UNITED STATES 


NUCLEAR REGULATORY COMMISSION 


WASHINGTON, D.C. 20555-0001 

March 12, 2012 

All Power Reactor Licensees and 

Holders of Construction Permits in 

Active or Deferred Status 

SUBJECT: 	 REQUEST FOR INFORMATION PURSUANT TO TITLE 10 OF THE CODE OF 

FEDERAL REGULATIONS 50.54(f) REGARDING RECOMMENDATIONS 2.1,2.3, 

AND 9.3, OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM 

THE FUKUSHIMA DAI-ICHI ACCIDENT 

This letter is being issued in accordance with the provisions of Sections 161.c, 103.b, and 182.a 

of the Atomic Energy Act of 1954, as amended (the Act), and the U.S. Nuclear Regulatory 

Commission (NRC or Commission) regulation in Title 10 of the Code of Federal Regulations 

(10 CFR), Part 50, Section 50.54(f). Pursuant to these provisions of the Act or this regulation, 

you are required to provide further information to support the evaluation of the NRC staff 

recommendations for the Near-Term Task Force (NTTF) review of the accident at the 

Fukushima Dai-ichi nuclear facility. The review will enable the staff to determine whether the 

nuclear plant licenses under your responsibility should be modified, suspended, or revoked. For 

combined license (COL) holders under 10 CFR Part 52, the issues in NTTF Recommendation 

2.1 and 2.3 regarding seismic and flooding reevaluations and walkdowns are resolved. 

Therefore, COL holders are not required to respond to Enclosures 1 through 4 of this letter. 

Similarly, information requests in Enclosures 3 and 4 are not applicable to holders of 

construction permits under 10 CFR Part 50. Operating power reactor licensees under 

10 CFR Part 50 are required to respond to all of the information requests. 

BACKGROUND 

Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the 

March 11, 2011, Great T5hoku Earthquake and subsequent tsunami, the NRC established the 

NTTF in response to Commission direction. The NTTF Charter, dated March 30, 2011, tasked 

the NTTF with conducting a systematic and methodical review of NRC processes and 

regulations and determining if the agency should make additional improvements to its regulatory 

system. Ultimately, a comprehensive set of recommendations contained in a report to the 

Commission (dated July 12, 2011, SECY-11-0093 (Agencywide Documents Access and 

Management System (ADAMS) Accession No. ML 111861807)) was developed using a decision 

rationale built around the defense-in-depth concept in which each level of defense-in-depth 

(namely prevention, mitigation, and emergency preparedness (EP)) is critically evaluated for its 

completeness and effectiveness in performing its safety function. 

The current regulatory approach, and the resultant plant capabilities, gave the NTTF and the 

NRC the confidence to conclude that an accident with consequences similar to the Fukushima 

accident is unlikely to occur in the United States (U.S.). The NRC concluded that continued 

plant operation and the continuation of licensing activities did not pose an imminent risk to 

public health and safety. 
- 2 ­

On August 19, 2011, following issuance of the NTTF report, the Commission directed the NRC 

staff in staff requirements memorandum (SRM) for SECY-11-0093 (ADAMS Accession No. 

ML 112310021), in part, to determine which of the recommendations could and should be 

implemented without unnecessary delay. 

On September 9,2011, the NRC staff provided SECY-11-0124 to the Commission (ADAMS 

Accession No. ML 11245A 158). The document identified those actions from the NTTF report 

that should be taken without unnecessary delay. As part of the October 18, 2011, SRM for 

SECY-11-0124 (ADAMS Accession No. ML 112911571), the Commission approved the staff's 

proposed actions, including the development of three information requests under 

10 CFR 50.54(f). The information collected would be used to support the NRC staff's evaluation 

of whether further regulatory action was needed in the areas of seismic and flooding design, 

and EP. 

On December 23, 2011, the Consolidated Appropriations Act, Public Law 112-074, was Signed 

into law. Section 402 of the law also requires a reevaluation of licensees' design basis for 

external hazards, and expands the scope to include other external events, as described below: 

The Nuclear Regulatory Commission shall require reactor licensees to 

re-evaluate the seismic, tsunami, flooding, and other external hazards at their 

sites against current applicable Commission requirements and guidance for such 

licensees as expeditiously as possible, and thereafter when appropriate, as 

determined by the Commission, and require each licensee to respond to the 

Commission that the design basis for each reactor meets the requirements of its 

license, current applicable Commission requirements and guidance for such 

license. Based upon the evaluations conducted pursuant to this section and 

other information it deems relevant, the Commission shall require licensees to 

update the design basis for each reactor, if necessary. 

Reevaluation of the design basis with respect to other external events will be requested later as 

a separate action from this letter. However, licensees are encouraged to consider this when 

performing the Recommendation 2.3 walkdowns for flooding. 

In the context of Recommendation 2.1 of this 10 CFR 50.54(f) letter, the NRC staff definition of 

vulnerability1 is broad enough to capture both prevention and mitigation aspects and also 

include features of protection such as hardware, procedures, temporary measures, and 

potentially available off-site resources. Such a definition allows both licensees and the NRC 

staff to assess plant response to a natural hazard event as an integrated system providing 

consideration for all available resources. Information resulting from such an evaluation will help 

the staff decide upon the most appropriate regulatory action focusing on the most beneficial 

safety enhancements. 

1 For the purpose of this document, plant-specific vulnerabilities are defined as 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. 
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ACTION 

The NRC has concluded that it requires the information requested in the enclosures to this letter 

to verify the compliance with your plant's design basis and to determine if additional regulatory 

actions are appropriate. Therefore, you are required, pursuant to Section 182(a) of the Atomic 

Energy Act of 1954, as amended, and 10 CFR 50.54(f), to submit a response to this letter. You 

must confirm receipt of this letter within 30 days, however, each attachment contains a 

topic-specific schedule for response. Your response must be written and signed under oath or 

affirmation. 

The NRC has provided information in each enclosure on acceptable approaches for responding 

to the information requests. Alternate approaches with appropriate justification will be 

considered. 

This request contains information collection requirements that are subject to the Paperwork 

Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections were approved 

by the Office of Management and Budget (OMB) under an expedited clearance, approval 

number 3150-0211, which expires September 30,2012. Prior to the expiration date, the NRC 

will submit the collection to OMB for renewal. 

The burden for these information collections is estimated to average 13,300 hours per 

response, as detailed in Table 1. This estimate includes the time for reviewing instructions, 

searching existing data sources, gathering data, performing necessary analyses, and 

completing and reviewing the collection of information. These estimates represent the average 

level of effort per plant; actual levels of effort may vary depending upon the results of the hazard 

analyses. Send comments regarding this burden estimate or any other aspect of these 

information collections, including suggestions for reducing the burden, to the Information 

Services Branch (T-5 F53), U.S. Nuclear Regulatory CommiSSion, Washington, DC 

20555-0001, or by email to INFOCOLLECTS.RESOURCE@NRC.GOV; and to the Desk Officer, 

Office of Information and Regulatory Affairs, NEOB-10202, (3150-0211), Office of Management 

and Budget, Washington, DC 20503. 

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for 

information or an information collection requirement unless the requesting document displays a 

currently valid OMB control number. 

Table 1 Burden Estimate (hours) 

Hazard 

Evaluation 

RiSk/Integrated 

Assessment Walkdowns 

EP 

Communications 

EP 

Staffing 

Enclosure 1 1700 3500 N/A N/A N/A 

Enclosure 2 1300 2700 N/A N/A N/A 

Enclosure 3 N/A N/A 2000 N/A N/A 

Enclosure 4 N/A N/A 2000 N/A N/A 

Enclosure 5 N/A N/A N/A 50 50 
-4­

In accordance with 10 CFR 2.390, "public inspections, exemptions, and requests for 

withholding," a copy of this letter and your response will be made available for inspection and 

copying at the NRC Website at www.nrc.gov, and/or at the NRC Public Document Room. If you 

believe that any of the information to be submitted meets the criteria in 10 CFR 2.390 for 

withholding from public disclosure, you must include sufficient information, as required by the 

subsection, to support such a determination. 

INFORMATION REQUEST JUSTIFICATION 

Hazard Reevaluations and Walkdowns 

Current NRC regulations and associated regulatory guidance provide a robust regulatory 

approach for the evaluation of site hazards associated with natural phenomena. However, this 

framework has evolved over time as new information regarding site hazards and the potential 

consequence has become available. As a result, the licensing basis, design, and level of 

protection from natural phenomena differ among the existing operating reactors in the U.S., 

depending on when the plant was constructed and licensed for operation. Additionally, the 

assumptions and factors that were considered in determining the level of protection necessary 

at these sites vary depending on a number of contributing factors. To date, the NRC has not 

undertaken a comprehensive re-establishment of the design basis for existing plants to reflect 

the current state of knowledge or current licensing criteria. 

Protection from natural phenomena is critical for safe operation of nuclear power plants. Failure 

to protect structures, systems, and components (SSCs) important to safety from natural 

phenomena with appropriate safety margins has the potential to result in common-cause 

failures with significant consequences, as was demonstrated at Fukushima. Additionally, the 

consequences of an accident from some natural phenomena may be aggravated by a 

"cliff-edge" effect, in that a small increase in the hazard (e.g., flooding level) may sharply 

increase the number of SSCs affected. 

As the state of knowledge of these hazards has evolved significantly since the licensing of many 

of the plants within the U.S., and given the demonstrated consequences from Fukushima, it is 

necessary to confirm the appropriateness of the hazards assumed for U.S. plants and their 

ability to protect against them. 

In accordance with Commission direction, the NRC staff is implementing the following: 

A hazard evaluation consistent with Recommendation 2.1 will be implemented in two phases as 

follows: 

• 	 Phase 1: Issue 10 CFR 50.54(f) letters to all licensees to request that they reevaluate 

the seismic and flooding hazards at their sites using updated seismic and flooding 

hazard information and present-day regulatory guidance and methodologies and, if 

necessary, to request they perform a risk evaluation. The evaluations associated with 

the requested information in this letter do not revise the design basis of the plant. This 

letter implements Phase 1. 
-5­

• 	 Phase 2: Based upon the results of Phase 1, the NRC staff will determine whether 

additional regulatory actions are necessary (e.g., update the design basis and SSCs 

important to safety) to provide additional protection against the updated hazards. 

The NRC staffs goal is to complete Phase 1 and collect sufficient information to make a 

regulatory decision for most plants within five years. It is anticipated that collection of this 

information for all plants will take no longer than seven years. 

Information collection on hazard protection walkdowns consistent with Recommendation 2.3 will 

be implemented in a single-phase. The results from these walkdowns are expected to capture 

any degraded, non-conforming conditions, and cliff-edge effects for flooding so that they are 

addressed by the licensee's corrective action program and will provide input to 

Recommendation 2.1. It is anticipated that this effort will be completed within approximately 

one year. 

Emergency Preparedness 

Further, if mitigation is not successful in preventing the release of radioactive materials from the 

plant, EP provides additional defense-in-depth to minimize exposure to radiation to the public. 

The accident at Fukushima reinforced the need for effective EP, the objective of which is to 

ensure the capability to implement effective measures to mitigate the consequences of a 

radiological emergency. The accident at Fukushima highlighted the need to determine and 

implement the required staff to fill all necessary positions responding to a multi-unit event. 

Additionally, there is a need to ensure that the communication equipment relied upon to 

coordinate the event response during a prolonged station blackout can be powered. 

The reevaluation and related analysis being conducted under this request are justified by the 

need to enhance those EP measures that support the prevention or mitigation of core damage 

and uncontrolled release of radioactive material. The justification in this letter, as well as the 

background and discussions in each of its enclosures, provide the reasoning and justification for 

this request. Moreover, the reevaluation and related analysis will serve to meet NRC's 

obligation under the Consolidated Appropriations Act, for 2012 (Pub Law 112-74), Section 402, 

and also affords licensees the opportunity to inform the NRC regarding safety-related decisions. 
- 6­

If you have any questions on this matter, please contact your NRC licensing Project Manager. 

Sincerely, 

~L~ Office of Nuclear Reactor Regulation 

Michael . Johnson, Director 

Office of New Reactors 

Enclosures: 

1. Recommendation 2.1: Seismic 

2. Recommendation 2.1: Flooding 

3. Recommendation 2.3: Seismic 

4. Recommendation 2.3: Flooding 

5. Recommendation 9.3: EP 

6. Licensees and Holders of Construction Permits 

cc: Listserv 
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 NRC's 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 
-2­

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 21 st 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 (lPEEE) for Severe Accident Vulnerabilities," (ADAMS 

Accession No. ML031150485). GL 88-20, referred to as the IPEEE program, requested that 

each I[censee 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 licensee's 

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 
-3­

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 beexamined.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 safetylrisk assessment stage of the GIP, the NRC staff reviewed and 

evaluated the new information received with the ESP/Cal 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. Ml1 02950263). 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 safetylrisk 

assessment. 

For the GI-199 safetylrisk 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 
-4­

in a quantitative assessment. In particular, the staff's 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 CFR1 00.23. "Geological and Sejsmic Siting Criteria" 

The seismic design bases for currently operating nuc,lear 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 vulnerabilities 1 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 vlIfnerability 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 nev/ly calculated hazard evaluation have not been shown to be 

capable of performing their intended safety functions, 
-5­

To implement NTTF Recommendation 2.1, the staff is utilizing the general process developed 

for GI-199 as presented in the draft GL for GI-199 (ADAMS Accession No. ML 11710783). 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 

Level1/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 

NRC's 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 
-6­

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 (Le., 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 w.hich 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 ($) 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 NTIF recommended an evaluation of the 

spent fuel pool (SFP) integrity. The addressee's 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 rang~ 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 elevatlon(s) 
-7­

(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 highfrequency 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 sse, the parameter values defining the seismic margin 

(e.g., the HCLPF capacity and any other parameter values such as the 

median acceleration capacity (Cso) 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 
-8­

(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, FussellVesely 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 
-9­

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 NRC's 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 licensee's 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 

licensee's 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 88 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 88 

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 (Le., 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, 10Hz, and 25 Hz and peak 

ground acceleration - PGA) and annual exceedance frequencies (1 x 10-6 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 SiteSpecific 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/CR6728, "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 
-2­

• 	 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 (1 x1 0·6 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 an'd 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 10Hz 

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 10Hz), the documentation should also include a 

confirmation that affected plant structures and equipment at variolls elevations will maintain their functions important to safety at the 

higher acceleration levels. 
-3­

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. 
-4­

Figure 1. Development of Requested Information and Its Use in Regulatory Analysis. 
-5­

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. ML 11272A 111, October 3, 2011. 

SECY 11-0124, "Recommended Action to be taken without Delay from the Near-Term Task 

Force Report," ADAMS Accession No. ML 11245A 158, September 9, 2011. 

