclear Regulatory Commision, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10CFR50.54(f)", Generic Letter No.

88-20.

2. NRC NUREG 1488, "Revised Livermore Seismic Hazard Estimates for 69 Sites East of the Rocky Mountains", Published in April 1994.

3. US Geological Survey USGS):

http://earthq uake. usqs~gov/hazards/products/conterminous/2008/update201001/curves

4. EPRI NP-6041, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin (revision 1)".

5. Indian Point 3 "Individual Plant Examination of External Events", IP3-RPT-UNSPEC-02182, September 1997.

6. Entergy, Indian Point Unit 3 Updated final Safety Analysis Report.

7. "Design of Nuclear Power Reactors Against Earthquakes", Proceedings of the Second World Conference on Earthquake Engineering", Vol. I, Japan, 1960.

8. Westinghouse, "Seismic Response Curves for Class I Buildings for Indian Point Generating Station Unit No. 3"

9. Teledyne TR-7366-4, "Digitized Data by Frequency - Design Response Spectra Database for Indian Point Unit 3"

10. EPRI Report TR-1 03959, "Methodology for Developing Seismic Fragilities", EPRI July 1994

11. American Society of Civil Engineers, Standard ASCE 4-98, "Seismic Analysis of Safety Related Nuclear Structures and Commentary".

Page 4 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03I Uniform Hazard Spectra 1.0 8

6 4

-J a

.4 2:

w

.. o. ,1 8

w 6" 4.

2.

0. 2 4 8 8 0.1 2 4 a ai.0 2 UNDAMPED NATURAL PERIOD T-*SC Figure 1: Site Response Spectra (Operational Basis Earthquake) 0.1g PGA [6]

8 "8.

4 0

-U .2

,15 0.1 lu 8:

0 2 4 . 6:. 8 0.1 ,2 .4 .6 ;8 1.0 2 UNDAMPED NATURAL PERIOo T-SEC Figure 2: Site response Spectra (Design Basis Earthquake) 0.15 PGA [61 Page 5 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 201 1-ISRT-03 Uniform Hazard Spectra

.4 t4 tl G2 1 10 lO0 Figure 3: LLNL 5,000 and 10,000 Return Uniform Hazard Spectra [5]

Page 6 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011 -ISRT-03 Uniform Hazard Spectra USGS Hazard Curves 1.OOE-03 8.OOE-04

=6.OOE-04 0.

o4.OOE-04 x I W2.00E-04 O.OOE+O0 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 Acceleration (g)

Figure 4: USGS IPEC Specific Hazard [3]

Page 7 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03 Uniform Hazard Spectra 8"SetrWA I Av A

I~ A- A 1 - AimM56 SCALEWTft.kOTUFW ACCELdRAflON

/ \ -%u&

/i MFAAN)Qr _W INTLMNACIPTOMP4~wOMR Figure 5: Floor Response Spectra Scaling [4]

Page 8 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011 -ISRT-03 IUniform Hazard Spectra 3J.3 I.. WMao ..

tcaeFa*wie The assumption of vertically propagating plane shear and comprnessional waves when performinsSSm" analysis us ly conse.rtn !ter* s of p*jdcing in-structu*relspnse.n Ithe absenceof W .alys to TABLE 3.-2. Reductions to Ground

,espams spe*tra Redudion Factor fr~Plan Frequency Dimonsown of (Hz) 0, ft 300 ft' 1'A10 .1.0 10.. 0.9 0A 25 .0.8 .0.6 Figure 6: Wave Incoherence Factors [11]

Page 9 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03 Uniform Hazard Spectra Table 1: LLNL UHS LLNL 10,000 year UHS Ground

Response

Frequency Accel (Hz) (g) 1 0.037 2.5 0.133 5 0.236 10 0.343 25 0.368 50 0.230 Page 10 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-1SRT-03I Uniform Hazard Spectra Table 2: USGS Seismic Hazard Curve 2008 USGS Seismic Hazard Curves for Indian Point As Used in the GI-199 Safety/Risk Assessment PGA 10 Hz 5 Hz 1 Hz annual annual annual annual exceedence exceedence exceedence exceedence accel frequency accel frequency accel frequency accel frequency (g's) (per year) (g's) (per year) Ws) (per year) (0) (per year) 0.007 8.02E-03 0.0075 1.20E-02 0.0075 1.47E-02 0.0037 1.27E-02 0.0098 5.72E-03 0.0113 8.43E-03 0.0113 9.73E-03 0.0056 8.13E-03 0.0137 4.04E-03 0.0169 5.86E-03 0.0169 6.37E-03 0.0084 4.98E-03 0.0192 2.82E-03 0.0253 3.98E-03 0.0253 4.07E-03 0.0127 2.89E-03 0.0269 1.95E-03 0.038 2.62E-03 0.038 2.52E-03 0.019 1.63E-03 0.0376 1.34E-03 0.057 1.69E-03 0.057 1.53E-03 0.0285 8.78E-04 0.0527 9.06E-04 0.0854 1.06E-03 0.0854 9.08E-04 0.0427 4.61E-04 0.0738 6.09E-04 0.128 6.49E-04 0.128 5.26E-04 0.0641 2.37E-04 0.103 4.07E-04 0.192 3.87E-04 0.192 2.97E-04 0.0961 1.21E-04 0.145 2.65E-04 0.288 2.23E-04 0.288 1.62E-04 0.144 6.10E-05 0.203 1.71E-04 0.39 1.44E-04 0.39 1.01E-04 0.216 3.04E-05 0.284 1.07E-04 0.432 1.24E-04 0.432 8.52E-05 0.324 1.47E-05 0.397 6.46E-05 0.649 6.45E-05 0.649 4.23E-05 0.4 9.79E-06 0.556 3.69E-05 0.8 4.49E-05 0.8 2.87E-05 0.487 6.57E-06 0.778 1.97E-05 0.973 3.13E-05 0.973 1.95E-05 0.73 2.68E-06 0.9 1.46E-05 1.46 1.37E-05 1.46 8.09E-06 0.9 1.61E-06 1.09 9.58E-06 2.19 5.12E-06 2.19 2.91E-06 1.09 9.81E-07 1.52 4.05E-06 3.28 1.42E-06 3.28 8.50E-07 1.64 3.08E-07 2.13 1.24E-06 5 1.42E-07 4.92 1.38E-07 2.46 8.23E-08 I 1 1 3.69 1.60E-08 Table 3: USGS UHS Page 11 of 43

Indian Point Energy Center Position Paper ; Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011 -ISRT-03 Uniform Hazard Spectra Attachment A In-structure Response Spectra for Control and Diesel Generator Building Page 12 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011 -ISRT-03I Uniform Hazard Spectra AI.0 Control & Diesel Generator Building (C&DG Building) Evaluation The Control and Diesel Generator Building dynamic model is as summarized in [8]. The resulting design response spectra from [8] have been digitized as summarized in [9].

The dynamic model is depicted in Figure Al and A2. The resulting dynamic response for the Control and Diesel Generator Building are summarized in Figure A3 and A4.

Based on the structure response, and using the scaling approach outlined above in Section 4.0, the resulting amplification factors and peak of the In-structure response Spectra for the UHS were derived:

EL32 ft EL 48 ft Direction Hazard 1st Design Hazard Incoh Amp Peak Peak Mode Freq GRS Response Factor Factor FRS (g) FRS (g)

(Hz)

N-S LLNL 11.720 0.188 0.340 0.950 1.723 0.593 0.820 E-W LLNL 13.180 0.186 0.348 0.900 1.680 0.496 0.740 HAZARD Resulting digitized values are listed in Table Al for the design spectra, and Table A2 for the hazard spectra. Applicable north-south and east-west spectra plots are also included, Figures A5 through A9.

Page 13 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03 Uniform Hazard Spectra

. . . . . . " -. . . , - I'*

. -. 2

~I ,

Figure Al: Control & Diesel Generator Building Dynamic Model (N-S) [8]

7 .T- wc-.s- ,* \n,\ OAK CQI-Jr/CL LUWMPEO t4ASS 000EL i~

P 1m 32 *m Al £ 1 lsl 1 3Ell

... 1 4Te aJl -J

/*0 Figure A2: Control & Diesel Generator Building Dynamic Model (E-W) 18]

Page 14 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 201 1-ISRT-03 Uniform Hazard Spectra AMVc AMore "a:~

p, 0AJr 419 ( OCO 4o 0 0~c 4-. '1 ii Figure A3: Control & Diesel Generator Building Dynamic Information (N-S) [8]

• =*,*u'---"-4:.':)*-  %%,=Qa . ,-_,.*_-% * -_*

I MA~S ~

Il poI~4r I. .1* I I j I

• I _

I 0

I~

.1 0

• I

• aZ4~, - .&~3t~ -.

• *2S~

3 *O~O' I.(J~ -S 4- - .1-1.3 1.oo II 7

S I.c~oc~ * .V2.(-"~. .4,4 -. a~49 Figure A4: Control & Diesel Generator Building Dynamic Information (E-W) [8]

Page 15 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011 -ISRT-03 Uniform Hazard Spectra IP3 Control & Diesel Generator Building -

Design Ground Response 0.600 ý 0.500 w

I

• 1 U

0.300 0.200 U

0.100 0.000 1.000 10.000 100.000 Frequency (Hz)

- DBE 5% Damping -- LLNL 10,000 yr UHS -USGS 10,000yr UHS Figure A5: C&DG Building Ground Response Page 16 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03 Uniform Hazard Spectra IP3 Control & Diesel Generator Building -

El. 32 ft Floor Response N-S 0.700 0.600 0.500 w OAOO w 0.300

'U 0.200 0.100 1.000 10.000 100.000 Frequency (Hz)

- DBE 5% Damping Grnd -DBE 32' 5% Damp N-S

- LLN L10,000 UHS Scaled N-S Figure A6: C&DG Building FRS IP3 Control & Diesel Generator Building -

El. 32 ft Floor Response E-W 0.600 0.400 0.200

,i.............

0.000 1.000 10.000 100.000 Frequency (Hz)

- DBE 5% Damping Grnd -DBE 32' 5% Damp E-W

- LLNL 10,000 UHS Scaled E-W Figure A7: C&DG Building FRS Page 17 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03 Uniform Hazard Spectra IP3 Control & Diesel Generator Building -

El. 48 ft Floor Response N-S 0.900 0.800 0.700 0.600 0.500

• 1

.3 U

0.400 ]

0.300 .0.....

0 ...

0.200 0 .100 J................

0 .0 00 ................

1.000 10.000 100.000 Frequency (Hz)

-DBE 5% Damping Grnd -DBE 48' 5% Damp N-S

- LLN L10,000 UHS Scaled N-S Figure A8: C&DG Building FRS IP3 Control & Diesel Generator Building -

El. 48 ft Floor Response E-W 0.800 I

0.600 J S.2400 0.20000 1.000 10,000 100.000 Frequency (Hz)

- DBE 5% Damping Grnd -DBE 48' 5% Damp E-W

- LLNL 10,000 UHS Scaled E-W Figure A9: C&DG Building FRS Page 18 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 201 1-ISRT-03 Uniform Hazard Spectra Table Al: C&DG Building Digitized Design Spectra [9]

Control Building SSE Control Building SSE Control Building SSE 5% Structural Control Building SSE 5% Control Building SSE 5% 5% Structural 5% Structural Damping Ground Structural Damping 32' Structural Damping 32' Damping 48' Damping 48' Elevation Elevation Elevation E-W Elevation N-S Elevation E-W N-S Accel Accel Frequency (g) Frequency Accel (g) Frequency Accel (g) Frequency (g) Frequency Accel (g) 1.024 0.153 1.002 0.154 1.006 0.153 1.038 0.153 1.023 0.152 1.065 0.153 1.064 0.152 1.066 0.153 1.067 0.152 1.065 0.152 1.090 0.161 1.091 0.161 1.091 0.163 1.090 0.161 1.091 0.161 1.113 0.174 1.111 0.173 1.108 0.172 1.107 0.171 1.097 0.166 1.138 0.183 1.139 0.182 1.137 0.183 1.141 0.182 1.113 0.174 1.249 0.183 1.252 0.184 1.251 0.183 1.250 0.184 1.135 0.180 1.284 0.198 1.284 0.200 1.283 0.199 1.284 0.200 1.164 0.182 1.352 0.198 1.371 0.200 1.359 0.199 1.371 0.200 1.189 0.177 1.432 0.207 1.431 0.208 1.429 0.208 1.416 0.206 1.281 0.177 1.613 0.208 1.559 0.208 1.508 0.208 1.670 0.213 1.314 0.198 1.670 0.212 1.672 0.212 1.592 0.210 1.787 0.225 1.416 0.198 1.798 0.223 1.800 0.225 1.669 0.214 1.912 0.232 1.471 0.207 2.003 0.235 1.902 0.230 1.770 0.223 2.186 0.252 1.530 0.207 2.096 0.241 2.005 0.237 2.167 0.248 2.272 0.252 1.665 0.213 2.273 0.245 2.166 0.246 2.264 0.250 2.375 0.261 1.998 0.240 2.375 0.254 2.268 0.248 2.370 0.258 2.621 0.261 2.125 0.249 2.619 0.254 2.377 0.256 2.624 0.258 2.776 0.267 2.264 0.252 2.780 0.260 2.626 0.256 2.778 0.267 2.946 0.285 2.374 0.261 2.963 0.276 2.780 0.263 2.932 0.281 3.333 0.285 2.612 0.261 3.337 0.276 2.943 0.281 3.328 0.281 3.537 0.301 2.754 0.268 3.571 0.289 3.331 0.281 3.564 0.307 4.168 0.301 2.936 0.283 4.181 0.289 3.569 0.295 4.144 0.307 4.558 0.280 3.310 0.283 4.575 0.258 4.179 0.295 4.558 0.274 5.543 0.280 3.573 0.321 5.556 0.258 4.562 0.270 5.528 0.274 6.207 0.265 4.161 0.321 6.215 0.235 5.531 0.270 6.196 0.263 6.859 0.265 4.560 0.288 7.062 0.235 6.215 0.252 7.032 0.263 10.081 0.275 7.138 0.286 8.190 0.208 7.107 0.252 8.190 0.344 12.610 0.440 8.237 0.476 9.747 0.193 9.862 0.262 9.737 0.344 19.724 0.440 11.507 0.476 12.195 0.186 12.346 0.279 16.502 -0.228 30.675 0.220 16.529 0.254 19.734 0.186 19.724 0.279 30.675 0.210 50.000 0.190 19.724 0.225 30.675 0.170 30.675 0.180 50.000 0.170 40.000 0.185 50.000 0.150 50.000 0.150 50.000 0.185 Grnd--A E-W--A N-S--A E-W--A N-S--A Page 19 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03 Uniform Hazard Spectra I Table A2: C&DG Building Digitized UHS LLNS Factored of Design USGS factor of Design LLNS Factored of Design FRS FRS FRS E-W EL 32 N-S EL 32 E-W EL 32 N-S EL 32 E-W EL 48 N-S EL 48 Freq Accel Freq Accel Accel Accel Freq Accel Freq Accel 1.0020 0.2592 1.0060 0.2636 0.3649 0.3877 1.0380 0.2562 1.0230 0.2610 1.0640 0.2557 1.0660 0.2636 0.3599 0.3877 1.0670 0.2557 1.0650 0.2610 1.0910 0.2703 1.0910 0.2799 0.3805 0.4118 1.0900 0.2704 1.0910 0.2772 1.1110 0.2913 1.1080 0.2963 0.4100 0.4359 1.1070 0.2878 1.0970 0.2853 1.1390 0.3059 1.1370 0.3154 0.4306 0.4640 1.1410 0.3064 1.1130 0.2999 1.2520 0.3086 1.2510 0.3154 0.4344 0.4640 1.2500 0.3086 1.1350 0.3104 1.2840 0.3358 1.2830 0.3433 0.4727 0.5051 1.2840 0.3363 1.1640 0.3139 1.3710 0.3358 1.3590 0.3433 0.4727 0.5051 1.3710 0.3363 1.1890 0.3046 1.4310 0.3491 1.4290 0.3575 0.4914 0.5259 1.4160 0.3465 1.2810 0.3046 1.5590 0.3491 1.5080 0.3575 0.4914 0.5259 1.6700 0.3576 1.3140 0.3416 1.6720 0.3568 1.5920 0.3619 0.5022 0.5324 1.7870 0.3781 1.4160 0.3416 1.8000 0.3771 1.6690 0.3685 0.5309 0.5421 1.9120 0.3897 1.4710 0.3573 1.9020 0.3859 1.7700 0.3848 0.5432 0.5662 2.1860 0.4228 1.5300 0.3573 2.0050 0.3983 2.1670 0.4277 0.5607 0.6293 2.2720 0.4228 1.6650 0.3671 2.1660 0.4134 2.2640 0.4314 0.5819 0.6346 2.3750 0.4381 1.9980 0.4141 2.2680 0.4168 2.3700 0.4450 0.5867 0.6546 2.6210 0.4381 2.1250 0.4289 2.3770 0.4305 2.6240 0.4450 0.6061 0.6546 2.7760 0.4485 2.2640 0.4345 2.6260 0.4305 2.7780 0.4598 0.6061 0.6764 2.9460 0.4787 2.3740 0.4500 2.7800 0.4418 2.9320 0.4841 0.6219 0.7121 3.3330 0.4787 2.6120 0.4500 2.9430 0.4715 3.3280 0.4841 0.6638 0.7121 3.5370 0.5056 2.7540 0.4618 3.3310 0.4715 3.5640 0.5292 0.6638 0.7785 4.1680 0.5056 2.9360 0.4873 3.5690 0.4959 4.1440 0.5292 0.6980 0.7785 4.5580 0.4700 3.3100 0.4873 4.1790 0.4959 4.5580 0.4713 0.6980 0.6934 5.5430 0.4700 3.5730 0.5535 4.5620 0.4530 5.5280 0.4713 0.6377 0.6934 6.2070 0.4450 4.1610 0.5535 5.5310 0.4530 6.1960 0.4536 0.6377 0.6673 6.8590 0.4450 4.5600 0.4956 6.2150 0.4228 7.0320 0.4536 0.5952 0.6673 10.0810 0.4616 7.1380 0.4923 7.1070 0.4228 8.1900 0.5929 0.5952 0.8723 12.6100 0.7396 8.2370 0.8198 9.8620 0.4394 9.7370 0.5929 0.6186 0.8723 19.7240 0.7396 11.5070 0.8198 12.3460 0.4688 16.5020 0.3921 0.6600 0.5768 30.6750 0.3696 16.5290 0.4379 19.7240 0.4688 30.6750 0.3618 0.6600 0.5322 50.0000 0.3192 19.7240 0.3873 30.6750 0.3024 50.0000 0.2929 0.4256 0.4308 40.0000 0.3187 50.0000 0.2520 0.0000 0.3547 0.0000 50.0000 0.3187 Page 20 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011 -ISRT-03 ,Uniform Hazard Spectra Attachment B In-structure Response Spectra for Primary Auxiliary Building Page 21 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03 Uniform Hazard Spectra B1.0 Primary Auxiliary Building (PAB) Evaluation The Primary Auxiliary Building (PAB) dynamic model is as summarized in [8]. The resulting design response spectra from [8] have been digitized as summarized in [9].

The dynamic model is depicted in Figure B1. The resulting dynamic response for the PAB is summarized in Figure B2 and B3.

Based on the structure response, and using the scaling approach outlined above in Section 4.0, the resulting amplification factors and peak of the In-structure response Spectra for the UHS were derived:

EL55 ft EL73 ft Direction Hazard 1st Design Hazard Incoh Amp Peak Peak Mode Freq GRS Response Factor Factor FRS (g) FRS (g)

(Hz)

N-S LLNL 14.300 0.186 0.350 1.000 1.879 0.571 0.697 E-W LLNL 14.680 0.186 0.350 0.950 1.793 0.538 0.663 HAZARD Resulting digitized values are listed in Table B1 for the design spectra, Table B2 for the hazard spectra. Applicable north-south and east-west spectra plots are also included, Figures B4 to B8.

Page 22 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011 -ISRT-03 IUniform Hazard Spectra i,

j.-

'1 iI

,i

~777oil

- I "i 1u/~Ps ,flIsSL. ,AI:DEL -oP bPTJ- 04S_77 k/t.CS7 4A~bP NL-,PT14 SCLJTN EA - IiZI Figure Bi: PAB Dynamic Model [8]

F -J1AA I,.I Pý p /

As'4 g A~ I I. .A..i~4 F-~i ~.rr 1 c~I o 100

• ep3 -. 2.4.3 4

'4- I.aa Lac., I 4'

'I a - ____

Figure 82: PAB Dynamic Information (N-S) [8]

Page 23 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03 Uniform Hazard Spectra PFMA ? UX*isI 'JL NC . -r- S,

,*, ASS, 5 . 5.o4. --. t 1.0*

4 l..oo I. oo - "

Figure B3: PAB Dynamic Information (E-W) [8]

PAB - Ground Response 0.600 0.500

- 0.400 0.200 "---:-K 0.100 0.000 1.000 10.000 100.000 Frequency (Hz)

-DBE 5% Damping - LLNL 10,000yr UHS - USGS 10,000yr UHS Figure B4: PAB Ground Response Page 24 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03 Uniform Hazard Spectra PAB - EL 55 ft N-S Floor Response 0.600 0.500

• 1 S

-i 0.400 0.300 U

0.200 .................

