ML12056A047

Enclosure 1 - Recommendation 2.1: Seismic

ML12056A047
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Quad Cities, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle
Issue date:	03/12/2012
From:	Office of New Reactors, Office of Nuclear Reactor Regulation
To:
Gratton C
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ML12056A046	List: ML12053A340 ML12056A047 ML12056A048 ML12056A049 ML12056A050 ML12056A051 ML12056A052
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RECOMMENDATION 2.1: SEISMIC PURPOSE The U.S. Nuclear Regulatory Commission (NRC or Commission) is issuing this information request for the following purposes:

• To gather information with respect to Near-Term Task Force (NTTF) Recommendation 2.1, as directed by staff requirements memoranda (SRM) associated with SECY-11-0124 and SECY-11-0137, and the Consolidated Appropriations Act, for 2012 (Pub Law 112-74), Section 402, to reevaluate seismic hazards at operating reactor sites

• To collect information to facilitate NRCs determination if there is a need to update the design basis and systems, structures, and components (SSCs) important to safety to protect against the updated hazards at operating reactor sites

• To collect information with respect to the resolution of Generic Issue (GI) 199 Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.54(f), addressees are required to submit a written response to this information request.

BACKGROUND The SSCs important to safety in operating nuclear power plants are designed either in accordance with, or meet the intent of Appendix A to 10 CFR Part 100 and Appendix A to 10 CFR Part 50, General Design Criteria (GDC) 2. GDC 2 states that SSCs important to safety at nuclear power plants must be designed to withstand the effects of natural phenomena such as earthquakes, tornados, hurricanes, floods, tsunami, and seiches without loss of capability to perform their intended safety functions. The design bases for these SSCs reflect consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area. The design bases also reflect margin to account for the limited accuracy, quantity, and period of time in which the historical data have been accumulated.

In response to the accident at the Fukushima Dai-ichi nuclear power plant caused by the March 11, 2011, Tohoku earthquake and subsequent tsunami, the Commission established a NTTF to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. The purpose of this letter is to gather information with respect to NTTF Recommendation 2.1 for seismic hazards. Recommendation 2.1, as amended by the SRMs associated with SECY-11-0124 and SECY-11-0137, instructs the NRC staff to issue requests for information to licensees pursuant to Sections 161.c, 103.b, and 182.a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). This information request is for licensees and holders of construction permits under 10 CFR Part 50 to reevaluate the seismic hazards at their sites against present-day NRC requirements and guidance. Based upon this information, the NRC staff will determine whether additional regulatory actions are necessary (e.g., update the design basis and SSCs important to safety) to protect against the updated hazards. In developing Recommendation 2.1, the NTTF recognized that the state of knowledge Enclosure 1

of seismic hazard within the United States (U.S.) has evolved and the level of conservatism in the determination of the original seismic design bases should be reexamined.

Since the issuance of GDC 2, the NRC has developed new regulations, regulatory guidance, and several regulatory programs aimed at enhancements for previously licensed reactors.

These regulatory programs for enhancements are described in Section 4.1.1 of the NTTF Report, Recommendations for Enhancing Reactor Safety in the 21st Century. Two recent programs are the individual plant examinations of external events (IPEEEs) and GI-199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, dated June 9, 2005 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML051600272). The following paragraphs summarize these two programs.

Individual Plant Examination of External Events:

On June 28, 1991, the NRC issued Supplement 4 to Generic Letter (GL) 88-20, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, (ADAMS Accession No. ML031150485). GL 88-20, referred to as the IPEEE program, requested that each licensee identify and report to the NRC all plant-specific vulnerabilities to severe accidents caused by external events. The IPEEE program included the following four supporting objectives:

(1) Develop an appreciation of severe accident behavior.

(2) Understand the most likely severe accident sequences that could occur at the licensees plant under full-power operating conditions.

(3) Gain a qualitative understanding of the overall likelihood of core damage and fission product releases.