10 CFR 50.54(f) - "Conditions of Licenses" 

"Recommendations for Enhancing Reactor Safety in the 21 st Century: The Near-term Task 

Force Review of Insights from the Fukushima Dai-ichi Accident," ADAMS Accession No. 

ML 111861807, 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 Nc. ML 102950263, October 29,2010. 
-6­

"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 Level1/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. 
-7­

NUREG-2117, "Practicallrnplementation 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" 
RECOMMENDATION 2.1: FLOODING 


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 amended 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 and flooding hazards at operating 

reactor sites 

• 	 To collect information to facilitate NRC's 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 to address Generic Issue (GI) 204 regarding flooding of nuclear 

power plant sites following upstream dam failures 

Pursuant to Sections 161.c, 103.b, and 182.a of the Atomic Energy Act of 1954, as amended, 

and Title 10 of the Code of Federal Regulations (10 CFR), Section 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 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 the 

NTTF to conduct a systematic review of NRC processes and regulations, and to make 

recommendations to the Commission for its policy direction. The NTTF developed a set of 

recommendations that are intended to clarify and strengthen the regulatory framework for 

protection against natural phenomena. The purpose of this letter is to gather information related 

to NTTF Recommendation 2.1 for flooding 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 letter requests licensees and 

holders of construction permits under 10 CFR Part 50 to reevaluate the flooding hazards at their 

sites against present-day regulatory guidance and methodologies being used for early site 

permits and combined license reviews (SECY-11-0124, Staff Recommendations 2 and 4 for 

Enclosure 2 
-2­

NTTF Recommendation 2.1). This request is consistent with and required by the Consolidated 

Appropriations Act for 2012 (Pub Law 112-74), Section 402. 

In developing Recommendation 2.1, the NTTF recognized that, "since the establishment of 

GDC 2, the NRC's requirements and guidance for protection from seismic events, floods, and 

other natural phenomena has continued to evolve," and that "as a result, significant differences 

may exist between plants in the way they protect against design-basis natural phenomena and 

the safety margin provided." 

Since the issuance of GDC 2 in 1971, the NRC has developed new regulations, regulatory 

guidance, and several regulatory programs aimed at enhancements for previously licensed 

reactors. A summary of these regulatory programs for enhancements are described in 

Section 4.1.1 of the NTTF report. From this summary, items of note with regard to flooding 

include the individual plant examination of external events (IPEEE) program, the new 

requirement in 10 CFR 100.20 for applications after January 10,1997, and efforts underway to 

update Regulatory Guide (RG) 1.59, "Design Basis Floods for Nuclear Power Plants." 

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," (Agencywide 

Documents Access and Management System (ADAMS) Accession No. ML031150485) to 

request 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 licensee's 

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. 

In most cases, licensees used a qualitative progressive-screening approach in lieu of a more 

quantitative approach to assess the flooding hazard. NUREG-1742, "Perspectives Gained from 

the Individual Plant Examination of External Events (IPEEE) Program," volumes 1 and 2 issued 

April, 2002 (ADAMS Accession Nos. ML021270070 and ML021270674) states that "given the 

SUbstantial uncertainties involved in developing site-specific flood hazard curves, a 

consideration of possible combinations of multiple effects causing a range of flood levels would 

have enhanced the robustness of some of the lice,nsee's analyses and lent greater confidence 

Enclosure 2 
-3­

to their findings." It should be noted that the term "vulnerability" was not defined in GL 88-20. 

Instead, GL 88-20 states that licensees should provide a discussion on how vulnerability is 

defined for each external event evaluated. NUREG-1742 notes that "as a result, the use of the 

term vulnerability varied widely among the IPEEE submittals ... Some licensees avoided the term 

altogether, other stated that no vulnerabilities existed at their plant without defining the word, 

and still others provided a definition of vulnerability along with a discussion of their findings." 

New Requirements for Evaluation of Dam Hazards in 10 CFR 100.20: 

The staff established a new requirement in 10 CFR 100.20, "Factors to be Considered when 

Evaluating Sites," in 1996. The requirement in 10 CFR 1 00.20(b) states that for applications 

submitted on or after January 10, 1997, the nature and proximity of man-related hazards must 

be evaluated to establish site parameters for use in determining whether a plant design can 

accommodate commonly occurring hazards, and whether the risk of other hazards is very low. 

A parenthetical statement in the new regulation specifically identifies dams as hazards to be 

evaluated at a plant site. 

Tsunami and Regulatory Guide 1.59 Updates: 

Following the Sumatra earthquake and its accompanying tsunami in December 2004, the NRC 

staff initiated a study to examine tsunami hazards at power plant sites. Study results are 

documented in NUREG/CR-6966, "Tsunami Hazard Assessment at Nuclear Power Plant Sites 

in the United States of America," which was published in March 2009. As the NTTF report 

notes, "while tsunami hazards are not expected to be the limiting flood hazard for operating 

plants sited on the Atlantic Ocean and the Gulf of Mexico, plants in these coastal regions do not 

currently include an analysis of tsunami hazards in their licensing basis." 

Regulatory Guide 1.59, "Design Basis Floods for Nuclear Power Plants," was originally issued in 

1973. The most recent version is Revision 2, published in 1977, including an errata dated 

July 1980, and a substitution of methods presented in Appendix A (ADAMS Accession 

No. ML003740388). NRC staff is in the process of updating RG 1.59 to address advances in 

flooding analysis in the 35 years since Revision 2 was published. Although the update to 

RG 1.59 update is not complete, NUREG/CR7046, "Design Basis Flood Estimation for Site 

Characterization at Nuclear Power Plants in the United States of America," was published in 

November 2011. This report documents present-day methodologies used by the NRC to review 

early site permits (ESPs) and combined license (COL) applications. 

GI-204: Flooding of Nuclear Power Plant Sites following Upstream Dam Failures: 

Page 28 of the NTrF report states that, "In August 2010, the NRC initiated a proposed GI 

regarding flooding of nuclear power plant sites following upstream dam failures." The NRC staff 

approved this generic issue as GI-204 on February 29, 2012. The staff notes that the flood 

hazard information gathered by this 10 CFR 50.54(f) request will be applicable to the resolution 

of GI-204. 
-4­

APPLICABLE REGULATORY REQUIREMENTS 


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

• 	 10 CFR 50.54, "Conditions of Licenses" 

• 	 Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, 

GDC 2, "Design Bases for Protection against Natural Phenomena" 

• 	 Appendix A, "Seismic and Geologic Siting Criteria for Nuclear Power Plants,~ to 10 CFR 

Part 100 

• 	 Subpart B, "Evaluation Factors for Stationary Power Reactors Site Applications On or 

After January 10, 1997," to 1 0 CFR Part 100 

In GDC 2 it 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 are to reflect appropriate consideration of the most severe of the 

natural phenomena that have been historically reported for the site and surrounding area. The 

design bases are also to reflect sufficient margin to account for the limited accuracy, quantity, 

and period of time in which the historical data have been accumulated. 

Present-day regulations for reactor site criteria (Subpart B to 10 CFR Part 100 for applications 

on or after January 10, 1997) states, in part, that the physical characteristics of the site, 

including hydrology, must be evaluated and site parameters established such that potential 

threats from such physical characteristics will pose no undue risk to the type of facility proposed 

to be located at the site (10 CFR 100.21(d». Factors to be considered when evaluating sites 

includes the nature and proximity of dams and other man-related hazards (10 CFR 100.20(b» 

and the physical characteristics ofthe site, including the hydrology (10 CFR 100.20(c». 

DISCUSSION 

The NTTF recommended that the Commission direct several actions to ensure adequate 

protection from natural phenomena, consistent with the current state of knowledge and 

analytical methods. These actions should be undertaken to prevent fuel damage and to ensure 

containment and spent fuel pool integrity. In particular, Recommendation 2.1 states, "Order 

licensees to reevaluate the seismic and flooding hazards at their sites against current NRC 

requirements and guidance, and ·if necessary, update the design basis and SSCs important to 

safety to protect against the updated hazards." 

Staff assessment of Recommendation 2.1 is discussed in SECY -11-0124. Staff noted that the 

assumptions and factors that were considered in flood protection at operating plants vary. In 

some cases, the design bases did not consider the effects from local-intense precipitation and 

related site drainage. In other cases, the probable maximum flood is calculated differently at 

units co-located at the same site, depending on the time of licensing, resulting in different 
-5­

design-basis flood protection. The NTTF and the staff noted that some plants rely on operator 

actions and temporary flood mitigation measures such as sandbagging, temporary flood walls 

and barriers, and portable equipment to perform safety functions. For several sites, the staff 

noted that not all appropriate flooding hazards are documented in the updated final safety 

analysis report. The NTTF and the staff also noted that flooding risks are of concern because of 

a "cliff-edge" effect, in that the safety consequences of a flooding event may increase sharply 

with a small increase in the flooding level. Therefore, the staff concluded that all licensees 

should confirm that SSCs important to safety are adequately protected from flooding hazards. 

!n the SRM to SECY-11-0124 the Commission approved the staff's proposed actions, which 

were to implement the NTTF recommendations as described in the SECY without delay. With 

regard to reevaluating flooding hazards, staff's approved actions are to: 

1. 	 Initiate stakeholder interactions to discuss application of present-day regulatory 

guidance and methodologies being used for ESP and COL reviews to the reevaluation of 

flooding hazards at operating reactors. 

2. 	 Develop and issue a request for information to licensees pursuant to 10 CFR 50.54(f) to: 

a. 	 reevaluate site-specific flooding hazards using the methodology discussed in Item 1 

above, and 

b. 	 identify actions that have been taken or are planned to address plant-specific 

vulnerabilities associated with the updated flooding hazards. 

The SRM to SECY-11-0124 also directed the NRC staff to do the following: 

• 	 For Recommendation 2.1, when the staff issues the requests for information to licensees 

pursuant to 10 CFR 50.54(f) to identify actions that have been taken or are planned to 

address plant-specific vulnerabilities associated with the reevaluation of seismic and 

flooding hazards, the staff should explain the meaning of "vulnerability." 

• 	 The staff should inform the Commission, either through an Information Paper or briefing 

of the Commissioners' Assistants, when it has developed the technical bases and 

acceptance criteria for implementing Recommendation 2.1, 2.3, and 9.3. 

Additionally, the Consolidated Appropriations Act, for 2012 (Pub Law 112-74), Section 402, 

directs the NRC to "require reactor licensees to reevaluate the seismic, tsunami, flooding, and 

other external hazards at their sites against current applicable Commission requirements and 

guidance for such licensees as expeditiously as possible, and thereafter, when appropriate, as 

determined by the Commission, and require each licensee to respond to the Commission that 

the design basis for each reactor meets the requirements of its license, current applicable 

Commission requirements and guidance for such license." These other external hazards can 

include meteorological and other natural phenomena that could reduce or limit the capacity of 

safety-related cooling water supplies. These other external hazards will be addressed 

separately from this information request. 
-6­

Following the Commission's direction to implement the staff's proposed actions without delay, 

the NRC staff will implement Recommendation 2.1 in two phases, as follows: 

• 	 Phase 1: Issue 10 CFR 50.54(f) letters to all licensees to reevaluate the seismic and 

flooding hazards at their sites against present-day regulatory guidance and 

methodologies used for ESP and Cal reviews. 

• 	 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 

This information request addresses only Phase 1; Phase 2 will be conducted after receiving 

responses to this request. 

The NRC staff will interact with industry and stakeholders to develop approaches that can be 

applied in a uniform and consistent manner across the different sites and plant conditions. This 

type of an integrated approach will allow the NRC and industry time to assess the significance 

of any new information related to the hazard evaluation in a systematic manner. This approach 

is also consistent with Commission direction to initiate stakeholder interactions. As such, 

responses to this request for information are expected in stages, as outlined in the Required 

Response section. 

Because of the experience gained by both the NRC and the industry in preparing and reviewing 

numerous ESPs and COls, present-day methodologies associated with evaluating flooding 

hazards at plant sites are well documented. It is anticipated that some interactions will be 

required with the industry and other stakeholders on particulars associated with implementing 

these methodologies for the existing plants (e.g., certain data collection activities are likely to be 

needed). However, the timeframe outlined in the requested response section takes this into 

account. General steps to develop the flooding hazard evaluation are discussed under the 

requested actions section below, and detailed steps are provided in Attachment 1. 

Information related to the identification of actions that will be taken or planned to be taken to 

address plant-specific vulnerabilities will inform staff's development of "acceptance criteria" 

necessary to conduct Phase 2, or to address other regulatory actions as necessary. The 

approaches and methodology used to develop this information requires multiple interactions 

between the NRC staff, industry, and other stakeholders. The timeframe discussed in the 

requested response section explicitly recognizes this aspect. 

REQUESTED ACTIONS 

Addressees are requested to perform a reevaluation of all appropriate external flooding sources, 

including the effects from local intense precipitation on the site, probable maximum flood (PMF) 

on stream and rivers, storm surges, seiches, tsunami, and dam failures. It is requested that the 

reevaluation apply present-day regulatory guidance and methodologies being used for ESP and 

Cal reviews including current techniques, software, and methods used in present-day standard 

engineering practice to develop the flood hazard. The requested information will be gathered in 
-7­

Phase 1 of the NRC staffs two phase process to implement Recommendation 2.1, and will be 

used to identify potential vulnerabilities 1 

• 

For the sites where the reevaluated flood exceeds the design basis, addressees are requested 

to submit an interim action plan that documents actions planned or taken to address the 

reevaluated hazard with the hazard evaluation. 

Subsequently, addressees should perform an integrated assessment of the plant to identify 

vulnerabilities and actions to address.them. The scope of the integrated assessment report will 

include full power operations and other plant configurations that could be susceptible due to the 

status of the flood protection features. The scope also includes those features of the ultimate 

heat sinks (UHS) that could be adversely affected by the flood conditions and lead to 

degradation of the flood protection (the loss of UHS from non-flood associated causes are not 

included). It is also requested that the integrated assessment address the entire duration of the 

flood conditions. 

REQUESTED INFORMATION 

The NRC staff requests that each addressee provide the following information. Attachment 1 

provides additional information regarding present-day methodologies and guidance used by the 

NRC staff performing ESP and COL reviews. The attachment also provides a stepwise 

approach for assessing the flood hazard that should be applied to evaluate the potential hazard 

from flood causing mechanisms at each licensed reactor site. . 