............... 10..0....0....

1.000 10.000 Frequency (Nz)

DBE 5% Damping Grnd - DBE 55' 5% Damp N-S -- LLN L55' UHS N-S Figure B5: PAB FRS PAB - EL 55 ft E-W Floor Response 0.600 OAOO-0.200 1.000 10.000 100.000 Frequency (Hz)

-DBE 5% Damping Grnd - DBE 5% Damping EL 55' E-W

- LLNS S' UHS E-W Figure B6: PAB FRS Page 25 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03 Uniform Hazard Spectra PAB - EL 73 ft N-S Floor Response 0.800 0.700 0.600 ..........

S 0.500 0.400 S

0 U

0.300 U

0.200 0.100 .

0.000 1.000 10.000 100.000 Frequency (Hz)

- DBE 5% Damping Grnd - DBE 73' 5% Damp N-S - LLNL 73' UHS N-S Figure B7: PAB FRS PAB - EL 73 ft E-W Floor Response 0.800------ ... ..........- ....... .. - *- ... . ....

0 .6 0 0 ....... .................

0.400 0.200 S 0.000 ....... .....

1.000 10.000 1 Frequency (Hz)

- DBE 5% Damping Grnd - DBE 5% Damping EL 73' E-A

-LLNS 73' UHS E-W Figure B8: PAB FRS Page 26 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011 -ISRT-03 Uniform Hazard Spectra Table Bi: PAB Digitized Design Spectra [9]

PAB SSE5%

PAB SSE 5% Structural Structural Damping PAB SSE5% Intake Structure SSE 5% PA SSE5% Structural Intake Structure SSE 5%

Damping Ground Ground Elevation Structural Damping Structural Damping 55' Damping 73 Elevation Structural Damping 73' Elevat*on (from E-W* (from N-S) 55'Elevation E-W Elevation N-S_ EW Elevation N-S Freq A Freg Accel (g) Freq Accel (g) Freq Accel (g) Freq Accel (g) Freg Accel (g) 1.029 0.152 1.009 0.153 1.024 0.154 1.036 0.154 1.025 0.152 1.036 0.152 1.065 0,152 1.067 0.153 1.067 0.154 , 1.064 0.154 1.065 0.152 1.066 0.152 1.088 0.160 1.089 0.161 1.091 0.163 1.089 0.162 1.088 0.160 1.090 0.161 1111 0.174 1.104 0.170 1.112 0.174 1.112 0.176 1.113 0.175 1.111 0.174 1.138 0.182 1.138 0.184 1.138 0.183 1.138 0.184 1.139 0.184 1.137 0.183 1.249 0.183 1.250 0.184 1.250 0.183 1.251 0.186 1.251 0.184 1.251 0.185 1.281 0.199 1.282 0.199 1.284 0.198 1.285 0.201 1.2.4 0.199 1.284 0.199 1.354 0.199 1.353 0.199 1.353 0-198 1.356 0.201 1.355 0.199 1.358 0.199 1.429 0.207 1.425 0.207 1.426 0.207 1.428 0.209 1430 0.208 1.428 0.208 1.613 0.208 1.514 0.207 1.509 0.207 1.514 0.209 1.505 0.208 1.506 0.208 1.668 0,211 1.607 0.209 1.669 0.212 1.606 0.211 1.620 0.211 1.669 0.214 1.789 0.222 1.665 0.212 1.768 0.222 1.666 0.214 1.669 0.214 1.742 0.222 1.902 0.228 1.717 0.216 1.899 0.231 1.706 0.218 2.002 0.240 1.829 0.227 1.992 0.234 1.781 0.224 2.171 0.248 1.777 0.226 2.159 0.251 1.920 0.234 2.088 0.240 1.984 0.233 2.265 0,249 1.848 0.229 2.24 0.252 2.062 0.246 2.194 0.245 2.058 0.239 2.380 0.258 1.932 0.235 2.379 0.260 2.178 0.253 2.252 0.245 2.166 0.244 2.620 0.258 2.168 0.251 2.621 0.260 2.251 0.253 2.374 0.253 2.266 0.246 2.773 0.263 2.249 0.251 2.776 0.267 2.364 0.261 2.610 0.253 2.374 0.254 2.943 0.282 2.372 0.259 2.946 0.287 2.625 0.261 2.777 0.259 2.618 0.254 3.343 0.282 2.624 0.259 3.324 0.287 2.762 0.266 2.943 0.276 2.765 0.259 3.575 0.300 2.769 0.265 3.565 0.311 2.930 0.288 3.326 0.276 2.934 0.276 4.171 0.300 2.934 0.284 4.172 0.311 3.319 0.288 3.563 0.288 3.338 0.276 4.552 0.271 3.327 0.284 4.558 0.298 3.556 0.313 4.158 0.288 3.582 0.288 5.498 0.271 3.565 0.304 5.519 0.290 4.143 0.313 4.552 0.258 4.191 0.288 6.188 0.264 4.136 0.304 7.047 0.290 4.519 0.301 5.552 0.258 4.591 0.257 14.535 0.264 4.543 0.276 9.970 0.311 5.467 0.295 6.231 0.234 5.543 0.257 30.675 0.160 5.501 0.276 12.516 0.370 7.072 0.295 7.072 0.234 6.234 0.234 50.000 0.150 6.112 0.271 18.416 0.370 8.217 0.334 8.258 0.207 7.102 0.234 7.042 0.271 30.675 0.170 9.950 0.371 9.833 0.192 8.137 0.209 9.881 0.292 50.000 0.170 18.416 0.371 12.579 0.186 9.699 0.193 _ 14.535 0.292 J 30.675 0.180 19.724 0.186 12.392 0.186 30.675 0.150 50.000 0.150 30.675 0.170 19.724 0.186 50.000 0.150 _ _

50,000 0.150 30.675 0.160i .....

Page 27 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011 -ISRT-03 Uniform Hazard Spectra I I I 1 50.000 1 0.160 II : I I .. I I Table 82: PAB Digitized UHS PAB Elevation 55 LLNS Factor of Design FRS USGS factor of LLNS Factor of Design FRS Design FRS E-W I4-S E-W N-S E-W N-S Freq Accel FFreq Accel Accel Accel Freq Accel Freq Accel 1.024 0.276 1.036 0.290 0.377 0.399 1.025 0.272 1.036 0.286 1.067 0.276 1.064 0.290 0.377 0.399 1.065 0.272 1.066 0.286 1.091 0.292 1.089 0.305 0.399 0.420 1.088 0.286 1.090 0.303 1.112 0.312 1.112 0.330 0.427 0.455 1.113 0.313 1.111 0.327 1.138 0.328 1.138 0.346 0.449 0.477 1.139 0.329 1.137 0.343 1.250 0.328 1.251 0.349 0.449 0.480 1.251 0.329 1.251 0.347 1.284 0.355 1.285 0.377 0.486 0.519 1.284 0.357 1.284 0.375 1.353 0.355 1.356 0.377 0.486 0.519 1.355 0.357 1.358 0.375 1.426 0.370 1.428 0.392 0.506 0.540 1.430 0.373 1.428 0.391 1.509 0.370 1.514 0.392 0.506 0.540 1.505 0.373 1.506 0.391 1.669 0.380 1.606 0.397 0.519 0.546 1.620 0.378 1.669 0.402 1.768 0.399 1.666 0.403 0.545 0.555 1.669 0.384 1.742 0.416 1.899 0.413 1.706 0.409 0.565 0.564 2.002 0.430 1.829 0.427 2.171 0.444 1.777 0.424 0.607 0.584 2.159 0.449 1.920 0.439 2.265 0.447 1.848 0.431 0.611 0.593 2.264 0.452 2.062 0.462 2.380 0.462 1.932 0.441 0.632 0.607 2.379 0.467 2.178 0.475 2.620 0.462 2.168 0.471 0.632 0.649 2.621 0.467 2.251 0.475 2.773 0.472 2.249 0.471 0.645 0.649 2.776 0.478 2.364 0.491 2.943 0.505 2.372 0.487 0.691 0.671 2.946 0.514 2.625 0.491 3.343 0.505 2.624 0.487 0.691 0.671 3.324 0.514 2.762 0.500 3.575 0.538 2.769 0.498 0.735 0.686 3.565 0.557 2.930 0.541 4.171 0.538 2.934 0.534 0.735 0.736 4.172 0.557 3.319 0.541 4.552 0.486 3.327 0.534 0.665 0.736 4.558 0.534 3.556 0.588 5.498 0.486 3.565 0.571 0.665 0.786 5.519 0.519 4.143 0.588 6.188 0.473 4.136 0.571 0.647 0.786 7.047 0.519 4.519 0.565 14.535 0.473 4.543 0.519 0.647 0.714 9.970 0.558 5.467 0.554 30.675 0.287 5.501 0.519 0.392 0.714 12.516 0.663 7.072 0.554 50.000 0.269 6.112 0.509 0.368 0.701 18.416 0.663 8.217 0.628 7.042 0.509 0.701 30.675 0.305 9.950 0.697 9.881 0.550 0.757 50.000 0.305 18.416 0.697 14.535 0.550 0.757 30.675 0.338 30.675 0.282 0.388 50.000 0.282 50.000 0.282 0.388 Page 28 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03 Uniform Hazard Spectra I Attachment C In-structure Response Spectra for Intake Structure Page 29 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 201 1-ISRT-03 Uniform Hazard Spectra C1.0 Intake Structure Evaluation The Intake Structure (IS) dynamic model is as summarized in [8]. The resulting design response spectra from [8] have been digitized as summarized in [9]. The dynamic model is depicted in Figure C1. The resulting dynamic response for the IS is summarized in Figure C2 and C3.

Based on the structure response, and using the scaling approach outlined above in Section 4.0, the resulting amplification factors and peak of the In-structure response Spectra for the UHS were derived:

EL 15 ft Direction Hazard 1st Design Hazard Incoh Amp Peak Mode Freq GRS Response Factor Factor FRS (g)

(Hz)

N-S LLNL 11.720 0.186 0.346 1.000 1.857 1.016 E-W LLNL 13.180 0.186 0.357 1.000 1.920 1.022 HAZARD Resulting digitized values are listed in Table Cl for the design spectra, Table C2 for the hazard spectra. Applicable north-south and east-west spectra plots are also included, Figures C4 to C6.

Page 30 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-1SRT-03I Uniform Hazard Spectra 27 I -

1 :1

':LUMPe5C MASS /401Z'eL o~ e~~rA~ J~&cWPE~

4 .r.JO. ~4JW

.EAPTa -I c~UA J4 4 I..

Figure Cl: IS Dynamic Model [8]

1 -.-.

4 T A K.S 5T~QCTuP.~. -~ -

IAc~~ ~4.IAPE~S I

MACS 9

'I

~ra.z~1i 1c~ 4~ 1f U

/o

.A I I I I

0 II 1.

0 I.

a I

I.oo -J.~,o I' I i I I!

I I I I . . . .

Figure C2: IS Dynamic Information (N-S) [8]

Page 31 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-1SRT-03 Uniform Hazard Spectra AA5S .4 -,

FI I MI "Z 4 Figure C3: IS Dynamic Information (E-W) [8]

IP3 Intake Structure - Ground Response 0.600 T 0.500 i i 0 .3 0 0 .....................

0o.300o --

0.200 0.100 _

0.000 - ----------- _ _ _

1.000 10.000 100.000 Frequency(Hz)

- DBE 5% Damping - LLNL l0,O00yr UHS - USGS 10,O00yr UHS Figure C4: IS Ground Response Page 32 of 43

Indian Point Energ Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-1SRT-03 Uniform Hazard Spectra IP3 Intake Structure - EL 15 ft N-S Floor Response 1.200 T 1.0001 I_ _i _ _

0.8001 101600]

0.200 1i 0.000 1.000 10.000 100.000 Frequetcy (Hz)

-OBE 5%Damping Grnd -D8E15.5% Damp N-S -LINt 15' VHS N-S Figure C5: IS FRS IP3 Intake Structure - EL 15 ft E-W Floor Response 1.500 - ----

S U 1.0001

• 1 I

El I*ll ll l or 0.0004- _. -_

1.000 10.000 100.000 Frequency (Hz)

-DOE 5% Damping Grnd - DBE 5% Damping EL 15' E-W

-LNSI151UHS E-W Figure C6: IS FRS Page 33 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 201 1-ISRT-03 Uniform Hazard Spectra Table Cl: IS Digitized Design Spectra [9]1 Intake Structure SSE 5% Intake Structure SSE 5% Intake Structure SSE 5%

Structural Damping Ground Structural Damping 15' Structural Damping 15' Elevation (from N-S) Elevation E-W Elevation N-S Frequency Accel (g) Frequency Accel (g) Frequency Accel (g) 1.051 0.153 1.030 0.153 1.033 0.154 1.065 0.153 1.065 0.153 1.067 0.154 1.090 0.161 1.091 0.163 1.089 0.160 1.113 0.174 1.111 0.175 1.108 0.172 1.139 0.183 1.142 0.183 1.139 0.184 1.252 0.183 1.165 0.183 1.164 0.184 1.284 0.199 1.188 0.179 1.191 0.178 1.359 0.199 1.236 0.178 1.283 0.178 1.427 0.206 1.282 0.178 1.318 0.200 1.614 0.207 1.318 0.200 1.425 0.200 1.670 0.210 1.430 0.200 1.474 0.211 1.783 0.222 1.470 0.208 1.562 0.211 1.861 0.226 1.554 0.208 1.671 0.215 2.005 0.233 1.665 0.213 1.908 0.235 2.104 0.241 1.796 0.226 2.007 0.244 2.271 0.246 1.894 0.231 2.211 0.256 2.386 0.254 1.990 0.238 2.277 0.256 2.636 0.254 2.111 0.247 2.382 0.265 2.783 0.259 2.263 0.251 2.650 0.265 2.951 0.276 2.364 0.260 2.779 0.271 3.337 0.276 2.603 0.260 2.942 0.287 3.578 0.289 2.754 0.266 3.334 0.287 4.186 0.289 2.910 0.283 3.589 0.331 4.602 0.257 3.286 0.283 4.195 0.331 5.559 0.257 3.526 0.298 4.537 0.319 6.285 0.233 4.092 0.298 5.618 0.305 7.102 0.233 4.478 0.281 7.112 0.305 8.362 0.206 5.423 0.281 8.278 0.547 10.010 0.191 6.101 0.261 14.556 0.547 12.516 0.186 6.959 0.261 18.416 0.240 19.724 0.186 9.690 0.277 30.675 0.210 30.675 0.170 12.121 0.532 50.000 0.200 50.000 0.150 19.724 0.532 30.675 0.250 50.000 0.190 Page 34 of 43

-~ Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011 -ISRT-03 Uniform Hazard Spectra I GRNDo I I EW__ lI I __A I Table C2: IS Digitized UHS LLNS Factored of Design FRS E-W N-S Freq Accel Freq Accel 1.030 0.294 1.033 0.285 1.065 0.294 1.067 0.285 1.091 0.313 1.089 0.298 1.111 0.336 1.108 0.320 1.142 0.351 1.139 0.341 1.165 0.351 1.164 0.341 1.188 0.344 1.191 0.331 1.236 0.341 1.283 0.331 1.282 0.341 1.318 0.372 1.318 0.384 1.425 0.372 1.430 0.384 1.474 0.391 1.470 0.399 1.562 0.391 1.554 0.399 1.671 0.400 1.665 0.410 1.908 0.436 1.796 0.434 2.007 0.453 1.894 0.443 2.211 0.475 1.990 0.457 2.277 0.475 2.111 0.474 2.382 0.492 2.263 0.483 2.650 0.492 2.364 0.499 2.779 0.504 2.603 0.499 2.942 0.534 2.754 0.511 3.334 0.534 2.910 0.544 3.589 0.614 3.286 0.544 4.195 0.614 3.526 0.573 4.537 0.593 4.092 0.573 5.618 0.566 4.478 0.539 7.112 0.566 5.423 0.539 8.278 1.016 6.101 0.500 14.556 1.016 6.959 0.500 18.416 0.446 9.690 0.532 30.675 0.390 12.121 1.022 50.000 0.371 19.724 1.022 30.675 0.480 50.000 0.365 Page 35 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03 Uniform Hazard Spectra Attachment D In-structure Response Spectra for Vapor Containment Inner Structure Page 36 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03I Uniform Hazard Spectra D13.0 VC Inner Structure Evaluation The Vapor Containment (VC) Inner Structure (VC-IS) dynamic model is as summarized in [8]. The resulting design response spectra from [8] have been digitized as summarized in [9]. The dynamic model is depicted in Figure D1 and D2. The resulting dynamic response for the VC-IS is summarized in Figure D3 and D4.

Based on the structure response, and using the scaling approach outlined above in Section 4.0, the resulting amplification factors and peak of the In-structure response Spectra for the UHS were derived:

EL 94 ft Direction Hazard 1st Design Hazard Incoh Amp Peak Mode Freq GRS Response Factor Factor FRS (g)

(Hz)

N-S LLNL 17.130 0.188 0.355 0.900 1.699 1.225 E-W LLNL 32.840 0.167 0.283 0.900 1.524 0.443 HAZARD Resulting digitized values are listed in Table D1 for the design spectra, Table D2 for the hazard spectra. Applicable north-south and east-west spectra plots are also included, Figures D5 to D6.

Page 37 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011 -ISRT-03 Uniform Hazard Spectra C4N1AL.

I L UMPE A.ASS' *O. r T? 5

-.oJ

^.0T4.0J*- --AP~r1/-UAxv Figure D1: VC-IS Dynamic Model (N-S) [8]

.1 MA 7-62 ILUQMPEL. MASS .4C6L. OF 74

• A A.4 EA ST - WCE 5T 5rb~

Page 38 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 201 1-ISRT-03 Uniform Hazard Spectra Figure D2: VC-IS Dynamic Model E-W [8]

I qo"70~-TAI4 sJýfl C.UI 4~i5I1E CPot NT I.

0 a C0 0

~.515* - I ~a~)

.450 I.C:)

00(

• Il-i

-. ~ 11 ~2.O2. -.

.154-

'7 ~.00 Lo~~

- 311 - .2o2..

Figure D3: IS Dynamic Information (N-S) [8]

f C - . ... ..- . o.. ._-

4~

CS2. 0

.5 '414

i. 00.

1.00 I Oqq~4-Figure D4: IS Dynamic Information (E-W) [8]

Page 39 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 201 1-ISRT-03 Uniform Hazard Spectra VC Inner Containment Structure -

Ground Response 0.600 T 0.500 II OAOO-0.300 01.00 0.100 0.000 - I I 1.000 10.o00 100.000 Frequency (Hz)

-DBE 5% Damping - LLNL 10,000 yr UHS - USGS 10,000 yr UHS Figure D5: VC-IS Ground Response Page 40 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 201 1-ISRT-03 Uniform Hazard Spectra I VC Inner Containment Structure - EL 94 ft N-S Floor Response 1.400 1.200 --------- j I0.800 i 1.000 0.400 0.200 0.000 1.000 10.000 100.000 Frequency (Hz)

- DBE 5% DampingGrnd - DBE 94' 5% Damp N-S - LLNL94' UHS N-S Figure D6: VC-IS FRS VC Inner Containment Structure - EL 94 ft E-W Floor Response 0.500 --------- ---

• .*0.400 0300 0200 0.100 ................