(4) Reduce, if necessary, the overall likelihood of core damage and radioactive material releases by modifying, where appropriate, hardware and procedures that would help prevent or mitigate severe accidents.

The external events to be considered in the IPEEE were: seismic events; internal fires; high winds, floods, and other external initiating events, including accidents related to transportation or nearby facilities and plant-unique hazards.

NUREG-1742, Perspectives Gained from the Individual Plant Examination of External Events (IPEEE) Program, issued April, 2002 (ADAMS Accession Nos. ML021270070 and ML021270674), provides insights gained by the NRC from the IPEEE program. Almost all licensees reported in their IPEEE submittals that no plant vulnerabilities were identified with respect to seismic risk (the use of the term vulnerability varied widely among the IPEEE submittals). However, most licensees did report at least some seismic anomalies, outliers, or other concerns. In the few submittals that did identify a seismic vulnerability, the findings were comparable to those identified as outliers or anomalies in other IPEEE submittals. Seventy percent of the plants proposed improvements as a result of their seismic IPEEE analyses. In several responses, neither the IPEEE analyses nor subsequent assessments documented the

potential safety impacts of these improvements, and in most cases, plants have not reported completion of these improvements to the NRC.

Generic Issue 199:

In support of early site permits (ESPs) and combined licenses (COLs) for new reactors, the NRC staff reviewed updates to the seismic source and ground motion models provided by applicants. These seismic updates included new Electric Power Research Institute models to estimate earthquake ground motion and updated models for earthquake sources in the Central and Eastern United States (CEUS), such as those around Charleston, SC, and New Madrid, MO. These reviews identified higher seismic hazard estimates than previously assumed, which may result in an increased likelihood of exceeding the safe-shutdown earthquake (SSE) at operating facilities in the CEUS. The staff determined that based on the evaluations of the IPEEE program, seismic designs of operating plants in the CEUS do not pose an imminent safety concern. At the same time, the staff also recognized that because the probability of exceeding the SSE at some currently operating sites in the CEUS is higher than previously understood, further study was warranted. As a result, the staff concluded on May 26, 2005 (ADAMS Accession No. ML051450456), that the issue of increased seismic hazard estimates in the CEUS should be examined under the Generic Issues Program (GIP).

Generic Issue (GI)-199 was established on June 9, 2005 (ADAMS Accession No. ML051600272). The initial screening analysis for GI-199 suggested that estimates of the seismic hazard for some currently operating plants in the CEUS have increased. The NRC staff completed the initial screening analysis of GI-199 and held a public meeting in February 2008, (ADAMS Accession Nos. ML073400477 and ML080350189) concluding that GI-199 should proceed to the safety/risk assessment stage of the GIP.

Subsequently, during the safety/risk assessment stage of the GIP, the NRC staff reviewed and evaluated the new information received with the ESP/COL submittals, along with 2008 U.S. Geological Survey seismic hazard estimates. The staff compared the new seismic hazard data with the earlier evaluations conducted as part of the IPEEE program. The NRC staff completed the safety/risk assessment stage of GI-199 on September 2, 2010 (ADAMS Accession No. ML100270582), concluding that GI-199 should transition to the regulatory assessment stage of the GIP. The safety/risk assessment also concluded that (1) an immediate safety concern did not exist and (2) adequate protection of public health and safety was not challenged as a result of the new information. The NRC staff presented this conclusion at a public meeting held on October 6, 2010 (ADAMS Accession No. ML102950263). Information Notice 2010-018, Generic Issue 199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, dated September 2, 2010 (ADAMS Accession No. ML101970221), summarizes the results of the GI-199 safety/risk assessment.