1. 	 Hazard Reevaluation Report 

Perform a flood hazard reevaluation. Provide a final report documenting results, as well 

as pertinent site information and detailed analysis. The final report should contain the 

following: 

a. 	 Site information related to the flood hazard. Relevant SSCs important to safety 

and the UHS are included in the scope of this reevaluation, and pertinent data 

concerning these SSCs should be included. Other relevant site data includes the 

following: 

i. 	 detailed site information (both designed and as-built), including 

present-day site layout, elevation of pertinent SSCs important to safety, 

site topography, as well as pertinent spatial and temporal data sets 

ii. 	 current design basis flood elevations for all flood causing mechanisms 

iii. 	 flood-related changes to the licensing basis and any flood protection 

changes (including mitigation) since license issuance 

iv. 	 changes to the watershed and local area since license issuance 

1 A definition of vulnerability in the context of this enclosure is as follows: Plant-specific vulnerabWties 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 functions. 
-8­

v. current licensing basis flood protection and pertinent flood mitigation 

features at the site 

vi. additional site details, as necessary, to assess the flood hazard (i.e., 

bathymetry, walkdown results, etc.) 

b. 	 Evaluation of the flood hazard for each flood causing mechanism, based on 

present-day methodologies and regulatory guidance. Provide an analysis of 

each flood causing mechanism that may impact the site including local intense 

precipitation and site drainage, flooding in streams and rivers, dam breaches and 

failures, storm surge and seiche, tsunami, channel migration or diversion, and 

combined effects. Mechanisms that are not applicable at the site may be 

screened-out; however, a justification should be provided. Provide a basis for 

inputs and assumptions, methodologies and models used including input and 

output files, and other pertinent data. 

c. 	 Comparison of current and reevaluated flood causing mechanisms at the site. 

Provide an assessment of the current design basis flood elevation to the 

reevaluated flood elevation for each flood causing mechanism. Include how the 

findings from Enclosure 4 of this letter (Le., Recommendation 2.3 flooding 

walkdowns) support this determination. If the current design basis flood bounds 

the reevaluated hazard for all flood causing mechanisms, include how this finding 

was determined. 

d. 	 Interim evaluation and actions taken or planned to address any higher flooding 

hazards relative to the design basis, prior to completion of the integrated 

assessment described below, if necessary. 

e. 	 Additional actions beyond Requested Information item 1.d taken or planned to 

address flooding hazards, if any. 

2. 	 Integrated Assessment Report 

For the plants where the current design basis floods do not bound the reevaluated 

hazard for all flood causing mechanisms, provide the following: 

a. 	 Description of the integrated procedure used to evaluate integrity of the piant for 

the entire duration of flood conditions at the site. 

b. 	 Results of the plant evaluations describing the controlling flood mechanisms and 

its effects, and how the available or planned measures will provide effective 

protection and mitigation. Discuss whether there is margin beyond the 

postulated scenarios. 

c. 	 Description of any additional protection and/or mitigation features that were 

installed or are planned, including those installed during course of reevaluating 
-9­

the hazard. The description should include the specific features and their 

functions. 

d. 	 identify other actions that have been taken or are planned to address 

plant-specific vulnerabilities. 

REQUIRED RESPONSE 

Within approximately 60 days of the date of this information request, NRC staff will determine 

the priority for each reactor site to complete the hazard reevaluation report. The site priority will 

determine the submittal date for addressees to provide written responses to Requested 

Information item 1 (Hazard Reevaluation Report). 

In accordance with Sections 161.c, 103. b, and 182.a of the Atomic Energy Act of 1954, as 

amended, and 10 CFR 50.54(f), an addressee must respond as described below. 

1. 	 Within 60 days of the date of the NRC's issuance of guidance on implementation details 

of the Integrated Assessment Report, including criteria for identifying vulnerabilities, 

submit an approach for developing an Integrated Assessment Report including criteria 

for identifying vulnerabilities2

. 

2. 	 In accordance with the NRC's prioritization plan, within 1- to 3-years from the date of this 

information request, submit the Hazard Reevaluation Report. Include the interim action 

plan requested in item 1.d, if appropriate. 

3. 	 Within 2 years following submittal of the Hazard Reevaluation Report to the NRC, any 

addressee who is requested to complete an Integrated Assessment should submit 

written responses to Requested Information item 2. 

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 prioritization described above will be based on information from COL and ESP applications, 

updated hazard levels if new information exists, and site-specific circumstances. This 

prioritization scheme is intended to use both the NRC and industry resources most effectively. 

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 

2 The NRC staff will develop the implementation details ot the Integrated Assessment Report. including criteria for identifying 

vulnerabilities This information is scheduled to be developed by November 30.2012 and tile NRC staff will interact with 

stakeholders. as appropriate during this process. 
-10­

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. 
Attachment 1 to Recommendation 2.1: Flooding Enclosure 2 

PROCEDURE 


The steps shown in Figure 1 of this attachment represent an acceptable approach to perform 


the reevaluation of the flood hazard and integrated assessment. The flood hazard reevaluation 


should address all flood causing mechanisms that are pertinent to the site based on the 


geographic location and interface of the plant with the hydrosphere. The reason for omitting any 


of these flood causing mechanisms should be clearly discussed in the final report. A discussion 


of typical flood causing mechanisms is included below. Many types of flood causing 


mechanisms are included in that discussion, but it is important to note that each site should 


address unique characteristics and any additional flood causing mechanisms identified. 


Step 1: 


All licensees should review information concerning the current flooding hazard against that for 


which the plant is designed. This information will be used in the following steps for reevaluation 


of the flood hazard. Pertinent information includes, but is not limited to, the following: 


• 	 Current design basis flood hazard 

• 	 Fiood elevations and other effects considered in the flood protection3 for all flood causing 

mechanisms. 

• 	 Changes in licensing basis since initial licensing including site drainage characteristic 

and modification, watershed changes, new dam construction, revision of dam operations 

• 	 New information pertinent to the hydrologic characteristics including changes to dam 

operation, new flood studies and changes to meteorological basis (e.g., maximum 

precipitation studies) 

• 	 Pertinent information from site-related or watershed-related studies 

• 	 Site changes since issuance of the operating license (new barriers, openings, revised 

drainage systems, new structures, etc.) 

• 	 Flood protection mechanisms and identifying characteristics (e.g., structures and 


procedures) 


• 	 Pertinent features identified in site walkdowns 

Step 2: 

Reevaluate the flood hazard based on present day regulatory guidance and methodologies for 

each flood causing mechanism. Using any new site-related information and site details 

identified in Step 1, evaluate all possible flood causing mechanisms. Documentation of all 

methodologies should be discussed. This step of the process reiterates the current hierarchical 

hazard assessment (HHA) used by U.S. Nuclear Regulatory Commission (NRC) staff. The 

HHA is described as a progressively refined, stepwise estimation of the site-specific hazards 

that evaluates the safety of the site with the most conservative plausible assumptions consistent 

with available data. 

3 Examp'es of other effects include dynamic wave effects. scouring, and debris transportation 
-2­

(a) 	 Select one flood causing mechanism to be reanalyzed 

(b) 	 Develop a conservative estimate of the site related parameters using simplifying 

assumptions for a flood causing mechanism and perform the reevaluation. 

(c) 	 Determine if the reevaluated flood hazard elevation (from Step 2b) is higher than 

the original design flood elevation for the selected flood causing mechanism. If 

not, use this flood elevation for this causal mechanism in Step 3. 

(d) 	 Determine if the site-related parameters can be further refined. If yes, perform 

reevaluation (repeat step 2c). If no, use this flood elevation for this causal 

mechanism in Step 3. 

(e) 	 Determine if all flood causing mechanisms have been addressed. If yes, 

continue to Step 3. If no, select another flood causing mechanism (Step 2a). 

Step 3: 

For each flood causing mechanism, compare the final flood elevations from the hazard 


reevaluation against the current design basis flood elevations. Using this comparison, 


determine whether the design basis flood bounds each reevaluated hazard from Step 2. If it is 


determined that the current design basis flood bounds all of the reevaluated hazards, proceed to 


Step 4. If not all of the reevaluated hazards are bounded by the current design basis flood, 


proceed to Step 6 for additional analysis. 


Step 4: 


Submit a report in accordance with Requested Information item 1, Hazard Reevaluation Report. 


It is anticipated that activities associated with the NTIF Recommendation 2.3 are completed 


and form a partial basis for the information requested. 


Step 5: 

No further action is required. This step demonstrates termination of the process for resolution of 

NTIF Recommendation 2.1. 

Step 6: Submit a report in accordance with the Requested Information item 1, Hazard 

Reevaluation Report, including any relevant information from the results of plant walkdown 

activities related to NTIF Recommendation 2.3. Also, provide plans for conducting further 

analysis (steps 7 through 9) and submitting the final report identified in Requested Information 

item 2. 

Step 7: 

For the flood causing mechanisms that were not bounded, or for a controlling flood causing 

mechanism, perform an integrated assessment using the procedures developed in interactions 

with the NRC staff. The purpose of the integrated assessment is to determine the effectiveness 

of the existing design basis and any other planned or installed features for the protection and 

mitigation of flood conditions for the entire duration of the flood. 

Step 8: 

Identify vulnerabilities, if any, as a result of the assessment conducted in Step 7. Also, identify 

any planned actions or actions that were already taken to address these vulnerabilities. 
-3­

Step 9: 

Submit a report in accordance with the Requested Information item 2. Include a brief summary 

of the flood causing mechanisms and the associated parameters that were used in the 

assessment. 

Step 10: 

The information provided in Step 9 will be evaluated by the NRC in Phase 2 to consider any 

additional regulatory actions. 

FLOOD CAUSING MECHANISMS 

The NRC regulations require that structure, systems and components (SSCs) important to 

safety of a nuclear power plant are adequately protected from the adverse effects of flooding. 

The NRC staff discusses the approach for determining the flood hazard for new reactors in its 

current guidance documents, NUREG-0800 and NUREG/CR-7046. 

As part of analyzing the flood hazard, it is important to list all plausible flood causing 

mechanisms that are capable of generating a severe flood at the site and to recognize that 

several scenarios of a particular flood causing mechanism can affect the site. For example, 

extreme precipitation can cause flooding in adjacent rivers, near-by tributaries, and on-site 

drainage facilities. Similarly, flood causing mechanisms that are not plausible at a particular site 

may also be ruled out. Present day NRC staff guidance applies the HHA (see 

NUREG/CR-7046) to each pertinent flood causing mechanism at a site. 

The following is a list of flood causing mechanisms that should be addressed in a flood hazard 

analysis. Site specific characteristics may warrant review of other mechanisms in addition to 

those listed here. 

1. Local Intense Precipitation 

Local intense precipitation is a measure of the extreme precipitation at a given location. 

Generally, local intense precipitation values are developed using methods called Probable 

Maximum Precipitation (PMP) based on the methods developed by the federal government and 

published in hydrometeorological reports (HMR) by the National Weather Service. For extreme 

precipitation, localized precipitation values are developed using methods in HMR 52 (eastern 

areas ofthe United States (U.S.)) as well as regionalized reports within the HMR publication 

series. 

The elevation of the site is not relevant for mitigation of flooding from local intense precipitation. 

The runoff carrying capacity of the site grading design and the performance of any active or 

passive drainage systems would determine the depth and velocity of surface runoff at the site. 

Typically, any active drainage system should be considered non-functional at the time of local 

intense precipitation event. Generally, runoff losses should be ignored during the local intense 

precipitation event to maximize the runoff. Hydraulic parameters that affect the depth and 

velocity of flow should be chosen carefully and should be consistent with values used in 

standard engineering practice. 
-4­

2. Flooding in Streams and Rivers 

The probable maximum flood (PMF) in rivers and streams adjoining the site should be 

determined by applying the PMP to the drainage basin in which the site is located. The PMF is 

based on a translation of PMP rainfall on a watershed to flood flow. The estimation of PMP for 

regional areas within the U.S. is based on HMRs and the appropriate regional report should be 

used to develop the PMP for a given site and watershed. The PMP is a deterministic estimate 

of the theoretical maximum depth of precipitation that can occur at a time of year of a specified 

area. A rainfall-to-runoff transformation function, as well as runoff characteristics, based on the 

topographic and drainage system network characteristics and watershed properties are needed 

to appropriately develop the PMF hydrograph. The PMF hydrograph is a time history of the 

discharge and serves as the input parameter for other hydraulic models which develop the flow 

characteristics including flood flow and elevation. The U.S. Army Corps of Engineers hydrologic 

and hydraulic methods are widely accepted in engineering practice. However, unique 

characteristics or preference of the analysis may dictate use of other models. Appropriate 

justification for selection of methods, data and models would depend on site-specific 

circumstances. 

3. Dam Breaches and Failures 

Flood waves resulting from the breach of upstream dams, including domino-type or cascading 

dam failures should be evaluated for the site. Water storage and water control structures (such 

as onsite cooling or auxiliary water reservoirs and onsite levees) that may be located at or 

above SSCs important to safety should also be evaluated. Additional effects for earthen 

embankments, such as sediment, should also be considered. Models and methods used to 

evaluate the dam failure and the resulting effects should be applicable to the type of failure 

mechanism and should be appropriately justified. Recent analyses completed by State and 

Federal agencies with appropriate jurisdiction for dams within the watershed may be used. 

4. Storm Surge 

Storm surge is the rise of offshore water elevation caused principally by the shear force of the 

hurricane or tropical depression winds acting on the water surface. Technical reports, from the 

National Oceanic and Atmospheric Administration (NOAA), provide guidance on developing 

wind fields for a probable maximum hurricane. The wind field parameter is input to coastal 

hydrodynamics simulation model that predict water surface rise based on the shear forces 

imparted by the wind. However, appropriate justification for selection of methods, data, and 

models depends on site-specific circumstances. 

5. Seiche 

A seiche is an oscillation of the water surface in an enclosed or semi-enclosed water body 

initiated by an external cause. If a seiche is determined to be possible at the site, then 

appropriate numerical modeling may be needed. For bays and lakes with irregular geometries 

and variable bathymetries, numerical longwave hydrodynamics modeling may be the only viable 

technique to determine hazard. 
-5­

6. Tsunami 

A tsunami is a series of water waves generated by a rapid, large scale disturbance of a water 

body due to seismic, landslide or volcanic tsunamingenic sources. An assessment with respect 

to tsunami can include a stepwise approach addressing: the susceptibility of the site's region 

subject to tsunami, the susceptibility of the plant site affected by tsunami, and specific hazards 

of the site posed to safety of the plant by tsunami. 

7. Ice Induced Flooding 

Ice jams and ice dams can cause flooding by impounding water upstream of a site and 

subsequently collapsing or downstream of a site impounding and backing up water. There is no 

method to assess a probable maximum ice jam or ice dam, therefore, historical records are 

generally accessed to determine the most severe historical event in the vicinity of the site. This 

method is based on an observed historical observation and reasonable margin should be 

considered. 

8. Channel Migration or Diversion 

Flood hazard associated with channel diversion is due to the possible migration either toward 

the site or away from it. For natural channels adjacent to the site, historical and geomorphic 

processes should be reviewed for possible tendency to meander. For man-made channels, 

canals or diversions used for the conveyance of water located at a site, possible failure of these 

structures should be considered. 