1.000 10.000 100.000 Ftrequency(Hz)

- DBE 5% Damping Grnd -- DBE 5% Damping EL 94' E-W

-LLNS 94' UHS E-W Figure D7: VC-IS FRS Page 41 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 2011-ISRT-03 Uniform Hazard Spectra I Table DI: IS Digitized Design Spectra [9]

VC ICSSE5%Structural VCIC SSE 5% Structural VCICSSE 5% VCIC SSE5%

Damping Ground Damping Ground Structural Damping Structural Damping Elevation (from E-W) Elevation (from N-S) 94' Elevation E-W 94' Elevation N-S Freg Acce (g) Fre Accel (g) Freq Accel Fre Accel (g) 1.001 0.154 1.003 0.154 1.001 0.153 1.000 0.157 1.020 0.154 1.069 0.154 1.023 0.152 1.023 0.154 1.066 0.153 1.094 0.163 1.065 0.152 1.063 0.154 1.092 0.162 1.114 0.175 1.089 0.161 1,111 0.176 1.106 0.172 1.138 0.183 1.108 0.172 1.150 0.183 1.138 0.184 1.252 0.184 1.139 0.184 1.211 0.177 1.251 0.184 1.284 0.199 1.253 0,184 1.278 0.179 1.284 0.200 1.358 0.199 1.286 0.199 1.314 0.198 1.356 0.200 1.432 0.206 1.352 0.199 1.427 0.201 1.432 0.209 1.619 0.209 1.434 0.206 1.471 0.210 1.595 0.209 1.673 0.212 1.565 0.208 1.665 0.224 1.671 0.212 1.800 0.224 1.675 0.213 1.803 0.236 1.792 0.224 2.005 0.237 1.781 0.223 1.993 0.248 1.910 0.230 2.165 0.245 1.925 0.230 2.172 0.261 2.004 0.235 2.270 0.248 2.015 0.235 2.490 0.270 2.190 0.246 2.378 0.257 2.202 0.246 1 2.901 0.282 2.264 0.246 2.639 0.257 2.272 0.246 3.085 0.296 2.381 0.254 2.782 0.262 2.384 0.255 3.549 0.301 2.634 0.254 2.943 0.278 2.632 0.255 3.791 0.312 2.773 '0.260 3.337 0.278 2.787 0.261 4.021 0.305 2.938 0,276 3.597 0.291 2.951 0.277 4.924 0.296 3-337 0.276 4.191 0.291 3.337 0.277 5.992 0.289 3.570 0.289 4.568 0.259 3.584 0.291 8.006 0.289 4.153 0.289 5.543 0.259 4.174 0.291 9.671 0.300 4,545 0.258 6.258 0.234 4.566 0.260 12.195 0.721 5.546 0.258 7.112 0.234 5.556 0.260 19.724 0.721 6.227 0.234 8.197 0.210 6.262 0.237 30.675 0.307 7.087 0.234 9.009 0.200 7.052 0.237 50.000 0.220 8.137 0.210 10.000 0.192 8.382 0.211 9.940 0.192 19.724 0.187 12.547 0.205 _

12.195 0.187 30.675 0.179 19.724 0.205 19.724 0.187 50.000 0.150 30.675 0.170 _

30.675 0.170 _ _ 50.000 0.150 50.000 0.150 1 Page 42 of 43

Indian Point Energy Center Position Paper Indian Point 3 In-Structure Response Spectra for LLNL & USGS 201 1-ISRT-03 Uniform Hazard Spectra

" G RND E-W --

A G RND N-S --A I EL55 E -W --

A EL55 A N-S --

Table D2: IS Digitized UHS LLNS Factor of Design FRS E-W N-S Freq Accel Freq Accel 1.001 0.232704 1 0.26613 1.023 0.231332 1.023 0.26239 1.065 0.231332 1.063 0.26239 1.089 0.245353 1.111 0.29876 1.108 0.262268 1.15 0.31151 1.139 0.279946 1.211 0.30097 1.253 0.279946 1.278 0.30403 1.286 0.302805 1.314 0.337 1.352 0.302805 1.427 0.34158 1.434 0.314539 1.471 0.35654 1.565 0.317434 1.665 0.37999 1.675 0.32414 1.803 0.40089 1.781 0.340293 1.993 0.42163 1.925 0.350808 2.172 0.44423 2.015 0.358123 2.49 0.45919 2.202 0.374429 2.901 0.47975 2.272 0.374429 3.085 0.50337 2.384 0.388145 3.549 0.5117 2.632 0.388145 3.791 0.5309 2.787 0.398203 4.021 0.51765 2.951 0.422433 4.924 0.5032 3.337 0.422433 5.992 0.49147 3.584 0.443463 8.006 0.49147 4.174 0.443463 9.671 0.51 4.566 0.396679 12.195 1.22495 5.556 0.396679 19.724 1.22495 6.262 0.361628 30.675 0.52121 7.052 0.361628 50 0.37387 8.382 0.320939 12.547 0.311796 19.724 0.311796 30.675 0.259068 50 0.228589 Page 43 of 43

-Entergy ENG REPORT IP-IPT-11-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency Attachment 4 Review of the Concerns of the Unit I Stack

sin orIIndian Point SiteIndian I2011-ISRT-04 Point Energy Center Seismic Review of the Concern of the Unit 1 Stack I

Indian Point Site Seismic Review of the Concern of the Unit I Stack Prepared by: Richard S. Drake PE Reviewed by: Dragos Nuta, PE, William Henries, PE, and Paul Bruck 1.0 Purpose The purpose of this paper is to summarize the Design Basis condition, the concerns of the Unit 1 Stack on Indian Point Energy Center and its ability to withstand seismic events. The report will also discuss the Unit 1 Superheater Stack hazard assessment related to the SCDF along with its current condition and potential future modifications.

2.0 Description of the structure:

The Superheater Building and the Stack structure were originally designed to serve an oil-fired boiler for electric production for Unit 1 however, at present the stack serves only the house service boiler at the Indian Point Generating Station Unit 2 and the vent from Unit 1. The stack consists of a 254'-5" high riveted steel structure which is supported on the building superstructure. A 48" diameter vent from unit 1 is attached to the southeast side of the Stack exterior with support brackets approximately every 4 '. The Unit 1 vent stack enters into the main stack at the 228'-2" level and continues through the center of the main stack terminating at the 264'-5" level, 10'-0" higher than the exit elevation of the main stack. The riveted steel shell represents the structural component resisting wind loads, etc.; while the interior is protected with a 2-1/2" thick gunite lining designed to resist chemical attack from flue gases. Appurtenances on the stack include anchor bolts, an offset-caged ladder (running the full height), a service platform at the top, navigational obstruction lights and a ring angle cap. The Original boiler outlet ductwork has been removed and at present the only active component is the Unit 1 vent which enters on the east side, approximately 10' above the stack base.

2.1 Design Basis - Seismic and Tornado Design Summary The Superheater Stack is located on top of the Unit 1 Superheater Building. The failure of the stack could result in damage to safety related structures including the Unit 2 Control Building and the Emergency Diesel Generator Building. The Superheater Stack was shortened 80 feet to assure that the stack does not pose any danger to the nearby Seismic Category I structures and components. Evaluation of the shortened stack showed that the structure has enough capacity to resist the IP2 DBE and a 360 mph wind.

The Indian Point Unit 1 Superheater Stack has been analyzed for seismic, tornado, and vortex shedding wind load, Reference 4.0.1.

Page 1 of 4

Indian Point Energy Center 201 1-ISRT-04 Indian Point Site Seismic Review of the Concern of the Unit 1 Stack A spectrum response analysis was performed for the Superheater building considering the design basis earthquake (DBE - Also commonly referred to as the SSE or Safe Shutdown Earthquake),

which has a maximum horizontal ground motion of 0.15g. A damping coefficient equal to seven percent was assumed for all modes.

Tornado loads were based on a 360-mph wind using the shape factors for a rectangular building as defined in ASCE Paper 3269. It was assumed that 20-percent of the wall area of the building was still intact as a reaction surface for the wind in addition to the total surface area of major equipment and the stack at its existing height. On the basis of this analysis, the building has approximately the same resistance capacity to a 360-mph tornado wind as it does for the 0.1 5g earthquake.

Unit 1 was retired and the Superheater and associated equipment have been removed from certain areas of the Superheater Building and the areas refurbished to provide permanent administrative facilities. These areas do not contain any safety-related equipment. The total loading on the Superheater building has been reduced from the original design loading due to the removal of Superheater-associated equipment. Therefore, the administrative facilities will not adversely affect the response of the Superheater Building during a safe-shutdown earthquake.

2.2 Independent Safety Evaluation Concern During the Independent Safety Evaluation performed at IPEC in 2008 several concerns were raised about the Unit 1 Stack. The three main issues identified with the Unit 1 Stack were:

1. The Unit 1 Stack is exhibiting signs of age related degradation,

2. Nuclear, radiological and industrial safety concerns with the stack debris in the hopper.

3. The ISE recommended the removal of the stack before the license renewal period.

2.3 Inspections and reports:

The Superheater Building and the Superheater Stack are inspected periodically to ensure they are capable of maintaining their function. These inspections include the Maintenance Rule inspections as well as other periodic and special inspections. The Maintenance Rule inspections were last performed in 2004 (IP-RPT-05-00443) and 2007 (IP-RPT-07-00110, EC 28581). In addition, the Unit 1 SAFSTOR Team has performed periodic inspections of all Unit 1 structures in 2007 and 2009. The 2009 SAFSTOR Report is IP-RPT-09-00039.

Finally, as a result of the ISE report a special Inspection was performed of the Unit 1 Stack as a follow up to the last Stack inspection of 1995. International Chimney Corp, inspected the stack in 2009 (EC19587) and ABS Consulting performed a structural assessment in 2008, Reference 4.0.8.

This assessment was expanded by Lucius Pitkin Inc (LPI) who performed a Fitness-for-Service inspection of the Stack in 2010 and prepared estimates for the repair or dismantlement of the Stack.

The SAFSTOR & National Chimney inspections identified corrosion and degradation of the Superheater Building support steel. The latest inspection stated that the Stack is structurally acceptable based on the visual and UT inspections performed. The Structural stack inspections have noted continuing degradation and did recommend painting and maintenance repairs.

• Vent stack base structural anchorage and support framing is corroding Page 2 of 4

Indian Point Energy Center Position Paper 201 1-ISRT-04 Indian Point Site Seismic Review of the Concern of the Unit 1 Stack

• Ash hopper is full of acidic, contaminated ash, and stack liner material shows signs of through wall corrosion.

" Horizontal and vertical cross-bracings have degraded.

• National Chimney notes partial collapse of the inner stack (i.e. Unit 1 RCA vent).

" Industrial Safety concern due to rusting of handrails on the top platforms and Superheater catwalks.

The other concerns with the stack and its degradation are:

• Given the seismic fragility for the stack is below 0.30g, the Seismic Margin Review Earthquake (SME) for IP2, the stack must be postulated to fail.

• Failure of the stack adversely affects safety related structures, including the Control Building and Emergency Diesel Generator Building.

• As a result, the Unit 1 Stack is a large contributor to the seismic core damage frequency for IP2.

3.0 Conclusions The Unit 1 Superheater Stack and supporting steel work have been inspected and found to be structurally acceptable at this time but requires maintenance, industrial safety repairs, and re-painting to provide continued corrosion protection.

The vent line on the stack itself is in satisfactory condition throughout the entire height from elevation 135'-7" to 228' where it enters the main stack. The vent line inside the main stack is in poor condition and the protective coating has deteriorated exposing 90% of the steel shell. The vent line has surface rust and the stay rods supporting the vent stack are delaminating. Most of the Appurtenances on the Unit 1 Stack are satisfactory but have surface rust. The exceptions to this are the top service platform that has a handrail that is in poor condition with the vertical posts almost fully deteriorated at their connection points. The Ash Hopper is full of ash and gunite debris to the top. The hopper has considerable deterioration which will require repair or removal of the hopper. The gunite lining is in poor condition throughout the majority of the stack and needs to have the loose material removed. The Unit 1 Stack coating has areas that are blistering and have rust staining. The Unit 1 Stack needs to be re-painted.

Management has reviewed the cost of performing the required repairs and repainting versus the cost of full or partial dismantling of the stack down to the 167' elevation. Partial removal would allow not having to modify the service boiler inlets. The stack is currently scheduled to be removed in 2012 but the final decision is still under management review.

4. 0

References:

1. Unit 2 and Unit 3 FSAR, Final Safety Analysis Reports.

2. Unit 2 and Unit 3 TS, Technical Specifications.

Page 3 of 4

Indian Point Energy Center SPosition Paper 201 1-ISRT-04 Indian Point Site Seismic Review of the Concern of the Unit 1 Stack

3. Unit 2 Design Basis Document, Seismic Structures and Devices, SSD DBD, section 4.3.2

4. Unit 1 Superheater Building Maintenance Rule inspection Report 2004, IP-RPT-05-00443

5. Unit 1 Superheater Building Maintenance Rule inspection Report 2007, IP-RPT-07-001 10, EC 28581

6. Unit 1 SAFSTOR Team inspection report IP-RPT-09-00039.

7. International Chimney Corp, inspection report May 2009, ICC-CE-39094 (IP-RPT-10-00006, EC19587)

8. ABS Superheater Assessment December 2008, IP-RPT-1 0-00007.

9. International Chimney Corp, inspection report, 1995. (IP-RPT-05-00212)

10. Seismic and Tornado analysis of the Superheater and Stack for Indian Point Unit 1, IP-RPT-04-0021 0.

11.REQUEST FOR QUOTES FOR REPAIR OR DEMOLITION OF THE IPEC UNIT 1 VENT STACK, prepared by Lucuis Pitkin Inc., REPORT No. A10337-R-001.

Page 4 of 4

Alk I-ýEnter&y ENG REPORT IP-RPT-I 1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency Attachment 5 Develop Methodology that emulates NRC's use of IPEEE Parameters to Calculate Seismic Core Damage Frequency (SCDF) Values

Indian Point Energy Center Position Paper I Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05I to Calculate Seismic Core Damage Frequency (SCDF) Values Develop Methodology that Emulates NRC's use of IPEEE Parameters to Calculate Seismic Core Damage Frequency (SCDF) Values Prepared by: Dragos Nuta, PE Reviewed by: Paul Baughman EXECUTIVE

SUMMARY

• Full understanding of methodology used by NRC to calculate SCDF values was achieved.

• The NRC Report results were duplicated, validating our understanding.

" The methodology is used to develop more realistic estimates of risk In developing the SCDF estimates, the NRC report (Ref. 1) indicates that the IP 3 Seismic PRA (SPRA) was used to establish plant-level fragility curves parameters. These parameters were then convolved with the hazard curves derived by United States Geological Survey (USGS) for the IPEC coordinates, and resulted in the estimates of the SCDFs based on the peak ground acceleration (PGA) hazard curve, as well as the 10 Hz, 5 Hz, and 1 Hz hazard curves.

In trying to identify the areas of conservatism that resulted in the high NRC calculated estimates for the SCDF, a logical plan of action would have been to revise the fragilities of structures or components shown to be high contributors to risk, and then substitute these values in the SPRA model and rerun the analyses to derived more realistic estimates for the IP 3 SCDF. However, the SPRA model is not functional, and revised fragilities cannot be reflected in new analyses to establish a more realistic SCDF.

As a result, it was decided that the only way available for developing more realistic seismic hazard estimates is to fully understand the basis for the NRC developed parameters and methodology used, after which the NRC SCDF calculation methodology would be used to reflect the revised fragilities and calculate more realistic SCDF estimates. Since we did not have a clear understanding of the various parameters used by NRC in their work, our understanding of the NRC methodology had to be verified by duplicating the results obtained by the NRC.

Activities undertaken to achieve verification of the NRC SCDF estimates included:

1. Document input information referred to in the NRC report (Ref. 1)

2. Extract all information from the NRC report describing how various parameters were established.

3. Document the methodology used to calculate the SCDF estimates.

4. Obtain USGS Hazard Curves (which were used in the NRC Report)

5. Derive parameters and develop mathematical formulation for all terms used in the SCDF calculations.

6. Use the methodology and derived parameters and perform risk analyses that duplicate the NRC Report SCDFs for IP 3.

Page 1 of 22

Indian Point Energy Center Position Paper I Develop Methodology that Emulates NRC' use of IPEEE Parameters 201 1-SRT-05 to Calculate Seismic Core Damage Frequency (SCDF) Values A discussion of the work performed in support of all activities described above and the results obtained is provided below.

1. Document input information referred to in the NRC report (Ref. 1)

Page C-5 of the NRC Report (Ref. 1) states that the C50 and Pc parameters were obtained from the SPRA probability plot of the reported plant-level fragility data.

Table C.1. Bases for Establishing Plant-Level Fragility Curves Parameters From IPEEE Information.

Basis% Source. Parameters*

la SPRA C5 and P3 c determined by probability plot of the reported Applicability of the l a basis is provided in Table C-2 from which an excerpt is provided below.

Indeed, the IP 3 IEEE Report (Ref. 2) developed a plant level fragility (that reflects not only the seismic robustness of the IP 3 Structures, Systems, and Components (SSCs), but also the effects of random and other non-seismic related failures. The Mean Capacity at the plant level was given as 0.34g.

The NRC work used this 0.34 g median capacity, together with a composite coefficient Pc, representing randomness and uncertainty, of 0.34 in their risk assessment work.

The parameters are introduced in Table C-2 (Page C-7) of the NRC report, as extracted below:

Table C-2. Plant4.Level Fragility Data.

Docket PGA Fragility Spectral Ratios Numbe HCL C5 P3c 10 5 1 Bas Plant r IPEEE Method PF *_Hz Hz Hz is Indian Point 2 050002 seismic PRA 0.6 0.4 1.6 1-2 0.4 lb

, 47 , _ _ j8 2 3 1 Indian Point 3. .050002 seismic PRA 0.3 0.3 1.5 1.6 0.8 Ia 86 4 4 6 1 1 It should be noted that the C 50 and P~c parameters, as well as the Spectral Ratios will be discussed below.

2. Extract all information from the NRC report describing how various parameters were established.

In using plant level fragilities based on the IPEEE Review Level Earthquake (RLE), which was based on the Lawrence Livermore National Laboratories (LLNL) hazard data and ground response spectra, adjustments to the fragilities had to be made so that they may be used in conjunction with the hazard curves, and ground spectra, based on 2008 USGS data.

Page 2 of 22

Indian Point Energy Center Position Paper Develop Methodology that Emulates NRC' use of ,PEEE Parameters 2011-ISRT-05 to Calculate Seismic Core Damage Frequency (SCDF) Values The NRC captures the adjustments in Figure C-I (Page C-4) of the NRC Report (Ref. 1) as follows:

MSCSO in a m10CSO specir.al shape used InSPRA or SMA defirkes the.spectral ratios iN, 1

m10 HCLPF H.CLPF mHCPF ,,,ýHCLPF +-.P(Wliure);;0.01 1HKZ S.Hz 10 Hz Pga i Definition of Spectral Fragility Terms of the PGA-Based Fragility and C-I.

FigureThe in Figure C-1. The Definition of Spectral Fragil Iity In Terms of the PGA-Based Fragility and the Review-Level Response Spectrum.

The median seismic capacity C50 is calculated as follows:

The High Confidence of a Low Probability of Failure (HCLPF) based on the work performed using the LLNL data, given the 0.34g median capacity (Am) and the 0.34 composite coefficient, is:

HCLPF = Am e -(2.3264 x Pc) = 0.34g x e -(2.3264 x 0.34) 0. 15g The HCLPF is the ground acceleration that corresponds to a 1% probability of core damage. The HCLPF is related to the median seismic capacity C50 to be used in the work reflecting the USGS hazard curves as follows:

C50 = HCLPF e (2.3264 x 13c) = 0.15g x 2.2339 z 0.34g.

The ml, Ms, and mlo spectral ratios were obtained from the LLNL based RLE as the ratio between the spectral amplitudes at 1 Hz, 5 Hz, and 10 Hz, and the peak ground acceleration for the spectrum.

The calculated spectral ratios are provided in Table C-2 from which excerpts were provided above.

The IP 3 spectral ratios for 1 Hz, 5 Hz, and 10 Hz are given as 0.81, 1.61, and 1.56, respectively.

Page 3 of 22

Indian Point Energy Center Position Paper Develop Methodology that-Emulates NRC' use of IPEFE Parameters 2011-ISRT-05 to Calculate Seismic Core Damage Frequency (SCDF) Values

3. Document the methodology used to calculate the SCDF estimates.

Appendix A of the NRC Report (Ref.1 ) contains an extensive explanation as to how the SCDF is calculated using the Hazard Curve for a certain frequency and the fragilities for a certain SSC or, in our case, the plant level fragility (consisting of the median capacity and composite uncertainty). The mathematical formulations, as shown below in the excerpt from Appendix A:

aj dPct0 (a)

SCDF(ai,ai-1) f H (a) daHj(a)PcD(a) (A-1 2)

I da I is solved numerically.

To allow an understanding of exactly what was done as part of this task that validated our understanding of the NRC parameters used and methodology by duplicating the SCDF estimates published by NRC for Indian Point 3, we present a plain language description of the steps involved in estimating a SCDF.

The steps involved in obtaining a seismic core damage frequencyestimate are as follows:

a) Obtain the hazard curves for various frequencies of interest, including the PGA.

b) Tabulated values of accelerations and their annual exceedance frequency are usually provided so data needs not be read off of the Hazard Curve. The data is arranged in order of increasing accelerations.

c) Between each ascending pair of accelerations (interval), establish the interval acceleration and interval frequency as described below.

d) Calculate the interval z-value for a normal distribution, given the interval acceleration, the median capacity of the item for which the SCDF is calculated, and the composite P3c, which represents the randomness and uncertainty in the median capacity, as described below.

e) Calculate the conditional probability of failure f) Convolve the interval frequency with the conditional probability of failure and obtain the probability of failure for each interval.

g) The summation of the probabilities of failure for all acceleration intervals represents the total probability of failure or the SCDF for the item having a median capacity Am and the associated Pe, when the hazard curve for the selected frequency is considered.

Page 4 of 22

Indian Point Energy Center Position Paper Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05 to Calculate Seismic Core Damage Frequency (SCDF) Values The steps discussed above under b) through g) were incorporated into an EXCEL spreadsheet as shown below:

A B C D E F G H Hazard Curve Freg. Interval Interval Interval Z Conditional Probability Probability Acceleration Acceleration (a) Frequency ln(a/Am)/13c of Failure of Failure (A_ i+Ai+1 )/2 (Bi-Bi+l) NORMDIST Z = (Fi) ExGj where "i" is the row number under consideration A discussion of the various steps, with headings captured under Columns A through H in the EXCEL spreadsheet, is presented below.

a) Obtain the hazard curves for various frequencies of interest, including the PGA.

Hazard Curves are developed by ground motion experts. After establishing all seismic zones of known seismic activity that could affect a particular location, each seismic zone is associated with a magnitude M that may occur at the specific seismic zone, including its recurrence interval, the distance from the seismic zone to the site where the Hazard Curves must be established, as well as other parameters that deal with variability and wave propagation.