For the GI-199 safety/risk assessment, the NRC staff evaluated the potential risk significance of the updated seismic hazards on seismic core damage frequency (SCDF) estimates. The changes in SCDF estimate in the safety/risk assessment for some plants lie in the range of 10-4 per year to 10-5 per year, which meet the numerical risk criteria for an issue to continue to the regulatory assessment stage of the GIP. However, as described in NUREG-1742, there are limitations associated with utilizing the inherently qualitative insights from the IPEEE submittals

in a quantitative assessment. In particular, the staffs assessment did not provide insight into which SSCs are important to seismic risk. Such knowledge is necessary for the NRC staff to determine, in light of the new understanding of seismic hazards, whether additional regulatory action is warranted.

APPLICABLE REGULATORY REQUIREMENTS

• Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, GDC 2, Design Bases for Protection against Natural Phenomena

• 10 CFR 50.54, Conditions of Licenses

• 10 CFR 50.34(a)(1), (a)(3), (a)(4), (b)(1), (b)(2), and (b)(4)

• Appendix A, Seismic and Geologic Siting Criteria for Nuclear Power Plants, to 10 CFR Part 100, Reactor Site Criteria

• 10 CFR 100.23, Geological and Seismic Siting Criteria The seismic design bases for currently operating nuclear power plants were either developed in accordance with, or meet the intent of GDC 2 and 10 CFR Part 100, Appendix A. Although the regulatory requirements in Appendix A to 10 CFR Part 100 are fundamentally deterministic, the NRC process for determining the seismic design basis ground motions for new reactor applications after January 10, 1997, as described in 10 CFR 100.23, requires that uncertainties be addressed through an appropriate analysis such as a probabilistic seismic hazard analysis.

DISCUSSION Recommendation 2.1, as amended by the SRMs associated with SECY-11-0124 and SECY-11-0137, instructs the NRC staff to issue requests for licensees to reevaluate the seismic hazards at their sites using present-day NRC requirements and guidance, and identify actions that are planned to address plant-specific vulnerabilities1 associated with the updated seismic hazards. Recommendation 2.1 for seismic hazards will be implemented in two phases as follows:

• Phase 1: Issue 10 CFR 50.54(f) letters to all licensees to reevaluate the seismic hazard at their sites using updated seismic hazard information and present-day regulatory guidance and methodologies and, if necessary, to perform a risk evaluation.

• Phase 2: If necessary, and based upon the results of Phase 1, determine whether additional regulatory actions are necessary (e.g., update the design basis and SSCs important to safety) to protect against the updated hazards.

1 A definition of vulnerability in the context of this enclosure is as follows: Plant-specific vulnerabilities are those features important to safety that when subject to an increased demand due to the newly calculated hazard evaluation have not been shown to be capable of performing their intended safety functions.

To implement NTTF Recommendation 2.1, the staff is utilizing the general process developed for GI-199 as presented in the draft GL for Gl-199 (ADAMS Accession No. ML11710783). This process, described in Attachment 1, asks each addressee to provide information about the current hazard and potential risk posed by seismic events using a progressive screening approach. Depending on the comparison between the reevaluated seismic hazard and the current design basis, the result is either no further risk evaluation or the performance of a seismic risk assessment. Risk assessment approaches acceptable to the staff include a seismic probabilistic risk assessment (SPRA), or a seismic margin assessment (SMA).

Present-day NRC requirements and guidance with respect to characterizing seismic hazards use a probabilistic approach in order to develop a risk-informed performance-based ground motion response spectrum (GMRS) for the site. This approach is described in Regulatory Guide (RG) 1.208, A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion. RG 1.208 recommends the use of the Senior Seismic Hazard Analysis Committee (SSHAC) approach for treatment of expert judgment and quantifying uncertainty in order to develop seismic source and ground motion models for a Probabilistic Seismic Hazard Analysis used to develop the GMRS for a site.

The SMA approach should be the NRC SMA approach (e.g., NUREG/CR-4334, An Approach to the Quantification of Seismic Margins in Nuclear Power Plants, issued in August 1985 (ADAMS Accession No. ML090500182) as enhanced for full-scope plants in NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities). Part 10 of the American Society of Mechanical Engineers/American Nuclear Society standard (ASME/ANS), RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, provides an acceptable approach for determining the technical adequacy of the SMA approach used to respond to this information request. The SMA approach should include both core damage (accident prevention) and large early release (accident mitigation).

