RS-20-133, Response to Request for Additional Information Re License Amendment Request to Renewed Facility Operating Licenses to Adopt 10 CFR 50.69

Response to Request for Additional Information Re License Amendment Request to Renewed Facility Operating Licenses to Adopt 10 CFR 50.69

ML20290A791
Person / Time
Site:	LaSalle
Issue date:	10/16/2020
From:	Demetrius Murray Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
EPID L-2020-LLA-0017, RS-20-133
Download: ML20290A791 (45)

Text

Exelon Generation(,

4300 Winfield Road Warrenville, IL 60555 www.exeloncorp.com 10 CFR 50.90 10 CFR 50.69 RS-20-133 October 16, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374

Subject:

Response to Request for Additional Information regarding LaSalle License Amendment Request to Renewed Facility Operating Licenses to Adopt 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors" (EPID L-2020-LLA-0017)

References:

1. Letter from D. Murray (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Application to Adopt 10 CFR 50.69, "Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors," dated January 31 , 2020 (ML20031E699)

2. Letter from B. Vaidya (Project Manager, U.S Nuclear Regulatory Commission) to B. Hanson (Exelon Generation Company, LLC), "Lasalle County Station, Unit Nos. 1 And 2 - Request For Additional Information Regarding License Amendment Request To Renewed Facility Operating Licenses To Adopt 10 CFR 50.69, "Risk-Informed Categorization And Treatment Of Structures, Systems, And Components For Nuclear Power Reactors' (EPID L-2020-LLA-0017)," dated September 3, 2020 (ML20240A218)

3. Letter from B. Vaidya (Project Manager, U.S Nuclear Regulatory Commission) to B. Hanson (Exelon Generation Company, LLC), "Lasalle County Station, Units 1 And 2 - Correction To Request For Additional Information Regarding License Amendment Request To Renewed Facility Operating Licenses To Adopt 10 CFR 50.69, "Risk-Informed Categorization And Treatment Of Structures, Systems, And Components For Nuclear Power Reactors"' (EPID L-2020-LLA-0017), dated September 17, 2020 (ML20253A342)

In Reference 1, Exelon Generation Company, LLC (EGC) submitted a request to the U.S.

Nuclear Regulatory Commission (NRC) for an amendment to Renewed Facility Operating License Nos. NPF-11 and NPF-18 for LaSalle County Station, Units 1 and 2 (LSCS).

October 16, 2020 U.S. Nuclear Regulatory Commission Page 2 EGC's proposed license amendment request (LAR) would modify the licensing basis by the addition of a license condition to allow for the implementation of the provisions of Title 10 of the Code ofFederal Regulations (10 CFR), Section 50.69, "Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors."

On September 3, 2020, the NRC provided a Request for Additional Information (RAI)

(Reference 2) to support its continued review of Reference 1. On September 17, 2020, the NRC amended its Reference 2 letter to revise the due dates for its requests for additional information (Reference 3). to this letter contains the NRC's request for additional information along with EGC's response to questions 3, 4, 5, 6, 7, 9, and 10. Attachment 2 contains the EPRI report markups. Attachment 3 contains a List of Acronyms used in this letter.

EGC has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the NRC in Reference 1. The supplemental information provided in this letter does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. Furthermore, the supplemental information provided in this letter does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

There are no regulatory commitments contained in this letter.

Should you have any questions regarding this letter, please contact Jason Taken at 630-806-9804.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 16th day of October 2020.

Dwi M ray Sr. Manager, Licensing Exelon Generation Company, LLC : Response to Request for Additional Information : EPRI Report Markups : List of Acronyms cc: NRC Regional Administrator- Region Ill NRC Senior Resident Inspector- LaSalle County Station NRC Project Manager, NRR - LaSalle County Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 Response to Request for Additional Information

ATTACHMENT 1 Response to Request for Additional Information APLC 50.69 RAI No. 3:

Section 3.2.3 of the enclosure to the LAR includes a summary and flowchart (Figure 3-2 in the LAR) of the process to be followed for additional evaluations under the proposed approach . Additional information on the process is provided in Section 2.3.1 of the EPRI report. The discussion of the "screening" in Step 3 of Section 2.3.1 of the EPRI report states that "[t]hese screening decisions .. . likely will involve cost/benefit decisions in terms of how best to complete the sensitivity study." Information on what "cost/benefit decisions" would be included, how such considerations would alter the implementation of the proposed approach, and the justification for such decisions as well as the resulting changes is unavailable either in the LAR or the EPRI report.

Step 3a of Section 2.3.1 of the EPRI report mentions inherently rugged components, cites Appendix H of Nuclear Energy Institute (NEI) 12-06, Revision 4 (ADAMS Accession No. ML163548421 ), and lists certain SSCs as being inherently rugged. Appendix H of NEI 12-06, Revision 4, refers to other documents. It is unclear whether the list under step 3a of Section 2.3.1 of the EPRI report encompasses the inherently rugged components that are included in other documents referenced in Appendix H of NEI 12-06, Revision 4. The basis of consideration of such components as inherently rugged is also not provided.

Step 2 of Section 2.3.1 of the EPRI report discusses component classification and cites two different documents as resources for such classification . It is unclear to the NRG staff whether the classification(s) in the two cited documents consistent and the impact of any potential differences on the implementation of the proposed approach.

a. Discuss the justification for, intent of, and implementation of the consideration of the "cost/benefit decisions" in the screening process for the proposed alternate seismic approach. Include an explanation of how "cost/benefit decisions" would change the licensee's proposed approach and justify such changes .

EGC RESPONSE 3a:

Screening steps 3a , 3b, and 3c in the Seismic Correlated Failure Assessment flowchart are elective and can be used to reduce the scope of SSCs considered in the correlation assessment. The screening steps are intended to be applied similar to the screening processes employed in seismic PRAs (SPRAs) per Part 5 (Seismic PRA) of PRA standards ASME/ANS RA-S-2009a endorsed by NRC Reg Guide 1.200 Rev. 2, ASME/ANS RA-S-2013b endorsed by NRG (ML12319A074) for use with EPRI 1025287 (SPID), and ASME/ANS RA-S CASE 1 (aka "SPRA Code Case") accepted by NRC letter (ML18017A964). The screening steps result in only those SSCs that have the appropriate seismic risk considerations being included in the sensitivity study.

The phrase "cost/benefit decisions" will be removed from the EPRI report and the first paragraph describing Step 3 in the report will be revised to focus on the degree that the effort to screen out any SSC is less than the effort to keep the SSC on the list to be considered in the correlated failure assessment. EPRI report mark-ups are provided in Attachment 2. The EPRI mark-ups cited for this RAI response and in other RAI responses herein will be incorporated into a subsequent EPRI publication.

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ATTACHMENT 1 Response to Request for Additional Information

b. Provide the list of inherently rugged SSCs that will be used in Step 3a of Section 2.3.1 of the EPRI report. Include the basis of consideration of SSCs in the list as inherently rugged.

EGC RESPONSE 3b:

The EPRI Seismic Alternative report (3002017583 1 , 2018) in item 3a of Section 2.3.1 cites NEI 12-06, Revision 4, App. H, Seismic Mitigating Strategies Assessment (S-MSA, 2016) for a list of inherently rugged SS Cs since it was current at that time. However, the text in the EPRI report for step 3a of Section 2.3.1 will be revised per mark-ups provided in Attachment 2 to cite the more recent EPRI 3002012994, Seismic Fragility and Seismic Margin Guidance for SPRAs (2018). EPRI 3002012994 is a compilation of many prior industry documents and its Table 6-2 contains an updated list of inherently rugged SSCs including a basis for each SSC type.

The following information is added to Section 6.0, References of the original LAR for clarification.

• Reference 63: Seismic Fragility and Seismic Margin Gwdance for Seismic Probabilistic Risk Assessments. EPRI , Palo Alto, CA: 2018 . 3002012994.

c. Provide the list of component classification groups that will be used in Step 2 of Section 2.3.1 of the EPRI report.

EGC RESPONSE 3c:

Section 6.1.3 in EPRI 3002012994 provides a list of equipment classes that should be used in Step 2 of the Seismic Correlated Failure Assessment. The EPRI report will be revised per mark-ups provided in Attachment 2 to use that reference for the equipment classes.

Reference 1: Alternative Approaches for Addressing Seismic Risk in 10 CFR 50. 69 Risk-Informed Categorization. EPRI , Palo Alto, CA: 2020. 3002017583.

APLC 50.69 RAI No. 4:

Section 2.3.1 of the EPRI report provides the process for additional evaluation in the licensee's proposed alternate seismic approach . Step 8 of the process states that the loss-of-offsite power (LOOP) and small loss-of-coolant accident (SLOCA) initiators in the Full Power Internal Event (FPIE) PRA will be used , in conjunction with surrogate events, to determine the impact of seismic-specific failures of SSCs following the walkdown. The basis for limiting the evaluation to only the LOOP and SLOCA initiators is not provided. It appears that the "success path" approach for seismic margins analysis (SMA) used in the Individual Plant Examination for External Events (IPEEE) is used to identify the initiators. However, the proposed approach is based on insights from seismic PRAs and not SMAs. Seismic PRAs include several initiators in addition to LOOP and SLOCA.

1 EPRI 3002017583 is a February 2020 technical update (see Reference 1) that provides additional clarification into EPRI 3002012988. LaSalle will use the technical update report EPRI 3002017583 for this and all future RAI responses for the 10 CFR 50.69 application .

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ATTACHMENT 1 Response to Request for Additional Information The initiating frequencies for the LOOP and SLOCA initiators in step 8 of Section 2.3.1 of the EPRI report are provided as 1.0 and 1E-2 per year, respectively. The assumptions used for the initiating frequency of SLOCA in the proposed approach are unclear and the basis for its generic applicability is not provided. The insights from the Plant A, C, and D, case studies that support the licensee's proposed approach do not appear to inform the selection of the initiators and initiating frequencies . In addition , step 8a of Section 2.3 .1 of the EPRI report states "[o]ther appropriately justified values for small LOCA frequency may be used" but does not provide any information about the basis for such values. Step 8c indicates that the sequence to be included in the evaluation is the SLOCA-LOOP sequence (i.e., SLOCA with conditional LOOP) rather than the SLOCA sequence.

The discussion in step 8 of Section 2.3.1 of the EPRI report does not provide any information about including changes to key internal events modeling assumptions due to seismic-specific impacts. Examples of such impacts include, but are not limited to, non-recovery of offsite power and non-recovery of direct current power. Such impacts contribute to the results from an SPRA and, therefore, are expected to be contributing factors to the insights from the plant's A, C, and D, case studies in the EPRI report. It is unclear why such impacts are not being considered in the sensitivity study for the licensee's proposed approach.

a. Discuss how other seismically induced failures (e.g., medium-sized (MLOCA), modeling assumptions (e.g., non-recovery of LOOP), and all impacts of the seismically induced failure of the SSC being categorized were considered for inclusion in the FPIE SLOCA and LOOP event trees . If the sensitivity study excludes such failures and assumptions, justify their exclusion from the proposed approach.

EGC Response 4a:

Selection of LOOP and SLOCA Sequences Since the EPRI Tier 2 plants do not have a seismic PRA, a pseudo-deterministic approach is applied using insights from EPRI Tier 2 seismically-biased adjustments and quantification of the plant FPIE PRA. Insights from the EPRI Tier 3 plant SPRAs as well as other industry analyses (e.g . EPRI SPRA Implementation Guide and NTTF 2.1 seismic PRA submittals) were used to determine which seismic accident sequences would capture the important insights of seismic impacts on SSC categorization. The LOOP and SLOCA accident sequences were selected as the focus of the EPRI Tier 2 guidance for the following reasons:

1. Offsite power fragility (primarily due to ceramic insulator failure) is one of the lowest fragilities in a NPP SPRA. Typically, 95-100% of the calculated SCDF and SLERF of SPRAs involve seismic-induced site LOOP.

2. The LOOP and SLOCA accident sequences challenge the majority of critical safety functions (CSFs). NOTE: The seismic SLOCA accident sequences are actually seismic LOOP-SLOCA sequences (i.e., offsite AC power failed as well) .

3. Seismic-induced medium and large LOCAs use many of the same CSFs as Small LOCA accidents (refer to comparison table later in this response). In addition, the seismic fragilities of these other LOCA sizes are often modeled (in past and present NTTF 2.1 SPRAs) with the same or higher seismic capacities as small LOCA seismic fragility (thus equal or lower likelihood of occurrence). As such, modeling the potential Page 3 of 26

ATTACHMENT 1 Response to Request for Additional Information break size spectrum of seismic LOCA in the EPRI Tier 2 adjusted PRA model as a Small LOCA with the adjusted higher occurrence frequency would capture the most safety functions of the LOCA spectrum. Note: Excessive LOCA (e.g., seismic induced RPV support failure) is also typically a high seismic capacity fragility and offers no useful risk insights with respect to 50.69 component risk categorization as such scenarios are typically modeled in PRAs as leading directly to core damage.