From each seismic zone, wave propagation formulations are used to propagate seismic waves of various frequencies to the site where the Hazard Curves are established. The frequencies considered include, as a minimum, 1 Hz, 5 Hz, 10 Hz, and PGA. Attachment 1 contains an example of the seismic zones considered by USGS for CEUS.

As mentioned above, Hazard Curves are calculated for various frequencies. An excerpt from the NRC report (Ref. 1) shows PGA and 1 Hz Hazard Curve plots:

-L g(1N3-PGA USM. jzoos~.PGA Io1E424sA-I .-..

I-OE.03 0.01 0-1 1 to The X-axis contains acceleration levels, while the Y-axis contains annual exceedance probabilities, or annual exceedance frequencies. A point on a Hazard Curve is associated with an acceleration and an annual exceedance probability.

Page 5 of 22

Indian Point Energy Center Position Paper Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05 to Calculate Seismic Core Damage Frequency (SCDF) Values While the pairs of acceleration and exceedance probability may be read off of the Hazard Curves, a tabulation of accelerations and their annual exceedance probabilities is usually provided and contains a minimum of 10-12 pairs of values. Given the summation process used to emulate an integral, the more points considered, the more accurate the representation of the Hazard Curve in calculating the SCDF.

c) For every interval, establish the interval acceleration and frequency.

Given a tabulation that has acceleration values increasing in magnitude, and corresponding annual exceedance frequencies that decrease in magnitude, the interval from one acceleration value to the one immediately below, has an interval acceleration and frequency established as follows:

• The interval acceleration is the average of the top acceleration and the one immediately below. In the tabulation excerpted above, having columns A through H, the interval acceleration for the ith row is (Ai + A i+l) / 2. The result is placed in Column D of the excerpt.

• The interval frequency is calculated as the difference between the annual exceedance frequencies for two accelerations. In the tabulation above, the frequency for the ith row is the difference between the exceedance probability shown on row "i" and row "i + 1". This value is entered in Column E of the excerpted tabulation.

d) Calculate the z-value for the acceleration interval The z-value is associated with a normal distribution that has the mean value of zero and an area under the distribution curve of 1.0. It is calculated as Z = (In (a/Am)) / Pc where

" a = acceleration value for the interval. It is captured under Column D of the excerpted EXCEL spreadsheet above.

• Am is the median capacity of the SSC considered or, in our case, the point estimate fragility developed for the plant. When the plant Am value established using the LLNL Hazard Curves is used in conjunction with Hazard Curves developed by USGS, the median capacity for the plant is introduced as C 50 when the PGA Hazard curve is considered, and mi C 50 when the frequency "i" Hazard Curve is considered. Both C50 and the "m" factors were discussed under 2.

above.

• f3. is the composite uncertainty calculated from the randomness and uncertainty parameters.

Page 6 of 22

Indian Point Energy Center Position Paper Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05I to Calculate Seismic Core Damage Frequency (SCDF) Values For normal distributions, the z-value represents the number of standard deviations away from the mean (which is the center of the symmetrical distribution curve).

e) Calculate the conditional probability of failure.

The conditional probability of failure for a normal distribution is the area from the left tail of the distribution to the calculated z-value. For these calculations, the conditional probability of failure was calculated using the NORMDIST Z function in EXCEL and the z-value calculated as discussed above (and present in Column F of the tabulated excerpt).

The conditional probability of failure may be calculated by hand using tabulations of Z values and probabilities (area under the curve), remembering that the given area is always given from the mean to the absolute value of "z". Thus, probability for a negative z-value (z-value is to the left of the mean) will be obtained by reading the area (probability) from the normal distribution table, and subtracting from 0.5 (which is the area for each half of the normal distribution plot) to obtain the area from the left tail of the distribution to the negative z-value. The probability for a positive z-value (z-value is to the right of the mean) will be obtained by reading the area (probability) from the normal distribution table, and adding it to 0.5 (which is the area for each half of the normal distribution plot) to obtain the area from the left tail of the distribution to the positive z-value.

d) Multiply the interval (annual exceedance) frequency and the conditional probability of failure to obtain the probability of failure for each interval.

Values captured under Columns E and G of the excerpted tabulation, respectively for each acceleration interval are multiplied and the result are entered under Column H of the excerpted tabulation.

e) Calculate the total probability of failure (SCDF)

The total probability of failure for an SSC having a median capacity Am and P3c composite uncertainty, when considering a Hazard Curve representing a specific frequency, is obtained by adding the probabilities of failure for all acceleration intervals used to discretize the Hazard Curve.

The values of probability of failure in Column H were summed to provide the Total Probability of Failure. When the IP 3 point estimate fragilities are used, as discussed under d) above, the summation represents the SCDF.

As discussed below, the NRC Report (Ref. 1) provides SCDF using the 1 Hz, 5 Hz, 10 Hz, and PGA USGS Hazard Curves. The NRC reported SCDF results were duplicated using the methodology described above, as described under 6, below.

4. Obtain USGS Hazard Curves (which were used in the NRC Report)

Hazard curves for various wave frequencies are obtained from USGS developed module 2008.

CEUS.OplO.2000.txt.gz by selecting the coordinates for a specific location.

As an example, using the Indian Point Coordinates of 41.2691 latitude and -73.9527 longitude, the information for the 10 Hz hazard curve, is obtained as follows:

Page 7 of 22

Indian Point Energy Center I Position Paper Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05 to Calculate Seismic Core Damage Frequency (SCDF) Values Ampis--> 2.50E-03 6.OOE-03 9.80E-03 1.37E-02 1.92E-02 2.69E-02 3.76E-02 41.3 -74 3.32E-02 1.64E-02 1.08E-02 8.04E-03 5.88E-03 4.23E-03 2.99E-03 41.3 -73.9 3.31E-02 1.64E-02 1.08E-02 8.02E-03 5.87E-03 4.22E-03 2.99E-03 41.2 -74 3.23E-02 1.61E-02 1.07E-02 8.01E-03 5.92E-03 4.31E-03 3.09E-03 41.2 -73.9 3.22E-02 1.61E-02 1.07E-02 7.99E-03 5.90E-03 4.30E-03 3.08E-03 The first two columns provide the latitude and longitude for the Hazard Curves. Top row represents the acceleration value and, for each of the four locations, the exceedance probability is provided for each acceleration. The tabulation extends into higher acceleration levels, and lower exceedance probabilities, but was truncated here at an acceleration of 0.0376g, as the tabulation is too long.

USGS Hazard Curves used for the work discussed herein, were obtained from the NRC and are as follows:

PGA 10 Hz 5 Hz 1 Hz Annual Annual Annual Annual exceedance exceedance exceedance exceedance Acceleration frequency Acceleration frequency Acceleration frequency Acceleration frequency W's) (per year) W's) (per year) W's) (per year) Ws) (per year) 0.007 8.02E-03 0.0075 1.20E-02 0.0075 1.47E-02 0.0037 1.27E-02 0.0098 5.72E-03 0.0113 8.43E-03 0.0113 9.73E-03 0.0056 8.13E-03 0.0137 4.04E-03 0.0169 5.86E-03 0.0169 6.37E-03 0.0084 4.98E-03 0.0192 2.82E-03 0.0253 3.98E-03 0.0253 4.07E-03 0.0127 2.89E-03 0.0269 1.95E-03 0.038 2.62E-03 0.038 2.52E-03 0.019 1.63E-03 0.0376 1.34E-03 0.057 1.69E-03 0.057 1.53E-03 0.0285 8.78E-04 0.0527 9.06E-04 0.0854 1.06E-03 0.0854 9.08E-04 0.0427 4.61 E-04 0.0738 6.09E-04 0.128 6.49E-04 0.128 5.26E-04 0.0641 2.37E-04 0.103 4.07E-04 0.192 3.87E-04 0.192 2.97E-04 0.0961 1.21E-04 0.145 2.65E-04 0.288 2.23E-04 0.288 1.62E-04 0.144 6.10E-05 0.203 1.71E-04 0.39 1.44E-04 0.39 1.01E-04 0.216 3.04E-05 0.284 1.07E-04 0.432 1.24E-04 0.432 8.52E-05 0.324 1.47E-05 0.397 6.46E-05 0.649 6.45E-05 0.649 4.23E-05 0.4 9.79E-06 0.556 3.69E-05 0.8 4.49E-05 0.8 2.87E-05 0.487 6.57E-06 0.778 1.97E-05 0.973 3.13E-05 0.973 1.95E-05 0.73 2.68E-06 0.9 1.46E-05 1.46 1.37E-05 1.46 8.09E-06 0.9 1.61 E-06 1.09 9.58E-06 2.19 5.12E-06 2.19 2.91E-06 1.09 9.81E-07 1.52 4.05E-06 3.28 1.42E-06 3.28 8.50E-07 1.64 3.08E-07 2.13 1.24E-06 5 1.42E-07 4.92 1.38E-07 2.46 8.23E-08 3.69 1.60E-08 As discussed under 6. below, the USGS Hazard Curves, which were plotted in Attachment 2, will be further discretized to provide better estimates of the SCDFs that may be compared to those presented in the NRC Report (Ref. 1).

5. Derive parameters and develop mathematical formulation for all terms used in the SCDF calculations.

Parameters and mathematical formulations were discussed under Item 3. above, as the methodology used to calculate the SCDF was documented. Below, we cover all these parameters and formulations and provide, when appropriate, quantifications for use in the SCDF analyses for comparison against the NRC results.

Page 8 of 22

Indian Point Energy Center IPosition Paper IDevelopCalculate Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05 to Seismic Core Damage Frequency (SCDF) Values As mentioned above, the methodology used to calculate the SCDF was incorporated into an EXCEL spreadsheet that contains, in Columns A through H, the parameters and formulations used. It is duplicated below:

A B C D E F G H Annual Hazard Curve Exceedance Interval Interval Interval Z Conditional Probability Probability Acceleration Freq Acceleration (a) Frequency ln(a/Am)/Pc of Failure of Failure a) Hazard Curve acceleration and annual exceedance frequency - The tabulation above contains in Columns A and B pairs of acceleration values and their corresponding annual exceedance frequency. While a "sample" run using this "raw" data will be presented and discussed below, the curves will be discretized further when calculating SCDFs for comparison with the NRC calculated values.

b) Acceleration interval, the interval acceleration (a) and the interval frequency are calculated values using the Hazard Curve information, as discussed under 3. above.

c) Z-Value - This is a calculated value based on the quantification of the following parameters:

• The interval acceleration (a) - This value is calculated as discussed under 3.

• The Median Capacity (Am) - When performing analyses using the USGS Hazard Curves and USGS spectra, the median capacities calculated using the LLNL ground spectra must be modified. As discussed under 2. above, the median capacities to be used for the 1 Hz, 5 Hz, 10 Hz, and PGA SCDF analyses are expressed in terms of C5 0 as follows:

Calculated al Hazard Curve Median C Frequency Spectral Ratio Capacity Capacity Formula Cpi (Hz) (g)

PGA I I x C50 0.34 1 Hz 0.81 0.81 x C50 0.2754 5 Hz 1.61 1.61 x C50 0.5474 10 Hz 1.56 1.56 x C5o 0.5304 The Composite Uncertainty coefficient (0c) - Pc is taken as 0.34, as discussed under 2.

above.

d) Conditional Probability of Failure and Probability of Failure - These values are calculated for each of the acceleration intervals (into which the Hazard Curve was discretized), and are contained under Columns G and H of the EXCEL spreadsheet header shown above.

e) Total Probability of Failure (SCDF) - This is a calculated value, equal to the summation of the probabilities of failure calculated for each acceleration interval.

An example of the use of these parameters and formulations for calculating the SCDF using the raw 1 Hz USGS Hazard Curve presented above is presented below:

Page 9 of 22

Indian Point Energy Center Position Paper Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05I to Calculate Seismic Core Damage Frequency (SCDF) Values Am - 0.34 x 0.81 0C f IP3 NRC Probability of Failure Solution 0.2754 0.34 1Hz A B C D E F G H Annual Hazard Curve Exceedance Interval Interval Interval Z Conditional Probability Probability Acceleration Freq. Acceleration (a) Frequency ln(a/Am)/13c of Failure of Failure (A12 + A13)/2 (B12 - B13) NORMDIST Z = (F12) E12xG12 0.0037 1.27E-02 .0037 to .0056 0.00465 0.0045903 -12.00399225 1.69281E-33 7.77052E-36 0.0056 8.13E-03 .0056 to .0084 0.007 0.0031563 -10.80092482 1.70377E-27 5.37761E-30 0.0084 4.98E-03 .0084 to .0127 0.01055 0.0020823 -9.594408019 4.22001E-22 8.78733E-25 0.0127 2.89E-03 .0127 to .019 0.01585 0.0012673 -8.397220841 2.28585E-17 2.89686E-20 0.019 - 1.63E-03 .019 to .0285 0.02375 0.00074781 -7.207770752 2.84376E-13 2.12659E-16 0.0285 8.78E-04 .0285 to .0427 0.0356 0.00041653 -6.017291025 8.868E-10 3.69379E-13 0.0427 4.61 E-04 .0427 to .0641 0.0534 0.00022454 -4.824746589 7.00907E-07 1.57382E-10 0.0641 2.37E-04 .0641 to .0961 0.0801 0.00011638 -3.632202154 0.000140506 1.63521E-08 0.0961 1.21E-04 .0961 to .144 0.12005 0.000059537 -2.442106658 0.007300916 4.34675E-07 0.144 6.10E-05 .144 to .216 0.18 0.000030573 -1.250787457 0.105506016 3.22564E-06 0.216 3.04E-05 .216 to .324 0.27 0.000015771 -0.058243021 0.476777526 7.51926E-06 0.324 1.47E-05 .324 to .4 0.362 4.8736E-06 0.804175369 0.789352149 3.84699E-06 0.4 9.79E-06 .4 to .487 0.4435 3.2154E-06 1.40139181 0.91945153 2.9564E-06 0.487 6.57E-06 .487 to .73 0.6085 3.8906E-06 2.331683312 0.990141321 3.85224E-06 0.73 2.68E-06 .73to .9 0.815 1.07E-06 3.191069197 0.999291263 1.06914E-06 0.9 1.61E-06 .9 to 1.09 0.995 6.29E-07 3.777994561 0.999920952 6.2858E-07 1.09 9.81E-07 1.09 to 1.64 1.365 6.73E-07 4.707897416 0.999998749 6.72639E-07 1.64 3.08E-07 1.64 to2.46 2.05 2.26E-07 5.904030841 0.999999998 2.25942E-07 2.46 8.23E-08 2.46 to 3.69 3.075 6.63E-08 7.096575276 1 6.6337E-08 3.69 1.60E-08 3.69 3.69 1.60E-08 7.632815149 1 1.5951E-08 Total Prob of Failure 2.45303E-05 A discussion of the plant level SCDF obtained using the 1 Hz USGS seismic Hazard Curve is presented hereafter, together with a "manual" check of a few of the calculated conditional probabilities Page 10 of 22

Indian Point Energy Center I Position Paper 2011-ISRT-05 I Develop Methodology that Emulates NRC' use of IPEEE Parameters to Calculate Seismic Core Damage Frequency (SCDF) Values I The upper right hand of the tabulation indicates for the 1 Hz analysis, the median capacity used was Am = 0.2754g and the composite Oc was 0.34.

Columns A and B, containing the Hazard Curve Accelerations and Annual Exceedance Frequency are identical to the values presented here under 4. above.

Given the Z value, EXCEL calculated the Conditional Probabilities of failure, tabulated in Column G, using the NORMDIST Z function. A verification of two of the entries is presented below.

Where the table entries represent the area under the curve from the mean to the Z value:

Sauw= R~ediickMomuwsmd Robwt &.c ~wkt, S~wdySwUiksu, Table A-il (Readin&Mn~4L: AddisoWmey, 19?3)Rapinlodvitb 0 Z For the Z value of 0.80, which is entered for the row showing the acceleration of 0.324g, the area in the table above is 0.2881. Therefore, the conditional probability of failure, which is the area starting from the left tail of the normal distribution to the Z-value, is 0.5, the area under the left half of the distribution plus the tabulated value of 0.2881 which equals 0.7881. This compares favorably with the NORMDIST Z value of 0.789352149.

For the Z value of 1.40, shown in the row for acceleration 0.4g, the area from the Normal Distribution Table above is shown as 0.4192. Thus, the conditional probability of failure, (i.e.,

the area under the normal distribution curve from the left tail to the Z-value) is 0.5 + 0.4192 =

0.9192, which compares favorably to the NORMDIST Z value calculated as 0.91945153.

Page 11 of 22

Indian Point Energy Center I Position Paper Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05 to Calculate Seismic Core Damage Frequency (SCDF) Values On a different note, the summation of all Probability of Failure entries calculated under Column H for all accelerations under Column A., or 2.45E-05, represents the Total Probability of Failure, or SCDF, for Indian Point No. 3 when considering the USGS 1 Hz Hazard Curve. Given the limited number of discretization points for the Hazard Curve, the value differs from the SCDF calculated in the NRC Report (Ref 1) which, as shown in the excerpt below, is 2.3E-05.

Table D-1. Seismic Care-Damage Frequencies Using 2008 USGS Seismic Hazard Curves, Updated USGS IPEEE. weakest Docket controtting simple '"weighted link Plant Name Number PGA 10 HZ 5 Hz 1 Hz max curve average average model Indian Point 3 05000286 916E-05 I.OE-04 6.6E-05 2.3E.05 _10E-04 10 Hz 7,0E-05 8TE-05 *1.0E.04

6. Use the methodology and derived parameters and perform risk analyses that duplicate the NRC Report SCDFs for IP 3.

The methodology developed for estimating SCDF given Hazard Curves for different frequencies and the parameters derived under 5. above, was used in conjunction with the USGS Hazard Curves for which the acceleration interval was refined using a log-log interpolation between the raw acceleration data points for a 0.001g interval. Excerpts from the SCDF analyses for the PGA, 10 Hz, 5 Hz, and 1Hz Hazard curves showing the 0.001g discretization and the Total Probability of Failure for each Hazard Curve are presented in Attachment 3.

A summary of the results obtained using the methodology described above, against those reported in the NRC Report shows full agreement:

Indian Point No. 3 Comparison of Seismic CDF Results with NRC Estimates Hazard Median Seismic CDF Curve Capacity NRC Simplified Frequency (g) (Table D-1) Methodology PGA 0.34 9.1E-05 9.1E-05 10Hz 0.5304 1.OE-04 1.OE-04 5Hz 0.5474 6.6E-05 6.6E-05 1Hz 0.2754 2.3E-05 2.3E-05 Excerpts from the EXCEL analyses using the USGS Hazard Curves for PGA, 10 Hz, 5 Hz, and 1 Hz are provided in Attachment 3. Due to the rather small acceleration interval of 0.001g, the number of rows is excessive and, as such, only a limited number of acceleration values are excerpted.

Page 12 of 22

Indian Point Energy Center IPosi tion Paper IDevelop 2011-ISRT-05 Methodology that Emulates NRC' use of IPEEE Parameters to Calculate Seismic Core Damage Frequency (SCDF) Values The heading of the tabulations list the frequency of the Hazard Curve used (which was discretized to 0.00Ig intervals), the calculated SCDF for the specific Hazard Curve, and the plant level Median Capacity and P3c used.

References:

1. NRC, "Safety/Risk Assessment Results for Generic Issue (GI) 199, Implications of Updated Probabilistic Hazard Estimates in Central and Eastern United States on Existing Plants," dated September 2, 2010 (ADAMS Accession Number ML100270582)

2. Report IP3-RPT-UNSPEC-02182, "New York Power Authority, Indian Point Three Nuclear Power Plant, Individual Plant Examination of External Events," September 1997.

Page 13 of 22

Indian Point Energy Center Position Paper Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05 to Calculate Seismic Core Damage Frequency (SCDF) Values Attachment 1 Excerpt from USGS CEUS Seismic Zone Aggregation File "run.CEUS.2007"

1. regional hazard, CEUS, with updates to atten models for 2007

1. Also update to NMSZ model, JUNE 1 2007. Updates to Mmax late May, 2007.

1. Update to charleston broad zone Sept 2007

1. !/bin/csh echo This script runs 2007 CEUS hazard with the clustered src of NMSZ echo for gldscv or gldpnw set path = ( /export/harmsen/Hazard $path echo You can run specific subsets by making arg 1 charles narrow test or faults

1. map contrib separates into B & J in version 3 if ($1 charles ) goto charles if ($1 narrow ) goto narrow if ($1 == faults ) goto faults if ($1 NMSZ) goto NMSZ if ($1 == nmc ) goto nmc echo Gridded CEUS hazard with 4 Mmax branches for J and AB M-conversions.