The NRC staff recommends that the SPRA approach at least be a Level 1 with an estimate of large early release frequency (LERF). By including containment performance and extending to Level 2 (including LERF) additional mitigation features that may be under consideration can be incorporated into the analyses. One acceptable approach for determining the technical adequacy of the SPRA is described in RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, (ADAMS Accession No. ML090410014) and ASME/ANS RA-Sa-2009). Consistent with the NRCs probabilistic risk assessment (PRA) policy statement, the technical adequacy of the methods used to develop the requested information must be sufficient to provide confidence in the results, such that the seismic risk information can be used in regulatory decision-making.

REQUESTED ACTIONS Addressees are requested to perform a reevaluation of the seismic hazards at their sites using present-day NRC requirements and guidance to develop a GMRS. Recently, new consensus seismic source models for the CEUS (NUREG-2115, Central and Eastern United States Seismic Source Characterization for Nuclear Facilities), referred to as the Central and Eastern United States Seismic Source Characterization, have been completed using a SSHAC Level 3 process. Addressees whose plants are located in the CEUS will be able to use this new seismic

source model to characterize the hazard for their plants. Addressees whose plants lie in the Western United States (WUS) are requested to develop seismic source and ground motion models to characterize their regional and site-specific seismic hazards. Consistent with current practice for 10 CFR Part 52, new reactor licensing, WUS addressees should perform a SSHAC Level 3 study to develop a probabilistic seismic hazard analysis.

Addressees are requested to submit, along with the hazard evaluation, an interim evaluation and actions planned or taken to address the reevaluated hazard where it exceeds the current design basis.

While the seismic hazard reevaluation is being performed, NRC staff and stakeholders will continue interacting to develop strategies for screening, prioritization, and potential interim actions as well as implementation guidance for the risk evaluation. For plants where the reevaluated hazard exceeds the current design basis, addressees may opt to perform an SPRA.

In addition, an SPRA, rather than a SMA, may be necessary for cases where the SMA screening tables are not usable due to a higher reevaluated hazard (i.e., GMRS). For all other plants where the reevaluated hazard exceeds the current design basis, the NRC will provide guidance on when an SMA option can be used. Factors that the staff will consider to determine whether an SPRA or an SMA is appropriate are (1) the extent to which the reevaluated hazard (GMRS) exceeds the current design basis (SSE), (2) the absolute seismic hazard based on an examination of the probabilistic seismic hazard curves for the site, and (3) previous estimates of plant capacity (e.g., IPEEE insights). The priority for the subsequent completion of the risk assessments by the addressees will also be based on the above factors. For example, as part of the GI-199 safety/risk assessment, the NRC staff found that assuming a factor of 1.3 times the SSE, combined with updated seismic hazard curves, distinguished between plants with lower and higher risk estimates.

Along with an assessment of reactor integrity, the NTTF recommended an evaluation of the spent fuel pool (SFP) integrity. The addressees evaluation should consider all seismically induced failures that can lead to draining of the SFP. The evaluation should consider SFP walls, liner, penetrations (cooling water supplies or returns, drains), transfer gates and seals, seals and bellows between the SFP, transfer canal, and reactor cavity, sloshing effects (including loss of SFP inventory, wave-induced failures of gates, and subsequent flooding),

siphon effects caused by cooling water pipe breaks, and other relevant effects that could lead to a significant loss of inventory of the SFP.