4. The NRC GL 88-20 Supplement 4 IPEEE seismic margins analysis (SMA) and the development of the seismic shutdown equipment list (SSEL) also focus on LOOP and SLOCA accident sequences. The most common SMA guideline used for IPEEE SMAs, EPRI NP 6041, states " The most likely sequence of events following a margin earthquake will be similar to those of the loss of offsite power (LOOP) event ..[and] ...

at least one success path be capable ofmitigating a small-break /oss-of-coo!ant-acc1dent"

5. The Tier 2 adjusted model quantification also includes the sequence transfers (and associated accident sequences) resulting from further probabilistic complications in the LOOP and SLOCA accident progression (i.e., SBO, LOOP/SBO-ATWS, and SLOCA-ATWS). Including a "failure to scram" fragility is not one of the Tier 2 guidance base adjustments to the FPIE PRA model but this is because of the typical high capacity of reactor internals and the fail safe aspect of reactor scram (i.e., typically low likelihood of seismic-induced failure to scram). However, control rod seismic surrogate correlated failure event development would be captured , in theory, by EPRI Tier 2 step 5a during 50.69 assessment of a system boundary including the control rods ; however, the control rods and reactor internals almost assuredly are already defined as HSS for all plants and thus screened out in EPRI Tier 2 Step 3c.

6. Desire to develop a sensitivity study that captures the dominant seismic risk while also keeping the sensitivity study from becoming overly complicated.

CSFs vs LOOP and LOCA Accident Sequences As mentioned above, the LOOP and SLOCA accident sequences challenge the most critical safety functions and their associated SSCs. BWR and PWR CSFs versus LOOP and LOCA accident sequence types are summarized in the table below.

CSF LOOP SLOCA MLOCA LLOCA (BWR) Reactivity Control x x x x (BWR) RCS Pressure Control x x Note (1) Note (1)

(BWR) RCS Inventory Control (High Pressure) x x Note (2) Note (3)

(BWR) RCS Depressurization x x Note (2) Note (3)

(BWR) RCS Inventory Control (Low Pressure) x x x x Page 4 of 26

ATTACHMENT 1 Response to Request for Additional Information CSF LOOP SLOCA MLOCA LLOCA (BWR) Containment Control (isolation, decay heat removal) x x x x (PWR) Reactivity Control x x x x (PWR) RCS Pressure Control x x Note (1) Note (1)

(PWR) RCS Inventory Control x x Note (2) Note (4)

(PWR) Heat Removal/Core Cooling x x Note (4) Note (4)

(PWR) Containment Integrity x x x x Notes to Table:

(1) The Medium and Large LOCA size breaks in the RCS in effect fulfill the initial RCS pressure control function and no relief valve operation is necessary.

(2) For Medium LOCAs in a BWR, operation of high pressure coolant makeup systems (e.g., High Pressure Coolant Injection (HPCI)) fulfill the high pressure inventory control function as well as the RCS Depressurization. Active RCS Depressurization methods are required if the high pressure inventory control function is failed . Likewise for the RCS Inventory Control function for PWRs with respect to the High Pressure Safety Injection (HPSI) system.

(3) For Large LOCAs in a BWR, the approximately immediate depressurization of the RCS by the large break bypasses the applicability of the high pressure inventory control and RCS depressurization safety functions.

(4) For Large LOCAs in a PWR, the HPSI system may or may not be used in the core cooling CSF as the RCS is depressurized such that the Low Pressure Safety Injection (LPSI) system would be used. Likewise, for the larger Medium LOCA breaks where the RCS would be depressurized. But the seismic capacity of these large RCS breaks is typically much higher than for the Small LOCAs and therefore, their contributions were considered to be very low for this study. It is noted that some plants may credit the LPSI system for Small LOCAs after RCS depressurization.

Seismic-Induced Impacts As discussed in step 8 of the flow chart in Figure 2-3 of the EPRI report, the FPIE PRA model is modified as follows and quantified for CDF and LERF to reflect typical seismic impacts included in probabilistic seismic risk assessments:

• Set site LOOP initiating event frequency to 1.0 (including the consequential site LOOP basic event)

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ATTACHMENT 1 Response to Request for Additional Information

• Set small LOCA initiating event frequency to 1.0E-02

• Set all other initiating event frequencies to 0

• Set basic events for offsite power recovery to 1

• Set basic events for other functional recoveries (for example, EOG recovery, DC power recovery) to 1

• Set basic events for new surrogates added to the FPIE to 1.0E-04

• Set Emergency AC (EAC) power surrogate runs times to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Step 8 of the flow chart in Figure 2-3 of the EPRI report is revised per mark-ups provided in Attachment 2 to reflect that some plants' FPIE PRA models include recovery of loss of DC power and that restoration credit should not be taken in the FPIE PRA models in this process since recovery of DC power after a seismic event is not generally credited in a seismic event.

b. Provide justification for the selected failure probability and the selected initiating frequencies for the LOOP as well as the SLOCA initiator using the insights from seismic PRAs, including the insights from the Plant A, C, and D, case studies used to develop the proposed approach.

EGC Response 4b:

The approach to applying surrogate events representing correlated or interaction failures to the full power internal events (FPIE) PRA is described in Section 2.3.1 of the EPRI document. Step 7 of the process involves adding new seismic surrogate events to the FPIE PRA logic model for the potential seismically correlated conditions identified in the previous steps (i.e., during the walkdowns). The recommended value in Step 7 is 1.0E-

04. Step 8.a of Section 2.3.1 states that the recommended event frequency for the LOOP initiator is 1.0. It should be noted that from a purely calculation standpoint, this is what is being performed in the EPRI process. However, what this really represents is setting the frequency of the LOOP initiator to a value of 1.0E-04/yr and setting the probability of correlated failures at that Annual Exceedance Probability (AEP) to 1.0. While these values appropriately represent the LOOP frequency and a conservative surrogate event failure probability (i.e. a probability of 1.0), they cannot be used directly in the proposed process because if the surrogate events were assigned a probability of 1.0, then they would have a Risk Achievement Worth (RAW) value of 1.0. To avoid this issue, the LOOP frequency of 1.0E-04 and the surrogate failure probability of 1.0 are switched to allow importance data to be generated for the surrogate events. As such, the discussion below addresses the justification for setting the LOOP initiator frequency to 1.0E-04 and setting the SLOCA frequency to 1.0E-02, based on the insights from the Plant A, C, and D, case studies.

LOOP Frequency:

An analysis was performed for the use of the 1.0E-04 value for the LOOP frequency Page 6 of 26

ATTACHMENT 1 Response to Request for Additional Information (combined with the probability of the surrogate failures, which is conservatively assumed to be 1.0) to show that the 1.0E-04 value is consistent with the actual values from case study plants. Based on the data developed for the case study plants, the seismic hazard (PGA vs AEP) was convolved with a single component having the seismic capacity for a seismic-induced LOOP. The result is the AEP (i.e. interval midpoint) for each PGA multiplied by the conditional failure probability of a LOOP at that PGA. Table 1 shows the HCLPFs of the seismic-induced LOOPs in the three case study plants along with the sum of all PGA intervals of the convolved hazard with the LOOP failure probability.

Table 1 - Seismic LOOP Frequency Plant LOOP HCLPF Am Br Bu LOOP Freq. (/yr)

A 0.09 0.3 0.30 0.45 7.00E-05 c 0.10 0.3 0.40 0.27 4.84E-04 D 0.09 0.3 0.30 0.45 5.58E-06 Average 0.09 0.3 0.33 0.39 1.86E-04 Note: Average of plant's A and D is approximately 3.8E-05/yr From the table it can be seen that for each of the case study plants the total seismic hazard exceedance frequency is less than 1.0E-04, with the exception of plant C, which has the highest seismic hazard of the three Tier 3 plants. The seismic hazard for tier 2 plants is lower than that for the tier 3 plants and therefore, the convolution results are expected to be bounded by the 1E-04 value used in this study.

SLOCA Frequency:

A similar analysis was performed for the use of the 1.0E-02 value for the SLOCA frequency to show that the 1.0E-02 value is consistent with the actual values from case study plants. Using the same approach as was used in the LOOP convolution above, the plant seismic hazard was convolved with the SLOCA HCLPF for each of the case study plants. Table 2 shows the SLOCA HCLPF along with the sum of all PGA intervals of the convolved hazard with the SLCOA failure probability.

Table 2 - Seismic SLOCA Frequency Plant SLOCA HCLPF Am Br Bu SLOCA Freq.

A 0.78 2.69 0.35 0.4 9.29E-07/yr c 0.90 2.54 0.3 0.33 4.35E-07/yr D 1.6 3.65 0.24 0.26 3.03E-09/yr Average 1.09 2.96 0.30 0.33 4.56E-07/yr From the table it can be seen that the SLOCA frequency for the case study plants, the Page 7 of 26

ATTACHMENT 1 Response to Request for Additional Information total seismic hazard exceedance frequency is less than 1.0E-06, for each of the three plants, with the average SLOCA frequency for the case study plants being approximately 5E-07/yr.

Based on the above values, the use of the 1.0E-02/yr as a surrogate for the seismically-induced SLOCA frequency, when combined with the 1.0E-04 probability of the surrogate events is considered to be appropriate and conservative for typical Tier 2 plants, based on the seismic hazard convolutions of the case study plants in the EPRI report.

References :

Seismic hazard curves are for the three case study plants are from references 1 - 3 below. The LOOP and SLOCA seismic capacities are from the SPRA submitted to the NRC in response to NTTF 2.1 seismic re-evaluation in references 4 - 6 below.

1. Plant A Seismic Hazard - ADAMS Report ID ML18240A065 (Table 3.1.1-1)

2. Plant C Seismic Hazard - ADAMS Report ID ML14092A019 (Table A-1 a)

3. Plant D Seismic Hazard -ADAMS Report ID ML14098A478 (Table A-1a (page E4-A2))

4. Plant A NTTF 2.1 SPRA Report --ADAMS Report ID ML18240A065 (Table 5.4-2)

5. Plant C NTTF 2.1 SPRA Report-ADAMS Report ID ML17088A130 (Table 5.4-4)

6. Plant D NTTF 2.1 SPRA Report-ADAMS Report ID ML17181A485 (Table 5.4-3)

c. Explain the basis for use of "[o]ther appropriately justified values for small LOCA frequency initiating frequency" SLOCA or SLOCA-LOOP sequences and how the proposed categorization process will ensure that the use of the "other" value is accepted by the NRC staff for the proposed alternate seismic approach (i.e., how the response to item (c) can be provided for such "other" value(s)).

EGC Response 4c:

The proposed approach of using the 1E-02/yr value for the SLOCA frequency is considered to provide conservative results. More realistic seismic-induced SLOCA frequency may be developed using accepted industry fragility analyses. But for this study, the 1E-02 will be used and other fragility analyses for this SLOCA frequency will not be allowed. Step Ba in the EPRI report has been revised per mark-ups provided in Attachment 2 to remove the sentence that allowed use of other values.

The proposed approach utilizes walkdowns to identify correlated failure and interaction concerns and then use a quantitative sensitivity calculation to determine their impact on the categorization of SSCs. The sensitivity calculation is a key step in the proposed alternate seismic approach and the resulting input to the categorization of SSCs. The proposed approach uses a failure probability for the surrogate events of 1E-4. Section 2.3.1 of the EPRI report states that the value is based a "typical" total seismic hazard exceedance frequency above which SPRAs would typically model loss of offsite power and for which correlated Page 8 of 26

ATTACHMENT 1 Response to Request for Additional Information failures may be likely. The licensee's proposed approach also states that the failure probability can be "justified based on the hazard." Further, it appears that the same surrogate event failure probability is applied regardless of the SSC and seismic-specific failure mode that the event is supposed to represent.

d. Demonstrate that the proposed surrogate sensitivity study using a single failure probability for the surrogate event, the two initiating events, the corresponding initiating event frequencies, and any changes due to response to items (a) through (c) above, results in categorization input that is equivalent, or conservative compared to corresponding results from a SPRA. An example of such a demonstration is to use PRAs for one of the plants in the case studies in the EPRI report and compare categorization outcome from the SPRA for that plant and the surrogate sensitivity study using the FPIE for that plant. The NRC staff recognizes that the demonstration is not an actual categorization.