9. Combined Effect Flood 

For flood hazard associated with combined events, American Nuclear Society (ANS) 2.8-1992 

provides guidance for combination of flood causing mechanisms for flood hazard at nuclear 

power reactor sites. In addition to those listed in the ANS guidance, additional plausible 

combined events should be considered on a site specific basis and should be based on the 

impacts of other flood causing mechanisms and the location of the site. 
-6­

Use site-specific data to 

refine analysis 

No 

To Page 2 .. 1111 R. 

Yes 

Figure 1. Development of Requested Information and Its Use in Regulatory Analysis. Page 1 of2 
From Page 1 

Yes 

No 

Figure 1. Development of Requested Information and Its Use in Regulatory Analysis. Page 2 of 2 

-8­

Enclosure 2 Reference List 

Sections 161.c, 103.b, and 182.a of the Atomic Energy Act of 1954, as amended 

SECY 11-0124, "Recommended Actions To Be Taken Without Delay from the Near-Term Task 

Force Report," Agencywide Documents and Management System (ADAMS) Accession No. 


ML 11245A158, September 9,2011. 


SECY 11-0137, "Prioritization of Recommended Actions to Be Taken in Response to 


Fukushima Lessons Learned," ADAMS Accession No. ML 11272A 111, October 3, 2011. 


"Recommendations for Enhancing Reactor Safety in the 21 st Century: The Near-term Task 


Force Review of Insights from the Fukushima Dai-ichi Accident" ADAMS Accession No. 


ML 111861807, July 12, 2011. 


10 CFR 50.54(f) - "Conditions of Licenses" 


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


Appendix A to 10 CFR Part 50, General Design Criteria 2 


"Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities" (Volume 60, 


page 42622, of the Federal Register (60 FR 42622)) 

Supplement 4 to GL 88-20, "Individual Plant Examination of External Events (IPEEE) for Severe 


Accident Vulnerabilities," ADAMS Accession No. ML031150485, June 28, 1991. 


10 CFR 100.20, "Factors to be Considered when Evaluating Sites," 


NUREG/BR-0058, Revision 4, "Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory 


Commission," ADAMS Accession No. ML042820192, September 30,2004. 


SRM SECY 11-0124, "Recommended Actions To Be Taken Without Delay from the Near-Term 


Task Force Report," ADAMS Accession No. ML 112911571, October 18, 2011. 


SRM SECY 11-0137, "Prioritization of Recommended Actions to Be Taken in Response to 


Fukushima Lessons Learned," ADAMS Accession No. ML 113490055, dated December 15, 


2011. 


10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities" 


NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear 


Power Plants: LWR Edition - Site Characteristics and Site Parameters (Chapter 2)," ADAMS 


Accession No. ML070400364, March 2007. 

-9­

NUREG/CR-7046, PNNL-20091, "Design-Basis Flood Estimation for Site Characterization at 


Nuclear Power Plants in the United States of America." ADAMS Accession No. ML 11321A 195, 


November 2011. 


RG 1.29, "Seismic Design Classification," Revision 4, ADAMS Accession No. ML070310052, 


March 2007. 


RG 1.59, "Design Basis Floods for Nuclear Power Plants," Revision 2, ADAMS Accession No. 


ML003740388, August 1977. 


RG 1.102, "Flood Protection for Nuclear Power Plants," Revision 1, ADAMS Accession No. 


ML003740308, September 1976. 


NOAA Hydrometeorological Report No. 52, "Application of Probable Maximum Precipitation 


Estimates - United States East of the 105th Meridian," U.S. Department of Commerce, National 


Oceanic and Atmospheric Administration, U.S. Department of the Army, Corps of Engineers, 


Washington, DC, August 1982. 


NOAA Hydrometeorological Report No. 51, "Probable Maximum Precipitation Estimates - United· 


States East of the 105th Meridian," U.S. Department of Commerce, National Oceanic and 


Atmospheric Administration, U.S. Department of the Army, Corps of Engineers, Washington, 


DC, 1978. 


NOAA Hydrometeorological Report No. 53, "Seasonal Variation of 10-square mile Probable 


Maximum Precipitation Estimates - United States East of the 105th Meridian," U.S. Department 


of Commerce, National Oceanic and Atmospheric Administration, U.S. Department of the Army, 


Corps of Engineers, Washington, DC, 1980. 


ANS 2.8-1992, "Determining Design Basis Flooding at Power Reactor Sites," 1992. 

RECOMMENDATION 2.3: 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 

2.3, as amended by staff requirements memorandum (SRM) associated with 

SECY-11-0124 and SECY-11-0137, 

• 	 To request licensees to develop a methodology and acceptance criteria for seismic 

walkdowns to be endorsed by the NRC staff, 

• 	 To request licensees to perform seismic walkdowns using the NRC-endorsed walkdown 

methodology, as defined herein, 

• 	 To identify and address degraded, nonconforming, or unanalyzed conditions through the 

corrective action program, and 

• 	 To verify the adequacy of licensee monitoring and maintenance procedures. 

Pursuant to Title 10 of the Code ofFederal Regulations (10 CFR), Section SO.S4(f}, addressees 

are required to submit a written response to this information request. 

BACKGROUND 

Structures, systems, and components (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 SO, General Design Criteria (GOG) 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 are to reflect appropriate consideration of the most severe of the natural 

phenomena that have been historically reported for the site and surrounding area. The design 

bases are also to reflect sufficient 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 the 

NTTF to conduct a systematic review of NRC processes and regulations and to make 

recommendations to the Commission for its policy direction. The NTTF developed a set of 

recommendations that are 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 2.3 for seismic hazards. Recommendation 2.3, and the 

SRMs associated with SECY-11-0124 and SECY-11-0137, instructs the NRC staff to issue 

requests for information to licensees pursuant to 10 CFR SO.S4(f}. This information request is 

for licensees to develop a methodology and acceptance criteria for seismic walkdowns to be 

endorsed by the staff following interaction with external stakeholders. It is requested that 

licensees perform the seismic walkdowns to identify and address plant-specific vulnerabilities 

(through its corrective action program) and verify the adequacies of monitoring and 

maintenance procedures. 

Enclosure 3 
-2­

In developing Recommendation 2.3, the NTTF recognized the need to verify the adequacy of 

features that play an integral role in the defense-in-depth approach for protection from natural 

phenomena. NTTF Recommendation 2.3 and SECY-11-0124 and SECY-11-0137 states that 

recent plant inspections have been conducted by NRC staff and industry in response to the 

Fukushima Dai-ichi accident and that these activities should be used to inform the 

implementation of this recommendation. Ongoing inspections of the Fukushima Dai-ichi and 

Dai-ni nuclear power stations may also provide inSights useful for this recommendation. 

Furthermore, recent lessons learned from the earthquake, near the North Anna Power Station 

should also be used to inform the development of the walkdown procedure(s). 

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" 

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. Appendix A 

requires that safety-related SSCs remain functional if the safe shutdown earthquake (SSE) 

occurs. 

DISCUSSION 

The NTTF recommended that the Commission direct several actions to ensure adequate 

protection from natural phenomena. The actions should be taken to prevent fuel damage, 

ensure containment integrity and the functionality of SSCs that support the spent fuel pool 

(SFP). In particular, NTTF Recommendation 2.3 states that the Commission should "Order 

licensees to perform seismic and flood protection walkdowns to identify and address plantspecific vulnerabilities and verify the adequacy of monitoring and maintenance for protection 

features such as water tight barriers and seals in the interim period until longer term actions are 

completed to update the design basis for external events." However, in the context of this letter, 

the NRC staff is focusing on degraded, nonconforming, or unanalyzed conditions. 

The NRC staff's assessment of NTTF Recommendation 2.3 is discussed in SECY-11-0124. 

The NRC staff agreed with the NTrF Recommendation 2.3 findings and noted that various 

walkdown guidance exists and that recent plant inspections by staff in accordance with 

Temporary Instruction (TI) 2515/183, "Follow-up to the Fukushima Dai-ichi Nuclear Station Fuel 

Damage Event," and licensees' plant inspections in response to the Fukushima Dai-ichi 

accidents should help inform the implementation of this recommendation. Results of the NRC 

staff's evaluation of the recent earthquake near North Anna Power Station may also provide 

insights. 
-3­

In its SRM to SECY-0124, the Commission approved the staff's proposed actions to implement 

without delay the NTTF recommendations as described in the SECY paper. With regard to 

Recommendation 2.3, the NRC staff-approved actions are to develop and issue a request for 

information to licensees pursuant to 10 CFR 50.54(f) to develop a methodology and acceptance 

criteria for seismic walkdowns to be endorsed by the NRC staff following interactions with 

external stakeholders, perform seismic walkdowns to identify and address plant-specific 

degraded, nonconforming, or unanalyzed conditions (through the corrective action program) and 

verify the adequacy of monitoring and maintenance for protective features, and inform the NRC 

staff of the results of the walkdowns and corrective actions taken or planned. 

The TI2515/183 was issued by the NRC on March 23, 2011. Inspection activities were 

completed by April 29, 2011, and NRC inspection reports were issued by May 13, 2011. The 

NRC developed a Summary of Observations report to encapsulate the performance of TI 

2515/183 (see http://www.nrc.gov/NRRlOVERSIGHT/ASSESS/foliow-up-rpts.html). The 

summary report states that while individually, none of the observations posed a significant 

safety issue, they indicate a potential industry trend of failure to maintain equipment and 

strategies required to mitigate some design basis events. Regarding the licensees' capability to 

mitigate large fires or flooding coincident with seismic activity, the report notes that some 

equipment used to mitigate fires or station blackout was stored in areas that were not 

seismically qualified or that could be flooded. 

As outlined in the SECY-11-0124, the NRC staff intends to work with the industry and other 

stakeholders to endorse a procedure(s) to develop acceptance criteria, conduct walkdowns, and 

identify degraded, nonconforming, or unanalyzed conditions. It is antiCipated that the walkdown 

procedure will be developed by modifying various existing NRC and industry processes, 

including the recent inspections described above in accordance with TI 2515/183. Other 

guidance for seismic protection walkdowns include Electric Power Research Institute (EPRI) 

report NP-6041-SL, Revision 1, "A Methodology for Assessment of Nuclear Power Plant 

Seismic Margin," Seismic Qualification Utility Group procedure, "Generic Implementation 

Procedure (GIP) for Seismic Verification of Nuclear Power Plant Equipment," and International 

Atomic Energy Agency NS-G-2.13, "Evaluation of Seismic Safety for EXisting Nuclear 

Installations." Additional details of attributes of a walkdown procedure are described in the 

Requested Action below. 

The technical approach and methods used to develop the requested information should be 

integrated such that it accounts for design, physical barriers, procedures, temporary measures, 

and planned or installed mitigation measures to deal with external hazards. This type of an 

integrated approach will allow the NRC and industry to assess the significance of any new 

information related to the hazard in a systematic manner. 

REQUESTED ACTIONS 

In response to NTTF Recommendation 2.3, the Commission requests all licensees to perform 

seismic walkdowns in order to identify and address plant specific degraded, nonconforming, or 

unanalyzed conditions and verify the adequacy of strategies, monitoring, and maintenance 

programs such that the nuclear power plant can respond to external events. The walkdown will 

verify current plant configuration with the current licenSing basis, verify the adequacy of current 

strategies, maintenance plans, and identify degraded, nonconforming, or unanalyzed conditions. 
-4­

The walkdown procedure should be developed and submitted to the NRC. The procedure may 

incorporate current plant procedures, if appropriate. Prior to the walkdown, licensees should 

develop acceptance criteria, collect appropriate data, and assemble a team with relevant 

technical skills. Improvements made as part of the licensees' response to the individual plant 

examination of external events (lPEEE) program for seismic issues should be reported. 

If any condition identified during the walkdown activities represents a degraded, nonconforming, 

or unanalyzed condition (i.e., noncompliance with the current licensing basis) for an SSC, 

describe actions that were taken or are planned to address the condition using the guidance in 

Regulatory Issues Summary 2005-20, Revision 1, Revision to NRC Inspection Manual Part 

9900 Technical Guidance, "Operability Conditions Adverse to Quality or Safety," including 

entering the condition in the corrective action program. Reporting requirements pursuant to 

10 CFR 50.72 should also be considered. Additionally, these findings should be considered in 

the Recommendation 2.1 hazard evaluations, as appropriate. 

REQUESTED INFORMATION 

1. 	 The NRC requests that each addressee confirm that they will use the industrydeveloped, NRC-endorsed, seismic walkdown procedures1 or provide a description of 

plant-specific walkdown procedures that include the following characteristics: 

a. Determination of the seismic walkdown scope and any combined effects 

b. Consideration of NUREG-1742, EPRI Report NP-6041, GIP, and common issues 

and findings discussed in the responses to TI 2515/183 

c. Pre-walkdown actions (e.g., data collection, review of drawings and procedures, 

identification of the plant licensing basis, identification of current seismic 

protection levels) 

d. Identification of SSCs requiring seismic protection and used in the protection of 

the reactor and spent fuel pool, including the ultimate heat sink (UHS) 

e. Description of the walkdown team composition and qualifications 

f. Details of the information to be collected during the walkdown including 

equipment access considerations 

g. Documentation and peer review requirements 

2. 	 Following the NRC's endorsement of the walkdown procedure, addresses are requested 

to conduct the walkdown and submit the final report which includes the following: 

a. 	 Information on the plant-specific hazard licensing bases and a description of the 

protection and mitigation features considered in the licensing basis evaluation 

b. 	 Information related to the implementation of the walkdown process 

c. 	 A list of plant-specific vulnerabilities (including any seismic anomalies, outliers, or 

other findings) identified by the IPEEE and a description of the actions taken to 

eliminate or reduce them (including their completion dates) 

d. 	 Results of the walkdown including key findings and identified degraded, 

nonconforming, or unanalyzed conditions. Include a detailed description of the 

1 NRC staff are currently engaged with industry and other external stakeholders to develop NRC-endorsed procedures. The NRC 

staff anticipates completing this activity by May, 2012. 
-5­

actions taken or planned to address these conditions using the guidance in 

Regulatory Issues Summary 2005-20, Revision, 1, Revision to NRC Inspection 

Manual Part 9900 Technical Guidance, "Operability Conditions Adverse to 

Quality or Safety," including entering the condition in the corrective action 

program 

e. 	 Any planned or newly installed protection and mitigation features 

f. 	 Results and any subsequent actions taken in response to the peer review 

REQUIRED RESPONSE 

In accordance with 10 CFR 50.54(1), an addressee must respond as described below. The 

submission of the requested information is in stages to allow adequate time for further 

interactions with the stakeholders to provide clarifications, to develop implementation 

procedures and processes, and to develop the associated guidance as needed. 