1. On most computers each of these gridded haz runs takes several hrs, 3 to 4 hr on Solaris machines.

echo First 8 runs are for gridded hazard, standard mag corr.

hazgridXnga2 CEUS.2007.ABl.in > ceus.bl.log &

hazgridXnga2 CEUS.2007.AB2.in > ceus.b2.log hazgridXnga2 CEUS.2007.AB3.in > ceus.b3.log &

hazgridXnga2 CEUS.2007.AB4.in > ceus.b4.log hazgridXnga2 CEUS.2007.Jl.in > ceus.jl.log &

hazgridXnga2 CEUS.2007.J2.in > ceus.j2.log hazgridXnga2 CEUS.2007.J3.in > ceus.j3.log &

hazgridXnga2 CEUS.2007.J4.in > ceus.j4.log if ($1 == cy ) exit 0

1. mag cor = yes new terminology from Mueller nov 15 nmc:

echo Next 8 runs are for gridded hazard, no mag corr.

hazgridXnga2 CEUS.2007a.ABl.in > ceus.bln.log &

hazgridXnga2 CEUS.2007a.AB2.in > ceus.b2n.log hazgridXnga2 CEUS.2007a.AB3.in > ceus.b3n.log &

hazgridXnga2 CEUS.2007a.AB4.in > ceus.b4n.log hazgridXnga2 CEUS.2007a.Jl.in > ceus.jln.log &

hazgridXnga2 CEUS.2007a.J2.in > ceus.j2n.log hazgridXnga2 CEUS.2007a.J3.in > ceus.j3n.log &

hazgridXnga2 CEUS.2007a.J4.in > ceus.j4n.log if ( $1 == grid ) exit 0 echo completed ceus gridded hazard challenge, no mag cor if ( $1 == nmc ) exit 0 charles:

echo Next 8 runs are for Charleston South Carolina source zones echo Charleston broad extended further offshore Sept 2007 hazgridXnga2 CEUSchar.broad.in > char.broad.log &

hazgridXnga2 CEUScharA.broad.in> charA.broad.log hazgridXnga2 CEUScharB.broad.in>charB.broad.log &

hazgridXnga2 CEUScharC.broad.in> charC.broad.log Page 14 of 22

- Indian Point Energy Center Position Paper Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05 to Calculate Seismic Core Damage Frequency (SCDF) Values narrow:

hazgridXnga2 CEUScharn.in> char.narrow.log hazgridXnga2 CEUScharnA.in> charnA.narrow.log &

hazgridXnga2 CEUScharnB.in> charnB.narrow.log &

hazgridXnga2 CEUScharnC.in> charnC.narrow.log if ($1 == charles ) exit 0 echo Narrow zones concluding 2007 Charleston hazard runs if ($1 == narrow ) exit 0

1. fault runs. As of june 2007, using hazFXnga7c for psha analysis faults:

hazFXnga7c CEUScm.in > ch meers.log &

NMSZ:

echo Running modified 2002 unclustered NMSZ model with 5 virtual fault branches echo NMSZ is explicitly run with 500 and 1000 yr recur. intervals hazFXnga7c NMSZnocl.500yr.5branch.in > NMSZ5c.2007.log hazFXnga7c NMSZnocl.1000yr.5branch.in > NMSZlm.2007.log echo Next run NMSZ cluster models, 500 750 and 1500 year cluster recurrence time hazFXnga7c newmad.500.cluster.in > c500.log &

hazFXnga7c newmad.750.cluster.in > c750.log &

hazFXnga7c newmad.1000.cluster.in > cl000.1og &

hazFXnga7c newmad.1500.cluster.in > c1500.log &

exit 0 Page 15 of 22

Indian Point Energy Center Position Paper Develop Methodology that Emulates NRC' use of IPEEE Parameters 201 1-ISRT-05 to Calculate Seismic Core Damage Frequency (SCDF) Values Attachment 2 USGS Hazard Curves for Indian Point Site Page 16 of 22

Indian Point Energy Center Position Paper Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05I to Calculate Seismic Core Damage Frequency (SCDF) Values USGS Hazard Curves 1.80E-03 1.70E-03 1.60E-03 1.50E-03 1.40E-03 1.30E-03 1.20E-03 1.10E-03

--- PGA CL 1.OOE-03

---

10 Hz 9.OOE-04 5 Hz 8.OOE-04 1 Hz LU 7.OOE-04 6.OOE-04 5.OOE-04 4.OOE-04 3.OOE-04 2.OOE-04 1.OOE-04 O.OOE+O0 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 Acceleration (g)

Page 17 of 22

Indian Point Energy Center Position Paper Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05 to Calculate Seismic Core Damage Frequency (SCDF) Values I ATTACHMENT 3 Excerpts from SCDF Calculations for Various USGS Hazard Curve Frequencies Page 18 of 22

Indian Point Energy Center IPosition Paper I Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05 to Calculate Seismic Core Damage Frequency (SCDF) Values PGA USGS Hazard Curve 0.001g Intervals Total Seismic Core Damage Frequecy 9.iOE-05 Plant Median Capacity 0.34 Plant Composite Uncertainty, PC 0.34 Seismic Level (g) Frequency of Failure Delta Risk at Exceedance Probability g level 0.007 8.02E-03 1,67E-29 1.68E-32 0.008 7.01E-03 1.00E-27 7.85E-31 0.009 6.23E-03 3.40E-26 1.74E-29

.0...5.72E-03 .7,42E-25 4,78E-28 0.011 5.07E-03 1.13E-23 4.96E-27 0.012 4.64E-03 1.30E-22 4.79E-26 0.013 4.27E-03 1.17E-21 2.63E-25 0.014 4.04E-03 8.5 7E-21 3.20E-24 0.015 3.67E-03 5.30E-20 1.29E-23 0.016 3.42E-03 2.82E-19 6.06E-23 0.017 3.21E-03 1.32E-18 2.51E-22 0.018 3.02E-03 5.53E-18 1.11E-21 0.019 2.82E-03 2.10E-17 2.59E-21 0.020 2.70E-03 7.26E-17 1.02E-20 0.021 2.55E-03 2.33E-16 2.96E-20 0.022 2.43E-03 6.94E-16 8.02E-20 0.023 2.31E-03 1.94E-15 2.05E-19 0.024 2.21 E-03 5.13E-15 4.96E- 19 0.025 2.11E-03 1.28E-14 1.14E-18 0.026 2.02E-03 3.06E-14 2.27E- 18 0,027 1.95E-03 TOOEl-14 6.OOE- 18 0.028 1.86E-03 1.53E-13 1.11E-17 0.029 1.79E-03 3.24E- 13 2.17E- 17 0.030 1.72E-03 6.62E-13 4.13E-17 0.031 1.66E-03 1.31E-12 7.61E-17 0.032 1.60E-03 2.51E-12 1.37E-16 0.033 1.55E-03 4.68E-12 2.39E-16 0.034 1.50E-03 8.52E-12 4.09E-16 0.035 1.45E-03 1.5 IE-1I 6.83E-16 0.036 1.40E-03 2.63E-1I 1.12E-15 0.037 1.36E-03 4.46E-11 1.09E-15 0.038 1.34E-03, 7,43E- I 4.0,9E-15 0.039 1.28E-03 1.22E-10 4.47E- 15 Last rows of the EXCEL spreadsheet are below:

2.126 1.25E-06 1.00E+00 2.05784E-09 2.127 1.25E-06 1.00E+00 2.05349E-09 2.128 1.25E-06 1.00E+00 2.04915E-09 2.129 1.24E-06 1.OOE+00 2.04482E-09 2.130 ~ 1.2423E-06 1.OOE+00 1.2423E-06 Page 19 of 22

Indian Point Energy Center Position Paper I Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05I to Calculate Seismic Core Damage Frequency (SCDF) Values I 10 H~z USGS Hazard Curve 0.001g Intervals Total Seismic Core Damage Frequency= IO1E-04 Plant Median Capacity 0.5304 Plant Composite Uncertainty, C 0.34 Seismic Level (g) Frequency of Failure Delta Risk at Exceedance Probability g level 0.008 1.20E-02 2.62E-34 4.58E-37 0.009 1.03E-02 1.36E-32 1.21E-35 0.010 9.37E-03 4.34E-31 4.08E-34 0.011 8,43E-03 9.39E-30 4,19E-33 0.012 7.99E-03 1.48E-28 8.24E-32 0.013 7.43E-03 1.78E-27 8.58E-31 0.014 6.95E-03 1.72E-26 7.22E-30 0.015 6.53E-03 1.37E-25 5.07E-29 0.016 6.16E-03 9.25E-25 2.75E-28 0.017 '5.8613-03 5.42E-24 :1.87E-27 0.018 5.52E-03 2.80E-23 7.82E-27 0.019 5.24E-03 1.30E-22 3.26E-26 0.020 4.99E-03 5.44E-22 1.24E-25 0.021 4.76E-03 2.09E-21 4.33E-25 0.022 4.55E-03 7.40E-21 1.41 E-24 0.023 4.36E-03 2.44E-20 4.26E-24 0.024 4.19E-03 7.55E-20 1.56E-23 0.025 3,98E-03 2.20E-19 2.41 E-23 0.026 3.87E-03 6.08E-19 8.92E-23 0.027 3.72E-03 1.60E- 18 2.17E-22 0.028 3.59E-03 4.0lE-18 5.08E-22 0.029 3.46E-03 9.66E- 18 1.14E-21 0.030 3.34E-03 2.24E-17 2.47E-21 0.031 3.23E-03 5.OOE-17 5.17E-21 0.032 3.13E-03 1.08E-16 1.05E-20 0.033 3.03E-03 2.26E- 16 2.06E-20 0.034 2.94E-03 4.59E-16 3.95E-20 0.035 2.85E-03 9.09E- 16 7.37E-20 0.036 2.77E-03 1.75E-15 1.34E-19 0.037 2.70E-03 3.30E-15 2.39E- 19 0.038~ 2.62E-03> 6.07E-15 4.44E- 19 0.039 2.55E-03 1.09E-14 7.58E-19 Last rows of the EXCEL spreadsheet are shown below 4.996 1.43E-07 1 1.56078E-10 4.997 1.43E-07 1 1.55876E-10 4.998 1.42E-07 1 1.55675E-10 4.999 1.42E-07 1 1.55473E-10 5.000 j1 .4205E-07 I' 1.4205E-07 Page 20 of 22

Indian Point Energy Center Position Paper I Develop Methodology that Emulates NRC' use of IPEEE Parameters 2011-ISRT-05 I to Calculate Seismic Core Damage Frequency (SCDF) Values 5 Hz USGS Hazard Curve 0.001g Intervals Total Seismic Core Damage Frequency = 6.59E-05 Plant Median Capacity 0.5474 Plant Composite Uncertainty, PC 0.34 SesmcLee () Frequency of Failure Delta Risk at Sesmcevl g) Exceedance Probability g level 0.009 1.22E-02 4.48E-33 5.48E-36 0.010 1.10E-02 1.47E-31 1.86E-34 0.012 9.14E-03 5.24E-29 3.87E-32 0.013 8.40E-03 6.45E-28 4.07E-31 0.014 7.77E-03 6.35E-27 3.46E-30 0.015 7.22E-03 5.15E-26 2.45E-29 0.016 6.75E-03 3.54E-25 1.34E-28 0076.37E-03 2,11 E-24 9.09E-28 0.018 5.94E-03 1.11E-23 3.83E-27 0.019 5.59E-03 5.199E-23 1.61 E-26 0.020 5.28E-03 2.21 E-22 6.15E-26 0.021 5.OOE-03 8.58E-22 2.16E-25 0.022 4.75E-03 3.08E-21 7.05E-25 0.023 4.52E-03 1.03E-20 2.15E-24 0.024 4.31 E-03 3.22E-20 7.90E-24 0.025' '4.07E-03 9.48E-20~ 1.22E-23 0.026 3.94E-03 2.65E-19 4.51 E-23 0.027 3.77E-03 7.02E-19 1.11 E-22 0.028 3.61E-03 1.78E-18 2.59E-22 0.029 3.47E-03 4.33E-18 5.85E-22 0.030 3.33E-03 1.OIE-17 1.27E-21 0.031 3.21E-03 2.28E-17 2.67E-21 0.032 3.09E-03 4.96E-17 5.43E-21 0.033 2.98E-03 1.05E-16 1.07E-20 0.034 2.88E-03 2.15E-16 2.06E-20 0.035 2.78E-03 4.28E-16 3.86E-20 0.036 2.69E-03 8.31E-16 7.07E-20 0.037 2.60E-03 1.57E-15 1.26E-19 0.038 2.52E-03 2.92E-15 2.32E-19 0.039 2.44E-03 5.29E-15 3.98E-19 Last rows of the EXCEL spreadsheet are shown below:

4.996 1.29E-07 1 1.15686E-10 4.997 1.29E-07 1 1.1556E-10 4.998 1.29E-07 1 1.15433E-10 4.999 1.29E-07 I -9.52273E-09 5.000 1.3826E-07 1I 1.3826E-07 Page 21 of 22

Indian Point Energy Center Position Paper Develop Methodology that Emulates NRC' use of IPEEE Parameters 201 1-ISRT-05 to Calculate Seismic Core Damage Frequency (SCDI) Values I 1 Hz USGS Hazard Curve 0.001g Intervals' Total Seismic Core Damage Frequen = 2.4 Plant Median Capacity 1 0.2754 Plant Composite Uncertainty, Pc 0.34 Seismic Level (g) Frequency of Failure Delta Risk at Exceedance Probability g level 0.004 1.27E-02 525E-34 I1 86E-36 0.005 9.19E-03 5.86E-31 6.20E-34 0.006 8.13E-03 1.55E-.28 '2.98E-31 0.007 6.21E-03 1.52E-26 1.87E-29 0.008 4.98E-03 7.29E-25 3,14E-28 0.009 4.54E-03 2.03E-23 1.19E-26 0.010 3.96E-03 3.68E-22 1.71 E-25 0.011 3.49E-03 4.78E-21 1.80E-24 0.012 3.12E-03 4.70E-20 1.05E-23 0.,03 . '.2.89E 3 .. 3.,6E-197 1..39E-22<

0.014 2.52E-03 2.38E-18 5.63E-22 0.015 2.28E-03 1.30E-17 2.62E-21 0.016 2.08E-03 6.21E-17 1.07E-20 0.017 1.91E-03 2.62E-16 3.92E-20 0.018 1.76E-03 9.91E-16 1.30E-19 Q0,09 1.63E-03 3.41E-15 <4.16E-19 0.020 1.50E-03 1.08E-14 1.16E- 18 0.021 1.40E-03 3.18E-14 3.03E-18 0.022 1.30E-03 8.74E-14 7.43E-18 0.023 1.22E-03 2.26E- 13 1.72E-17 0.024 1.1 4E-03 5.54E-13 3.80E-1 7 0.025 1.07E-03 1.29E-12 8.01 E- 17 0.026 1.01E-03 2.88E-12 1.62E-16 0.027 9.53E-04 6.15E-12 3.15E-16 0.028 9.02E-04 1.27E- I1 3.03E-16 0,,0029 8.78E-04 2.51E-1 I 1.73E-5~

0.030 8.09E-04 4.83E- 1I 1.99E- 15 0.031 7.68E-04 9.02E- 11 3.41E-15 0.032 7.30E-04 1.64E-10 5.71E-15 0.033 6.95E-04 2.89E-10 9.33E-15 0.034 6.63E-04 4.99E-10 1.49E- 14 0.035 6.33E-04 8.43E-10 2.34E-14 0.036 6.05E-04 1.39E-09 3.59E-14 0.037 5.80E-04 2.25E-09 5.43E-14 0.038 5.56E-04 3.58E-09 8.06E-14 0.039 5.33E-04 5.60E-09 1.18E-13 I ast rows of the y(P cnrPeisheet sXF ar s~hown h-Iow:

3.686 1.60E-08 1 1.75759E-11 3.687 1.60E-08 I 1.75519E-11 3.688 1.60E-08 1 1.75279E- 11 3.689 1.60E-08 1 1.75039E-11 3.690 1 1.60E~-08 I 1.60E-08 Page 22 of 22

ItEntterg ENG REPORT IP-RPT- 1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency Attachment 6 Revised SCDF for IP3 using USGS & EPRI Hazard Curves

Indian Point Energy Center Position Paper I Revised SCDF Estimates for IP3 using USGS & EPRI Hazard Curves I201 1-ISRT-06 II Prepared by: Dan Nuta, PE Reviewed by: John Bretti Assessment of the Effect of Identified Conservatisms on the SCDF Estimates Obtained Using the USGS Hazard Curves.

Seismic core damage frequency analyses using USGS Hazard Curves used by NRC for 1 Hz, 5 Hz, 10 Hz, and PGA discretized to acceleration intervals of 0.001g were performed using fragilities adjusted for identified conservatisms. The analyses consist of convolving the USGS Hazard Curves for different frequencies with the various fragilities.

A description of the analyses performed is as follows:

Analyses Series 1 - NRC fragility values; i.e., Am = 0.34g and P3c = 0 .34.

Analyses Series 2 - Analyses with "adjusted" NRC fragility values (i.e., Am = 1.3 x 0.34g, and the same 13c = 0.34). This fragility removes the conservatism caused by double-counting the peak to valley variability in both the hazard curves and fragilities estimates.

Analyses Series 3 - Analyses using the surrogate fragility, Am = 0.75g and 13c = 0.3. Based on results of the re-evaluation of the low fragility components (high contributors to risk), the median capacities are equal to, or higher than, the conservative 0.75g median capacity selected for the robust, screened-out SSCs. These analyses do not account for double-counting of the peak to valley variability.

Analyses Series 4 - Analyses that reflect the higher fragilities for the SSCs previously found to have low fragilities and the removal of the peak to valley decrease in the median capacities by using an "adjusted" surrogate fragility of Am = 1.3 x 0.75g = 0.98g, and Pc = 0.3.

Analyses Series 5 -: Analyses using Am = 0.98g and 13c = 0.40 (Using different Am and P3c values that give the same HCLPF of 0.50g)

Analyses Series 6 - Analyses using Am = 1.3 x 0.98 = 1.27g, and 13c = 0.4 (Same as Series 5. but also removing the double-counting for peak to valley variability)

Results from the analyses are presented below.

Analysis Series - 1 2 3 4 5 6 Hazard Curve NRC Reported Entergy SCDFs Entergy SCDFs Entergy Entergy SCDFs NRCR r Etr SC3 Ente g SCDFs Entergy Entergy SCDF Am =.34 1.3 x .34 A =75g Am = 1.3 x 0.75g Am = 0.98g Am = 1.3 x 0.98g c3c.30 3c =.40 P3c.40 PGA 9.1E-05 9.1OE-05 6.11 E-05 2.41E-05 1.44E-05 1.59E-05 9.12E-06 10 Hz 1.OE-04 1.01E-04 6.65E-05 2.51E-05 1.47E-05 1.65E-05 9.39E-06 5 Hz 6.6E-05 6.59E-05 4.19E-05 1.46E-05 8.27E-06 9.47E-06 5.23E-06 1 Hz 2.3E-05 2.34E-05 1.45E-05 4.88E-06 2.74E-06 3.16E-06 1.74E-06 Page 1 of 2

Indian Point Energy Center Position Paper Revised SCDF Estimates for IP3 using USGS & EPRI Hazard Curves Assessment of the Effect of Identified Conservatisms on the SCDF Estimates Obtained Using the EPRI Updated Hazard Curves.

Seismic core damage frequency analyses using updated hazard curves developed by EPRI for 1 Hz, 10 Hz, and PGA discretized to acceleration intervals of 0.001 g were performed using fragilities adjusted for identified conservatisms. The analyses consist of convolving the EPRI updated hazard curves for different frequencies with the various fragilities A description of the analyses performed is as follows:

Analyses Series 1 - NRC fragility values; i.e., Am = 0.34g and P3c = 0 .34.

Analyses Series 2 - Analyses with "adjusted" NRC fragility values (i.e., Am = 1.3 x 0.34g, and the same P3c = 0.34). This fragility removes the conservatism caused by double-counting the peak to valley variability in both the hazard curves and fragilities estimates.

Analyses Series 3 - Analyses using the surrogate fragility, Am = 0.75g and P3c = 0.3. Based on results of the re-evaluation of the low fragility components (high contributors to risk), the median capacities are equal to, or higher than, the conservative 0.75g median capacity selected for the robust, screened-out SSCs. These analyses do not account for double-counting of the peak to valley variability.

Analyses Series 4 - Analyses that reflect the higher fragilities for the SSCs previously found to have low fragilities and the removal of the peak to valley decrease in the median capacities by using an "adjusted" surrogate fragility of Am = 1.3 x 0.75g = 0.98g, and 3c = 0.3.

Analyses Series 5 -: Analyses using Am = 0.98g and P3c = 0.40 (Using different Am and P3c values that give the same HCLPF of 0.50g)

Analyses Series 6 - Analyses using Am = 1.3 x 0.98 = 1.27g, and P3c = 0.4 (Same as Series 5. but also removing the double-counting for peak to valley variability)

Results from the analyses are presented below.

Analysis Series --* 1 2 3 4 5 6 Hazard Curve NRCSCDF Reported Entergy SCDFs EtrySDs Entergy negyCDsSCDFs Entergy SCDFsEnegEtry Entergy SCDFs HzrCuv CFAm = .34, 1.3 x .34g, SDs Am - 1.3 x 0.75g, negEtry Frequency USGS Based A=.34 1. =.34 AM75 Am =.30 Am = 0.98g Am = 1.3 x 0.98g Estimates 34c =.30 1cc =c.40 3c= .40 PGA 9.1E-05 3.94E-05 3.1OE-05 1.61 E-05 1.06E-05 1.12E-05 7.08E-06 10 Hz 1.OE-04 4.12E-05 3.21E-05 1.60E-05 1.02E-05 .1OE-05 6.77E-06 5 Hz 6.6E-05 n/a n/a n/a n/a n/a n/a 1 Hz 2.3E-05 2.83E-06 1.68E-06 4.87E-07 2.53E-07 3.05E-07 1.56E-07 Page 2 of 2

_-Entergy ENG REPORT IPRPTI 1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency Attachment 7 Revised Calculations for Identified Low Capacity Components

Indian Point Energy Center Position Paper Identify Low Capacity Components and Prioritize List for Review I201 1-ISRT-07 II Identify Low Capacity Components and Prioritize List for Review Prepared By: A. Unsal I D. Gaynor Reviewed By: P. Baughman Purpose The purpose of this task is to identify the low seismic capacity components for the sequences, described below, that are significant contributors to risk in order to prioritize the effort to determine a more realistic plant-level fragility. This will be done by a review of IPEEE report Table 3.1.5.1, which lists the HCLPFs of the components in the plant model.