EGC Response 4d:

In response to RAI APLC 4(d) on the LaSalle 50.69 LAR, Exelon RM has performed the requested demonstration of the EPRI 3002012988 Tier 2 alternative seismic approach for a selected system SSC 50.69 categorization and has compared those results to the results produced by a seismic PRA (SPRA) model. The conclusion from this demonstration analysis is that for the selected system analyzed the EPRI 3002012988 Tier 2 alternative seismic approach produces equivalent results (in terms of number of system components identified as HSS from a seismic risk perspective) than would otherwise be produced from the plant SPRA. This demonstration analysis is discussed further below.

Based on the August 18, 2020 clarity teleconference between Exelon and the NRC, the NRC noted:

• The LaSalle demonstration during the June 2020 audit provided much information on the EPRI 3002012988 process

• The response to RAI APLC 4d may use insights from Steps 1 thru 5 of the LaSalle SBLC demonstration to assist the Plant A demonstration

• This requested Plant A based demonstration is acknowledged not to be an actual 50.69 categorization As such, similar to the June 2020 audit demonstration for the LaSalle SBLC system, the response to this RAI 4d uses the Plant A Standby Liquid Control (SLC) system as well as insights from Steps 1 through 5 of the June 2020 audit demonstration.

Tier 2 Demonstration Analysis for Plant A SLC System Based on consideration of the seismic surrogate groups identified in the LaSalle June 2020 audit demonstration and the Plant A SLC system configuration (no walkdowns or other detailed investigations are performed for this demonstration), six (6) seismic surrogate events (4 correlated and 2 interaction) are identified for the Plant A SLC system and summarized in Table APLC4d-1.

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ATTACHMENT 1 Response to Request for Additional Information The seismic surrogate basic events shown in Table APLC4d-1 were manually inserted into the latest official version of the Plant A FPIE PRA single-top fault tree logic at the appropriate logic gates to fail the intended aspects of the SLC system. Additional PRA modeling adjustments were made consistent with EPRI 3002012988 Tier 2 Step 8 guidance, as follows:

• Adjust LOOP and Small LOCA IE Frequencies to 1.0 and 1E-2. Respectively: Plant A FPIE PRA initiator basic events adjusted are %LOOP-GRID, Grid Centered LOOP Initiating Event, which is modeled in the PRA as causing a site LOOP condition and the %S2, Small LOCA Initiating Event.

• Other IE Frequencies Set to 0: All other initiators in the Tier 2 adjusted PRA logic model are set to a frequency value of 0.0.

• Set Conditional LOOP Probability to 1: Plant A basic event EPHOSPTRIPO, Consequential Loss of Offsite Power Given Plant Trip, set to 1.0. This is the conditional LOOP basic event used in small LOCA accident sequences.

• Disable Offsite AC and Other Functional Recoveries: All such recoveries in the Plant A PRA (offsite power recoveries, EOG recoveries) are set to a logic TRUE value (i.e.,

directly failed) for the Tier 2 quantification. No other functional recoveries beyond these two types are modeled in the Plant A PRA LOOP/SBO and small LOCA accident sequence logic.

• Surrogate Run Times Increased to PRA Mission Time: All such basic events in the Plant A PRA (fifteen EOG Failure to Run basic events) were adjusted to increase the probability exposure time to 24 hrs for the Tier 2 quantification.

The Tier 2 adjusted PRA model is quantified in the same manner as the Plant A FPIE PRA single-top quantification (i.e., same Boolean engine quantifier, same 1E-12 quantification truncation level, same mutually-exclusive file, and then the adjusted flag and recovery files). Consistent with NEI 00-04 component risk importance calculations, Tier 2 component risk information is based the various applicable failure modes, i.e., the seismic surrogate basic events added into adjusted FPIE PRA model as well as applicable non-seismic failure modes. These Tier 2 based component risk importances are compared against the NEI 00-04 high safety significance, HSS, thresholds. As can be seen from Table APLC 4d-1, no Plant A U-2 SLC component is identified as HSS from the EPRI 3002012988 Tier 2 method in this demonstration.

Comparison of Tier 2 Based Component Importances to SPRA Based Importances Per the NRC RAI request, risk results from the SPRA model used for the Plant A NTTF 2.1 Seismic submittal are used for comparison to the risk results produced by the EPRI 3002012988 Tier 2 approach. The risk importance results (FV and RAW) from the Plant A SPRA are based on the EPRI ACUBE software output post-processing of the SPRA results.

Table APLC4d-2 provides a summary comparison of the Plant A Unit 2 SLC component 50.69 risk significance determination based on the EPRI 3002012988 Tier 2 method and the NEI 00-04 method using the Plant A Unit 2 at-power SPRA. As can be seen from Page 10 of 26

ATTACHMENT 1 Response to Request for Additional Information Table APLC4d-2, all Plant A Unit 2 SLC components are determined to be seismic Low Safety Significant regardless of whether the EPRI 3002017583 Tier 2 method is used or whether a full SPRA is used. In the case of the Plant A Unit 2 SLC system, the EPRI 3002012988 Tier 2 alternative seismic approach produces equivalent results (in terms of number of system components identified as HSS from a seismic risk perspective) than would otherwise be produced from the plant SPRA.

Table APLC4d-1 SLC Seismic Surrogate Events Added to Plant A U-2 FPIE PRA Basic Event ID Basic Event Description Probability Comment SEVS-14XCPl2 Seismic Surrogate 1E-4 Same component types (Correlated): Seismic FTO of located nearby one another SLC Injection Squib Valves 2- on the same elevation, RB 11-14Aand B 195 ft.

SPMSP40XCSI Seismic Surrogate 1E-4 Same component types 2 (Correlated): Seismic Failure located nearby one another to Operate of SLC Pumps 2A, on the same elevation, RB BP40 195 ft.

SRVS- Seismic Surrogate 1E-4 Same component types 39XCOl2 (Correlated): Seismic Failure located nearby one another (Failed Open) of SLC Pump on the same elevation, RB Relief Valves 2-11-39A and B 195 ft.

SACS-76XCFl2 Seismic Surrogate 1E-4 Same component types (Correlated): Seismic Failure located nearby one another (Rupture) of SLC Pump on the same elevation, RB Discharge Pulsation 195 ft.

Dampeners 2AT76A(B)

SPMSP40XCRI Seismic Surrogate 1E-4 No such seismic interaction 2 (Interaction): Seismic Failure actually exists for Plant A but, of SBLC Test Tank Impacts similar to LS June 2020 audit, and Disrupts SLC Pump interaction assumed for Suction demonstration.

SPMSP40XCW Seismic Surrogate 1E-4 No such seismic interaction 12 (Interaction): Seismic Failure actually exists for Plant A but, of Block Wall Impacts SLC similar to LS June 2020 audit, Pumps 2A, BP40 interaction assumed for demonstration.

Notes to Table APLC4d-1:

(1) A seismic surrogate correlation group is not identified for the SLC system heating equipment. Failure of the SLC heating function at t=O does not fail the Page 11of26

ATTACHMENT 1 Response to Request for Additional Information SLC function. The SLC system would be used within t<2 hrs and the sodium pentaborate will still be within solution at that time.

(2) Check valves and manual valves in the LS SLC system have been screened as inherently rugged (per EPRI Tier 2 step 3a) and no seismic surrogate correlated groups are identified for these. In addition, unlike LaSalle, the Plant A SLC system does not contain MOVs and thus no seismic surrogate correlated groups are identified for MOVs.

Table APLC4d-2 Comparison of EPRI Tier 2 Based vs SPRA Based Plant A U-2 SLC Component Seismic Risk Importance Determinations High Safety Significant (Seismic)

EPRI Tier2 Plant A Plant A Unit 2 SLC Component Step 9 <1> SPRA SLC Injection Explosive Valve 2-11-14A No No SLC Injection Explosive Valve 2-11-148 No No SLC Pump 2AP40 No No SLC Pump 28P40 No No SLC Pump Relief Valve 2-11-39A No No SLC Pump Relief Valve 2-11-398 No No SLC Discharge Pulsation Damper 2AT76A No No SLC Discharge Pulsation Damper 2AT768 No No SLC Train A Discharge Manual Valve 2-11-13A No (2) No SLC Train 8 Discharge Manual Valve 2-11-138 No (2) No SLC Train A Discharge Check Valve 2-11-43A No (2) No SLC Train 8 Discharge Check Valve 2-11-438 No (2) No SLC Train A Suction Manual Valve 2-11-12A No (2) No SLC Train 8 Suction Manual Valve 2-11-128 No (2) No SLC Injection Manual Valve 2-11-18 No (2) No LC Injection Manual Valve 2-11-15 No (2) No SLC Main Tank Manual Valve 2-11-11 No (2) No SLC Main Tank 20T18 No (2) No Page 12 of 26

ATTACHMENT 1 Response to Request for Additional Information Table APLC4d-2 Comparison of EPRI Tier 2 Based vs SPRA Based Plant A U-2 SLC Component Seismic Risk Importance Determinations High Safety Significant (Seismic)

EPRI Tier2 Plant A Plant A Unit 2 SLC Component Step 9 <1> SPRA SLC Injection Check Valve 2-11-16 No (2) No SLC Injection Check Valve 2-11-17 No (2) No SLC Handswitch 11A-S001 No (2) No Other SLC Components (Note 3) No (3) No (3)

Notes to Table APLC4d-2 :

(1) The component risk importances are based on the seismic surrogate basic events and applicable non-seismic failure mode basic events (summation of FV values and maximum of RAW values). The list of components for which seismic risk importances are produced is limited to those components appearing in seismic surrogate groups.

(2) These Plant A Unit 2 SLC components are screened from the EPRI 3002012988 Tier 2 quantification due to assessed low seismic risk significance.

(3) Other SLC components not shown in this table are not explicitly modeled in the PRA and for the purposes of this Tier 2 process demonstration are reasonably assessed as Low Safety Significance.

Page 13 of 26

ATTACHMENT 1 Response to Request for Additional Information

e. Explain how failure probability/ies other than a fixed value, determined based on the response to item (d), will be developed and "justified based on the hazard" for implementation in the proposed approach and how the use of the "other" value is accepted by the NRC staff for the proposed alternate seismic approach (i.e., how the response to item (e) can be provided for such "other" value(s)).

EGC Response 4e:

The proposed approach of using a seismic-induced failure probability of 1E-04 for the surrogate event, although relatively simplistic, is considered to provide conservative results. More realistic seismic-induced failure probabilities may be developed for the surrogates using accepted industry fragility analyses (e.g. Separation of Variables or the Hybrid approach) as described in Appendix B of the EPRI report. The EPRI report Appendix B was revised per mark-ups provided in Attachment 2 to include clarification on the accepted fragility methods that can be used . In addition, step 5c in section 2.3.1 of the EPRI report was revised per mark-ups provided in Attachment 2 to include clarification that instead of screening SSCs out, the SSC may remain in the modified FPIE model but with a seismic-induced failure probability from a plant-specific fragility analysis.

The proposed alternate seismic approach includes addition of surrogates to the FPIE PRA under the appropriate areas in the logic model (Step 7 of Section 2.3 .1 of the EPRI report) .

The approach likens the surrogates to a common-cause failure mode. The FPIE PRA usually includes common-cause failure events under a particular random failure mode for an SSC (e.g., common-cause failure of pumps to run). However, the seismic correlated or interaction failures fail the SSC independent of a random failure mode of the SSC (i.e. , the seismic failure is under the so-called top gate for the SSC).

The proposed alternate seismic approach compares the results of the sensitivity for each surrogate event to the results to the F-V and RAW HSS criteria for common cause failure in the FPIE PRA from NEI 00-04 (that is, F-V > 0.005 or RAW> 20). The proposed approach and the use of sensitivity for seismic-specific failure modes is based on the insights from the case studies in the EPRI report. It appears to the NRC staff that the objective of the sensitivity is to obtain categorization inputs like those that would be obtained from a seismic PRA for seismic-specific failure modes. However, the categorization using the seismic PRAs in the case studies in the EPRI report are not based on the common cause failure but rather the individual SSC failure . As an example, Section 3.4.2 .2 of the EPRI report, which discusses the identification of HSS SSCs from the Plant C SPRA, states that "[c]omponents are considered HSS if the group F-V is greater than 0.005 or if the group RAW is greater than 2.0 for core damage frequency (CDF) or large early release frequency (LERF)." Therefore, it appears that using different thresholds from those used for the case studies, which provide the insights supporting the proposed approach , is (i) not supported by the basis for the proposed approach, and (ii) can result in categorization inputs different those that would be obtained from a seismic PRA for seismic-specific failure modes, which is contrary to the purpose of the proposed approach.

f. Justify the use of the criteria for common cause failure for determination of HSS SSCs from the proposed sensitivity study given that the case studies providing the underlying insights for the proposed approach used different F-V and RAW criteria and that the proposed sensitivity study includes both correlated as well as interaction failure modes.