REQUESTED INFORMATION The NRC requests that each addressee provide the following information (see Attachment 1 for additional details):

Seismic Hazard Evaluation (1) site-specific hazard curves (common fractiles and mean) over a range of spectral frequencies and annual exceedance frequencies (2) site-specific, performance-based GMRS developed from the new site-specific seismic hazard curves at the control point elevation(s)

(3) SSE ground motion values including specification of the control point elevation(s)

(4) comparison of the GMRS and SSE (if the GMRS is completely bounded by the SSE, an interim action plan or a risk evaluation is not necessary. However, if the GMRS exceeds the SSE only at higher frequencies information related to the functionality of high-frequency sensitive SSCs is requested. Attachment 1 provides further details)

(5) additional information such as insights from NTTF Recommendation 2.3 walkdown and estimates of plant seismic capacity developed from previous risk assessments to inform NRC screening and prioritization (6) interim evaluation and actions taken or planned to address the higher seismic hazard relative to the design basis, as appropriate, prior to completion of the risk evaluation described below (7) selected risk evaluation approach (if necessary)

Seismic Risk Evaluation (8) SMA or SPRA (depending on criteria discussed above)

A. For plants that perform a SMA, the following information is requested:

(1) description of the methodologies used to quantify the seismic margins of high confidence of low probability of failure (HCLPF) capabilities of SSCs, together with key assumptions (2) detailed list of the SSC seismic margin values with reference to the method of seismic qualification, the dominant failure modes, and the source of information (3) for each analyzed SSC, the parameter values defining the seismic margin (e.g., the HCLPF capacity and any other parameter values such as the median acceleration capacity (C50) and the logarithmic standard deviation or beta values) and the technical bases for the values (4) general bases for screening SSCs (5) description of the SMA, including the development of its logic models, the seismic response analysis, the results of the evaluation of containment performance, the results of the screening analysis, the results of the plant seismic walkdown, the identification of critical failure modes for each SSC, and the calculation of HCLPF capacities for each SSC included in the SMA logic model

(6) description of the process used to ensure that the SMA is technically adequate, including the dates and findings of peer reviews (7) identified plant-specific vulnerabilities and actions planned or taken B. For plants that perform a SPRA, the following information is requested:

(1) list of the significant contributors to SCDF for each seismic acceleration bin, including importance measures (e.g., Risk Achievement Worth, Fussell-Vesely and Birnbaum)

(2) a summary of the methodologies used to estimate the SCDF and LERF, including the following:

i. methodologies used to quantify the seismic fragilities of SSCs, together with key assumptions ii. SSC fragility values with reference to the method of seismic qualification, the dominant failure mode(s), and the source of information iii. seismic fragility parameters iv. important findings from plant walkdowns and any corrective actions taken

v. process used in the seismic plant response analysis and quantification, including the specific adaptations made in the internal events PRA model to produce the seismic PRA model and their motivation vi. assumptions about containment performance (3) description of the process used to ensure that the SPRA is technically adequate, including the dates and findings of any peer reviews (4) identified plant-specific vulnerabilities and actions that are planned or taken (9) SFP Evaluation A. description of the procedures used to evaluate the SFP integrity B. results of the evaluation C. identified actions that have been taken or that will be taken to address vulnerabilities associated with the SFP integrity

REQUIRED RESPONSE In accordance with 10 CFR 50.54(f), an addressee must respond as described below:

1. Within 60 days of the date of the NRCs issuance of guidance on screening and prioritization criteria, and the implementation details of the risk assessment, each addressee is requested to submit: (1) its intention to follow the NRC-developed guidance2, or (2) an alternative approach, including acceptance criteria.

2. Within 1.5 years of the date of this information request, each CEUS addressee is requested to submit a written response consistent with the requested information, seismic hazard evaluation, items 1 through 7 above. Within approximately 30 days of receipt of the last addressee submittal, the NRC staff will have determined the acceptability of the licensees proposed risk evaluation approach, if necessary, and priority for completion.

3. Within 3 years of the date of this information request, each WUS addressee is requested to submit a written response consistent with the requested information, seismic hazard evaluation, items 1 through 7 above. Within approximately 30 days of receipt of the last addressee submittal, the NRC staff will have determined the acceptability of the licensees proposed risk evaluation approach, if necessary, and priority for completion.