Evaluation IPEEE report Table 3.1.5.1 was reviewed with the help of the Entergy PRA group. In IP3 IPEEE report Section 3.1.5.5 [1], the mean core damage frequency (CDF) due to a seismic event is calculated to be 4.4 x 10 5 /year. The dominant core damage sequence is a seismic-induced loss-of-offsite power (LOSP) and the subsequent loss of on-site AC power from all three emergency diesel generators. This sequence contributes 43.5% of the total mean seismic CDF (1.92 x 10-5/year).

The second highest sequence contributing to the seismic CDF involves loss of secondary side cooling due to depletion of the condensate storage tank, and failure of RHR shutdown cooling due to the seismic event, which contributes 7.47 x 10-6/year (16.9% of the total mean seismic CDF).

The third highest sequence contributing to the seismic CDF involves a loss of component cooling water (CCW) due to a seismic event, which leads to an RCP seal LOCA and loss of decay heat removal. This sequence contributes 7.33 x 10 6/year (16.6% of the total CDF). This sequence includes success of the containment fan coolers, which was included in the sequence for use in the Level 2 analysis. An additional sequence, with a mean seismic CDF contribution of 2.84 x 10 /year, differs only in as much as it includes failure of the fan cooler units. Since these are essentially the same with regard to core damage, the overall contribution is the sum of these two sequences, or 1.02 x 10 5/year (23.1% of the total mean seismic CDF).

Components were selected for further investigation based on:

1) The potential for their failure to cause an SBO by disabling the emergency onsite AC power (EDGs),

2) Selected components that impact other dominant sequences, and Page 1 of 4

Indian Point Energy Center Position Paper 2011-SRT-07I Identify Low Capacity Components and Prioritize List for Review I

3) Having a seismic HCLPF capacity below that of the "surrogate" element (i.e., having a HCLPF 84 value below 0.5g).

Results The following components were identified as resulting in or potentially contributing to the Station Blackout and other dominant sequences in the IP3 IPEEE.

The follow-on task will be to review the calculations which determined the HCLPFs and median capacities for the selected components and determine whether opportunities exist to develop, consistent with industry-accepted criteria, more realistic values. If existing calculations cannot be located, then new calculations will need to be performed to estimate the HCLPFs. The results of this follow-on effort will be presented in Attachment 7 of this report.

References

[1] IP31PEEE Report, IP3-RPT-UNSPEC-02182, September 1997.

[2] Entergy Drawing 9321-F-33853, "Electrical Distribution and Transmission System" Median Component Description Capacity HCLFP4 Impact Notes (g ) (g)_I mp a ct Note s Battery Chargers 31 & 32 0.51 0.24 SBO Note 1 Motor Control Centers 36A, 36B & 0.62 0.24 I SBO Note 2 37 Station

&6 Service Transformers 2, 3, 5 0.67 06 0.31

.1 SBO SO Note Nt 3 Switchgear 31 & 32 (Contains vital 0.67 0.31 SBO 480V buses 2A, 3A, 5A and 6A)

Service Water System Pumps 31, 32, 33, 34, 35 & 36 (provide cooling 0.69 0.31 SBO for EDGs)

Battery Banks 31, 32, & 34 0.88 0.41 SBO Note 4 EDG 31, 32 & 33 Control Panels 0.88 0.41 SBO EDG Current Transformers 31, 32 & 0.99 0.46 SBO 33 0.99 0.46 SBO Supervisory Panel 0.52 0.24 Note 5 Central Control Room Racks 0.67 0.31 Note 5 Page 2 of 4

Indian Point Energy Center I Position Paper I 2011-ISRT-07 I Identify Low Capacity Components and Prioritize List for Review I

Median HCLF m Component Description Capacity (g) Impact Notes (g) (gmac_ oe Control Room Flight Panel 0.67 0.31 Note 5 Seal LOCA Component Cooling Water (CCW) 0.41 0.19 with Loss Surge Tanks 31 & 32 of Decay Heat Removal Seal LOCA with Loss 0.52 0.24 oftDecay CCW Heat Exchangers 31 & 32 of Decay Heat Removal Loss of RHR Heat Exchangers 31 & 32 0.49 0.23 Shutdown Cooling Loss of RHR Pumps 31 & 32 0.62 0.29 Shutdown Cooling Motor Control Centers 36D, 36E N/A N/A Note 6 Note 1: DC buses 31 and 32 support EDGs 32 and 33. Loss of Battery Chargers 31 and 32 would eventually result in loss of those DC buses (following depletion of the batteries) and failure of the associated EDGs.

Battery Charger 33 supports EDG 31 and has a HCLPF greater than the surrogate value of 0.5g. However, since it appears that the success criteria for the essential service water system at the time of the IPEEE was two out of three SW pumps, loss of two EDGs would result in not meeting this success criteria. This would then cause a consequential failure of the third EDG due to loss of cooling from the essential service water header, and result in an SBO.

Note 2: MCCs 36A, 36B and 37 are associated with 480V Buses 5A and 6A, which are fed from EDGs 33 and 32, respectively. MCC 36C is associated with 480V Bus 2A, which is fed from EDG 31, and has a HCLPF of 0.4g. Although these MCCs provide power to many plant safety systems, since they do not directly fail the EDGs or the SW pumps, their failure was likely not binned to SBO.

Note 3: Failure of the Station Service Transformers would not, by themselves preclude powering the 480V buses from the EDGs. They are physically Page 3 of 4

Indian Point Energy Center Position Paper Identify Low Capacity Components and Prioritize List for Review 2011-ISRT-07 I I connected to Switchgear 31 and 32, however, and have a shared anchorage.

Note 4: Battery banks 31, 32 and 34 are associated with three of the four 480V buses. Battery Bank 33 supports EDG 31. However, should Battery Banks 31 and 32 fail immediately, they would preclude starting two of the three EDGs. Similar to the discussion in Note 1, the loss of two EDGs would result in not meeting the essential SW success criteria.

This would then cause a consequential failure of the third EDG due to loss of cooling from the essential service water header, and result in an SBO.

Note 5: The CCR panels and racks are not modeled as SBO contributors in the IP3 IPEEE.

Note 6: EDG ventilation fan power supplies have been changed since the IPEEE. All fans were previously powered by MCC36A and MCC36B.

The power sources were subsequently modified and unitized. Currently, fans 314 and 315 (for EDG 31) are powered from MCC36C, fans 316 and 317 (for EDG 32) are powered from MCC36D, and fans 318 and 319 (for EDG 33) are powered from MCC36E. MCC36C has a HCLPF of 0.40g but MCCs 36D and 36E, which are located in the 480V switchgear room, do not appear on the IPEEE component list.

Page 4 of 4

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: Supervisory Panel Indian Point Unit 3 HCLPF Review . Page I of 2 By: Paul Baughman 4-4-2011 Front bottom of panel is welded to channel anchored to floor. This takes shear Reviewed: Paul Bruck 4-5-2011 on bottom. Back top has angle bolted to wall (modification). Angle is not bolted to top of panel. Consider horizontal shear taken by weld at front of Component: SUPERVISORY PANEL base. Overturning moment and vertical force taken by top angle in bearing.

Structure: Control Building EL 53 Due to long length of panel, only front-to-back force with vertical needs to be checked. Welds are stronger than anchor bolts at top - anchor bolts control

Reference:

GIP SEWS Mod package DC#94-3-163CPR 9321-F-30523, IP3-CALC-STR-01157 HCLPF based on LLNL spectra PGA (g) 0.23 horizontal 0.15 vertical is two-thirds of horizontal FRS (g) 0.82 N-S direction y-direction CB EL 53 Use 48 Assume N-S is y (short) direction 0.74 E-W direction x-direction 0.15 vertical is same as ground because building is stiff Axes x = width y = depth z = height Weight 4737.96 lb Use actual 738 ib/ft weight from mod calculation with 6.42' length of critical angle length z cg 55 inches 110" High/2, IP-CALC-STR-01157, Pg 1 1 1 1 x anc inches This is the E-W direction Neglect overturning moment in long direction y anc 72 inches This is the N-S direction Pivot about frontof cabinet, Same Ref as z cg f'c 3000 psi Rx Reaction due to x accel Small due to panel length Ry 2967.806 Reaction due to y accel Lb, reaction on top angle due to overturning moment Rz 355.347 Reaction due to z accel Lb, reaction on top angle due to vertical acceleration Rdl -2368.98 Reaction due to weight Lb, reaction at top angle due to weight ...

Supervisory Panel HCLPF Calculation Rev O.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: Supervisory Panel TV 865.6099 Tension due to Ry Tension per bolt due to uplift Ry, 3"bolts. Arm of Ry =1.75", distance bolt from Page 2.of 2 Tz 103.6429 Tension due to Rz angle edge = 2" per mod calc._

Tdl -690.9525 Tension due to Rdl Vy 989.2685 Shear due to y accel Shear per bolt - 3 bolts Vz 118.449 Shear due to z accel Vdl -789.66 Shear due to weight Tsy 907.0671 lOOyt4Oz Tsz 449.8869 40y+lOOz Ts 907.0671 controlling Ts Vs 1036.648 lOOy+40z Tnom 2290 GIP FS = 3 1/2" bolt Vnom 2380 GIP FS = 3 1/2" bolt RFp 0.75 fc/4,000 psi RFs 0.95 f'c/lO,000 psi + 0.65 Tred 1717.5 Reduced for 3000 psi concrete Vred 2261 Reduced for 3000 psi concrete Tc 2146.875 adjusted for NP-6041 factor of safety = 2.4 Vc 3391.5 adjusted for NP-6041 factor of safety = 2.0 Scale Factors Bi-linear shear-tension interaction FS1 3.128575!

FS2 2.42449 FS 2.424491 HCLPF 0.557633 HCLPF = FS*PGA Beta c 0.46 assumed Beta c Am 1.62864 7-__

Supervisory Panel HCLPF Calculation Rev O.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: CCR Racks Indian Point Unit 3 HCLPF Review F'age, 1 of 2 By: Paul Baughman 3-31-2011 Reviewed: A. Unsal 04-06-2011 Component: CCR Racks 10 of 11 cabinets are welded with fillets. Most are 2"x3/16".

I Two per cabinet, both sides. Based on Rack B-1 HCLPF based on LLNL spectra PGA (g) 0.23 horizontal ....

0.15 vertical is two-thirds of horizontal ............

FRS (g) 0.82 N-S direction y-direction CB EL 53 Use 48 Assume N-S is y (short) direction 0.74 E-W direction x-direction_

0.15 vertical is same as ground because building is stiff Axes x =width y = depth z = height Weight 16500 lb 11 cabinets at 1500 lb each z cg 46 inches half of height plus l" base x anc 220 inches This is the E-W direction Use pivot distance neglecting end bolted cabinet y anc 26 inches This is the N-S direction Pivot about back edge of cabinet - use average depth to welds Tx 638.25 Tension due to x accel 4 welds in end cabinet Ty 1196.885 Tension due to y accel 20 welds on one side of cabinets Tz 61.875 Tension due to z accel 40 welds Tdl -412.5 Tension due to weight 40 welds Vx 338.25 Shear due to x accel 40 welds_

Vy 305.25 Shear due to y accel 40 welds CCR Racks HCLPF Calculation Rev 0.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: CCR Racks Tsx 1141.754 $OOx+40Y+40z _________ .... Page.2of 2 Tsy 1476.935 40x+I00y+40z Tsz 795.9288 40x÷4Ov+lOOz Ts 1476.935 controlling Ts Vs 359.6129 lOOx+40y Ts + Vs 1520.085 vector sum Cap 4056.413 GIP 30600 psi Capacity of one weld Scale Factor FS 2.93991; HCLPF 0.676179 HCLPF = FS*PGA .....

Beta c 0.46 assumed Beta c Am 1.9748711 CCR Racks HCLPF Calculation Rev O.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: Cur Trans Enclos Indian Point Unit 3 HCLPF Review Page 1 of 2 By: . UnsaT 4/4/2011 Reviewed: P. Baughman 4/5/2011, Component: CURRENT TRANSFORMER ENCLOSUREs 31, 32, 33 (BF4, BFS, BF6)

HCLPF based on LLNL spectra

_Comments PGA (g) 0.23 horizontal 0.15 vertical is two-thirds of horizontal Based on fundamental mode and building configuration.

FRS (g) 0.365 N-S direction Y-direction Peak of FRS used in GIP evaluation 0.365 E-W direction X-direction 0.15 vertical is same as ground because building is stiff Axes x = width y = depth z = height Weight 6300 lb From GIP assessment z cg 18.5 inches From GIP assessment x anc 32.5 inches This is the E-W direction Consider rotation about CL of anchor bolts y anc 38.0 inches This is the N-S direction Consider rotation about CL of anchor bolts fc 3000 psi Tx 436.32 Tension due to x accel Ty 559.75 Tension due to y accel Tz 118.125 Tension due to z accel TdI -787.5 Tension due to weight ......

Vx 287.44 Shear due to x accel Vy 287.44 Shear due to y accel Tsx 707.461 lOx+40y+40z .

Current Transformers 313233 HCLPF Calculation Rev 0.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-LSRT-07 Indian Point 3 HCLPF Review Component: Cur Trans Enclos Tsy 781.52 40x+lOOy+40z _ 'age 2 of 2 Tsz 516.55 40x+40y+100z Ts 781.52 controlling Ts Vs 309.58 100x+40Y Tnom 1460 GIP FS = 3 3/8" bolt From GIP assessment Vnom 1420 GIP FS = 3 3/8" bolt (Anchors have been grouted per PID 15056)

RFp 0.75 f'c/4,000 psi RFs 0.95 f'c/10,000 psi + 0.65 Tred 1095.0 Reduced for 3000 psi concrete Vred 1349.0 Reduced for 3000 psi concrete Tc 1368.8 adjusted for NP-6041 factor of safety = 2.4 Vc 2023.5 adjusted for NP-6041 factor of safety = 2.0 Scale Factors Bi-linear shear-tension interaction FS1 2.76 FS2 2.54 FS 2.54 HCLPF 0.58 HCLPF = FS*PGA Beta c 0.46 assumed Beta c i Am 1.70 1 1 Current Transformers 31_32_33 HCLPF Calculation Rev 0.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: EDG Control Panel Indian Point Unit 3 HCLPF Review -"' Page 1 OT By: Paul Baughman 03-31-2011 Reviewed: A. Unsal 04-06-2011 Component: EDG Control Panel HCLPF based on LLNL spectra El 15' DG BLDG Comments PGA (g) 0.23 horizontal 0.15 vertical is two-thirds of horizontal FRS (g) 0.365 N-S direction y-direction PSA ....

0.365 E-W direction x-direction PSA 0.15 Use PGA for vertical Axes x = width y = depth z =height _z = height Weight 2500 lb From GIP assessment z cg 60 inches From GIP assessment x anc 44 inches Assume cab width twice the A-46 cg _Consider rotation about equip edges y anc 18 inches Assume cab depth twice the A-46 cg [Consider rotation about equip edges Note says cab is bolted through steel plate,.plate is welded to structural steel Tx 622.1591 Tension due to x accel Ty 1520.833 Tension due to y accel Tz 93.75 Tension due to z accel Tdl -625 Tension due to weight Vx 228.125 Shear due to x accel Vy 228.125 Shear due to y accel Tsx 1267.992 100x+40y+40z EDG Control Panel HCLPF Calculation Rev 0.xIsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-LSRT-07 Indian Point 3 HCLPF Review Component: EDG Control Panel Tsy 1807.197 40x+100y+40z __age o ,

Tsz 950.947 40x+40y+lO0z Ts 1807.197 controlling Ts Vs 245.6981 lOOx+40y Tc 6672.5 1/2" A307 bolt Used AISC allowable times 1.7 Vc 3336.3 1/2" A307 bolt Used AISC allowable times 1.7 Scale Factors Bi-linear shear-tension interaction FS1 4.038021 FS2 4.047974 FS 4.038021 HCLPF 0.928745 HCLPF = FS*PGA Beta c 0.46 assumed Beta c Am 2.712522 EDG Control Panel HCLPF Calculation Rev 0.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: Flight Panel Indian Poin t Unit 3 HCLPF Review Page 1 of 2 By: Paul Baughman 3-31-2011 Reviewed: A. Unsal 04-06-2011 Component: FUGHT PANEL 10 of 11 cabinets are welded with fillets. Most are 2"x3/16".

Two per cabinet, both sides.

HCLPF based on LLNL spectra PGA (g) 0.23 horizontal 0.15 vertical is two-thirds of horizontal_

FRS (g) 0.82 N-S direction y-direction CB EL 53 Use 48 Assume N-S is y (short) direction 0.74 E-W direction x-direction I 0.15 vertical is same as ground because building is stiff Axes x = width y = depth z = height Weight 16500 lb 11 cabinets at 1500 lb each z cg 46 inches half of height plus 1" base x anc 220 inches This is the E-W direction Use pivot distance neglecting end bolted cabinet y anc 26 inches This is the N-S direction Pivot about back edge of cabinet - use average depth to welds Tx 638.25 Tension due to x accel 4 welds in end cabinet Ty 1196.885 Tension due to y accel 20 welds on one side of cabinets Tz 61.875 Tension due to z accel 40 welds Tdl -412.5 Tension due to weight 40 welds Vx 338.25 Shear due to x accel 140 welds Vy 305.25 Shear due to y accel _40 welds Flight Panel HCLPF Calculation Rev O.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: Flight Panel Tsx 1141.754 100x+40y+40z. Page 2 of 2 Tsy 1476.935 40x+lOOy+40z Tsz 795.9288 40x+40+ylO0z Ts 1476.935 controlling Ts Vs 359.6129 10Ox+40y" _

Ts + Vs 1520.085 vector sum Cap 4056.413 GIP 30600 psi Capacity of one weld Scale Factor FS 2.93991 HCLPF 0.676179 HCLPF = FS*PGA Beta c 0.46 assumed Beta c Am 2.9748711 Flight Panel HCLPF Calculation Rev 0.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: MCC 36AB, 37 Paul Bruck Date: 4/412011 1: Paul Baugihman Date: 4/5/2011 IMCC 36 A, 368, 37 A, 5 ft Elevation of Primar Auxiliary Building MCC SEWS Sheets, Drwg 9321-F-13123.1 Unistrut Catalog (General & Nuclear) _!

Estimate of PAB In-structure Response Spectra HCLPF based on LLNL spectra _

X1 PGA (g) 0.23 horizontal _ -

0.15 vertical is two-thirds of horizontal1 0.57 N-S direction X-direction <-P 0.541 E-W direction jY-direction] <-P 0.15 vertical is same as Inputs (Refer to SEWS):

x=width 'Y= depth z = heii Anchorages will be evaluated: 1. Weld to floor Frame, 2. Hold-down into unistrut embeds, 3. Unistrut embed capacity it 800 Ib Consider 1 Unit 1From GIP assessment !-

7 45 inches From GIP assessment _ __

54 inches Z cg + Z1. ZI = 9 in (see 9321-F-12123) I _

16 inches This is the N-S direction (Xl) 1 Consider rotation about equip edges MCC-36A 36B 37 HCLPF Calculation Rev 0.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: MCC 36AB, 37

1. Weld Loading MCC to C Beams & C Beams to WF: (See SEWS and 9321-F-13123) I age z ot 4 I I Weld between MCC and base (see SK-019 of SEWS) will control vs Weld Tension Loading: between base and beams (9321-F-13123) 7_

Tx 160.3125 Tension due to x accel Consider 8in of weld effective at resisting this (per side - 3/16" Fillet) - Limiting Ty 0 Tension due to y accel Consider line-up as unit, uplift from overturning is negligible Tz 7.5 Tension due to z accel Consider 16 in of weld effective Total (both sides - 3/16" Fillet)

Tdi -50 Tension due to weight Consider 16 in of weld effective Total (both sides - 3/16" Fillet)

Shear Loading: 2.Shadut x a lsrif df[o b ie 1" t Vx 28.5 Shear due to x accel Consider 16 in of weld effective Total (both sides - 3/16" Fillet)

Vy 27 Shear due to y accel Consider 16 in of weld effective Total (both sides - 3/16"n Fillet) ____

Load Combinations: (Using the 100/40/40 rule):

Tsx 163.3125 lOOx+40y+40z Tsy 67.125 40x+100y+40z Tsz 71.625 40x+40y+100z Ts 163.3125 controlling Ts Vs 30.4777 100x+40y Controlling Vs Weld Comb 166.1321 SRSS of ((Ts+TdI) + (Vs))

Weld Allow 30600 psi GIP (0.3 x 60,000 x 1.7) for E60XX electrode Weld Leg 0.1875 in 9321-F-13123 limiting 3/16 weld I Weld Cap 4056.413 lb/in 0.707 (Throat) x Weld Leg x Weld Allow FS 23.41679

2. Beam Anchorage to Unistruts (See SEWS and 9321-F-13123)

WF below MCC is attached to Unistruts using bolts.