Alternately, discuss any updated F-V and RAW criteria for use with the proposed sensitivity study and discuss their consistency with the case studies in the EPRI report as Page 14 of 26

ATTACHMENT 1 Response to Request for Additional Information well as NEI 00-04.

EGC Response 4f:

The trial characterizations in the EPRI report were undertaken to determine the seismic related categorization insights in support of the report. Section 2.3.2 of the EPRI reports states that the trial cases identified a small number of instances where unique seismic insights were identified. These were associated with seismically correlated failures or interaction failures, and led to SSCs that were uniquely HSS from the SPRA. Since the trial evaluations were performed to determine any seismic related categorization insights rather than perform an actual 50.69 SSC categorization, it was decided that a reduced RAW threshold of 2 should be used. This was done to provide added sensitivity to the trial evaluations to ensure that all risk insights that might be provided by a seismic PRA were identified in these evaluations. Had these evaluations used a criterion of RAW >20, it is possible that fewer risk insights would have been identified and evaluated in the EPRI report. As stated in the EPRI document, even using the significantly lower RAW criteria of 2 for all events, the only insights gained were that correlated and interaction failures need to be evaluated. Using this lower criterion for correlated events in the seismic PRA trial evaluations, whose sole intention was to identify seismic-specific risk insights, does not in any way invalidate the use of the appropriate RAW threshold of 20, specified in NEI 00-04 for common cause events, when performing the evaluation of the insights gained from the trial evaluations.

For a system being categorized under 50.69, the approach developed in Section 2.3.1 of the EPRI report identifies the conditions that would be treated as seismically correlated fragilities or interaction consequences if an SPRA were being performed. As noted in the EPRI report, "In a FPIE PRA common cause evaluations behave similarly to seismically correlated conditions in that they model the impacts of multiple SSC failures; therefore, the common cause methods can be employed to determine the necessary insights for 50.69 categorization." This is the basis for the EPRI Tier 2 methodology (i.e., treating correlated events and interaction events, which fail more than one component as common cause events). This is consistent with the process outlined in NEI 00-04 Section 5.3 that would be applied if an SPRA was available. Therefore, where an SPRA is used, correlated events and interaction events would use the HSS RAW threshold of 20, since Section 5.1 of the NEI document states that "the RAW for common cause events reflects the relative increase in CDF/LERF that would exist if a set of components or an entire system was made unavailable". It goes on to say that "As a result, the risk significance of the RAW values of common cause basic events is considered separately from the basic events that reflect an individual component". Based on this, the RAW values for correlated events and interaction events are by definition RAW values of common cause events, and should be treated as such . This is the justification for using a RAW threshold of 20 for high safety significance in the Tier 2 approach described in the EPRI document.

APLC 50.69 RAI No. 5 Step 10 in Section 2.3.1 of the EPRI report states that if the importance measure criteria are met by the surrogate basic events, then the corresponding SSC should be considered HSS.

However, it is unclear whether the comparison against the importance measure criteria and the consequent categorization as HSS will be performed if the results from either of the sensitivities (i.e. , for LOOP and for SLOCA) show the criteria being met or both the Page 15 of 26

ATTACHMENT 1 Response to Request for Additional Information sensitivities need to show that the criteria is met or some combination shows the criteria is met. Further, the basis for the use of any approach (i.e., individual , combined, etc.) for comparison against the criteria is not provided.

Explain how the results from the sensitivity will be used for comparison against the importance measure criteria in NEI 00-04 (i.e., individually, combined, etc.). The explanation should include a justification of how the proposed comparison approach is consistent with the guidance in NEI 00-04 and reflective of the insights from the case studies in the EPRI report.

EGC RESPONSE 5:

Step 10 in Section 2.3.1 of the EPRI report states:

"For each seismic surrogate event, compare the results to the F-V and RAW HSS criteria for common cause components in the FPIE PRA from NEI 00-04 (that is, F-V > 0.005 or RAW >

20). For seismic surrogate events, if the F-V or RAW criteria are met, all SSCs modeled by that surrogate event should be considered HSS."

To clarify, the sensitivity study will be a combined study with the initiators set to the values specified (i.e. LOOP set to 1.0 and SLOCA set to 1E-2). The F-V and RAW importance results will be compared to the importance thresholds as indicated in Step 10 (Step 9 in newer EPRI 3002015783) . These thresholds are consistent with the thresholds identified in NEI 00-04 for common cause in NEI 00-04 section 5.1. Since the correlated failures are common cause type failures, common cause criteria was determined to be appropriate.

For all SSCs that are included in a surrogate event (either a correlated failure event or an integrated failure event) the F-V value for the surrogate event will be added to the F-V values for the non-seismic failure events for that SSC. The resulting F-V will be compared to the criterion for SSCs in the FPIE PRA from NEI 00-04 (that is F-V > 0.005). The maximum RAW value of the surrogate event and any other non-seismic common cause failure events involving that SSC will be compared to the criterion for common cause SSCs in the FPIE PRA from NEI 00-04 (that is RAW> 20). The maximum RAW value for any non-seismic-related, non-common cause failure events of that SSC will be compared to the PRA criterion from NEI 00-04 (RAW> 2). If the SSC exceeds the threshold for any of these three criteria, it will be categorized as HSS.

APLC 50.69 RAI No. 6 Figure 2-3 in the EPRI report depicts the steps and decisions made as part of the licensee's proposed alternate seismic approach . Certain steps in the flowchart in Figure 2-3 are unclear in their intent and scope. The accompanying discussion does not include information to address the lack of clarity in Figure 2-3.

Step 3b of Section 2.3.1 of the EPRI report states that if the SSCs under consideration is not used in safety functions that support mitigation of core damage or containment performance, it can be screened . The EPRI report also provides an example of screening based on this criterion. However, step 4 of Section 2.3.1 includes additional evaluation of SSCs where correlation and interaction considerations can impact the function of the SSCs. It is unclear whether the impact of the failure of SSCs that are screened out per step 3b of Section 2.3.1 of the EPRI report on SSCs with functions supporting mitigation is considered . For the Page 16 of 26

ATTACHMENT 1 Response to Request for Additional Information example provided in the EPRI report, a chiller system that is not part of the safety function of the tank, the consideration of seismic-specific failure modes of the chiller system (e.g.,

correlated or interaction failure of the heat exchangers and or piping resulting in a drain path for the tank) can change the screening decision. In addition , Section 4 of NEI 00-04 includes discussion of categorization of interfacing systems. It is unclear how the screening criterion in step 3b will interface with the guidance in Section 4 of NEI 00-04 for any interfacing systems.

a. Step 3b indicates a decision related to the SSC function for mitigation of core damage or containment performance. It is unclear if the mitigation function being questioned here are hazard-specific. Clarify whether the question in Step 3b includes SSCs functions for mitigation of core damage or containment performance for seismically-induced design basis and severe accidents. Include justification if the functions exclude mitigation of seismically-induced design basis and severe accidents.

EGC RESPONSE 6a:

The mitigation functions that are supported by an SSC are not hazard specific. The mitigation functions that are required to support maintaining key safety functions (e.g .,

Inventory Control, Containment Heat Removal) to mitigate core damage and large, early release frequency (i.e., a typical figure of merit for containment performance) apply to all evaluated hazards (e.g., Internal Events, Fire, Seismic). The LaSalle PRA models evaluate SSC functions for mitigation of core damage and large, early release frequency (LERF) for design-basis as well as severe accidents (e.g ., beyond design basis events).

For the purposes of this seismic correlation sensitivity study, the SSCs supporting the mitigation functions modeled in the FPIE PRA are assumed to be bounding with respect to the mitigation functions required in response to other hazard events. The LaSalle PRA model documentation identifies if specific SSCs are not credited for the mitigation of core damage or LERF when the SSCs do not meet the PRA success criteria.

For example, the Standby Liquid Control (SBLC) test tank does not typically support mitigation of any design basis accidents or beyond design basis accidents. The SBLC test tank is not credited in the LaSalle PRA, nor is it credited in any typical industry BWR PRA model, for the mitigation of core damage or LERF. The SBLC test tank does not have sufficient capacity as a water supply source to support the PRA success criteria for the mitigation of core damage or LERF scenarios in the LaSalle PRA model. Therefore, the SBLC test tank is screened as a seismic "target". However, the SBLC test tank is still evaluated as a potential seismic "source" (e.g., seismic interactions that may impact SSCs credited in the PRA).

b. The discussion for Step 3b in Section 2.3.1 of the EPRI report includes an example of a chiller system which would be screened out from consideration in the proposed alternate seismic approach. However, neither Step 3b nor the example in the discussion in Section 2.3.1 of the EPRI report considers the indirect impact of seismically-induced failures of SSCs on mitigation of core damage or containment performance. The case study for Plant C demonstrates such an impact. Table 3-9 of the EPRI report shows failures of two SSCs, one of them a chiller, (S_CB-CHLR-NCSW-FLOOD and S_ 1FC-ACU-FLD) which results in flooding and consequently loss of mitigation. Step 3b does not appear to include such impacts and would therefore, potentially screen out such SSCs. Clarify whether the question in Step 3b includes the indirect impacts of seismically-induced failures of SSCs on mitigation of core damage or containment performance. Include justification if such impacts are excluded given the insights from the Plant C case study.

Page 17 of 26

ATTACHMENT 1 Response to Request for Additional Information EGC RESPONSE 6b:

Figure 2-3 of the EPRI report will be revised per mark-ups provided in Attachment 2 to clarify that SSCs can be screened out in Step 3 from consideration of a correlated functional failure of the SSC itself but still need to be considered for potential seismic interaction failures that could affect multiple SSCs. This is accomplished by revising the text in Step 4 to say "Screened out from functional seismic correlation review" and by adding a line from Step 1 into Step Sb to show that all system SSCs need to be considered in the seismic interaction review.

c. Step Sa through Sc in Figure 2-3 and the accompanying guidance in Appendices A and B of the EPRI report do not discuss the consideration and treatment of the indirect impacts of seismically-induced failures of SSCs on mitigation of core damage or containment performance discussed in item (b) above. If indirect impacts of seismically-induced failures of SSCs on mitigation of core damage or containment performance are included in Step 3b in response to item (b ), explain how such SSCs will be considered in the walkdowns in Step Sa through Sc.

EGC RESPONSE 6c:

The intent of Step 3 of the flowchart is to screen SSCs which: 1) are seismically rugged such that their core damage mitigation or containment performance function would not be directly impacted by the seismic event (step 3a) such that they would not be in a correlation group; 2) have no core damage mitigation or containment performance function during a design basis or beyond design basis seismic event (step 3b) such that even if they were included in a correlation group they would have no impact; or 3) have already been determined to be high safety significant (HSS) from some other aspect of the categorization process (step 3c) such that correlating other SSCs with them will not change their HSS designation . Note that for step 3c, this does not exclude these SSCs from being included in a correlation group developed in step Sa for an LSS SSC that did not screen to box 4 in step 3.

The criteria in steps 3a, 3b and 3c are not intended to screen these SSCs from consideration as the "source" of a seismically induced interaction failure of other components . This has been reflected in the updated flowchart where it shows that SSCs for which the answer is 'Yes' to either box 3a, 3b or 3c; the evaluation only screens from the functional correlation review in box 4. All SSCs in the system being categorized proceed from box 1 to box Sb where their interaction is evaluated regardless of the answer to box 3 screening.

During the walkdowns, SSCs in the system being categorized will be evaluated to determine if they could physically impact one or more other components, either within the system being categorized or within another system which has a core damage mitigation or containment performance function . If so, an interaction failure basic event for the component would be created. Note that although technically possible, it is highly unlikely that an SSC that meets the criterion in step 3a would cause an interaction failure of another SSC, since it is seismically rugged and unlikely to cause the failure of another SSC. However, these SSCs will still be reviewed during the interaction review in step Sb.

Additionally, the criteria in step 3a is not intended to screen these SSCs from being the Page 18 of 26

ATTACHMENT 1 Response to Request for Additional Information "target" of an seismic interaction failure. Even though an SSC is seismically rugged, it could still be failed by the interaction with the failure of another non-seismically rugged SSC. Some examples would be seismically rugged SSCs which are failed due to flooding from a seismically induced flood, or which are failed by being impacted by another SSC which is failed by the seismic event (e.g., falling components or block walls).