4. For hazard reevaluations that the NRC determines demonstrate the need for a higher priority, addressees are requested to complete the risk evaluation (items 8B and 9 above) over a period not to exceed 3 years from the date of the prioritization.

5. For hazard reevaluations that the NRC determines do not demonstrate the need for a higher priority, addressees are requested to complete the risk evaluation (items 8A or 8B and 9 above) over a period not to exceed 4 years from the date of the prioritization.

If an addressee cannot meet the requested response date, the addressee must provide a response within 90 days of the date of this information request and describe the alternative course of action that it proposes to take, including the basis of the acceptability of the proposed alternative course of action and estimated completion dates.

The required written response should be addressed to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, 11555 Rockville Pike, Rockville, MD 20852, under oath or affirmation under the provisions of Sections 161.c, 103.b, and 182.a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). In addition, addressees should submit a copy of the response to the appropriate regional administrator.

2 The NRC staff will develop screening and prioritization criteria, and the implementation details of the risk assessment, including criteria for identifying vulnerabilities. This information is scheduled to be developed by November 30, 2012 and the NRC staff will interact with stakeholders, as appropriate during this process.

Attachment 1 to Seismic Enclosure 1 Introduction This Attachment describes an acceptable process for developing the information requested by the U.S. Nuclear Regulatory Commission (NRC). Figure 1 illustrates the process, which is based on a progressive screening approach. The following paragraphs provide additional discussion about each individual step in Figure 1.

Step 1. Addressees should develop site-specific base rock and control point elevation hazard curves (i.e., corresponding to fractile levels of 0.05, 0.16, 0.50, 0.84, and 0.95 and the mean) over a range of spectral frequencies (0.5 Hz, 1 Hz, 2.5 Hz, 5 Hz, 10 Hz, and 25 Hz and peak ground acceleration - PGA) and annual exceedance frequencies (1x10-6 and higher) determined from a probabilistic seismic hazard analysis (PSHA) as follows:

• Addressees of plants located in the Central and Eastern United States (CEUS) are expected to use the CEUS Seismic Source Characterization (CEUS-SSC) model (NUREG-2115, Central and Eastern United States Seismic Source Characterization for Nuclear Facilities) and the appropriate Electric Power Research Institute (2004, 2006) ground motion prediction equations. Regional and local refinements of the CEUS-SSC are not necessary for this evaluation.

• Addressees of plants located in the Western United States (Columbia, Diablo Canyon, Palo Verde, and San Onofre) should develop an updated, site-specific PSHA. Any new or updated seismic hazard assessment should consider all relevant data, models, and methods in the evaluation of seismic sources and ground motion models. Consistent with Regulatory Guide (RG) 1.208, A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion, addressees should use a Senior Seismic Hazard Analysis Committee (SSHAC) study, as described in NUREG/CR-6372, Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts. Consistent with current practice, as described in NUREG-2117, Practical Implementation Guidelines for SSHAC Level 3 and 4 Hazard Studies, a SSHAC Level 3 study should be performed.

• To remove non-damaging lower-magnitude earthquakes, addressees should either use a lower bound magnitude cutoff of moment magnitude (Mw) 5 or the cumulative absolute velocity (CAV) filter for the PSHA. The CAV filter should be limited to Mw less than or equal to 5.5.

• Addressees should use site response methods 2 or 3, as described in NUREG/CR-6728, Technical Basis for Revision of Regulatory Guidance on Design Ground Motions:

Hazard- and Risk-consistent Ground Motion Spectra Guidelines. The dynamic site response should be determined through analyses based on either time history or random vibration theory. The subsurface site response model, for both soil and rock sites, should extend to sufficient depth to reach the generic rock conditions as defined in the ground motion models used in the PSHA. In addition, a randomization procedure should be used that appropriately represents the amount of subsurface information at a

• given site. In addition, the randomization procedure should accommodate the variability in soil depth (including depth to generic rock conditions), shear-wave velocities, layer thicknesses, and strain dependant nonlinear material properties at the site. Generally, at least 60 convolution analyses should be performed to define the mean and standard deviation of the site response. Site amplification curves should be developed over a broad range of annual exceedance frequencies (1x10-6 and higher) to facilitate estimation of seismic core damage frequency.