1. 	 Within 120 days of the date of this information request, the addressee will confirm that 

they intend to use the NRC-endorsed seismic walkdown procedures, or provide to the 

NRC a description of the process that will be used to conduct the walkdowns and to 

. develop the needed information. 

2. 	 Within 180 days of the NRC's endorsement of the walkdown process, each addressee 

will submit its final response. This response should include a list of any areas that are 

unable to be inspected due to inaccessibility and a schedule for when the walkdown will 

be completed. 

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(1). In addition, addressees should submit 

a copy of the response to the appropriate regional administrator. 
-6­

Enclosure 3 Reference List 

SECY 11-0124, "Recommended Actions to be taken without Delay from the Near-Term Task 

Force Report," Agencywide Documents Access and Management System (ADAMS) Accession 

No. ML 11245A 158, September 9, 2011. 

SECY 11-0137, "Prioritization of Recommended Actions to be Taken in Response to Fukushima 

Lessons Learned," ADAMS Accession No. ML 11272A 111, October 3, 2011. 

SRM SECY 11-0124, "Recommended Action to be taken without Delay from the Near-Term 

Task Force Report," ADAMS Accession No. ML112911571, dated October 18, 2011. 

SRM SECY 11-0137, "Prioritization of Recommended Actions to Be Taken in Response to 

Fukushima Lessons Learned," ADAMS Accession No. ML113490055, dated December 15, 

2011. 

10 CFR 50.54 - "Conditions of Licenses" 

Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, GOC 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) 

10 CFR 50.72, "Immediate Notification Requirements for Operating Nuclear Power Reactors" 

Appendix A, "Seismic and Geologic Siting Criteria for Nuclear Power Plants," to 

10 CFR Part 100, "Reactor Site Criteria" 

Temporary Instruction 2515/183, "Follow-up to the Fukushima Dai-ichi Nuclear Station Fuel 

Damage Event" 

Summary of Observations report to encapsulate the performance of TI 2515/183 

(http://www.nrc.gov/NRRlOVERSIGHT/ASSESS/foliow-up-rpts.html). 

Electric Power Research Institute (EPRI) report NP-6041-SL Revision 1, "A Methodology for 

Assessment of Nuclear Power Plant Seismic Margin," August 1991. 

Seismic Qualification Utility Group (SQUG) procedure: "Generic Implementation Procedure 

(GIP) for Seismic Verification of Nuclear Power Plant Equipment," 

International Atomic Energy Agency (IAEA) NS-G-2.13, "Evaluation of Seismic Safety for 

Existing Nuclear Installations." 
RECOMMENDATION 2.3: FLOODING 


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 

2.3, as amended by staff requirements memorandum (SRM) associated with 

SECY-11-0124 and SECY-11-0137, 

• 	 To request licensees to develop a methodology and acceptance criteria for flooding 

walkdowns to be endorsed by the NRC staff, 

• 	 To request licensees to perform flooding walkdowns using an NRC-endorsed walkdown 

methodology, as defined herein 

• 	 To identify and address degraded, nonconforming, or unanalyzed conditions through the 

corrective action program 

• 	 To identify and address cliff-edge effects through the corrective action program 

• 	 To verify the adequacy of licensee monitoring and maintenance procedures. 

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section SO.S4(f), addressees 

are requi"red to submit a written response to this inforrnation request. 

BACKGROUND 

Structures, systems, and components (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 

SO, 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 are to reflect 

appropriate consideration of the most severe of the natural phenomena that have been 

historically reported for the site and surrounding area. The design bases are also to reflect 

sufficient 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 DaHchi nuclear power plant caused by the 

March 11,2011, Tohoku earthquake and subsequent tsunami, the Commission established the 

NTTF to conduct a systematic review of NRC processes and regulations, and to make 

recommendations to the Commission for its policy direction. The NTTF developed a set of 

recommendations that are intended to clarify and strengthen the regulatory framework for 

protection against natural phenomena. The purpose of this letter is to gather information related 

to NTTF Recommendation 2.3 for flooding hazards. Recommendations 2.3, and the SRMs 

associated with SECY-11-0124 and SECY-11-0137, instructs the NRC staff to issue requests 

for information to licensees pursuant to 10 CFR SO.S4(f). This information request is for 

licensees to develop a methodology and acceptance criteria for flooding walkdowns to be 

endorsed by the NRC staff following interaction with external stakeholders. Licensees are 

requested to perform flood protection walkdowns to identify and address plant-specific 

Enclosure 4 
-2­

degraded, nonconforming, or unanalyzed conditions and cliff-edge effects (through the 

corrective action program) and verify the adequacy of monitoring and maintenance procedures. 

In developing Recommendation 2.3, the NTTF observed that, "some plants have an 

overreliance on operator actions and temporary flood mitigation measures such as 

sandbagging, temporary flood walls and barriers, and portable equipment to perform safety 

functions." The NTTF report also states that, "the Task Force has concluded that flooding risks 

are of concern due to a 'cliff-edge' effect, in that the safely consequences of a flooding event 

may increase sharply with a small increase in the flooding level. Therefore, it would be very 

beneficial to safety for all licensees to confirm that SSCs important to safely are adequately 

protected from floods." 

The NRC, in the past, has developed regulatory programs aimed at identifying plant-specific 

vulnerabilities to external flooding hazards. In June of 1991, the NRC issued Supplement 4 to 

Generic Letter (GL) 88-20, "Individual Plant Examination of External Events (lPEEE) for Severe 

Accident Vulnerabilities, 10 CFR SO.S4(f)." This GL requested that "each licensee perform an 

individual plant examination of external events to identify vulnerabilities, if any, to severe 

accidents and report the results together with any licensee determined improvements and 

corrective actions to the Commission." Flood.,related hazards were considered in the IPEEE 

program as one of the high winds, floods, and other (HFO) external initiating-event hazards. Of 

the 70 IPEEE submittals, most indicated some type of walkdown was performed for the HFO 

events. However, NUREG-1742 states, "the [HFO walkdown] submittals usually did not provide 

detailed descriptions of the walkdown procedures and results." NUREG-1742 also states that, 

"A few licensees proposed flood-related countermeasures that may be optimistic. For example, 

one licensee took credit for sandbagging up to a level of 9 feet. In several other submittals, 

flood barriers made of various construction materials, such as logs or concrete berms, were 

credited with being effective for preventing flooding, but the submittals did not discuss whether 

the licensees performed confirmatory testing to verify the effectiveness of certain of these 

mitigating actions." 

In late December 1999, a severe storm induced flooding at Le Blayais nuclear power plant site 

in France. Lessons learned from this flooding event are documented in World Association of 

Nuclear Operators Significant Event Report (SER) 2000-3, "Severe Storm Results in Scram of 

Three Units and Loss of Safety System Functions due to Partial Plant Flooding," and in Institute 

of Nuclear Power Operations (INPO) SER 1-01, with the same title. Both reports list significant 

aspects and important lessons learned from the flooding event. On March 11, 2010, Electricite 

de France presented lessons learned from the 1999 Blayais flood at the NRC's Regulatory 

Information Conference 

(http://www.nrc.gov/public-involve/conference-symposia/ric/pastl2010/slides/th3Sdefraguierepv. 

pdf). Lessons learned discussed in this presentation were: (1) cable openings and trenches 

were an unrecognized common-mode vulnerability requiring review of existing protective 

measures, (2) difficulty in detecting water in affected rooms and an inadequate warning system, 

and (3) the flood's effects on support functions and surrounding areas were not adequately 

accounted or were inappropriate for the weather conditions. 
-3­

APPLICABLE REGULATORY REQUIREMENTS 

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

• 	 10 CFR 50.54, "Conditions of Licenses" 

• 	 Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, 

GDC 2, "Design Bases for Protection against Natural Phenomena" 

• 	 Appendix A, "Seismic and Geologic Siting Criteria for Nuclear Power Plants," to 10 CFR 

Part 100 

The flooding 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 (seismically 

induced floods and water waves). 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 are to reflect appropriate 

consideration of the most severe of the natural phenomena that have been historically reported 

for the site and surrounding area. The design bases are also to reflect sufficient margin to 

account for the limited accuracy, quantity, and period of time in which the historical data have 

been accumulated. 

DISCUSSION 

The NTrF recommended that the Commission direct several actions to ensure adequate 

protection from natural phenomena. These actions should be taken to prevent fuel damage and 

to ensure containment and spent fuel pool integrity. In particular, Recommendation 2.3 states 

that the Commission should "Order licensees to perform seismic and flood protection 

walkdowns to identify and address plant-specific vulnerabilities and verify the adequacy of 

monitoring and maintenance for protection features such as water tight barriers and seals in the 

interim period until longer term actions are completed to update the design basis for external 

events." However, in the context of this letter, the NRC staff is focusing on degraded, 

nonconforming, or unanalyzed conditions and cliff-edge effects. 

The NRC staff's assessment of NTIF Recommendation 2.3 is discussed in SECY-11-0124. 

The NRC staff agreed with the NTIF Recommendation 2.3 findings and noted that some plants 

rely on operator actions and temporary flood mitigation measures such as sandbagging, 

temporary flood walls and barriers, and portable equipment to perform safety functions. Results 

of staff's inspections at nuclear power sites in accordance with Temporary Instruction (TI) 

2515/183 identified potential issues and observations regarding mitigation measures. Recent 

flooding at the Fort Calhoun site showed the importance of temporary flood mitigation 

measures. The NRC staff also noted that guidance should be developed for flooding 

walkdowns with external stakeholder involvement to ensure consistency. 
-4­

In its SRM to SECY-11-0124, the Commission approved the NRC staff's proposed actions to 

implement without delay the NTIF recommendations as described in the SECY. With regards 

to Recommendation 2.3, NRC staff's approved actions are to develop and issue a request for 

information to licensees pursuant to 10 CFR 50.54(f) to develop a methodology and acceptance 

criteria for flooding walkdowns to be endorsed by the NRC staff following interaction with 

external stakeholders, perform flood protection walkdowns to identify and address plant-specific 

degraded, nonconforming, or unanalyzed conditions and cliff-edge effects (through the 

corrective action program) and verify the adequacy of monitoring and maintenance for 

protection features, and inform the NRC of the results of the walkdowns and corrective actions 

taken or planned. 

The TI2515/183 was issued by the NRC on March 23, 2011. Inspection activities were 

completed by April 29, 2011, and NRC inspection reports were issued by May 13, 2011. The 

NRC developed a Summary of Observations report to document the performance of TI 

2515/183 (see http://www.nrc.gov/NRR/OVERSIGHT/ASSESS/follow-up-rpts.html). The 

summary report states that while individually, none of the observations posed a significant 

safety issue, they indicate a potential industry trend of failure to maintain equipment and 

strategies required to mitigate some design basis events. Regarding the licensee's capability to 

mitigate. design bases flooding events, the report notes that some equipment (mainly pumps) 

would not operate when tested, or lacked test acceptance criteria, and that some discrepancies 

were identified with barrier and penetration seals. 

Additional review of Section 03.03 of the responses to TI2515/183 indicates that several sites 

were susceptible to water accumulation that submerged safety-related cables. Issues were 

noted with cracks in penetrations, evidence of water infiltration, and groundwater intrusion. 

Individual TI inspection reports noted that a few licensee-proposed flood-related 

countermeasures may not achieve the intended mitigative effect. Flood barriers made of 

various construction materials were credited with being effective for preventing flooding, but the 

confirmatory testing to verify the effectiveness of certain of these mitigating actions was not 

conclusive. It should be noted that these findings are consistent with findings documented in 

the "Perspectives Gained" section of the IPEEE program report (NUREG-1742). 

The Advisory Committee on Reactor Safeguards (ACRS) in its letter dated October 13, 2011, 

requested that the Commission consider that "site-specific external hazards, vulnerabilities, and 

consequences need to be evaluated in an integrated context. For example. tornadoes and 

hurricanes may cause extended loss of offsite power with coincident physical damage to 

nonsafety structures or equipment at multiple units that has not been fully evaluated. Damage 

from severe storms or other site-specific hazards may also disable external essential cooling 

water supplies. Vulnerabilities to those hazards and subsequent damage may not be identified 

from assessments that focus only on design-basis seismic and flooding events." The ACRS 

further requested that "Near-term actions related to NTIF Recommendation 2.3 should be 

expanded to assure that the walkdowns address the integrated effects of severe storms as well 

as seismic and flooding events. The walkdowns and associated assessments should confirm 

that the identified hazards and vulnerabilities remain bounded by the current plant licensing 

basis." 
-5­

The NRC staff will interact with industry and stakeholders to develop a methodology and 

acceptance criteria for flooding walkdowns. These walkdowns should integrate the External 

Flood results in NUREG-1742, common issues and findings discussed in Section 03.03 of the 

responses to TI 2515/183, and the Significant Aspect findings discussed INPO SER 1-01. It is 

anticipated that the walkdown procedure will be developed or modified using various existing 

NRC- and industry-developed procedures. As mentioned in SECY-11-0124, recent flood events 

such as those at Fort Calhoun should also provide valuable insights. Additional attributes of the 

walkdown procedure are described in the Requested Action section below. The technical 

approach used to develop the needed information should be holistic and integrated to account 

for the site-specific design, physical barriers, procedures, temporary measures, and planned or 

installed mitigation measures to deal with the potential flooding scenarios. 

As stated earlier, the NRC staff will interact with industry and other stakeholders to develop an 

approach, which can be applied in a uniform and consistent manner across the different sites 

and plant conditions. An integrated approach will allow the NRC and industry to assess the 

significance of any new information related to flooding hazards in a systematic manner. During 

these interactions, the NRC staff will also work with industry and stakeholders to identify 

efficiencies and strategies to ensure that responses and reviews are timely and support the 

Commission guidance on the overall schedule. 

As mentioned in the cover letter, other external events (e.g., extreme winds and its effects) will 

be covered as a separate action from this letter. It would be prudent for addressees to consider 

the inclusion of other external events in these walkdown procedures due to the potential efficient 

use of similar resources to perform these walkdowns. 

REQUESTED ACTIONS 

The NRC requests that each addressee confirm that they will use the industry-developed, NRCendorsed, flood walkdown procedures 1 or provide a description of plant-specific walkdown 

procedures. The requested actions include the following: 

(1) Perform flood protection walkdowns using an NRC-endorsed walkdown methodology, 

(2) Identify and address plant-specific degraded, nonconforming, or unanalyzed conditions, 

as well as, cliff-edge effects through the corrective action program, and consider these 

findings in the Recommendation 2.1 hazard evaluations, as appropriate, 

(3) Identify any other actions taken or planned to further enhance the site flood protection, 

(4) Verify the adequacy of programs, monitoring and maintenance for protection features, 

and, 

(5) Report to the NRC the results of the walkdowns and corrective actions taken or planned. 