If it is determined during the walkdown that this is the case, an interaction basic event will be created.

d. Explain how the consideration of indirect impacts of seismically-induced failures of SSCs on mitigation of core damage or containment performance discussed in items (a) through (c) above interfaces with the categorization of interfacing SSCs discussed in Section 4 of NEI 00-04. If the proposed alternate seismic approach introduces changes to the approach in Section 4 of NEI 00-04, justify such changes.

EGC RESPONSE 6d:

In the Seismic Correlated Failure Assessment, a separate review for potential interaction concerns is performed in Step 5b, and SSCs screened in Step 3b would not be excluded from the assessment of the impact of their failure on SSCs with functions supporting mitigation, as discussed in the response to RAI 6.c. This interaction review will address SSCs within the system being categorized that could cause an interaction failure of other SSCs, either within the system being categorized or in other systems with a core damage mitigation or containment performance , as well as SSCs in other systems which could cause an interaction failure of SSCs within the system being categorized. In neither case will the interaction review change the system boundary (which defines the boundary between the categorized system and any interfacing systems). Section 4 of NEI 00-04 will still be used to identify the system boundary, system functions, and mapping components to functions. The Seismic Correlated Failure Assessment is not intended to change those system engineering assessments in NEI 00-04.

APLC 50.69 RAI No. 7:

The steps in Section 2.3.1 of the EPRI report discuss the process for performing the sensitivity study. The surrogates used for the sensitivity study are intended to reflect the impact of seismic-specific failure modes identified in SPRAs. The steps in Section 2.3.1 of the EPRI report do not provide any guidance on whether the surrogates for SSCs that have already been categorized will be retained in the licensee's FPIE PRA model for subsequent sensitivity studies (for other SSCs).

Explain, with justification, whether surrogates that were incorporated in the licensee's FPIE PRA for the characterization of an SSC will be retained and included in the sensitivity studies for subsequent (and distinct) SSC categorizations.

EGC RESPONSE 7:

The intent of the correlation study is to determine if potentially low safety significant (LSS)

SSCs are sensitive to correlation or seismic interaction failures. The surrogates are modeled to represent a common cause failure or seismic interaction failure with the given initiating events (LOOP and SLOCA) for components which were not screened out from steps 1-5.

The approach was intended to be on a system by system basis and not intended to be Page 19 of 26

ATTACHMENT 1 Response to Request for Additional Information retained for additional system categorizations.

Other programs using importance measures (e.g. MSPI and exclusion of flooding initiators) have shown that dominant contributors can reduce the importance of components, therefore, retaining surrogates from system to system may reduce the importance of components in one system . Performing the study on a system by system basis will result in more consistent results and be less likely to change as additional systems are categorized. The study is also limited to potentially low safety significant components within the system being categorized and does not include correlation for potentially high safety significant components or components which are otherwise screened out in steps 1-5 within the same system.

Therefore, limiting the study to LSS components within the same system and not retaining the surrogates in additional system categorizations meets the intent of the study and will provide consistent results.

APLC 50.69 RAI No. 9:

Appendix B of the EPRI report provides guidance on "capacity-based screening for high capacity SSCs." The guidance includes a screening value of the seismic core damage frequency and if a single SSC were to contribute that value or less, it could be screened from evaluation under the proposed alternate seismic approach . The proposed screening value appears to be an absolute value and therefore, not relative to the seismic risk for the licensee and the consideration of the importance measures thresholds used for categorization under 10 CFR 50.69. Depending on the seismic risk for the licensee, the proposed screening value can result in a non-trivial contribution relative to the seismic risk. Consequently, its use can result in screening of SSCs that would otherwise have to follow the other steps in Figure 3-2,"Seismic Correlated Failure Assessment for Tier 2 Plants," in the enclosure to the LAR (reproduced from Figure 2-3 in the EPRI report).

Appendix B of the EPRI report additionally provides guidance on the development of fragility values for SSCs to support the capacity-based screening. In addition to the state-of-practice approaches identified in Appendix B via relevant documents (e.g., representative values and conservative deterministic margins analysis [CDFM]), the guidance also discusses "more simplified and conservative approaches when justified by experienced engineers," "simplified approaches documented in ASCE 7," and "assessments made would have to be necessarily conservative ." In addition, Step 11 in Section 2.3.1 of the EPRI report discusses refinement to the fragility based on direction from the Integrated Decision-making Panel (IDP). SPRAs, such as those included in the case studies supporting the proposed alternate seismic approach, use state-of-practice approaches and undergo an NRC endorsed peer-review process for the implementation of such approaches. Further, based on reviews of several SPRAs that use the state-of-practice approaches, the NRC staff also has confidence in those analytical methods and their implementation by personnel with specialized knowledge about their use. The EPRI report and the LAR do not appear to include constraints on the type of approaches that can be used to achieve any "refinement" and therefore, such approaches can include those that have not been appropriately vetted even among the practitioners. It is unclear how the NRC staff can determine the acceptability of approaches that are not state-of-practice (either simplified or refined) and for which, as part of the proposed alternate seismic approach, an implementation peer-review will not be conducted.

It is expected that the fragility calculations require specialized knowledge and experience to implement the approaches as well as any caveats for the approaches in the corresponding Page 20 of 26

ATTACHMENT 1 Response to Request for Additional Information documents. The qualifications of the personnel performing the fragility development, even using state-of-practice approaches, has not been specified in either the LAR or the EPRI report.

Neither the LAR nor the EPRI report includes any information on the documentation of the fragility analysis. Such documentation appears to be necessary to support the conclusions of the analysis and for future regulatory processes, such as audits and inspections.

a. Justify the independence of the screening value of the seismic core damage frequency proposed in Appendix B of the EPRI report from the licensee's seismic risk and the importance measures thresholds used for categorization under 10 CFR 50.69.

Alternately, propose a screening approach relative to the licensee's seismic risk which considers the impact on the importance measures thresholds used for categorization under 10 CFR 50.69.

EGC RESPONSE 9a:

Appendix B of the EPRI report provides the basis for the generic seismic capacity HCLPF screening value recommended for use in the 50.69 categorization screening assessment.

As summarized in Appendix B, seismic capacity-based criteria were developed in EPRI 1025287 to determine a threshold HCLPF valued above which the screened SSCs are not expected to have significant impact on the result of the SPRA analyses, the ranking of accident sequences, the calculated sequence, or plant-level seismic CDF or LERF values .

As such, SSCs with capacities above that screening level would also not be expected to be high safety significant (HSS) components within the 50 .69 categorization process. For the 50.69 categorization screening assessment, the value in EPRI 1025287 was reduced by 50% down to 2.5E-7 to account for the more moderate seismic hazards that may occur at Tier 2 sites.

The option for licensees to choose another justified site-specific screening level will be removed from the EPRI report per mark-ups provided in Attachment 2.

b. Discuss how the calculation will be performed for comparison against the proposed screening threshold (or an updated threshold in response to item (a)). The discussion should indicate the seismic hazard curve (e.g., mean peak ground acceleration hazard curve) that will be used to perform the calculations.

EGC RESPONSE 9b:

See response to 9c (below).

c. Identify the analytical approaches and methods that are consistent with the state-of-practice and will be used for fragility calculations in the proposed alternate seismic approach.

EGC RESPONSE 9c:

RAls 9b and 9c are closely related and are answered together.

As described in Appendix B of the EPRI report, to perform the high capacity screening step (Sc) in the Seismic Correlated Failure Assessment, a seismic fragility must be developed for each SSC that is being assessed as part of the categorization process and compared to the HCLPF screening level. When the initial 50.69 Alternate Seismic Page 21of26

ATTACHMENT 1 Response to Request for Additional Information guidance was developed (3002012988), the EPRI fragility guidance was spread among several reports and the key references were provided in Appendix B. Since that time, EPRI fragility guidance has been consolidated into a single report, EPRI 3002012994.

Therefore, the current EPRI 50.69 Alternate Seismic report (3002017583) will be revised per mark-ups provided in Attachment 2 to remove the previous collection of fragility reports and refer to Section 3.3, Separation of Vadables Fragility Approach, and Section 3.4, Hybnd Fragility Approach, of EPRI 3002012994 for the appropriate fragility methods.

These fragility methods define the state of practice in use for SPRAs today and will serve to ensure consistency of fragility and screening methods within the 50.69 process.

Appendix B of the EPRI 50.69 Alternate Seismic report also notes that it may be necessary to use scaling methods to estimate in-structure response spectra as part of the fragility calculations. Sections 6.4, Special Considerations for High-Frequency Seismic Demands, and 6. 7, Response Analysis Scaling Methods, of EPRI 3002012994 also provide the most up to date criteria for scaling in-structure response spectra, incorporating the criteria currently referenced in the 50.69 Alternate Seismic report. Therefore, the Alternate Seismic report will be revised per mark-ups provided in Attachment 2 and updated to Sections 6.4 and 6.7 of EPRI 3002012994 for scaling criteria.

Lastly, the option of considering ASCE 7 criteria to estimate in-structure response spectra will be removed from the EPRI report per mark-ups provided in Attachment 2.

d. Discuss how the proposed approach will ensure NRC staff's review and acceptance of analytical approaches and methods for fragility calculations other than those identified in item (c) above prior to their use in the licensee's alternate seismic approach (e.g., through a LAR controlled by the license condition).

EGC RESPONSE 9d:

The responses to RAls 9b and 9c identify the updates to the EPRI report per mark-ups provided in Attachment 2 to require use of the current state of practice methods for fragility calculations documented in the most recent EPRI fragility report 3002012994. As currently proposed, no unreviewed methods would be applied, thereby alleviating the need for NRC staff's review and acceptance of new methods.

e. Discuss how the proposed approach will ensure that the qualifications of personnel that perform the fragility calculations, such as those using the methods identified in item (c),

above are sufficient to support the development of the fragilities.

EGC RESPONSE 9e:

Any fragilities performed as a result of Step Sc would be performed in accordance with the latest EPRI fragility guide (EPRI 3002012994). As with an SPRA, the personnel performing these fragility calculations would need to have the background/experience with the state-of-the-practice fragility methods in EPRI 3002012994. This can be accomplished by completing the EPRI training course for Seismic Fragilities or equivalent. As with all fragility calculations performed for an SPRA, these fragility calculations would be independently checked by another engineer with this same fragility background/experience.

NEI 00-04 Section 3.3, Use ofRisk Information, notes that when "risk information is used Page 22 of 26

ATTACHMENT 1 Response to Request for Additional Information to provide insights to the IDP, it is expected that the risk information will have been subject to quality measures." One of the quality measures listed is the "use of personnel qualified for the analysis." That requirement applies to the Seismic Alternative approach.

f. Discuss how the approach ensures that the fragility analyses performed for the proposed alternate seismic approach using the methods discussed in item (c) are documented to support the conclusions of the analyses and support future regulatory processes, such as audits and inspections.

EGC RESPONSE 9f:

NEI 00-04 Section 3.3, Use ofRisk Information, notes that when "risk information is used to provide insights to the IDP, it is expected that the risk information will have been subject to quality measures." Two of the quality measures listed are the use of "procedures that ensure control of documentation" and a requirement to "Provide documentation and maintain records in accordance with licensee practices." Section 11 of NEI 00-04, Program Documentation and Change Control, provides additional documentation criteria. All of these requirements apply to the evaluations performed in the Seismic Alternative approach.

APLC 50.69 RAI No. 10:

Table 3-1 of the enclosure to the LAR provides a "categorization evaluation summary." According to the table, the IDP can "change HSS to LSS [low safety significant]" for the "seismic" categorization step. Step 11 in Section 2.3.1 of the EPRI report that is incorporated by reference by the licensee (page 14 of the enclosure to the LAR includes Section 2.3.1 in the incorporation of the EPRI report in the LAR), states that the proposed approach (using the sensitivity study) is "pseudo-deterministic" and, therefore, the "seismic correlated group HSS designations should be treated similar to HSS designations using the IPEEE SMA SSEL [Safeguards Summary Event List] and in general, not be subjected to reconsideration by the Integrated Decision-Making Panel (IDP)." It appears that the LAR is inconsistent with the EPRI report which it is referencing and supposedly, following. In addition, the basis for the LAR's deviation from the EPRI report is also unclear.