Tension Loading: Consider 1 bolt per WF (The adjacent bolt carries the adjacent MCC Unit)

Tx 769.5 Tension due to x accel Consider 2 bolts effective, (1 per WF end) II Ty 358.8923 Tension due to y accel Consider 2 bolts effective, (2 per wide flange across width of MCC)

Tz 30 Tension due to z accel I Consider 4 bolts effective I______

MCC-36A 36B 37 HCLPF Calculation Rev 0.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: MCC 36AB, 37 Tdl -200 Tension due to weight Consider 4 bolts effective "age 5 OT Shear Loading: _

Vx 114 Shear due to x accel Consider 4 bolts effective Vy 108 Shear due to y accel Consider 4 bolts effective Load Combinations: (Using the 100/40/40 rule):

Tsx 925.0569 100x+40y+40z Tsy 678.6923 40x+100y+40z Tsz 481.3569 40x+40y+100z Ts 925.0569 controlling Ts Vs 121.9108 100x+40y Controlling Vs Unistrut Values per Unistrut Catalogs: General Enineering & Nuclear. Unistrut values utilize F-of-S =3 Unistrut bolts are A193B7 (per 9321-F-13123 bolt is 1/2"), Fy for A193B7 = 125 ksi Tallow 20001 lbs Unistrut allowables Tension (pull-out) F-of-S = 3 (Nuclear Catalog)

Vallow 1500 lbs Unistrut allowable Shear (slip-along) (F-of-S = 3)

FS1 2.378232 Tension loading FS2 12.30408 Shear slip-along FS 2.378232

3. PuU-out of Unistrut embedded (See SEWS and 9321-F-13123)

Unistrut Values per Unistrut Catalogs: General Enineering & Nuclear. Unistrut values utilize F-of-S = 3 Unistrut bolts are A193B7 (per 9321-F-13123 bolt is 1/2"), Fy for A193B7 = 125 ksi T Tension Load 1850.114 lbs 2 x Ts Tcapacity 2000 lbs/ft Unistrut allowables Tension allowable (pull-out) per foot of embed f'c reduction 0.866025 (f'c/4000)AO.5 IIII_

Embed L 2.666667 ft Spacing between beams, per 9321-F-13123 MCC-36A 36B 37 HCLPF Calculation Rev 0.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF.Review Component: MCC 36AB, 37 Tallow 4618.802 lbs Tcapacity x Lx f'c reduction ,age 4 or 4 FS1 2.712699 Tension loading ](Tallow + 2 x dwt)/Tension Load (Note 2 x dwt for adjacent cabinet)

FS 2.712699 Controlling F-of-S I FS all 2.378232 <-Controlling FS, for Anchorage item, 1, 2, 3 HCLPF 0.55 HCLPF = FS*PGA Beta c 0.46 assumed Beta c Am 11 MCC-36A 36B 37 HCLPF Calculation Rev O.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 201-LSRT-07 Indian Point 3 HCLPF Review Component: Service Water Pump Indian Point Unit 3 HCLPF Review ...... rage 1 or By: Paul Bruck Date: 4/1/2011 Reviewed: Paul Baughman Date: 4/1/2011 Component: Service Water Pump HCLPF based on LLNL spectra Comments PGA (g) 0.23 horizontal 0.15 vertical is two-thirds of horizontal Based on fundamental mode FRS (g) 1.02 N-S direction i 1and y-direction building configuration.

Peak of In-structure RS for LLNL & HCLPF 1.02 E-W direction x-direction_

0.15 vertical is same as ground because building is stiff Geometry Inputs (Refer to SEWS):

Axes x = width y = depth z = height Iy = depth z = height Weight 4397.5 lb Add 10% from Pump + Motor From GIP assessment z cg 43 inches I I From GIP assessment x anc 44 inches This is the E-W direction Consider rotation about equip edges y anc 44 inches This is the N-S direction Consider rotation about equip edges f'c 3000 psi Anchor Bolt Loading Upper Base Plate EL 15':

Tension Loading:

Tx 2191.754 Tension due to x accel Ty 2191.754 Tension due to y accel Tz 164.9063 Tension due to z accel Tdl -1099.375 Tension due to weight Service Water Pump HCLPF Calculation Rev 0.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: Service Water Pump Shear Loading: I _age z of Vx 1121.363 Shear due to x accel Vy 1121.363 Shear due to y accel Load Combinations: (Using the 100/40/40 rule):

Tsx 3134.418 100x+4Oy+40z Tsy 3134.418 40x+lOOy+40z Tsz 1918.309 40x+4Oy+lOOz Ts 3134.418 controlling Ts Vs 1207.744 lOOx+40y Controlling Vs Tnom 50400 GIP NOTE: should calculate using latest ACI 349 Vnom 25250 GIP I ... ..

Rep NOTE: should calculate using latest edition of ACI 349 - will result in higher capacity than REs used in SEWS. Edge distance is 7" - but this is not a straight edge - it is re-entrant corner.,

RF reduction will notbe as significant as reduction taken in original GIP F'c = 3000 psi Tc 38695 (Value & reduction per SEWS)

Vc 7936 (Value on SEWS incorectly double reduced capacity)

_ I___ _ _ I I___' _ _ __ __ _ _ _

Scale Factors Bi-linear shear-tension interaction FS1 12.69594 FS2 4.882469 FS 4.882469 HCLPF 1.122968 HCLPF = FS*PGA Beta c 0.46 assumed Beta c Am 3.279775 Service Water Pump HCLPF Calculation Rev O.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ASRT-07 Indian Point 3 HCLPF Review Component: Switchgear Indian Point Unit 3 HCPF Review P__age I oT By: Paul Baughman 3-31-2011 Reviewed: A. Unsal 04-06-2011 _

Component: Switchgear CB EL 15 HCLPF based on LLNL spectra _Switchgear part of multi-cabinet lineup, A-46 checked 4 cabinet group.

As-built sketches show weld to embeded channel in front, anchor bolts in back.

_Assume both sides bolted - this is conservative.

PGA (g) 0.23 horizontal I I 0.15 vertical is two-thirds of horizontal Based on fundamental mode I y. ._re11ion _ and building configuration.

FRS (g) 0.37 N-S direction y-direction Peak of FRS 0.37 E-W direction x-direction_

0.15 vertical Is same as ground because building is stiff Axes x = width y = depth z = height z = height Weight 9882 lb From GIP assessment z cg 44 inches From GIP assessment x anc 66 inches This is the E-W direction Consider rotation about equip edges y anc 73.5 inches This is the N-S direction Consider rotation about equip edges Nbx 8 Number bolts resisting uplift in x direction and only bolts in end cabinet Nby 4 Number bolts resisting uplift In y direction frc 3000 psi Tx 304.695 Tension due to x accel Ty 547.2073 Tension due to y accel Tz 92.64375 Tension due to z accel Tdl -617.625 Tension due to weight Vx 228.5213 Shear due to x accel V.y, 228.5213 Shear due to y accel Switchgear HCLPF Calculation Rev 0.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: Switchgear Page 2 of2 Tsx 560.6354 lOOx÷40y+40z Tsy 706.1428 40x+lOOy+40z Tsz 433.4047 40x+40y+lOOz Ts 706.1428 controlling Ts VS 246.1249 lOOx+40y Tnom 2290 GIP FS = 3 1/2" bolt Vnom 2380 GIP FS = 3 1/2" bolt RFp 0.75 f'c/4,000 psi RMs 0.95 f'c/1O,000 psi + 0.65 Tred 1717.5 Reduced for 3000 psi concrete Vred 2261 Reduced for 3000 psi concrete Tc 2146.875 adjusted for NP-6041 factor of safety = 2.4 Vc 3391.5 adjusted for NP-6041 factor of safety = 2.0 Scale Factors Bi-linear shear-tension interaction FS1 3.91493 FS2 3.967402 FS 3.91493 HCLPF 0.900434 HCLPF = FS*PGA Beta c 0.46 lassumed Beta c Am 2.629836 1 1 -

Switchgear HCLPF Calculation Rev Oxlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: Battery Banks Indian Point Unit 3 HCLPF Review Page 1 of 2 By: A. Unsal 1 4/1/2011 +/-

Reviewed: P. Baughman 4/4/2011 -

Component: BATTERY BANKS 31,32,34 HCLPF based on LLNL spectra

______ _______Comments_____ ____

Comments PGA (g) 0.23 horizontal 0.15 vertical is two-thirds of horizontal Based on fundamental mode FRS (g) 0.59 N-S direction E 1and x-direction building configuration.

Peak of FRS used in GIP evaluation 0.49 E-W direction y-direction 0.15 vertical is same as ground because building is stiff Axes x = width y = depth z = height Weight 3600 lb From GIP assessment z cg 29 inches (20"+38")/2 From GIP assessment - Average height x anc 48 inches This is the N-S direction _Consider rotation about CL of anchor bolts y anc 32.5 inches This is the E-W direction _Consider rotation about CL of anchor bolts f'c 3000 psi Tx 285.17 Tension due to x accel 50% additional capacity from middle anchors Ty 336.33 Tension due to y accel 56% additional capacity from middle anchors Tz 60 Tension due to z accel_

Tdl -400 Tension due to weight Vx 236 Shear due to x accel Vy 196 Shear due to y accel Tsx 443.70 lOOx+40y+40z Battery Banks 31, 32, 34 HCLPF Calculation Rev O.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: Battery Banks Tsy 474.40 40x+100y+40z ?age 2 of 2 Tsz 308.60 40x+40y+lOOz Ts 474.40 controlling Ts Vs 248.68 lOOx+40y Tnom 1460 GIP FS = 3 3/8" bolt Vnom 1420 GIP FS = 3 318" bolt RFp 0.75 fc/4,000 psi ........

RFs 0.95 f'c/lO,000 psi + 0.65 Tred 1095.0 Reduced for 3000 psi concrete Vred 1349.0 Reduced for 3000 psi concrete Tc 1368.8 adjusted for NP-6041 factor of safety = 2.4 Vc 202335 adjusted for NP-6041 factor of safety = 2.0 Scale Factors Bi-linear shear-tension interaction FS1 3.73 FS2 3.30 FS 3.30 HCLPF 0.76 HCLPF = FS*PGA ,,, ,,,,

Beta c 0.46 assumed Beta c Am 2.21 1 Battery Banks 31,32, 34 HCLPF Calculation Rev O.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities. 2011-iSRT-07 Indian Point 3 HCLPF Review Component: Battery Charger Indian Point Unit 3 HCLPF Review Page 1 of 2 By: Paul Baughman 3-31-2011 Reviewed: Paul Bruck 3-31-2011 Component: BATTERY CHARGER HCLPF based on LLNL spectra

_Comments_

PGA (g) 0.23 horizontal I I 0.15 vertical is two-thirds of horizontal Based on fundamental mode I -and building configuration.

FRS (g) 0.59 N-S direction y-direction Peak of FRS used in GIP evaluation 0.49 E-W direction x-direction 0.15 vertical is same as ground because building is stiff Axes x = width y = depth z = height Weight 3800 lb From GIP assessment z cg 42 inches From GIP assessment x anc 54 inches This is the E-W direction Consider rotation about equip edges y anc 24 inches This is the N-S direction Consider rotation about equip edges f'c 3000 psi Tx 724.1111 Tension due to x accel Ty 1961.75 Tension due to y accel ,

Tz 142.5 Tension due to z accel Tdl -950 Tension due to weight Vx 560.5 Shear due to x accel Vy 465.5 Shear due to y accel .....

Tsx 1565.811 lOOx+40y+40z Battery Charger HCLPF Calculation Rev O.xlsx 4/7/2011

Entergy Nuclear Operations, Inc. Seismic Component Fragilities 2011-ISRT-07 Indian Point 3 HCLPF Review Component: Battery Charger I. I. I V- I V Tsy 2308.394 140x+IOOY+i40z I -Page 2 of 2 Tsz 1216.844 40x+40y+lOOz Ts 2308.394 controlling Ts Vs 590.6189 lOOx+40y Tnom 4690 GIP FS = 3 3/4U bolt Vnom 5480 GIP FS = 3 3/4" bolt _

Tnom 6090 GIP FS = 3 7/8" bolt 7/8" Anchor Bolts on walkdown sheets Vnom 7700 GIP FS = 3 7/8" bolt _(GIP eval performed with 3/4" Anchors)

RFp 0.75 fc/4,000 psi RFs 0.95 f'c/10,000 psi + 0.65 Tred 4567.5 Reduced for 3000 psi concrete Vred 7315 Reduced for 3000 psi concrete Tc 5709.375 adjusted for NP-6041 factor of safety = 2.4 Vc 10972.5 adjusted for NP-6041 factor of safety = 2.0 Scale Factors Bi-linear shear-tension interaction FS1 2.884851 FS2 3.31447 FS 2.884851 IHCLPF 0.663516 HCLPF = FS*PGA Beta c 0.46 assumed Beta c Am 1.937885 _"

Battery Charger HCLPF Calculation Rev O.xisx 4/7/2011

Entergy Nuclear Operations Inc 2011-ISRT-07 Indian Point 3 HCLPF Evaluation of Residual Heat Removal Pumps 31 & 32 Prepared By: Paul Bruck 04-06-2011 Reviewed By: A. Unsal 04-07-2011 The residual heat removal (RHR) pumps are located at Elevation 15 feet in the Primary Auxiliary Building. This elevation is a basement elevation and is considered subjected to ground response. T he pump is supported vertically on a flat slide plate. Lateral seismic restraint is based on the suction an d discharge piping system welded to the pump. The pump has an offset m ass associated with the motor. The configuration of the piping at the pump is shown in Figure 1 below.

The piping system, including the pump and pump motor has been analyzed for the Design Basis Earthquake (DBE) response in-accordance with IP3 UFSAR design requirements. Piping at Indian P oint 3 is qualified considering damping of 1/2%. Per the IPEEE program, damping of 5% is appropriate for consideration. The design response spectra, associated with PGA of 0.15 g is depicted in Figure 2 be low. From this figure it can be seen that there is a significant reduction in spectral amplitude between the 0.5% and 5% damping curve between frequencies of 1 to 15 Hz, where piping response would be expected. This provides inherent margin in the piping analysis to accommodate increased seismic demand.

The pump and the offset mass of the motor are restrained by the piping on either side of the pump. Based on the pump drawing information (Refer to GIP SEWS, & Plant drawing IP3V-209-0041), the motor weight is 2,300 Ibs, with an offset of approximately 32 inches - to base, conservative for pipe). The peak of the LLNL ground response is estimated as 0.365 g's. Using this peak response would result in the following loading to the pipe:

Horizontal Moment = WT x Acceleration x Offset Horizontal Moment = 2,300 lbs x 0.365 g's x 32 in = 26,864 in-lbs Consider application of SRSS for two horizontal directions, Moment = 37,991 in-lbs The piping supporting each pump are line 10 suction and line 9 discharge. Line 9 and 10, per the IP2 line list (IP3-LIST-MULT-01177) are 8" and 14", respectively of piping class 601R. per the IP3 Piping Specification (SPEC TS-MS-024), both lines are Page 1 of 8

Entergy Nuclear Operations Inc 2011-ISRT-07 Indian Point 3 schedule 40. Conservatively considering the smaller pipe diameter, the section modulus for this pipe (Z = 16.809 in3). The resulting pipe stress from this contribution would be:

Fb = M/Z = 37,991 in-lbs / 16.809 in3 = 2,260 psi.

This stress is not considered a significant contribution. Based on the piping configuration, the stiffness of the suction and discharge piping and the configuration of the pipe supports, the ability of the piping to carry any potential overturning of the pump appears significantly rugged to re sist increased seismic loading. On this basis, it i s concluded that the pump HCLPF84 number above the surrogate element value of 0.50 g is applicable.

Figure 1: Configuration of RHR Pumps and Suction/Discharge Piping Local to Pump Page 2 of 8

Entergy Nuclear Operations Inc 2011-ISRT Indian Point 3 a

6 0 4 14r-2 r.is 0 2 4 6 8 0.1 ,2 ý4 6 .8 1.0 2 UNDAMPED NATURAL PERIOD T$-SC Figure 2: DBE 0.15g PGA Response Spectra Page 3 of 8 I

Entergy Nuclear Operations Inc 2011-ISRT-07--

Indian Point 3 HCLPF Evaluation of Residual Heat Removal Heat Exchanger 31 & 32 Prepared By: Paul Bruck 04-06-2011 Reviewed By: A. Unsal 04-07-2011 The RHR heat exchanger is located within the Vapor Containment, inner structure. The base of the heat exchanger is supported on a seismic support bracket at approximately elevation 68 feet, see Figure 1 below. The top of the heat exchanger is restrained by seismic rods, just below the elevation 94 ft operating deck level in the VC. The configuration of the heat exchanger is shown below in Figure 1 (based on drawing IP3V-0039-0002).

The seismic support at the base of the heat exchanger consists of steel frame support, as shown in Figure 2. The seismic support at the top of the heat exchanger consists of tie-rods as shown in Figure 3.

The heat exchanger and attached piping systems have been analyzed for the Des ign Basis Earthquake (DBE) response in-accordance with IP3 UFSAR design requirements.

Piping at Indian Point is qualified considering damping of 1/2%. Per the IPEEE program, damping of 5% is appropriate for consideration. The floor response spectra for the VC Inner Containment structure is applicable for evaluation of the heat exchanger and attached piping. There is a significant reduction in spectral amplitude between the 0.5% and 5% damping curve between frequencies of 1 to 15 Hz, where piping response would be expected. The design response spectra, associated with PGA of 0.15 g for the FRS at EL. 94 ft is depicted in Figure 4 and 5 (N-S and E-W).

From this figure it can be seen that there is a significant reduction in spectral amplitude between the 0.5% and 5% damping curve between frequencies of 1 to 15 Hz, where piping response would be expected. This provides inherent margin in the piping analysis to accommodate increased seismic demand. The design limits utilized in the piping analysis, per Plant UFSAR criteria, have additional margin beyond the acceptance criteria to accommodate increased seismic loading. Based on th e piping supply to from the heat ex changer providing ability to stabilize the heat exchanger, together with the seismic supports utilized in the design of the heat exchanger mounting, it appears the heat exchanger is significantly rugged to resist increased seismic loading. On th is basis, it is concluded that the heat exchanger HCLPF84 number will be above the surrogate element value of 0.50 g.

Page 4 of 8

Entergy Nuclear Operations Inc 2011-ISRT-07 Indian Point 3 Location of Upper Support Rods 1% 1 12" Inlet Nozzle I Location of Lower Support 8"O tlt1,1i 11 Figure 1: Configuration of Residual Heat Exchanger Page 5 of 8

Entergy Nuclear Operations Inc 2011-ISRT-07 Indian Point 3

• Jam% HEAT ExCHAQJGEI?*' r-LA,-rv-gA MFF.11 SPLAU QEL.(.#U-0

%CAILCAf,1' tive 10,%q W"OFl Figure 2: Seismic Support of heat Exchanger at Base (Plant Drawing 9321-F-13213)

Page 6 of 8

Entergy Nuclear Operations Inc 2011 -ISRT-07 Indian Point 3

-w Ia A

"tcome, VwM~gaUS "0 PMAU 4P %6,o 40-0 V04 1*24 cct* auý ecv**004o TIC- WD k WJALL LUCY PLAN Figure 3: Heat exchanger Seismic Support at Top of Heat Exchanger Page 7 of 8

Entergy Nuclear Operations Inc 2011-ISRT-07 Indian Point 3 Pamo(itwOM C."0 0,17 0.10 al0s II 1A Z10 60 1"O 20.0 Ga fWOCKY0cao0 FICURE CAAq Figure 4: DBE VC-IS FRS 0.15g PGA EL 94 ft (N-S) 0101 0.02 0.01 I

I.llI , ..... . t 3 I,, hIL

,wo FREO* M FICAJRE Ci3,5 Figure 5: DBE VC-IS FRS 0.15g PGA EL 94 ft (E-W)

Page 8 of 8

• -Entergy ENG REPORT IP-RPT4- 1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency Attachment 8 Calculate the CDF for SBO for surrogate level fragility using LLNL

Indian Point Energy Center Position Paper 201 1-ISRT-08 Calculate the CDF for SBO for surrogate level fragility using LLNL Calculate the CDF for SBO for surrogate level fragility using LLNL Prepared by: John Bretti Reviewed by: Doug Gaynor Purpose The purpose of Task 4.1 was to develop a simplified method to assess the potential impact on the Indian Point 3 seismic core damage frequency (SCDF) of increasing the calculated seismic capacity of components and structures, whose failure would lead to loss of the emergency onsite diesel generators. Since the seismic events of concern would render the offsite power sources and the Appendix R/SBO EDG unavailable due to their low seismic ruggedness, loss of the EDGs following such a "beyond design basis" seismic event would lead to an unrecoverable loss of all power (Station Blackout - SBO).

Since the seismic level associated with the "surrogate element" component fragility encompassed essentially all of the IP3 seismic risk contribution, this seismic level (0.75g) was chosen as the sensitivity value for this task.