A final report should be submitted to the NRC addressing items identified in the Requested 

Information section. 

1 NRC staff are currently engaged with industry and other external stakeholders to develop NRC·endorsed procedures, Tre NRC 

staff anticipates completing this activity by May, 2012. 
-6­

It is requested that the walkdown procedure verify that flood protection systems for the plant are 

available, functional, and implementable under a variety of site conditions. In particular, the 

walkdowns should confirm that: (1) cable and piping trenches and other penetrations to SSCs 

important to safety, including underground rooms, are not pathways for external ingress of 

water, (2) adequate water detection and warning systems are available, if credited in the current 

licensing basis, (3) the effects of elevated water levels and severe weather conditions would not 

impair support functions or would not impede performing necessary actions given the weather 

conditions, and (4) other factors at multi-unit sites (e.g., equipment availability and staffing) 

would not prevent implementation of flood protection measures. 

If any condition identified during the walkdown activities represents a degraded, nonconforming, 

or unanalyzed condition (i.e., noncompliance with the current licensing basis) for an SSC, 

describe actions that were taken or are planned to address the condition using the guidance in 

Regulatory Issues Summary 2005-20, Revisions 1, Revision to NRC Inspection Manual Part 

9900 Technical Guidance, "Operability Conditions Adverse to Quality or Safety," including 

entering the condition in the corrective action program. Reporting requirements pursuant to 

10 CFR 50.72 should also be considered. In addition, if any condition noted during the 

walkdown represents a cliff-edge effect, describe any measures taken or planned to address the 

condition(s) while the corrective action is being implemented. 

Along with an assessment of reactor integrity, the NTTF recommended an evaluation of spent 

fuel pools to assess the effectiveness of the flood protection. The approach should account for 

the site-specific design, physical barriers, procedures, temporary measures, and planned or 

existing mitigation measures. 

REQUESTED INFORMATION 

1. 	 The NRC requests that each addressee confirm that it will use the industry-developed, 

NRC-endorsed,flooding walkdown procedures or provide a descrrption of plant-specific 

walkdown procedures that include the following characteristics: 

a. 	 Address the NTTF Report's observations regarding "overreliance on operator actions 

and temporary flood mitigation measures" and the 'cliff-edge' effect regarding a sharp 

increase in flooding risks with a small increase in flooding level. 

b. 	 Integrate issues discussed in the External Flood Qualitative Results (Section 4.3.3) in 

NUREG-1742, common issues and findings discussed in Section 03.03 of the 

responses to TI 2515/183, and the Significant Aspect findings discussed in 

INPO SER 1-01. 

c. 	 Integrate insights from any new and relevant flood hazard information, as well as 

recent flood-related walkdowns such as the events at the Fort Calhoun site, as 

mentioned in SECY-11-0124. Additionally, relevant NRC inspection findings could 

provide additional insights. 

d. 	 Integrate the combined effects of flooding along with other adverse conditions, such 

as high winds, hail, lightning, etc., that could reasonably be expected to 

simultaneously occur. For example, steps in a flooding procedure that require 

manipulation of systems and components in outside areas of the plant site that could 

not be safely assessed because of storm conditions. 
-Te. 	 Identify pre-walkdown actions, such as the collection of current site topography 

including any changes since the original licensing (e.g., security improvements and 

temporary structures), sets of as-built drawings, review of the existing design basis 

flood level(s), review of any flood protection and pertinent flood mitigation features, 

such as exterior barriers, incorporated barriers, and temporary flood barriers. 

f. 	 Identify a list of pertinent elevations of Regulatory Guide 1.292 structures, systems, 

and components that should be designed to withstand the design basis hazard 

(similar to Table 1, i.e., 3.1.3 of American National Standards Institute/American 

Nuclear Society (ANSIIANS)-2.B-1992) 

g. 	 Identify the team composition and qualifications. 

h. 	 Verify that flood protection systems are available, functional, and implementable 

under a variety of site conditions by reviewing the following: 

i. 	 Operator availability, operator training, timeliness of response, equipment 

maintenance and operability, back-up availability, operator access under 

adverse site conditions3 

ii. 	 Methods and acceptance criteria to evaluate exterior barriers4 

iii. 	 Methods and acceptance criteria to evaluate incorporated barriers 

iv. 	 Methods and acceptance criteria to evaluate temporary flood barriers 

v. 	 Preparations in advance of adverse weather conditions 

i. 	 Identify programs in place that periodically verify the status and adequacy of flood 

mitigation strategies and equipment. 

j. 	 Develop a documentation template, including peer-review requirements, so that 

walkdown results can be efficiently and uniformly reviewed and evaluated. The 

template should also consider the reporting requirement discussed below. 

2. 	 Following NRC's endorsement of the walkdown procedure, conduct the walkdown and 

submit a final report which includes the following: 

a. 	 Describe the design basis flood hazard level(s) for all flood-causing mechanisms, 

including groundwater ingress. 

b. 	 Describe protection and mitigation features that are considered in the licensing basis 

evaluation to protect against external ingress of water into SSCs important to safety. 

c. 	 Describe any warning systems to detect the presence of water in rooms important to 

safety. 

d. 	 Discuss the effectiveness of flood protection systems and exterior, incorporated, and 

temporary flood barriers. Discuss how these systems and barriers were evaluated using 

the acceptance criteria developed as part of Requested Information item 1.h. 

2 Regulatory Guide 1.59, ·Design Basis Floods for Nuclear Power Plants·, and Regulatory Guide 1.102, Flood Protection for Nuclear 

Power Plants," both recommend the use of Regulatory Guide 1.29, "Seismic Design Classification" for identifying structures, 

systems, and components, that should be designed to withstand the conditions resulting from the design basis flood and remain 

functional. 

3 This may not be an all-inclusive list. 

4 See Regulatory Position 1 of Regulatory Guide 1.102, "Flood Protection for Nuclear Power Plants,» for definitions a(".ceptable to the 

NRC staff for exterior barriers, incorporated barriers, and temporary barriers 
-8­

e. Present information related to the implementation of the walkdown process (e.g., details 

of selection of the walkdown team and procedures,) using the documentation template 

discussed in Requested Information item 1.j, including actions taken in response to the 

peer review. 

f. Results of the walkdown including key findings and identified degraded, nonconforming, 

or unanalyzed conditions. Include a detailed description of the actions taken or planned 

to address these conditions using the guidance in Regulatory Issues Summary 2005-20, 

Revision 1, Revision to NRC Inspection Manual Part 9900 Technical Guidance, 

"Operability Conditions Adverse to Quality or Safety," including entering the condition in 

the corrective action program. 

g. Document any cliff-edge effects identified and the associated basis. Indicate those that 

were entered into the corrective action program. Also include a detailed description of 

the actions taken or planned to address these effects. 

h. Describe any other planned or newly installed flood protection systems or flood 

mitigation measures including flood barriers that further enhance the flood protection. 

Identify results and any subsequent actions taken in response to the peer review. 

REQUIRED RESPONSE 

In accordance with 10 CFR 50.54(f), an addressee must respond as described below. The 

submission of the requested information is in stages to allow adequate time for further 

interactions with the stakeholders to provide clarifications, to develop implementation 

procedures and processes, and to develop the associated guidance as needed. 

1. 	 Within 90 days of the date of this information request, the addressee will confirm that it 

intends to use the NRC-endorsed flooding walkdown procedures or provide the NRC a 

description of the process that will be used to conduct the walkdowns and to develop the 

needed information. 

2. 	 Within 180 days of NRC's endorsement of the walkdown procedure, each addressee will 

submit its final response for the requested information. This response should include a 

list of any areas that are unable to be inspected due to inaccessibility and a schedule for 

when the walkdown will be completed. 

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, A TIN: 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. 
-9­

Enclosure 4 References 

SECY-11-0124, "Recommended Actions to be taken without Delay from the Near-Term Task 

Force Report," Agencywide Documents Access and Management System (ADAMS) Accession 

No. ML 11245A 158, dated September 9, 2011. 

SECY-11-0137, "Prioritization of Recommended Actions to be Taken in Response to Fukushima 

Lessons-Learned," ADAMS Accession No. ML 11272A 111, October 3, 2011. 

"Recommendations for Enhancing Reactor Safety in the 21 st Century: The Near-term Task 

Force Review of Insights from the Fukushima Dai-ichi Accident," ADAMS Accession No. 

ML 111861807, July 12, 2011. 

10 CFR 50.54 - Conditions of Licenses 

10 CFR 50.72, "Immediate Notification Requirements for Operating Nuclear Power Reactors" 

Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants 

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

Temporary Instruction 2515/183, "Follow-up to the Fukushima Dai-ichi Fuel Damage Event," 

November 2011, ADAMS Accession No. ML 113220407. 

Energy and Water Development and Related Agencies Appropriations Act, 2012 

NUREG-0800, SRP Section 2.4 

NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of 

External Events (IPEEE) for Severe Accident Vulnerabilities," Final Report, ADAMS Accession 

No. ML063550238, June 1991. 

ASME/ANS RA-Sa-2009, American Society of Mechanical Engineers/American Nuclear Society 

standard, RA-Sa-2009, "Standard for Level1/Large Early Release Frequency Probabilistic Risk 

Assessment for Nuclear Power Plant Applications," 2009. 

INPO version, SER 1-01, "WANO Significant Event Report (SER) 2000-3, 'Severe Storm 

Results in Scram of Three Units and Loss of Safety System Functions Due to Partial Plant 

Flooding,'" February 2001 (Proprietary) 
RECOMMENDATION 9.3: EMERGENCY PREPAREDNESS 


Communications 


PURPOSE 

The U.S. Nuclear Regulatory Commission (NRC or Commission) is issuing this information 

request regarding the power supplies for communications systems to determine if additional 

regulatory action is warranted. This request is based upon Near-Term Task Force (NTTF) 

Recommendation 9.3 which proposed that facility emergency plans provide for a means to 

power communications equipment needed to communicate onsite (e.g., radios for response 

teams and between facilities) and offsite (e.g., cellular telephones and satellite telephones) 

during a prolonged station blackout. 

APPLICABLE REGULATORY REQUIREMENTS AND GUIDANCE 

Emergency plan communications requirements and detailed guidance on how to meet those 

requirements are contained in the following: 

1. 	 Title 10 of the Code of Federal Regulations (10 CFR) 50.47 (b)(6) states that provisions 

should be made for prompt communications among principal response organizations to 

emergency personnel and to the public. 

2. 	 Appendix E, "Emergency Planning and Preparedness for Production and Utilization 

Facilities," to 10 CFR Part 50, "Domestic Licensing for Production and Utilization 

Facilities," Section IV. E. 9, states that adequate provisions shall be made and described 

for emergency facilities and equipment, including "at least one onsite and one offsite 

communications system; each system shall have a backup power source." 

3. 	 NUREG-0696, "Functional Criteria for Emergency Response Facilities," issued 

February 1981, offers guidance on how to meet the requirements of Appendix E to 

10 CFR Part 50, and discusses the onsite and offsite communications requirements for 

the licensee's emergency operating facilities. 

DISCUSSION 

During the March 11, 2011, T5hoku earthquake and subsequent tsunami, the widespread 

destruction and loss of electrical power degraded communications capabilities onsite at 

Fukushima Dai-ichi and between the site and external stakeholders, such as local emergency 

response centers, the Japanese government, and corporate offices. Normal and emergency 

offsite communications systems lost power or were degraded by the earthquake and tsunami. 

Normal and emergency onsite communications were severely impacted by the loss of power to 

signal repeaters and depleted radio batteries. Accounts of the accident response refer to delays 

in repair activities caused by issues with the ability to effectively communicate between repair 

teams and the control rooms and the onsite emergency response center. 

Enclosure 5 
- 2 ­

The NRC requests that the following assumptions be made in preparing responses to this 

request for information: the potential onsite and offsite damage is a result of a large scale 

natural event resulting in a loss of all alternating current (ac) power. 

In addition, assume that the large scale natural event causes extensive damage to normal and 

emergency communications systems both onsite and in the area surrounding the site. It has 

been recognized that following a large scale natural event that ac power may not be available to 

cell and other communications infrastructures. 

REQUESTED ACTIONS 

It is requested that addressees assess their current communications systems and equipment 

used during an emergency event given the aforementioned assumptions. It is also requested 

that consideration be given to any enhancements that may be appropriate for the emergency 

plan with respect to communications requirements of 10 CFR 50.47, Appendix E to 

10 CFR Part 50, and the guidance in NUREG-0696 in light of the assumptions stated above. 

Also addressees are requested to consider the means necessary to power the new and existing 

communications equipment during a prolonged S80. 

REQUESTED INFORMATION 

1. 	 Addressees are requested to provide an assessment of the current communications 

systems and equipment used during an emergency event to identify any enhancements that 

may be needed to ensure communications are maintained during a large scale natural event 

meeting the conditions described above. The assessment should: 

• 	 Identify any planned or potential improvements to existing onsite communications 

systems and their required normal and/or backup power supplies, 

• 	 Identify any planned or potential improvements to existing offsite communications 

systems and their required normal and/or backup power supplies, 

• 	 Provide a description of any new communications system(s) or technologies that will be 

deployed based upon the assumed conditions described above, and 

• 	 Provide a description of how the new and/or improved systems and power supplies will 

be able to provide for communications during a loss of all ac power, 

2. 	 Addressees are requested to describe any interim actions that have been taken or are 

planned to be taken to enhance existing communications systems power supplies until the 

communications assessment and the resulting actions are complete, 

3. 	 Provide an implementation schedule of the time needed to conduct and implement the 

results of the communications assessment. 
- 3 ­

REQUIRED RESPONSE 


The addressee should respond to this request for information no later than 90 days from the 

date of issuance. 

If an addressee cannot meet the requested response date, the addressee must provide a 

response within 60 days of the date of this letter 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 date. 

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. 
Staffing 

PURPOSE 

The U.S. Nuclear Regulatory Commission (NRC or Commission) is issuing this information 

request to determine if additional regulatory action is warranted regarding the staff required to fill 

all necessary positions to respond to a multi-unit event. 

Single unit sites should provide the requested information as it pertains to an extended loss of 

all ac power, and impeded access to the site. 

APPLICABLE REGULATORY REQUIREMENTS AND GUIDANCE 

• 	 Title 10 of the Code of Federal Regulations (10 CFR) 50.47(b)(1) states, in part: ..... and 

each principal response organization has staff to respond and to augment its initial 

response on a continuous basis." 

(\ 	 10 CFR 50.47(b)(2) states, in part: ..... adequate staffing to provide initial facility accident 

response in key functional areas is maintained at all times, timely augmentation of 

response capabilities is available, and ..." 