Step 11 of the EPRI report states that "SSCs which are HSS solely due to surrogate events representing seismic induced interactions (such as block walls impacting equipment) may be downgraded to LSS by the IDP with appropriate justification". The discussion on consideration of seismic correlated group HSS designations similar to IPEEE SMA SSEL and the subsequent possibility of downgrade to LSS is unclear to the NRC staff.

a. Clarify whether the summary for seismic categorization in Table 3-1 is intended to be a deviation from the approach discussed in the EPRI report. If yes, provide the justification for the deviation based on the insights and guidance in the EPRI report for the proposed approach which is incorporated by reference in the LAR. If no, provide an updated version of Table 3-1 that clarifies the licensee's intent.

EGC RESPONSE 1Oa:

Table 3-1 in the LAR was not intended to be a deviation from the approach discussed in the EPRI report. An updated version of Table 3-1 is provided below that clarifies the intent for LaSalle.

Page 23 of 26

ATTACHMENT 1 Response to Request for Additional Information IDP Categorization Drives Change Element Step - NEI 00-04 Evaluation Level Associated HSSto Section Functions LSS Internal Events Base Case- Not Allowed Yes Section 5.1 Fire, Seismic and Other External Allowable No Risk (PRA Events Base Case Component Modeled)

PRA Sensitivity Allowable No Studies Integral PRA Assessment - Not Allowed Yes Section 5.6 Fire and Other Component Not Allowed No External Hazards -

Risk (Non-modeled) Seismic1 Function/Component Allowed No Shutdown - Section Function/Component Not Allowed No 5.5 Core Damage -

Function/Component Not Allowed Yes Defense-in- Section 6.1 Depth Containment -

Component Not Allowed Yes Section 6.2 Qualitative Considerations -

Function Allowable 2 N/A Criteria Section 9.2 Passive Passive - Section 4 Segment/Component Not Allowed No Notes:

1 This non-modeled seismic categorization element refers to LaSalle; in particular, use of EPRI 3002012988 (now EPRI 3002017583) and all related RA! responses for this 10 CFR 50. 69 LaSalle application.

2 The assessments of the qualitative considerations are agreed upon by the !DP in accordance with Section 9.2 In some cases, a 50. 69 categorization team may provide preliminary assessments ofthe seven considerations for the IDP's consideration, however the final assessments ofthe seven considerations are the direct responsibility ofthe!DP The seven considerations are addressedpreliminarily by the 50. 69 categorization team for at least the system functions that are not found to be HSS due to any other categorization step Each of the seven considerations requires a supporting justification for confirming (true response) or not confirming (false response) that consideration. If the 50. 69 categorization team determines that one or more ofthe seven considerations cannot be confirmed, then that function is presented to the !DP as preliminary HSS Conversely, 1fall the seven considerations are confirmed, then the function is presented to the !DP as preliminary LSS Page 24 of 26

ATTACHMENT 1 Response to Request for Additional Information The System Categorization Document, including the justifications provided for the qualitative considerations, is reviewed by the !DP The !DP is responsible for reviewing the preliminary assessment to the same level of detail as the 50 69 team (i e. all considerations for all functions are reviewed). The /DP may confirm the preliminary function risk and associatedjustification or may direct that it be changed based upon their expert knowledge. Because the Qualitative Cr!leria are the direct responsibility of the !DP, changes may be made from preliminary HSS to LSS or from preliminary LSS to HSS at the discretion of the /DP If the /DP determines any ofthe seven considerations cannot be confirmed (false response) for a function, then the final categorization of/hat function is HSS

b. Explain the apparent contradiction between two statements in the same step (step 11) in Section 2.3.1 of the EPRI report and clarify what is being proposed in terms of changes to HSS designations arising from the sensitivity study in the proposed approach by the IDP.

EGC RESPONSE 1Ob:

See response to 1Oc.

c. Explain , with examples, what would be "appropriate justification" by the IDP to downgrade an HSS determination arising from the sensitivity study in the proposed approach noting that a walkdown would have been conducted, the flowchart in Figure 2-3 of the EPRI report would have been followed, and a sensitivity would be performed before the HSS determination is made.

EGC RESPONSE 1Oc:

RAls 1Ob and 1Oc are closely related and are answered together.

The description of the final step in the Seismic Correlation Failure Assessment (Step 10) notes that the correlation assessment is a "pseudo-deterministic evaluation process rather than a full risk informed process" and that the "seismically correlated group HSS designations should be treated similar to HSS designations using the IPEEE SMA SSEL and in general, not be subject to reconsideration by the Integrated Decision-making Panel (IDP)."

The report goes on to identify one exception to the rule that leads to the "apparent contradiction" noted in the RAI question. That exception is if the HSS determination is due solely to a seismic interaction concern, and it could be shown that special treatment of the SSC would not change the likelihood of failure or its consequences, then there is no benefit to designating the SSC HSS. An example given in the EPRI report is seismic induced interactions on SSCs such as from block walls where the correlation sensitivity may show the RAW or FV greater than the acceptance criteria for HSS for the correlated equipment impacted by the block wall. If it could be shown that the failure of the block wall would fail the equipment regardless if the equipment was purchased, installed, and tested with special treatment or not, then the SSC could be categorized LSS. The other potential interaction condition would be seismic induced flooding or spray. Similarly, if it could be shown that special treatment of the SSC would not change the likelihood of failure or its consequences, then there is no benefit to designating the SSC HSS.

Therefore, a clearer statement of the criteria is that seismically correlated group HSS designations should not be subject to reconsideration by the Integrated Decision-making Page 25 of 26

ATTACHMENT 1 Response to Request for Additional Information Panel (IDP), unless the HSS determination is due solely to a seismic interaction concern , and it can be shown that special treatment of the SSC would not change the likelihood of failure or its consequences .

The EPRI report will be updated per mark-ups provided in Attachment 2 to incorporate these changes .

Page 26 of 26

ATTACHMENT 2 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 EPRI Report Markups

ATTACHMENT 2 Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization 39929115S3300201xxxx Technical Update, i;:serYaF)'Month , 202xQ Insert appropriate EPRI Title Page Auto Text entry here.

ATTACHMENT 2 Proposed Approach Another key insight from the four seismic categorization test cases is that seismic interactions (for example, seismic induced falling, deflections and flooding that affect nearby safety related components) are unique to seismic risk studies and can result in seismic unique insights potentially leading to HSS SSCs.

To better assess the importance of SSCs to seismic event response at Tier 2 plants, the internal events PRA may be used with some modifications and augmented by focused seismic walkdowns to obtain an indication of the importance of SSCs for mitigating seismic events. The process is depicted in Figure 2-3 and the steps are summarized. Note that this process is performed on a system basis, as the 50.69 categorization process is performed for a given system.

1. Identify the set of SSCs within the system to be categorized.

2. [G roupJthe SSCs within the system into the classes of equipment and distributed systems ( Commented [RJ1]: RA! Response 3c used for SPRAs. This format of grouping allows for an efficient assessment from a seismic perspective. Section 6.1.3 of EPRl 3002012994 [34] provides a list of equipmenr classes that should be used for this grouping.lnst1stl) SHlttments sweh as the IWRJ 3QQ2QQQ::'Q9 [18] ans IWRJ +/-'IP t3Q1 l 8ls [19] isentif) the list lfthese elasses. For example, separately group all manual valves, all check valves, all MOVs, all AOVs, all pumps, etc. This will make it easier to screen SSCs in the next step, as well as to see which SSCs already have CCF basic events modeled in the FPTE PRA.

3. efineJthe list of SSCs in the system being cate orized based on a series of screens to ( Commented [RJ2]: RA! Response 3b minimi!!e reduce the number of SSCs required to be evaluated as part of this correlation sensitivity study. Note that any/all of these screens can be incorporated into the process and that any order of implementing these three screens is acceptable. The screens are elective and can be implemented by the licensee to the degree that the effort to screen ou any SSC is less than the effort to keep the SSC on the list to be considered in the remaining steps of the correlated failure assessment. Tlrnse sereEming 8eeisiens ma) l:le 13lafl:t s13eeifie aHS lilrnl) "ill ill o ll o e esst/l:JeHefit Seeisisns i11 terms lf hs" eest ts ssmf!lets the ssnsiti it; srns; .

a. [S creenJout inherently rugged components. l>Jl!:I 12 Qi

• fltrnnsi i; II [12] Table 6.2 of _ ( Commented [RJ3]: RAJ Response 3b EPRI 3002012994 [34] provides the fulll ing list of inherently rugged components.

i. £t1aimHS ans small line illlMfl:teJ tarnos ii . Weisse ans "1Blte8 t1it1iHg iii. Hani,ial al SS, shsslt al SS, ans IH!'!ti,i1s sislts i, . Ps" e1 lfletate1 , al, es (P 10\'s an1 AOVs) nlt tefjttirn1 tl ehaRge state

b. Screen out SSCs that are not used in safety functions that support mitigation of core damage or containment performance. This will likely already be identified as part of the function definition within the 50.69 categorization process. An example would be a chiller system that maintains the temperature of water in a tank but is not part of the safety function of the tank (for example, provides core inventory). The SSCs in the chiller subsystem can be screened.

c. [Scree~ out ft:sffl H!Ftker s ahiatiBn the SSCs that have already been identified as ( Commented [RJ4]: RAJ Response 6c being HSS from either the FPIE PRA or HSS from the Integral Assessment. For this 2-7

ATTACHMENT 2 Proposed Approach screening step, the categorization based on the FPIE PRA and the Integral Assessment would need to have already been completed.

4. hose]SSCs screened out in Steps 3a, 3b, or 3c can be removed from consideration of [ Commented [RJS]: RA! Response 6c functional correlated seismic failures in Step 5atbs ssnsiti, it) stM8) . In a88itiBB , 88Cs S8HHm8S BHt lislB in 8t8fl ~8, e, Bl 9 san S818illB 8S fl Bill f.Mrthst ssismi@ @BnsisstatiBn.

5. For system components not screened out in Step 3, perform a seismic walkdown focused on the three activities listed below. The purpose of this step is to identify SSCs that could experience seismic correlated failures or could be subject to seismic interactions that would lead to failure of more than one SSC within the system being categorized. The following elements contribute to identifying these conditions.

a. Assess if the subject SSCs would likely experience correlated failures during a seismic event. Seismic correlation walkdown reviews are performed as part of an SPRA and the guidance associated with performing that correlation walkdown is documented in Appendix A.

b. [Assess] potential seismic interactions to identify conditions that could fail multiple [ Commented [RJ6]: RA! Response 4f components in the system !H~l8 Ii s ~rnats8 as ssismis sBrHlats8 I'aihu es ans thsrnrorn, SHBHls lis s alHats8 as sBmmB!l: simss I'ailHtss. Guidance for this seismic interaction walkdown review is also contained in Appendix A.

c. Screen out SSCs that are determined to be sufficiently rugged such that they would not be significant contributors to seismic risk in an SPRA. This screen focuses on the SSC seismic capacity associated with functional failures and anchorage. The screen can also be applied to identified seismic interactions provided the seismic capacity of the interacting item (for example, block wall) has a seismic capacity that meets the screening level. Appendix B contains a description of the approach recommended for this screening.

[As]an alternate to screening out an SSC, the SSC can be retained in the FPIE model ( Commented [RJ7]: RA J Response 4e with a seismic-induced failure probability developed for the surrogate by performing a plant-specific fragility analysis as described in Appendix B.

6. [SSCs]that are determined through the walkdown to be of high seismic capacity and not [ Commented [RJSJ : RA! Response 6c included in seismically correlated groups or correlated interaction groups can be screened out from further seismic considerations since their non-seismic failure modes are already addressed in the FPIE PRA and fire PRA. Those remaining components identified in step 5 with seismic correlated failures or seismic interaction failures of multiple components proceed forward for inclusion of associated seismic surrogate events in the Tier 2 adjusted PRA model.

2-8

ATTACHMENT 2 Proposed Approach

[ Commented [RJ9] : RA! Response 6c roup ~ In o eQulpMen d sses la s.sc *~

lnherenllV S mlc uag d Y~ <;SC not used In fuoc11on

- -No seep 3 ~ere-eris can

---

Mrfoffned--- I a*>~ oroer Id Ml y !UCs rv ....

=._j ld"11 ml*

tOr'ISl d I d ~l!n JC V !-----*I ..~--~ ont rart100 Iii ur cari led

/Sc SSC I~ High S!'i'llT11C C..p.J(lly .ar>d ---'f1ti.--------~

riot subJ D ntl!f'acc Seep 5 evah.1acoons ore performed as parr aj a i*umic w<rl dJ)wt1 I No SSCs Yes- tdenlJh d forcorrel lion - - - - - - -IN n-----t ar 111tl"l;ICt1on

._11ures IC Figure 2-3 Seismic Correlated Failure Assessment 2-9

ATTACHMENT 2 Proposed Approach

7. dd] new seismic surrogate events to the FPIE PRA logic model for the potential [ Commented [RJ10]: RAJ Response 6c seismically correlated and seismic interaction conditions identified in the previous steps.