• Addresses should document the low- and high-frequency controlling earthquakes at frequencies of 10-4 and 10-5 per year.

• Addressees should use the site-specific hazard curves to develop a performance-based ground motion response spectrum (GMRS) for the site, using the guidance in RG 1.208.

The site-specific GMRS should be determined and clearly specified at the same elevation as the design-basis safe shutdown earthquake (SSE) ground motion assuming a site profile with a free surface above the control point elevation.

Step 2. Addressees are requested to provide the new seismic hazard curves, the GMRS, and the SSE in graphical and tabular format. Addressees are also requested to provide soil profiles used in the site response analysis as well as the resulting soil amplification functions.

Step 3. If the SSE is greater than or equal to the GMRS at all frequencies between 1 and 10 Hz and at the PGA anchor point, then addressees may terminate the evaluation (Step 4)3 after providing a confirmation, if necessary, that SSCs, which may be affected by high-frequency ground motion, will maintain their functions important to safety.

Step 4. This step demonstrates termination of the process for resolution of NTTF, Recommendation 2.1 for plants whose SSE is greater than the calculated GMRS.

Step 5. Based on NRC screening criteria, addressees will be requested to perform a seismic margins analysis (SMA) or a seismic probabilistic risk assessment (SPRA). If addressees perform an SPRA, then they are requested to follow Steps 6a and 7a. If addressees perform an SMA, then they are requested to follow Steps 6b and 7b.

Step 6a. It is requested that addressees that perform an SPRA ensure that the SPRA is technically adequate for regulatory decision making and includes an evaluation of containment performance and integrity. RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, provides an acceptable approach for determining the technical adequacy of an SPRA used to respond to this information request.

Step 6b. It is requested that addressees that perform an SMA use a composite spectrum review level earthquake, defined as the maximum of the GMRS and SSE at each spectral frequency. The SMA should also include an evaluation of containment performance and 3

For plants with only a high frequency ground motion exceedance (above 10 Hz), the documentation should also include a confirmation that affected plant structures and equipment at various elevations will maintain their functions important to safety at the higher acceleration levels.

integrity. The American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) RA-Sa-2009 provides an acceptable approach for determining the technical adequacy of an SMA used to respond to this information request.

Step 7a. Document and submit the results of the SPRA to the NRC for review. The Requested Information section in the main body of Enclosure 1 identifies the specific information that is requested. In addition, addresses are requested to submit an evaluation of the SFP integrity.

Step 7b. Document and submit the results of the SMA to the NRC for review. The Requested Information section in the main body of Enclosure 1 identifies the specific information that is requested. In addition, addresses should submit an evaluation of the SFP integrity.

Step 8. Submit plans for actions that evaluate seismic risk contributors. NRC staff, industry, and other stakeholders will continue to interact to develop acceptance criteria in order to identify potential vulnerabilities.

Step 9. The information provided in Steps 6 through 8 will be evaluated in Phase 2 to consider any additional regulatory actions.

Develop new seismic hazard curves and GMRS 1

Submit new seismic hazard curves, GMRS, and interim actions 2

No GMRS > SSE?

3 Yes NRC Screening/Prioritization 5

SPRA SMA Develop SPRA Develop SMA 6a 6 Submit SPRA results and Submit SMA results and SFP evaluation SFP evaluation 7a 7 Submit proposed actions, if any, to evaluate seismic 8 risk contributors No further 4

action 9 Phase 2 Figure 1. Development of Requested Information and Its Use in Regulatory Analysis.