• 	 NUREG-0654/FEMA-REP-1, Revision 1, "Criteria for Preparation and Evaluation of 

Radiological Emergency Response Plans and Preparedness in Support of Nuclear 

Power Plants," Section B, Onsite Emergency Organization, states in part: 

Each licensee shall specify... functional areas of emergency activity ... 

These assignments shall cover the emergency functions in Table B-1 

entitled, 'Minimum Staffing Requirements for Nuclear Power Plant 

Emergencies.' The minimum on-shift staffing shall be as indicated in 

Table B-1. The licensee must be able to augment on-shift capabilities 

within a short period after declaration of an emergency. This capability 

shall be as indicated in Table B-1 ... 

DISCUSSION 

The events in Japan have highlighted the importance of responders during all phases of 

emergency event response. The regulations require emergency response capabilities during a 

broad spectrum of postulated reactor accidents. A natural event on the scale of the 2011 Great 

East Japan Earthquake and resulting tsunami could present new challenges to personnel and 

their safety. Specifically, the event stressed the existing regulatory framework and impacted the 

operator's capability to implement adequate protective measures to protect the public and plant 

staff. In light of the experience from the event, the unavailability of sufficient onsite staff during 

the initial phase of the emergency condition, the unavailability of staff designated to augment 

the onsite staff, the inability for offsite support to reach the site, and the unavailability and 

inability of relief staff to reach the site, the NRC recognizes that these in total could pose 

challenges to licensee response efforts. 
- 2 ­

A large scale natural event may alter the planned emergency framework by changing access 

routes (e.g., bridges washed out, debris blocking roadways, etc.). While several utilities have 

implemented a combined emergency operations facility that is capable of handling multi-unit 

events, the onsite technical support center and operational support center at sites with multiple 

reactors have been designed to handle any emergency at only one of the units. 

In conjunction with the Emergency Preparedness regulations Agencywide Documents Access 

and Management System (ADAMS) Accession No. ML 112070125 published on November 10, 

2011, the NRC published on December 5, 2011, in the Federal Register (76 FR 75771) interim 

staff guidance (lSG) in NSIRIDPR-ISG-01 ADAMS Accession No. ML 1113010523. Section 

IV.C of the ISG provides guidance on performing an on-shift staffing analysis, and identified 

Nuclear Energy Institute (NEI)-10-05, "Assessment of On-shift Emergency Response 

Organizations Staffing and Capabilities" ADAMS Accession No. ML 111751698, as an 

acceptable methodology for such an analysis. However, this methodology and guidance does 

not consider multiple unit events involving a large scale natural event with a loss of all 

alternating current (ac) power. 

This letter requests that addresses assess and provide the NRC with information regarding the 

ability to implement their emergency plan during a large scale natural event that results in the 

following: 

• 	 all units affected. 

• 	 extended loss of all ac power, and 

• 	 im peded access to the sites 

Addressees may find the capability for assessment activities, including repair team planning and 

preparation are particularly impacted. Therefore, it is requested that this assessment ensure 

that there is sufficient onsite staff and other resources to perform critical tasks until 

augmentation staff arrives to provide assistance and until other offsite resources become 

available. 

REQUESTED ACTIONS 

It is requested that addressees assess their current staffing levels and determine the 

appropriate staff to fill all necessary positions for responding to a multi-unit event during a 

beyond design basis natural event and determine if any enhancements are appropriate given 

the considerations of Near-Term Task Force (NITF) Recommendation 9.3. 

Single unit sites should provide the requested information as it pertains to an extended loss of 

all ac power, and impeded access to the site. 

REQUESTED INFORMATION 

1. 	 It is requested that addressees provide an assessment of the onsite and augmented staff 

needed to respond to a large scale natural event meeting the conditions described above. 

This assessment should include a discussion of the onsite and augmented staff available to 

implement the strategies as discussed in the emergency plan and/or described in plant 

operating procedures. The following functions are requested to be assessed: 
- 3 ­

• 	 How onsite staff will move back-up equipment (e.g., pumps, generators) from alternate 

onsite storage facilities to repair locations at each reactor as described in the Order 

regarding the NTIF Recommendation 4.2. It is requested that consideration be given to 

the major functional areas of NUREG-0654, Table 8-1, such as plant operations and 

assessment of operational aspects, emergency direction and control, 

notification/communication, radiological accident assessment, and support of operational 

accident assessment, as appropriate. 

• 	 New staff or functions identified as a result of the assessment. 

• 	 Collateral duties (personnel not being prevented from timely performance of their 


assigned functions). 


2. 	 Provide an implementation schedule of the time needed to conduct the onsite and 

augmented staffing assessment. If any modifications are determined to be appropriate, 

please include in the schedule the time to implement the changes. 

3. 	 Identify how the augmented staff would be notified given degraded communications 

capabilities. 

4. 	 Identify the methods of access (e.g., roadways, navigable bodies of water and dockage, 

airlift, etc.) to the site that are expected to be available after a widespread large scale 

natural event. 

5. 	 Identify any interim actions that have been taken or are planned prior to the completion of 

the staffing assessment. 

6. 	 Identify changes that have been made or will be made to your emergency plan regarding the 

on-shift or augmented staffing changes necessary to respond to a loss of all ac power, multiunit event, including any new or revised agreements with offsite resource providers (e.g., 

staffing, equipment, transportation, etc.). 

REQUIRED RESPONSE 

In accordance with Section 182.a of the Atomic Energy Act of 1954, as amended, and 10 CFR 

50.54(f), each addressee is requested to submit a written response consistent with the 

requested information. The response to requested information items 1 and 2 should be 

provided within 60 days of issuance of the ISG to be referenced in the NRC Order associated 

with NTIF Recommendation 4.2. The response to requested information items 3-6 should be 

provided within 90 days of the date of this letter. 

If an addressee cannot meet the requested response date, the addressee must provide a 

response within 60 days of the date of this letter 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 date. 

The required written response should be addressed to the U.S. Nuclear Regulatory 

Commission, ATIN: Document Control Desk, 11555 Rockville Pike, Rockville, MD 20852, under 
- 4 ­

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. 
POWER REACTOR LICENSEES AND HOLDERS OF 


CONSTRUCTION PERMITS IN ACTIVE OR DEFERRED STATUS 


Arkansas Nuclear One 

Entergy Operations, Inc. 

Docket Nos. 50-313 and 50-368 

License Nos. DPR-51 and NPF-6 

Mr. Christopher J. Schwarz 

Vice President, Operations 

Entergy Operations, Inc. 

Arkansas Nuclear One 

1448 S.R. 333 

Russellville, AR 72802 

Beaver Valley Power Station 

First Energy Nuclear Operating Co. 

Docket Nos. 50-334 and 50-412 

License Nos. DPR-66 and NPF-73 

Mr. Paul A. Harden 

Site Vice President 

FirstEnergy Nuclear Operating Company 

Mail Stop A-BV-SEB1 

P.O. Box 4, Route 168 

Shippingport, PA 15077 

Bellefonte Nuclear Power Station 

Tennessee Valley Authority 

Docket Nos. 50-438 and 50-439 

Construction Permit Nos. CPPR No. 122 and CPPR No. 123 

Mr. Michael D. Skaggs 

Senior Vice President, Nuclear Generation Development and Construction 

Tennessee Valley Authority 

6A Lookout Place 

1101 Market Street 

Chattanooga, TN 37402-2801 

Braidwood Station 

Exelon Generation Co., LLC 

Docket Nos. STN 50-456 and STN 50-457 

License Nos. NPF-72 and NPF-77 

Mr. Michael J. Pacilio 

President and Chief Nuclear Officer 

Exelon Nuclear 

4300 Winfield Road 

Warrenville,IL 60555 

Enclosure 6 
- 2 ­

Browns Ferry Nuclear Plant 

Tennessee Valley Authority 

Docket Nos. 50-259, 50-260 and 50-296 

License Nos. DPR-33, DPR-52 and DPR-68 

Mr. Preston D. Swafford 

Chief Nuclear Officer and Executive Vice President 

Tennessee Valley Authority 

3R Lookout Place 

1101 Market Street 

Chattanooga, TN 37402-2801 

Brunswick Steam Electric Plant 

Carolina Power & Light Co. 

Docket Nos. 50-325 and 50-324 

License Nos. DPR-71 and DPR-62 

Mr. Michael J. Annacone 

Vice President 

Carolina Power &Light Company 

Brunswick Steam Electric Plant 

P. O. Box 10429 

Southport, NC 28461 

Byron Station 

Exelon Generation Co., LLC 

Docket Nos. STN 50-454 and STN 50-455 

License Nos. NPF-37 and NPF-66 

Mr. Michael J. Pacilio 

President and Chief Nuclear Officer 

Exelon Nuclear 

4300 Winfield Road 

Warrenville, IL 60555 

Callaway Plant 

Union Electric Co. 

Docket No. 50-483 

License No. NPF-30 

Mr. Adam C. Heflin 

Senior Vice President and Chief Nuclear Officer 

Union Electric Company 

P. O. Box 620 

Fulton, MO 65251 
- 3 -


Calvert Cliffs Nuclear Power Plant 

Calvert Cliffs Nuclear Power Plant, LLC 

Docket Nos. 50-317 and 50-318 

License Nos. DPR-53 and DPR-69 

Mr. George H. Gellrich 

Vice President 

Calvert Cliffs Nuclear Power Plant, LLC 

Calvert Cliffs Nuclear Power Plant 

1650 Calvert Cliffs Parkway 

Lusby, MD 20657-4702 

Catawba Nuclear Station 

Duke Energy Carolinas, LLC 

Docket Nos. 50-413 and 50-414 

License Nos. NPF-35 and NPF-52 

Mr. James R. Morris 

Site Vice President 

Duke Energy Carolinas, LLC 

Catawba Nuclear Station 

4800 Concord Road 

York, SC 29745 

Clinton Power Station 

Exelon Generation Co., LLC 

Docket No. 50-461 

License No. NPF-62 

Mr. Michael J. Pacilio 

President and Chief Nuclear Officer 

Exelon Nuclear 

4300 Winfield Road 

Warrenville, IL 60555 

Columbia Generating Station 

Energy Northwest 

Docket No. 50-397 

License No. NPF-21 

Mr. Mark E. Reddemann 

Chief Executive Officer 

Energy Northwest 

MD 1023 

P.O. Box 968 

Richland, WA 99352 
- 4 ­

Comanche Peak Nuclear Power Plant 

Luminant Generation Co., LLC 

Docket Nos. 50-445 and 50-446 

License Nos. NPF-87 and NPF-89 

Mr. Rafael Flores 

Senior Vice President and Chief Nuclear Officer 

Luminant Generation Company, LLC 

Attn: Regulatory Affairs 

P. O. Box 1002 

Glen Rose, TX 76043 

Cooper Nuclear Station 

Nebraska Public Power District 

Docket No. 50-298 

License No. DPR-46 

Mr. Brian J. O'Grady 

Vice President - Nuclear and Chief Nuclear Officer 

Nebraska Public Power District 

72676 648A Avenue 

P.O. Box 98 

Brownville, f\lE 68321 

Crystal River Nuclear Generating Plant 

Florida Power Corp. 

Docket No. 50-302 

License No. DPR-72 

Mr. Jon A. Franke 

Vice President 

Attn: Supervisor, Licensing & Regulatory Affairs 

Progress Energy, Inc. 

Crystal River Nuclear Plant (NA2C) 

15760 West Power Line Street 

Crystal River, FL 34428-6708 

Davis-Besse Nuclear Power Station 

First Energy Nuclear Operating Co. 

Docket No. 50-346 

License No. NPF-3 

Mr. Barry S. Allen 

Site Vice President 

FirstEnergy Nuclear Operating Company 

c/o Davis-Besse NPS 

5501 N. State Route 2 

Oak Harbor, OH 43449-9760 
- 5­

Diablo Canyon Power Plant 

Pacific Gas & Electric Co. 

Docket Nos. 50-275 and 50-323 

License Nos. DPR-80 and DPR-82 

Mr. John T. Conway 

Senior Vice President - Energy Supply and Chief Nuclear Officer 

Pacific Gas and Electric Company 

Diablo Canyon Power Plant 

77 Beale Street, Mail Code B32 

San Francisco, CA 94105 

Donald C. Cook Nuclear Plant 

Indiana Michigan Power Co. 

Docket Nos. 50-315 and 50-316 

License Nos. DPR-58 and DPR-74 

Mr. Lawrence J. Weber 

Senior Vice President and Chief Nuclear Officer 

Indiana Michigan Power Company 

Nuclear Generation Group 

One Cook Place 

Bridgman, MI 49106 

Dresden Nuclear Power Station 

Exelon Generation Co., LLC 

Docket Nos. 50-237 and 50-249 

License Nos. DPR-19 and DPR-25 

Mr. Michael J. Pacilio 

President and Chief Nuclear Officer 

Exelon Nuclear 

4300 Winfield Road 

Warrenville, IL 60555 

Duane Arnold Energy Center 

NextEra Energy Duane Arnold, LLC 

Docket No. 50-331 

License No. DPR-49 

Mr. Peter Wells 

Site Vice President 

NextEra Energy 

Duane Arnold Energy Center 

3277 DAEC Road 

Palo, IA 52324-9785 
- 6 -

Edwin I. Hatch Nuclear Plant 

Southern Nuclear Operating Co. 

Docket Nos. 50-321 and 50-366 

License Nos. DPR-57 and NPF-5 

Mr. Dennis R. Madison 

Vice President 

Southern Nuclear Operating Company, Inc. 

Edwin I. Hatch Nuclear Plant 

11028 Hatch Parkway North 

Baxley, GA 31513 

Fermi 

Detroit Edison Co. 

Docket No. 50-341 

License No. NPF-43 

Mr. Jack M. Davis 

Senior Vice President and Chief Nuclear Officer 

Detroit Edison Company 

Fermi 2 - 210 NOC 

6400 North Dixie Highway 

Newport, MI 48166 

Fort Calhoun Station 

Omaha Public Power District 

Docket No. 50-285 

License No. DPR-40 

Mr. David J. Bannister 

Vice President and Chief Nuclear Officer 

Omaha Public Power District 

444 South 16th St. Mall 

Omaha, NE 68102-2247 

Grand Gulf Nuclear Station 

Entergy Operations, Inc. 

Docket No. 50-416 

License No. NPF-29 

Mr. Michael Perito 

Vice President, Operations 

Entergy Operations, Inc. 

Grand Gulf Nuclear Station, Unit 1 

7003 Bald Hill Road 

Port Gibson, MS 39150 
- 7 ­

H. B. Robinson Steam Electric Plant 

Carolina Power & Light Co. 