New seismic surrogate events should be added to the PRA under the appropriate areas in the logic model. For example, a new seismic surrogate basic event that models seismic correlated failure of two tanks would be added to the PRA logic model under the gates that model the individual tank failures . Seismic interaction surrogate events should be added to the model such that they fail the SSCs affected by the interaction. For example, a seismic interaction surrogate event that models a block wall falling onto two nearby pumps should be added to the PRA logic model under the gates that model the pumps.

The probability for the new seismic surrogate basic events should be set to a value equivalent to a "typical" total seismic hazard exceedance frequency above which SPRAs would typically model loss of offsite power and for which correlated failures may be likely. The recommended value is 1.0E-04, but other appropriately justified values may be used .

S. Quantify the FPIE PRA model for LOOP and small LOCA initiated accident sequences

~using the modified model containing the seismic wttft.surrogate events for the system being analyzed; and calculate the component-based risk importance measures for the system components encompassed by the seismic surrogate events. Since the majority of seismic risk in many SPRAs are the LOOP and small LOCA accident sequences, these events HflHSeRt are the most appropriate events for performing this correlation study.

The process is as follows.

a. tJ'he]recommended event frequency for the LOOP initiator is 1.0 and for the small [ Commented [RJ 11]: RAJ Response 4c LOCA initiator is l.OE-02 . The LOOP frequency value of 1.0 is recommended since the probability of the surrogate events (from Step 7) is the total seismic hazard exceedance frequency above which SPRAs would typically model loss of offsite power. The basis for the small LOCA frequency of l.OE-02 is that seismically-induced small LOCAs require failures that SPRAs show typically occur at much lower frequency than seismically-induced LOOP.

For multi-unit sites, the LOOP initiator selected for adjustment is the LOOP initiator that is modeled as a site LOOP condition (that is, loss of offsite power to all reactor units onsite).

The small LOCA initiator selected for adjustment is a water break small LOCA (that is, NSSS break below T AF). Otlm aflfll spriatel) justified alttes s11all IsOC 0 frejtt<iRe) ma) lie usea. The 111ajsrit) sf seismie rislt is frn111 the IsOOP/8~0 ans small IsOC

• aeeiseRt sejtteRees.

a.b. Set the frequency for all initiators other than LOOP and small LOCA initiators selected for adjustment in step Sa to 0. Note that many FPIE PRAs have multiple LOOP initiating events (for example, grid centered, switchyard centered, etc.) as well as multiple small LOCA initiators (small LOCAs above the top of active fuel (T AF) and small LOCAs below T AF) . Only one each of these needs to be set to 1.0 m as described in step Sa, above, while all the other initiators in the FPIE PRA shouldfeSt eatt be set to an occurrence frequency of O (or a logical value of FALSE).

2-10

ATTACHMENT 2 Proposed Approach

!F.c.Since a seismic event that causes a small LOCA is also assumed to cause a LOOP, update the PRA model to account for this. This is typically done by setting a conditional LOOP probability to 1.0, but can be done in any appropriate manner.

d. [M:anyJFPIE PRA models credit restoration of offsite power in the LOOP/SBO [ Commented [RJ12]: RAJ Response 4a accident sequences as well as other functional recoveries (for example, EDG recovery, DC power recovery). Th+sese credits should not be taken in this Tier 2 adjusted PRA quantification process since recovery of offsite or DC power after a seismic event is not generally credited in a seismic event. This is typically performe by setting the probabilities of the basic events in question to a value of 1 (or a logical value of TRUE). If the run time for emergency AC power (for example, EDGs, fuel oil pumps) was adjusted, the run time should be set to the PRA mission time.

lR fllaee 0f eHfllieit elfl, 0lt1ti0R m0seliRg 0f 0ffsite fll, er ree0, er; e11r es aRs eme1 geRe) 6 c fail11re rates, m8R) FPrn PRi" s 11se ii kat 81 e te1 mes Stll 1lgate Ft!R times (!er eHamfJle, ~ t0 8 krs~ fer emergeRe; "C fl El er eJ:11ifJmeHt (fer eJtamflle, l"lQCs, f110l lil traRsfer fHlfflflS~ easis e I @Rt fail!if@ flFlSaeilit; 0al01ilati8RS tl miHimii'ie esnse1 atism in the ealelilates PF:" rnslilts. Ci el'! that sffsite t'll et te8l i er; is ass11mes failes il'l this Tie! 3 fll leess, aR) s11eh Sllll 0gate !till tirnes eases lll ftttleti0Mal rne0, er; eresit slrn11ls Be ittereases t0 the PRA fall missi0H time (t;flieall; 34 k0H1s~ iH tke Tie1 3 asjHstes PR/ J:HaRtifieatilR. Tkis asj11stmeRt is t)flieall; flerlinmes B) rn, isiHg the rnM eiqi0sme time Hses in the flllBaeilit; eale11lati0H furm11la fut eaek 0fthe affuetes Basie e ents.

e. In addition to the LOOP and small LOCA accident sequences, the quantification of the Tier 2 adjusted PRA should also include the accident sequences transferring out of these event tree structures (for example, LOOP/SBO, LOOP/SBO-ATWS, small LOCA-ATWS).

f. The Tier 2 adjusted PRA model is quantified including the seismic surrogate events applicable only to the components in the system being categorized. Seismic surrogat

• events that may have been identified for in previous system categorizations 0ther s; stem esmfJsnents s11tsise the s; stem Being eategstii'ieS should not be added into th quantification or set to zero if already existing in the PRA model being used for the Tier 2 quantification.

e-:g.Consistent with the NEI 00-04 process, the risk importance information from the Tier ~2 adjusted quantification should be presented on a component basis and include the various applicable failure modes. In the case of the Tier 2 process, the list of components for which component risk information is produced is constrained to the system components encompassed by the identified and modeled seismic surrogate events. These component-based risk importances should include the seismic surrogate risk contributions as well as the non-seismic failure mode contributions used in the plant 50.69 categorization process.

9. o all SSCs that are included in a surro ate event (either a correlated failure event or an ( Commented [RJ13]: RAI Response 5 interaction failure event) the F-V value for the surrogate event will be added to the F-V values for the non-seismic failure events for that SSC. The resulting F-V will be compared to the criterion for SSCs in the FPIE PRA from NEI-00-04 (that is F-V >

0.005). The maximum RAW value of the surrogate event and any other non-seismic 2-11

ATTACHMENT 2 Proposed Approach common cause failure events involving that SSC will be compared to the criterion for common cause SSCs in the FPIE PRA from NEI-00-04 (that is RAW> 20). The maximum RAW value for any non-seismic-related, non-common cause failure events of that SSC will be compared to the PRA criterion from NET 00-04 (RAW> 2). If the SSC exceeds the threshold for any of these three criteria, it will be categorized as HSS. ~

i.rnmpE11rnnt llas8a Fish impE1rta1188 18sttlt prnatt88a in st8fl 8s8ismi8 stt¥rngat8 8 8nt, 81mparn drn rnsttlts tEI drn F "ana R Anr H88 8rit81 ia WI 81mm1n 8atl88 81mfl1118nts in the FPIE PR* frnm NEI QQ Q1 (that is, F " ' Q.QQ§ Ell R * " 2 fat nE1n CCF faihus meaes 1.ma RAW ' 3Q fet CCF fflihue meaes). The RAV:' ' 3Q 1188 e1ite1ien !tflfllies te Hie seisl'l'lie sttFFsgate failttFe l'l'lsaes aREI tke RBR seisl'l'lie CCP failt11 e l'l'lsaes. Tke R/ 'V '

3 II88 sritsFisn aflflliss te tke HeH ssismie Hen CCF failttrs R:1et:les.Fsr seisH1ie stm*sgate 8 8nts, iftlrn F v 1¥ RA m 8¥it8tia am m8t, all 88Cs m1a818a ll) that Stll11gat8 8 811t shsttl8 lle esHsiae1e8 II88

10. [SinceJthis process is a pseudo-deterministic evaluation process rather than a full risk Commented [RJ14]: RA! Responses !Ob informed process, these seismically correlated group HSS designations should be treated and !Oc similar to HSS designations using the lPEEE SMA SSEL. Therefore, seismically correlated group HSS designations should not be subject to reconsideration by the Integrated Decision-making Panel (IDP), unless the HSS determination is due solely to a seismic interaction concern, and it can be shown that special treatment of the SSC would not change the likelihood of failure or its consequences. In cases where seismic interactions (for example, seismically induced block wall failures, or flooding or spray failures) are the sole reason for an SSC to be designated HSS, the procurement, qualification, installation, maintenance, and testing applied to the SSC would not influence its failure probability , therefore the IDP may consider downgrading the SSC categorization to LSS. aHEi iH g8H8tal, HEit ll8 sttej88t tEI 18sE1Hsia81atiE1H ll) t¥18 IHt8g1at88 I)8sisiE1n making Pan81 (ll)Pj. 111 8 81, 88Cs hi8h am Il88 81181) Eitrn t1 s1t1rngat8 8 8nts F8fll8S8nting s8ismis inat188a int81a8tiE1ns (fur 8Jtampl8, llLElsk alls imflasting 8~[,li13ment) ma) se BB ng1aEieEI ts u;s B) t-He lf>P ith a13131 s131iate jHstif.ieatisn. In addition, the IDP may direct further engineering evaluation to refine any of the seismic evaluation insights.

+Q.:. 11. [s scsJscreened out in Steps Sc, 6, or 9 can be removed from further seismic ( Commented [RJ15]: RA! Response 6c consideration.

2.3.2 Technical Basis for Approach The test cases described in Section 3 showed that even for plants with high seismic ground motions compared to their design basis, there are very few if any SSCs that would be designated HSS for seismic unique reasons and the technical basis for the Tier 1 approach in Section 2.2.2 generally apply for the Tier 2 plants.

The test cases did identify a small number of instances where unique seismic insights associated with seismically correlated failures led to unique HSS SSCs. \Vhile these unique HSS SSCs would be unusual for moderate hazard plants, it is prudent to perform additional evaluations to identify the conditions where these correlated failures may occur, and determine their impact in the 50.69 categorization process.

2-12

ATTACHMENT 2 Proposed Approach For a system being categorized under 50.69, the process described in Section 2.3 .1 identifies the conditions that would be treated as seismically correlated fragilities or interaction consequences if an SPRA were being performed. It screens out SSCs that would either be very low contributors to seismic risk or would not be potential candidates for HSS, or were already categorized as HSS. These SSCs do not require additional seismic evaluations for considerations in the 50.69 categorization process. After that initial work, the evaluation follows a thought process similar to that in a SPRA, using a seismic walkdown to identify correlated conditions and potential seismic interaction and screen low seismic contributors.

The step 8 Tier 2 modeling and quantification adjustments are defined to reflect seismic induced conditions and how they would typically be modeled in a SPRA. For example, a site LOOP condition is a typical modeling assumption for seismic-induced loss of offsite power. In the absence of fragility calculations on specific piping segments, modeling seismic induced LOCAs as below T AF is a common SPRA approach as it is the more severe location compared to above TAF. Removing (or justifying continued credit) for functional recoveries is a typical SPRA approach and covered by a PRA Standard supporting requirement. Similarly, upon removal of credit for offsite power recovery the use of surrogate run time estimates in emergency AC equipment failure probabilities would produce optimistic risk results and thus SPRAs increase such surrogate run times to the PRA mission time (typically 24 hrs). The resulting unscreened SSCs are candidates for seismically correlated conditions that should be evaluated for their impacts on 50.69 categorization. In an FPIE PRA, common cause evaluations behave similarly to seismically correlated conditions in that they model the impacts of multiple SSC failures ;

therefore, the common cause methods can be employed to determine the necessary insights for 50.69 categorization.

The recommended methods and evaluation parameters in Section 2.3.1 for adding the surrogate seismic common cause events into the FPIE PRA and performing the sensitivity study serve to identify to necessary categorization insights. A common cause failure probability of l.OE-4 is used to align with the total seismic hazard exceedance frequency above which SPRAs would typically model loss of offsite power and for which correlated failures may be likely.

Seismically induced LOOP and small LOCA events represent the majority of seismic risk in many SPRAs and are the most appropriate events for performing this correlation study. The process assumes a LOOP occurs (frequency set to 1.0) and uses a small LOCA frequency of l.OE-02 because seismically-induced small LOCAs require failures that SPRAs show typically occur at much lower frequency than seismically-induced LOOP. This combination of events typically encompasses the majority of seismic risk.