Enclosure 1 Reference List Atomic Energy Act of 1954, as amended, Section 103.b, 161.c, and 182.a SECY 11-0137, Prioritization of Recommended Actions to be Taken in Response to Fukushima Lessons Learned, Agencywide Documents Access and Management System (ADAMS)

Accession No. ML11272A111, October 3, 2011.

SECY 11-0124, Recommended Action to be taken without Delay from the Near-Term Task Force Report, ADAMS Accession No. ML11245A158, September 9, 2011.

10 CFR 50.54(f) - Conditions of Licenses Recommendations for Enhancing Reactor Safety in the 21st Century: The Near-term Task Force Review of Insights from the Fukushima Dai-ichi Accident, ADAMS Accession No. ML111861807, July 12, 2011.

Generic Issue (GI)-199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, ADAMS Accession No. ML051600272, June 9, 2005.

NRC Generic Letter 1988-020, Supplement 4: Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), ADAMS Accession No. ML031150485, June 28, 1991.

NUREG-1742, Perspectives Gained from the Individual Plant Examination of External Events (IPEEE) Program - Final Report, ADAMS Accession Nos. ML021270070 and ML021270674, April 2002.

Identification of a Generic Seismic Issue, ADAMS Accession No. ML051450456, May 26, 2005.

Results of Initial Screening of Generic Issue 199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants. ADAMS Accession No. ML073400477, February 1, 2008.

02/06/2008 Summary of Category 2 Public Meeting with the Public and Industry to Discuss Generic Issue 199, Implications of Updated Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, ADAMS Accession No. ML080350189, February 8, 2008.

Results of Safety/Risk Assessment of Generic Issue 199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, ADAMS Accession No. ML100270582, September 2, 2010.

10/6/201 - Public Meeting Summary on Safety/Risk Assessment Results for Generic Issue 199, Implication of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, ADAMS Accession No. ML102950263, October 29, 2010.

NRC Information Notice 2010-018: Generic Issue 199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, ADAMS Accession No. ML101970221, September 2, 2010.

Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, GDC 2, Design Bases for Protection against Natural Phenomena 10 CFR 50.34(a)(1), (a)(3), (a)(4), (b)(1), (b)(2), and (b)(4), Contents of Applications; technical information.

Appendix A, Seismic and Geologic Siting Criteria for Nuclear Power Plants, to 10 CFR Part 100, Reactor Site Criteria Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities (Volume 60, page 42622, of the Federal Register (60 FR 42622)).

NUREG/BR-0058 Revision 4, Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission, ADAMS Accession No. ML042820192, September 30, 2004.

Draft NRC Generic Letter 2011-XX: Seismic Risk Evaluations for Operating Reactors, ADAMS Accession No. ML111710783, July 26, 2011.

NUREG/CR-4334, An Approach to the Quantification of Seismic Margins in Nuclear Power Plants, ADAMS Accession No. ML090500182, August 1985.

Part 10 of the American Society of Mechanical Engineers/American Nuclear Society standard, RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, ADAMS Accession No. ML090410014, March 2009.

Electric Power Research Institute (EPRI), CEUS Ground Motion Project Final Report, EPRI Technical Report 1009684, December 2004.

Electric Power Research Institute (EPRI), Program on Technology Innovation: Truncation of the Lognormal Distribution and Value of the Standard Deviation for Ground Motion Models in the Central and Eastern United States, Technical Report 1014381, Palo Alto, California, August 2006.

American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS)

RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, 2009.

NUREG/CR-6372, Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts, ADAMS Accession Nos. ML080090003 and ML080090004, April 30, 1997.

NUREG-2117, Practical Implementation Guidelines for SSHAC Level 3 and 4 Hazard Studies NUREG/CR-6728, Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Hazard- and Risk-Consistent Ground Motion Spectra Guidelines, ADAMS Accession No. ML013100232, October 2001.

Regulatory Guide 1.208, A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion, ADAMS Accession No. ML070310619, March 11, 2007.

NUREG-2115, Central and Eastern United States Seismic Source Characterization for Nuclear Facilities