Evaluation The evaluation consisted of three steps:

1. Develop the Simplified Method

2. Validate the Proposed Methodology

3. Use the Methodology to Calculate the Expected SBO Seismic CDF Contribution for the Surrogate Level Fragility (Capacity)

Step I -- Develop the Simplified Method The seismic hazard curve is expressed by an annual frequency of exceedance (per year) at various peak ground acceleration (pga) values. Table 3.1.2.1 of the IP3 IPEEE

[Reference 1] provides the mean annual frequency of exceedance at pga values ranging from 0.051g to 1.02g. This table is also included in Attachment 1.

The IP3 seismic component fragilities are characterized by a median capacity (m) and composite uncertainty (Pc). The current seismic component fragilities can be found in Table 3.1.5.1 of the IP3 IPEEE [Reference I]. By definition, the median capacity represents the seismic pga level at which the probability that the component fails is 0.50.

The component failure probability (p) can also be calculated at different pga levels using the following equation [Reference 2]:

Page 1 of 9

Indian Point Energy Center I Position Paper I 201 1-ISRT-08 I Calculate the CDF for SBO for surrogate level fragility using LLNL pAx) =(DLn /m Where p(x) = Component failure probability as a function of the seismic peak ground acceleration D[] = Standard normal cumulative distribution function x = Seismic peak ground acceleration (g) m = Plant/component median capacity (g)

Pc = Logarithmic composite standard deviation (uncertainty) of the response and fragility In addition to calculating component failure probabilities, the above equation can also be used to calculate the overall plant damage probability using known estimates of the plant median capacity and High Confidence Low Probability of Failure (HCLPF). This requires estimating the plant composite uncertainty (Pc), which can be calculated using the following equation:

• rgHCLPF1 PiC - 2.33 Once the seismic hazard frequency and component fragilities are known, the seismic risk frequency can be estimated by convoluting the hazard curve with the component failure probability using the following expression:

Risk = _fx)x -ffx).+l Where p(x) = Component failure probability evaluated at the average seismic pga level for the interval i to i+l fx) = Seismic frequency of exceedance for ith pga level f(x)i+l = Seismic frequency of exceedance for (i+l)th pga level In other words, the overall seismic risk frequency can be estimated by multiplying the plant/component failure probability at each discrete interval by the corresponding frequency of exceedance for that interval and summing the results over the entire interval.

This in turn can be used to estimate the results on a sequence-specific basis (e.g., station blackout) or at the plant level. When estimating the risk at the sequence level, the median capacity of the weakest component is used.

Step 2 -- Validate the Proposed Methodology Page 2 of 9

Indian Point Energy Center I Position Paper I I 201 1-ISRT-08 Calculate the CDF for SBO for surrogate level fragility using LLNL The proposed methodology was validated by generating an estimate of the IP3 seismic risk using the 1993 LLNL seismic hazard curve and current seismic plant/component fragilities contained in the IP3 IPEEE [Reference 1], and comparing it against the results of the IPEEE.

Two cases were run to perform this validation:

(a) SBO Contribution In the case of station blackout, the components with the lowest median capacity which can be attributed to failure of the emergency diesel generators (EDGs) are the battery chargers, based on the current IPEEE. As a result, the battery chargers are the dominant contributors to the seismic risk associated with station blackout. Based on Table 3.1.5.1 of the IP3 IPEEE [Reference 1], the median capacity of battery chargers 31 and 32 is 0.5 1g, with a composite uncertainty of 0.46. Using the above equations, the seismic SBO contribution attributed to the battery chargers can be estimated as 2.67E-5/yr (see Table 1 below).

Table I Seismic SBO Frequency Attributed to Battery Chargers 31 and 32 (LLNL Hazard)

(m = 0.51g and 13c= 0.46)

Seismic Level Frequency of Failure Delta Risk (g) Exceedance Probability at g level 0.051 1.15E-03 3.10E-06 1.54E-09 0.077 6.55E-04 5.94E-04 2.63E-07 0.153 2.12E-04 2.32E-02 3.14E-06 0.255 7.74E-05 9.70E-02 2.51 E-06 0.306 5.15E-05 2.19E-01 5.67E-06 0.408 2.56E-05 4.10E-01 4.68E-06 0.510 1.42E-05 6.20E-01 4.63E-06 0.663 6.74E-06 7.91 E-01 2.49E-06 0.816 3.58E-06 8.99E-01 1.65E-06 1.020 1.75E-06 9.34E-01 1.63E-06 Total 2.67E-05 Given that the chargers are the dominant contributors to SCDF due to SBO, the above estimate of the seismic SBO frequency compares favorably with the SCDF frequency of 1.92E-5/yr associated with SBO, which was estimated in the IP3 IPEEE.

(b) Overall Seismic CDF Using Plant Level Seismic Capacity The above evaluation was also performed at the plant level using the IPEEE plant level median capacity of 0.34g and HCLPF of 0.13g. This yields a composite Page 3 of 9

IPosition Paper I Indian Point Energy Center I

2011-ISRT-08 Calculate the CDF for SBO for surrogate level fragility using LLNL uncertainty (3c) of 0.41 and a total seismic core damage frequency of 6.1 IE-5/yr, as shown in Table 2.

Table 2 Total Seismic Core Damage Frequency (LLNL Hazard)

(m = 0.34a. HCLPF = 0,13a and B,= 0.41*

Seismic Level Frequency of Failure Delta Risk (g) Exceedance Probability at g level 0.051 1.15E-03 2.50E-05 1.24E-08 0.077 6.55E-04 4.25E-03 1.88E-06 0.153 2.12E-04 1.08E-01 1.46E-05 0.255 7.74E-05 3.21E-01 8.30E-06 0.306 5.15E-05 5.47E-01 1.42E-05 0.408 2.56E-05 7.67E-01 8.75E-06 0.510 1.42E-05 9.07E-01 6.78E-06 0.663 6.74E-06 9.70E-01 3.06E-06 0.816 3.58E-06 9.92E-01 1.82E-06 1.020 1.75E-06 9.96E-01 1.74E-06 Total 6.11E-05 The above estimate of the plant seismic risk compares favorably with the mean frequency of 4.4E-5/yr reported in the IP3 IPEEE.

(c) Comparison with NRC Seismic CDFUsing USGS Hazard Curve The 2008 USGS seismic hazard curves for the Indian Point site that were used in the GI-199 Safety/Risk Assessment were obtained from Martine Stutzke (NRC) and are shown in Attachment 2. Using the methodology described above and a plant median capacity of 0.34g (HCLPF = 0.13g), the total seismic CDF using the USGS hazard PGA curve was estimated as 9.87E-5/yr, as shown in Table 3 below.

Table 3 Total Seismic Core Damage Frequency (USGS Hazard)

(m = 0.34g, HCLPF = 0.13g and 13c= 0.41)

Seismic Level Frequency of Failure Delta Risk (g) Exceedance Probability at g level 0.007 8.02E-03 1.50E-19 3.46E-22 0.010 5.72E-03 1.74E-16 2.92E-19 0.014 4.04E-03 1.07E-13 1.31E-16 0.019 2.82E-03 3.46E-11 3.02E-14 0.027 1.95E-03 5.70E-09 3.48E-12 0.038 1.34E-03 4.97E-07 2.13E-10 0.053 9.06E-04 2.29E-05 6.79E-09 0.074 6.09E-04 5.48E-04 1.11E-07 0.103 4.07E-04 7.25E-03 1.03E-06 Page 4 of 9

Indian Point Energy Center Position Paper I201 1-ISRT-08 ICalculate the COF for SBO for surrogate level fragility using LLNL 0.145 2.65E-04 5.22E-02 4.93E-06 0.203 1.71E-04 2.09E-01 1.34E-05 0.284 1.07E-04 5.01E-01 2.13E-05 0.397 6.46E-05 7.93E-01 2.20E-05 0.556 3.69E-05 9.49E-01 1.63E-05 0.778 1.97E-05 9.86E-01 5.02E-06 0.900 1.46E-05 9.95E-01 5.04E-06 1.090 9.58E-06 9.99E-01 5.53E-06 1.520 4.05E-06 1.OOE+00 2.81E-06 2.130 1.24E-06 1.OOE+00 1.24E-06 Total 9.87E-05 The above estimate of the plant seismic risk compares favorably with the mean frequency of approximately 1E-4/yr reported by the NRC in the GI-199 Safety/Risk Assessment.

Step 3 -- Use the Methodology to Calculate the Expected SBO Seismic CDF Contribution for the Surrogate Level Fragility (Capacity)

This methodology was then used to assess the expected SCDF if the median capacity of the battery chargers and all other components that could fail the EDGs (including the EDGs themselves) could be shown to be at least equivalent to the surrogate element, which has a median capacity of 0.75g and an uncertainty (03c) of 0.30, as shown in Table 3.1.5.1 of the IPEEE.

Results Setting the median capacity of the battery chargers and all other components which could fail the EDGs (including the EDGs themselves) equivalent to the surrogate element, the seismic SBO frequency would decrease to 6.69E-6/yr, as shown in Table 4. However, it should be noted that if the composite uncertainty (Pc) does not change (i.e., remains 0.46), then the seismic SBO frequency only decreases to 1.0 1E-5/yr.

Table 4 Seismic SBO Frequency Using Surrogate Element Capacity (LLNL Hazard)

(m = 0.75g and 13c= 0.30)

Seismic Level Frequency of Failure Delta Risk (g) Exceedance Probability at g level 0.051 1.15E-03 1.05E-16 5.24E-20 0.077 6.55E-04 1.97E-10 8.72E-14 0.153 2.12E-04 7.17E-06 9.68E-10 Page 5 of 9

Indian Point Energy Center 201 1-ISRT-08 Calculate the CDF for SBO for surrogate level fragility using LLNL 0.255 7.74E-05 5.25E-04 1.36E-08 0.306 5.15E-05 6.70E-03 1.73E-07 0.408 2.56E-05 5.1OE-02 5.82E-07 0.510 1.42E-05 2.07E-01 1.54E-06 0.663 6.74E-06 4.82E-01 1.52E-06 0.816 3.58E-06 7.50E-01 1.38E-06 1.020 1.75E-06 8.48E-01 1.48E-06 Total 6.69E-06 Using this same approach but using the USGS hazard curve instead of the LLNL hazard curve decreases the seismic SBO frequency to 2.48E-5/yr, as shown in Table 5.

Table 5 Seismic SBO Frequency Using Surrogate Element Capacity (USGS Hazard)

(m = 0.75q and 13c= 0.30)

Seismic Level Frequency of Failure Delta Risk (g) Exceedance Probability at g level 0.007 8.02E-03 5.53E-51 1.27E-53 0.010 5.72E-03 6.01 E-44 1.01E-46 0.014 4.04E-03 1.95E-37 2.39E-40 0.019 2.82E-03 1.88E-31 1.64E-34 0.027 1.95E-03 4.88E-26 2.98E-29 0.038 1.34E-03 3.73E-21 1.60E-24 0.053 9.06E-04 8.38E-17 2.48E-20 0.074 6.09E-04 5.12E-13 1.03E-16 0.103 4.07E-04 9.91E-10 1.40E-13 0.145 2.65E-04 5.58E-07 5.27E-11 0.203 1.71E-04 8.85E-05 5.65E-09 0.284 1.07E-04 4.24E-03 1.80E-07 0.397 6.46E-05 6.53E-02 1.81 E-06 0.556 3.69E-05 3.48E-01 5.97E-06 0.778 1.97E-05 6.46E-01 3.29E-06 0.900 1.46E-05 8.27E-01 4.19E-06 1.090 9.58E-06 9.68E-01 5.35E-06 1.520 4.05E-06 9.98E-01 2.80E-06 2.130 1.24E-06 1.OOE+00 1.24E-06 Total 2.48E-05 References

1. "Indian Point Three Nuclear Power Plant Individual Plant Examination of External Events," IP3-RPT-UNSPEC-02182, New York Power Authority, September 1997.

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Indian Point Energy Center Position Paper 201 1-ISRT-08 Calculate the CDF for SBO for surrogate level fragility using LLNL

2. "Procedures for the External Event Core Damage Frequency Analyses for NUREG-1 150," NUREG/CR-4840, prepared for U.S. Nuclear Regulatory Commission by Sandia National Laboratories, October 1990.

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Indian Point Energy Center 21Position Paper I201 1-ISRT-08 I Calculate the ODE for SBO for surrogate level fragility using LLNL Attachment 1 LLNL Hazard Curves Values for IPEC Acceleration Frequency of Exceedance (/yr)

(g) 15% 50% 85% Mean 0.051 2.320E-04 7.650E-04 2.170E-03 1.152E-03 0.077 1.200E-04 4.170E-04 1.2 1OE-03 6.552E-04 0.153 2.930E-05 1.200E-04 4.OOOE-04 2.123E-04 0.255 8.200E-06 3.630E-05 1.470E-04 7.736E-05 0.306 4.850E-06 2.260E-05 9.720E-05 5.148E-05 0.408 1.810E-06 9.670E-06 4.910E-05 2.562E-05 0.510 7.340E-07 4.770E-06 2.640E-05 1.42 1E-05 0.663 2.220E-07 1.970E-06 1.2 1E-05 6.738E-06 0.816 7.780E-08 8.750E-07 6.090E-06 3.583E-06 1.020 2.230E-08 3.330E-07 2.970E-06 1.749E-06 Page 8 of 9

Attachment 2 2008 USGS Seismic Hazard Curves for Indian Point As Used in the GI-199 Safety/Risk Assessment PGA 10 Hz 5 Hz 1 Hz annual annual annual annual exceedence exceedence exceedence exceedence acceleration frequency acceleration frequency acceleration frequency acceleration frequency (g's) (per year) (g's) (per year) (g's) (per year) (g's) (per year) 0.007 8.02E-03 0.0075 1.20E-02 0.0075 1.47E-02 0.0037 1.27E-02 0.0098 5.72E-03 0.0113 8.43E-03 0.0113 9.73E-03 0.0056 8.13E-03 0.0137 4.04E-03 0.0169 5.86E-03 0.0169 6.37E-03 0.0084 4.98E-03 0.0192 2.82E-03 0.0253 3.98E-03 0.0253 4.07E-03 0.0127 2.89E-03 0.0269 1.95E-03 0.038 2.62E-03 0.038 2.52E-03 0.019 1.63E-03 0.0376 1.34E-03 0.057 1.69E-03 0.057 1.53E-03 0.0285 8.78E-04 0.0527 9.06E-04 0.0854 1.06E-03 0.0854 9.08E-04 0.0427 4.61 E-04 0.0738 6.09E-04 0.128 6.49E-04 0.128 5.26E-04 0.0641 2.37E-04 0.103 4.07E-04 0.192 3.87E-04 0.192 2.97E-04 0.0961 1.21E-04 0.145 2.65E-04 0.288 2.23E-04 0.288 1.62E-04 0.144 6.1OE-05 0.203 1.71E-04 0.39 1.44E-04 0.39 1.01E-04 0.216 3.04E-05 0.284 1.07E-04 0.432 1.24E-04 0.432 8.52E-05 0.324 1.47E-05 0.397 6.46E-05 0.649 6.45E-05 0.649 4.23E-05 0.4 9.79E-06 0.556 3.69E-05 0.8 4.49E-05 0.8 2.87E-05 0.487 6.57E-06 0.778 1.97E-05 0.973 3.13E-05 0.973 1.95E-05 0.73 2.68E-06 0.9 1.46E-05 1.46 1.37E-05 1.46 8.09E-06 0.9 1.61 E-06 1.09 9.58E-06 2.19 5.12E-06 2.19 2.91 E-06 1.09 9.81 E-07 1.52 4.05E-06 3.28 1.42E-06 3.28 8.50E-07 1.64 3.08E-07 2.13 1.24E-06 5 1.42E-07 4.92 1.38E-07 2.46 8.23E-08 I 1 _ 3.69 1.60E-08 Page 9 of 9

-- Entergy ENG REPORT P-RPT-I 1-00012 Reassessment of Indian Point 3 Seismic Core Damage Frequency Attachment 9 Plan for Adjusting SCDF without Re-Running SHIP Model

Indian Point Energy Center Position Paper 2011-ISRT-09 Plan for Adjusting SCDF without Re-Running SHIP Model Prepared by: Paul Baughman Reviewed by: Doug Gaynor Plan for Adjusting SCDF without Re-Running SHIP Model The plant SCDF from the IP3 IPEEE SPRA was calculated from the systems model using the SHIP code. The model incorporated seismic fragilities for components into event trees and fault trees. These organized into sequences where system induced failures could lead to core. The effect on plant SCDF of changing the fragility of a given component (structure, system or equipment) would normally be determined by changing the fragility parameters in the model and re-running it using the SHIP software. However, it is not possible, at this time, to change the inputs to the model. Therefore, the change in plant SCDF will be estimated using the following procedure.

In the IP3 SPRA seismic sequence quantification entailed the integration of the seismic hazard curve, component fragilities and the seismic system logic model to evaluate the frequency of system failure. The quantification was performed in three steps using the seismic component fragilities and SHIP computer code to determine the system, sequences and plant fragilities along with the seismic point estimate and mean core damage frequencies. The steps were:

1. Quantifying the seismic system logic model fragility

2. Quantifying seismic sequence fragility

3. Combining the system and sequence fragility and seismic hazards Three initiating events (seismic, loss-of-offsite power and small break LOCA) and 72 seismic sequences were identified and solved for the seismic PRA. The 72 seismic sequences were quantified using the LLNL hazard curves and best-estimate seismic fragilities and random failure frequencies. Of the 72 seismic-induced accident sequences, the ten dominant accident sequences loading to core damage contribute 99 percent of the total core damage frequency. All seismic sequences, ranked according to their contribution to core damage frequency, are listed below:

Sequence SCDF Contribution Loss of Offsite Power + Emergency AC Power - SBO 1.92E-05 43.5%

Loss of Offsite Power + RHR 7.47E-06 16.9%

Loss of Offsite Power + CCW 7.33E-06 16.6%

Failure of Surrogate Element 3.41 E-06 7.7%

Loss of Offsite Power + CCW + Cont Fan Coolers 2.84E-06 6.4%

Loss of AC Power + Failure to Insert Control Rods 1.96E-06 4.4%

Loss of Offsite Power + RHR + Cont Fan Coolers 9.54E-07 2.2%

Loss of Offsite Power + AFW + Failure to Establish F&B 2.65E-07 0.6%

Loss of Offsite Power + ATWS 1.16E-07 0.26%

Page 1 of 3

Indian Point Energy Center Position Paper 2011-ISRT-09 Plan for Adjusting SCDF without Re-Running SHIP Model Loss of Offsite Power + ATWS + Cont Fan Coolers Available 1.04E-07 0.24% I TOTAL 4.36E-05 98.8%

The total SCDF is the sum of the SCDFs for each sequence. The contribution of each sequence to the SCDF is equal to its SCDF divided by the total SCDF. The unrecoverable SBO sequence has the highest contribution, 43%.

If we assume that the fragility of the sequence is approximately equal to the fragility of the weakest component in the sequence, then we can compute the approximate SCDF by convolving the fragility of the weakest component with the seismic hazard. The resulting SCDF will not match the sequence SCDF exactly because it lacks the small contributions from other, stronger, components in the sequence and from random failures and operator actions. However, it should be close. A test run was done using the fragility of Battery Chargers 31 and 32. The resulting SCDF is with x% of the SCDF reported for the SBO sequence in the IP3 IPEEE submittal.

The battery charger fragility was derived from the HCLPF. The HCLPF as reported in the IPEEE submittal was 0.24g. If the HCLPF of the component was recalculated and found to be higher than 0.24g (because, say, of conservatism in the original HCLPF calculation), the SCDF of the sequence would be lower. The lower SCDF could be estimated by convolving the revised fragility (derived from the revised HCLPF) with the seismic hazard, assuming none of the other components have a HCLPF lower than the revised battery charger HCLPF. If this is the case, then this other component fragility should be used.

With the revised SCDF for the SBO sequence estimated, it can be added the SCDFs for the other sequences to estimate the revised plant SCDF considering the revised battery charge fragility.

The SCDF will still be based on the LLNL 1993 seismic hazard. To compute the revised SCDF based on the USGS hazard, it is necessary to develop a revised plant fragility. This can be done by specifying an appropriate composite uncertainty and finding the median capacity that will give the revised SCDF using the LLNL hazard (denoted as Method lb in NRC GI-199 report).

The revised fragility can then be convolved Steps in Revising the SCDF to Reflect Revised Component HCLPF

1. Determine the most significant accident sequence

2. Determine lowest HCLPF component in sequence

3. Compute the revised component HCLPF Page 2 of 3

W Indian Point Energy Center Position Paper 201 1-ISRT-09 Plan for Adjusting SCDF without Re-Running SHIP Model

4. Compute the revised component fragility from the HCLPF

5. Convolve the revised fragility with the LLNL eismic hazard to compute the revised SCDF for the sequence

6. Add the SCDFs for the other sequences to compute the revised plant SCDF

7. Compute the revised plant fragility using NRC Method lb

8. Convolve the revised plant fragility with the USGS hazard to compute the revised plant SCDF based on the USGS hazard

9. Compare the revised SCDF to the number reported in the NRC GI-199 report

10. Repeat for additional components and/or sequences Page 3 of 3