The component-based risk importance measures derived from this sensitivity study can then be used to identify the appropriate SSCs that should be HSS * *

• s@ismi@ int@Hrntisn rnlat@d *faihtt@s .

2.3.3 Summary For Tier 2 plants, the GMRS to SSE comparison is higher than Tier 1 plants but not high enough to be treated as Tier 3 plants. In Tier 2, there may be a limited number of unique seismic insights, most likely attributed to the possibility of seismically correlated failures, appropriate for consideration in determining HSS SSCs. The special sensitivity study recommended using common cause failures, similar to the approach taken in a FPIE PRA, can identify the 2-13

ATTACHMENT 2 Insert Appropriate Auto Text License Entry. If license is copyright, please delete References

33. NEI (K. Keithline) Letter to NRC (D. Skeen), White Paper Supporting Seismic Evaluation Guidance, "Criterion for Capacity-Based Selection of Structures , Systems and Components for Fragility Analyses in a Seismic Risk-based Evaluation", Project Code 689, November 28, 2012

31. &iB nie F r1gilit; 1f51Jlie "1tia 18 GMi 48 fs'j9 *fats. EPR-1, Pal0

• ltB, G * : 2QQ9. l Ql 92QQ.

3§. &eiainie ,v:, etgi!it; i4-ftj'Jlieeilia11 Gt1ivle. BPR-1, Pals Alts , CA: 2QQ2. 1QQ2988 3 e. ~kt~a w!agy Jj 98 s!tJpi 1g &iB nie K "1gi'iti88. EPIH, PalB A ltB , GA : 1991. TR l Q3 959

37. Ifaigl-i F1 8Jte81i8) 1Rr ag,,.a:n: 11fJtJliewliBri Gui-451 h38 Jro, 1C'..t1;iefB1ial G3:ef+: 1nYltiB1i aris/{i':Wigi!itj 4

E1aluatia 1. EPRT, Pal0 .6 lt0, c,*: 2QJ§. 3QQ2QQ1396

~34. ~eismiq Fragility and Seismic Margin Guidance for Seismic Probabilistic Risk ( Commented [RJ16]: RAI Response 3b Assessments. EPRI, Palo Alto, CA: 2018. 3002012994M~c.,.;;;/-~~.4tfitt¥!'1ttft~~~"6a@l ans A 88leiate8 Criteria fllf ~ttilsiHgs 8HB Other Strnetttfes," 2Qle.

,W,35. NUREG-1742, "Perspectives Gained from the Individual Plant Examination of External Events (IPEEE) Program, Final Report, Volumes 1 and 2," US NRC, April 2002 4G.36. NRC (W. Dean) Letter to All Power Reactor Licensees et. al. , Final Determination of Licensee Seismic Probabilistic Risk Assessments Under the Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(G") Regarding Recommendation 2.1 "Seismic" of the Near-Term Task Force Review of Insights from the Fukushima Dal-Ichi Accident, October 27, 2015 (ML15194A015) 4+:-37. NRC (M. Franovich) Letter to Duke Energy (E. Kapapoulous, Catawba Nuclear Station, Units 1 And 2, and McGuire Nuclear Station, Units 1 and 2, Screening and Prioritization Results Regarding Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review oflnsights from the Fukushima Dal-Ichi Accident, December 22, 2016 (ML16344A313).

5-3

ATTACHMENT 2 Insert appropriate Auto Text License Entry. If license is copyright, please delete.

B CRITERIA FOR CAPACITY-BASED SCREENING FOR HIGH CAPACITY SSCS Seismic risk insights from past SPRA and SMA studies have shown that high seismic capacity SSCs from the SPRA Seismic Equipment List (SEL) do not typically contribute to the seismic risk. Similarly, those seismic interaction scenarios (for example, block walls, falling objects, and displacements which cause impact with nearby elements) which can be demonstrated to have high seismic capacities, have also not resulted in significant risk contribution in past seismic studies. Therefore, these high seismic capacity SSCs and interactions are unlikely to be categorized as HSS and can be screened out from the S0.69 seismic categorization process. This high seismic capacity screening fits into Step Sc of the flow chart in Figure 2-3. The process for screening individual SSCs documented in EPRI 102S287 [11] (the SPID) will form the backbone for this screening approach . Following this approach, SSCs with a HCLPF capacity greater than the calculated screening level HCLPF could be categorized as low safety significant (LSS). I

[AsJan alternate to screening out an SSC, the SSC can be retained in the FPTE model using a ( Commented [RJ17]: RAJ Response 4e surrogate event with a seismic-induced failure probability based on specific fragility analyses.

B.1 Approach As part of the effort to develop the SPID [ 11 ], seismic capacity-based criteria were developed to determine which SSCs should have component specific calculated fragility values to ensure that proper focus was given to those SSCs with the potential to be risk-significant. These criteria were developed using a parametric/sensitivity study [33] which provided the basis for the SPID recommendations. SSCs with capacities above the calculated screening level are not expected to have significant impact on the result of the SPRA analyses, the ranking of accident sequences, or the calculated sequence- or plant-level seismic CDF or LERF values. As such, SSCs with capacities above that screening level would also not be expected to be high safety significant (HSS) components within the S0.69 categorization process.

Section 6.4.3 of the SPID [ 11] identifies the approach to develop a screening HCLPF value for these higher capacity fragilities. Following the SPID approach, a screening HCLPF value is calculated by convolving the fragility of a single element with the site-specific hazard curve such that the SCDF is at most about SE-7 per year. This can be done with trial and error runs using a quantification code or with a spreadsheet with an assttme8 estimated composite variability (for I example, ~c= 0.4) as described in [ 11 ]. This SE-7 screening criteria was developed for the higher seismic hazard plants where seismic typically has a corresponding higher resulting risk. For a medium to low seismic hazard site this screening level of SE-7 could potentially be unconservative, therefore an SCDF value of approximately Yi of the SPID value, or 2.SE-7 is judged to be more appropriate for purposes of this 50.69 categorization screening assessment.

[AsJan alternate to screening out an SSC, the SSC can be retained in the FPIE model by Commented [RJ18]: RAI Responses 4e and calculating a site-specific HCLPF as described below and convolving it with the site seismic 9a B-1

ATTACHMENT 2 Insert Appropriate Auto Text License Entry. /(license is copyright, please delete Criteria for Capacity-Based Screening/or High Capacity SSCs hazard curve to obtain seismic-induced failure probability that may be used rather than the 1.E-04 default value. Otho!! a)'!)'!Hl)'!riatel) jttstif'iea site S)'!eeif'ie sernening a~ttss rna) Ile ttsea.

tToJapply his approach, a seismic fragility must be developed for each SSC that is being assessed Commented [RJ19]: RAI Responses 4e and as part of the categorization process and compared to screening level developed as described 9c above. Criteria for performing fragility calculations are provided in Section 3.3, Separation of Variables Fragility Approach, and Section 3.4, Hybrid Fragility Approach, ofEPRI 3002012994

[34]. These fragility methods define the state of practice in use for SPRAs todayThe fragilit; rnethsaslsg) is olil estaelishea ana them am nt!rnellt!S rnfurenees in the literatt!Hl aeserieing ths rnsthsas. Fst!t RPR-1 Hl)'!Stts that sellseti sl; 88fltt!rn ths fiagilit) )'!rneess arn listed in Taels

~.

+aels !!a Ssismie i;ra!ijility Rsfsrsriess

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~sisl'l'lis i;"ra[3ility 0 i;ii;ilisatisFls Qi,ii1s !!PRI Rsi;isrt Hl19;;!QQ

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Ssisl'l'lie i;ra!ijilil)* oi;i13lieati0Fl Qi,iiss  !!PRI HlQ29!Hl (;;!QQ;;!l [;iii]

~ P1st"1s8sls§) fsr Qs 8ls13iA§ eSiSR'liS Qi,iilar:ies  !!PRI TR 1QA9!i9 (1991l [Ai]

i;ra!ijilitiss o P1sUrnlsls!iJ) fsr OsssssmsRI sf EH~I ~IP GQ11 Sb (1991) [19]

~lt!elsar Pia Rt Ssismie P1ar!iJiR For nuclear plants without existing SPRAs, one challenge will be to produce in-structure seismic responses for use in these fragilities. Development of finite element models and generation of new seismic response analyses using the current seismic hazard shape at the plant site is one option, however, more simplified and conservative approaches could be used when justified by experienced engineers within the structural dynamics field. Sections 6.4 and 6.7 ofEPRI 3002012994 [34] provide the most up to date criteria for scaling in-structure response spectra, consistent with the state of practice for SPRAs todayThese 8fl)'!l8aehes inelt!ae:

8ealiHg ef eHistittg )'!lallt seisrnie 1 BS)'!ense aHal; ses "he1 e the sha)'!BS ef the ttttife1 rn ha!lat El res)'JeHse s)'Jeetra (UIIRS) are sintila1 [JS, 19]

Esti11atisn sf high fHH}t!f.me; seismie 1es)'!snse t!Sing an 8fl)'!llaeh in EPIU 3QQ2QQ13% [37]

, hi eh aeseriees a )'!rneess ts estimate seismie res)'!enses ffir hara melt sites that ha, e gr01rna res)'!snse S)'!eetral )'leaks in the I igh fr@1t!@ne; )'!art sftl e rns)'!snse S)'!eetrnm Ill aaaitisn, it rna; alss Ile )'!SSsilll@ ffll fiagilit) anal; sts ts esnsBr i ati; el; estimate seismie aemanas ttsing si111)'!lifiea 8fl)'!rnaehes 80et1111entea in "SCB 7 [38] ferjt1stif) ing aaaitisnal SSCs that "et1l8 ha, e IICLPF ea)'!aeities alle, e the sernetting tltrnshsl1. "ssessmt.mts ma1e "et!la ha, e ts Ile neeessa1 ii; esnse1 , ati , e Ell iasea ts a1 as highe1 in strnetttrn rns)'!snse S)'!eetl a EI8R 8))

ans aeest111t ffir )'!Stential ariallilit; sfI8R8 rest1lts llasea sn the t1se efOrnse 8flflFSJtirnate rneOrnas .

B-2

ATTACHMENT 3 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 List of Acronyms

Attachment 3 List of Acronyms AC Alternating Current ADAMS Agencywide Documents Access and Management System AEP Annual Exceedance Probability Am Median Ground Acceleration Capacity (Fragility)

ANS American Nuclear Society APLA/B/C PRA Licensing Branch A/B/C within NRC Division of Risk Assessment ASCE American Society of Civil Engineers ASME American Society of Mechanical Engineers ATWS Anticipated Transient Without Scram Br Logarithmic Standard Deviation Representing Random Uncertainty Logarithmic Standard Deviation Representing Systematic or Modeling Bu Uncertainty BWR Boiling Water Reactor COF Core Damage Frequency CDFM Conservative Deterministic Failure Margin CSF Critical Safety Function DC Direct Current EAC Emergency AC Power EOG Emergency Diesel Generator (also shown as DIG or DG)

EPRI Electrical Power Research Institute FPIE Full Power Internal Events (also shown as IEPRA)

FV or F-V Fussell-Vesely FTO Failure to Open HCLPF High Confidence of Low Probability of Failure HPCI High Pressure Coolant Injection Page 1of3

Attachment 3 List of Acronyms HSS High Safety Significance IDP Integrated Decision-Making Panel IPEEE Individual Plant Examination of External Events LAR License Amendment Request LERF Large Early Release Frequency LOCA Loss of Coolant Accident LOOP Loss of Offsite Power LS or LAS LaSalle County Nuclear Generating Station LSS Low Safety Significance MLOCA Medium Break Loss of Coolant Accident NEI Nuclear Energy Institute NPP Nuclear Power Plant NRG Nuclear Regulatory Commission NTTF Near Term Task Force PGA Peak Ground Acceleration PRA Probabilistic Risk Assessment PWR Pressurized Water Reactor RAW Risk Achievement Worth RAI Request for Additional Information RCS Reactor Coolant System RPV Reactor Pressure Vessel SBLC or SLC Standby Liquid Control SBO Station Blackout SCDF Seismic Core Damage Frequency Page 2 of 3

Attachment 3 List of Acronyms SLERF Seismic Large Early Release Frequency Small Break Loss of Coolant Accident or Seismic-Induced Loss of SLOCA Coolant Accident SMA Seismic Margin Assessment SSEL Seismic Shutdown Equipment List or Safe Shutdown Equipment List SPID Screening, Prioritization and Implementation Details SPRA Seismic Probabilistic Risk Assessment SSC Structures, Systems and Component Page 3 of 3