Docket No. 50-261 

License No. DPR-23 

Mr. Robert J. Duncan II 

Vice President 

Carolina Power & Light Company 

3581 West Entrance Road 

Hartsville, SC 29550 

Hope Creek Generating Station 

PSEG Nuclear, LLC 

Docket No. 50-354 

License No. NPF-57 

Mr. Thomas Joyce 

President and Chief Nuclear Officer 

PSEG Nuclear LLC - N09 

P. O. Box 236 

Hancocks Bridge, NJ 08038 

Indian Point Energy Center 

Entergy Nuclear Operations, Inc. 

Docket Nos. 50-247 and 50-286 

License Nos. DPR-26 and DPR-64 

Mr. John Ventosa 

Vice President, Operations 

Entergy Nuclear Operations, Inc. 

Indian Point Energy Center 

450 Broadway, GSB 

P.O. Box 249 

Buchanan, NY 10511-0249 

James A. FitzPatrick Nuclear Power Plant 

Entergy Nuclear Operations, Inc. 

Docket 1'110. 50-333 

License No. DPR-59 

Mike Colomb 

Vice President, Operations 

Entergy Nuclear Operations, Inc. 

James A. FitzPatrick Nuclear Power Plant 

P.O. Box110 

Lycoming, NY 13093 
-8 

Joseph M. Farley Nuclear Plant 

Southern Nuclear Operating Co. 

Docket Nos. 50-348 and 50-364 

License Nos. NPF-2 and NPF-8 

Mr. Tom Lynch 

Vice President - Farley 

Southern Nuclear Operating Company, Inc. 

Joseph M. Farley Nuclear Plant 

7388 North State Highway 95 

Columbia, AL 36319 

Kewaunee Power Station 

Dominion Energy Kewaunee, Inc. 

Docket No. 50-305 

License No. DPR-43 

Mr. David A. Heacock 

President and Chief Nuclear Officer 

Dominion Energy Kewaunee, Inc. 

Innsbrook Technical Center 

5000 Dominion Boulevard 

Glen Allen, VA 23060-6711 

LaSalle County Station 

Exelon Generation Co., LLC 

Docket Nos. 50-373 and 50-374 

License Nos. NPF-11 and NPF-18 

Mr. Michael J. Pacilio 

President and Chief Nuclear Officer 

Exelon Nuclear 

4300 Winfield Road 

Warrenville, IL 60555 

Limerick Generating Station 

Exelon Generation Co., LLC 

Docket Nos. 50-352 and 50-353 

License Nos. NPF-39 and NPF-85 

Mr. Michael J. Pacilio 

President and Chief Nuclear Officer 

Exelon Nuclear 

4300 Winfield Road 

Warrenville, IL 60555 
- 9 ­

Millstone Nuclear Power Station 

Dominion Nuclear Connecticut, Inc. 

Docket Nos. 50-336 and 50-423 

License Nos. DPR-65 and NPF-49 

Mr. David A. Heacock 

President and Chief Nuclear Officer 

Dominion Nuclear Connecticut, Inc. 

Innsbrook Technical Center 

5000 Dominion Boulevard 

Glen Allen, VA 23060-6711 

Monticello Nuclear Generating Plant 

Northern States Power Company 

Docket No. 50-263 

License No. DPR-22 

Mr. Timothy J. O'Connor 

Site Vice President 

Northern States Power Company - Minnesota 

Monticello Nuclear Generating Plant 

2807 West County Road 75 

Monticello, MN 55362-9637 

Nine Mile Point Nuclear Station 

Nine Mile Point Nuclear Station, LLC 

Docket Nos. 50-220 and 50-410 

License Nos. DPR-63 and NPF-69 

Mr. Ken Langdon 

Vice President Nine Mile Point 

Nine Mile Point Nuclear Station, LLC 

P. O. Box 63 

Lycoming, NY 13093 

North Anna Power Station 

Virginia Electric & Power Co. 

Docket Nos. 50-338 and 50-339 

License Nos. NPF-4 and NPF-7 

Mr. David A. Heacock 

President and Chief Nuclear Officer 

Dominion Nuclear 

Innsbrook Technical Center 

5000 Dominion Boulevard 

Glen Allen, VA 23060-6711 
- 10­

Oconee Nuclear Station 

Duke Energy Carolinas, LLC 

Docket Nos. 50-269, 50-270 and 50-287 

License Nos. DPR-38, DPR-47 and DPR-55 

Mr. Preston Gillespie 

Site Vice President, Oconee Nuclear Station 

Duke Energy Carolinas, LLC 

7800 Rochester Highway 

Seneca, SC 29672 

Oyster Creek Nuclear Generating Station 

Exelon Generation Co., LLC 

Docket No. 50-219 

License No. DPR-16 

Mr. Michael J. Pacilio 

President and Chief Nuclear Officer 

Exelon Nuclear 

4300 Winfield Road 

Warrenville,IL 60555 

Palisades Nuclear Plant 

Entergy Nuclear Operations, Inc. 

Docket No. 50-255 

License No. DPR-20 

Mr. Anthony J. Vitale 

Site Vice President - Palisades 

Entergy Nuclear Operations, Inc. 

Palisades Nuclear Plant 

27780 Blue Star Memorial Highway 

Covert, Ml 49043 

Palo Verde Nuclear Generating Station 

Arizona Public Service Company 

Docket Nos. STN 50-528, STN 50-529 and STN 50-530 

License Nos. NPF-41, NPF-51 and NPF-74 

Mr. Randall K. Edington 

Executive Vice President Nuclear and Chief Nuclear Officer 

Arizona Public Service Co. 

P. O. Box 52034, MS 7602 

Phoenix, AZ 85072-2034 
- 11 ­

Peach Bottom Atomic Power Station 

Exelon Generation Co., LLC 

Docket Nos. 50-277 and 50-278 

License Nos. DPR-44 and DPR-56 

Mr. Michael J. Pacilio 

President and Chief Nuclear Officer 

Exelon Nuclear 

4300 Winfield Road 

Warrenville, IL 60555 

Perry Nuclear Power Plant 

First Energy Nuclear Operating Co. 

Docket No. 50-440 

License No. NPF-58 

Mr. Vito A Kaminskas 

Site Vice President - Nuclear - Perry 

FirstEnergy Nuclear Operating Company 

Perry Nuclear Power Plant 

10 Center Road, A290 

Perry, OH 44081 

Pilgrim Nuclear Power Station Unit No.1 

Entergy Nuclear Operations, Inc. 

Docket No. 50-293 

License No. DPR-35 

Mr. Robert Smith 

Vice President and Site Vice President 

Entergy Nuclear Operations, Inc. 

Pilgrim Nuclear Power Station 

600 Rocky Hill Road 

Plymouth, MA 02360-5508 

Point Beach Nuclear Plant 

NextEra Energy Point Beach, LLC 

Docket Nos. 50-266 and 50-301 

License Nos. DPR-24 and DPR-27 

Mr. Larry Meyer 

Site Vice President 

NextEra Energy Point Beach, LLC 

Point Beach Nuclear Plant, Units 1 &2 

6610 Nuclear Road 

Two Rivers, WI 54241-9516 
- 12 ­

Prairie Island Nuclear Generating Plant 

Northern States Power Co. Minnesota 

Docket Nos. 50-282 and 50-306 

License Nos. DPR-42 and DPR-60 

Mr. Mark A. Schimmel 

Site Vice President 

Northern States Power Company - Minnesota 

Prairie Island Nuclear Generating Plant 

1717 Wakonade Drive East 

Welch, MN 55089-9642 

Quad Cities Nuclear Power Station 

Exelon Generation Co., LLC 

Docket Nos. 50-254 and 50-265 

License Nos. DPR-29 and DPR-30 

Mr. Michael J. Pacilio 

President and Chief Nuclear Officer 

Exelon Nuclear 

4300 Winfield Road 

Warrenville, IL 60555 

R E. Ginna Nuclear Power Plant 

RE. Ginna Nuclear Power Plant, LLC 

Docket No. 50-244 

License No. DPR-18 

Mr. Joseph E. Pacher 

Vice President 

RE. Ginna Nuclear Power Plant, LLC 

RE. Ginna Nuclear Power Plant 

1503 Lake Road 

Ontario, NY 14519 

River Bend Station 

Entergy Operations, Inc. 

Docket No. 50-458 

License No. NPF-47 

Mr. Eric W. Olson 

Vice President, Operations 

Entergy Operations, Inc. 

River Bend Station 

5485 U.S. Highway 61 N 

St. Francisville, LA 70775 
- 13 ­

Salem Nuclear Generating Station 

PSEG Nuclear, LLC 

Docket Nos. 50-272 and 50-311 

License Nos. DPR-70 and DPR-75 

Mr. Thomas Joyce 

President and Chief Nuclear Officer 

PSEG Nuclear LLC - N09 

P. O. Box 236 

Hancocks Bridge, NJ 08038 

San Onofre Nuclear Generating Station 

Southern California Edison Co. 

Docket Nos. 50-361 and 50-362 

License Nos. NPF-10 and NPF-15 

Mr. Peter T. Dietrich 

Senior Vice President and Chief Nuclear Officer 

Southern California Edison Company 

San Onofre Nuclear Generating Station 

P.O. Box 128 

San Clemente, CA 92674-0128 

Seabrook 

NextEra Energy Seabrook, LLC 

Docket No. 50-443 

License No. NPF-86 

Mr. Paul Freeman 

Site Vice President 

NextEra Energy Seabrook, LLC 

clo Mr. Michael O'Keefe 

NextEra Energy Seabrook, LLC 

P.O. Box 300 

Seabrook, NH 03874 

Seguoyah Nuclear Plant 

Tennessee Valley Authority 

Docket Nos. 50-327 and 50-328 

License Nos. DPR-77 and DPR-79 

Mr. Preston D. Swafford 

Chief Nuclear Officer and Executive Vice President 

Tennessee Valley Authority 

3R Lookout Place 

1101 Market Street 

Chattanooga, TN 37402-2801 
- 14 -

Shearon Harris Nuclear Power Plant 

Carolina Power & Light Co. 

Docket No. 50-400 

License No. NPF-63 

Mr. Christopher L Burton 

Vice President 

Progress Energy Carolinas, Inc. 

Shearon Harris Nuclear Power Plant 

P. O. Box 165, Mail Zone 1 

New Hill, NC 27562-0165 

South Texas Project 

STP Nuclear Operating Co. 

Docket Nos. 50-498 and 50-499 

License Nos. NPF-76 and NPF-80 

Mr. Edward D. Halpin 

President, Chief Executive Officer and Chief Nuclear Officer 

STP Nuclear Operating Company 

South Texas Project 

P. O. Box 289 

Wadsworth, TX 77483 

St. Lucie Plant 

Florida Power & Light Co. 

Docket Nos. 50-335 and 50-389 

License Nos. DPR-67 and NPF-16 

Mr. Mano Nazar 

Executive Vice President and Chief Nuclear Officer 

NextEra Energy 

700 Universe Boulevard 

P. O. Box 14000 

Juno Beach, FL 33408-0420 

Surry Power Station 

Virginia Electric & Power Co. 

Docket Nos. 50-280 and 50-281 

License Nos. DPR-32 and DPR-37 

Mr. David A. Heacock 

President and Chief Nuclear Officer 

Dominion Nuclear 

Innsbrook Technical Center 

5000 Dominion Boulevard 

Glen Allen, VA 23060-6711 
15 -

Susquehanna Steam Electric Station 

PPL Susquehanna, LLC 

Docket Nos. 50-387 and 50-388 

license Nos. NPF-14 and NPF-22 

Mr. Timothy S. Rausch 

Senior Vice President and Chief Nuclear Officer 

PPL Susquehanna, LLC 

769 Salem Boulevard 

NUCSB3 

Berwick, PA 18603-0467 

Turkey Point 

Florida Power & Light Co. 

Docket Nos. 50-250 and 50-251 

License Nos. DPR-31 and DPR-41 

Mr. Mano Nazar 

Executive Vice President and Chief Nuclear Officer 

NextEra Energy 

700 Universe Boulevard 

P. O. Box 14000 

Juno Beach, FL 33408-0420 

Vermont Yankee Nuclear Power Station 

Entergy Nuclear Operations, Inc. 

Docket No. 50-271 

License No. DPR-28 

Mr. Christopher J. Wamser 

Site Vice President 

Entergy Nuclear Operations, Inc. 

Vermont Yankee Nuclear Power Station 

320 Governor Hunt Road 

Vernon, VT 05354 

Virgil C. Summer Nuclear Station 

South Carolina Electric & Gas Co. 

Docket No. 50-395 

License No. NPF-12 

Mr. Thomas D. Gatlin 

Vice President Nuclear Operations 

South Carolina Electric & Gas Company 

Virgil C. Summer Nuclear Station 

Post Office Box 88, Mail Code 300 

Jenkinsville, SC 29065 
- 16 -

Vogtle Electric Generating Plant 

Southern Nuclear Operating Co. 

Docket Nos. 50-424 and 50-425 

License Nos. NPF-68 and NPF-81 

Mr. Tom E. Tynan 

Vice President 

Southern Nuclear Operating Company, Inc. 

Vogtle Electric Generating Plant 

7821 River Road 

Waynesboro, GA 30830 

Vogtle Electric Generating Plant, Units 3 & 4 

Southern Nuclear Operating Co. 

Docket Nos. 52-025 and 52-026 

License Nos. NPF-91 and NPF-92 

Mr. B. L. Ivey 

Vice President, Regulatory Affairs 

Southern Nuclear Operating Company, Inc. 

40 Inverness Center Parkway 

Bin B022 

Birmingham, AL 35242 

Waterford Steam Electric Station 

Entergy Operations, Inc. 

Docket No. 50-382 

License No. NPF-38 

Ms. Donna Jacobs 

Vice President, Operations 

Entergy Operations, Inc. 

Waterford Steam Electric Station, Unit 3 

17265 River Road 

Killona, LA 70057-0751 

Watts Bar Nuclear Plant, Unit 1 

Tennessee Valley Authority 

Docket No. 50-390 

License No. NPF-90 

Mr. Preston D. Swafford 

Chief Nuclear Officer and Executive Vice President 

Tennessee Valley Authority 

3R Lookout Place 

1101 Market Street 

Chattanooga, TN 37402-2801 
-17 ­

Watts Bar Nuclear Plant. Unit 2 

Tennessee Valley Authority 

Docket No. 50-391 

Construction Permit No. CPPR No. 092 

Mr. Michael D. Skaggs 

Senior Vice President, Nuclear Generation Development and Construction 

Tennessee Valley Authority 

6A Lookout Place 

11 01 Market Street 

Chattanooga, TN 37402-2801 

William B. McGuire Nuclear Station 

Duke Energy Carolinas, LLC 

Docket Nos. 50-369 and 50-370 

License Nos. NPF-9 and NPF-17 

Mr. Regis T. Repko 

Vice President 

Duke Energy Carolinas, LLC 

McGuire Nuclear Site 

12700 Hagers Ferry Road 

Huntersville, NC 28078 

Wolf Creek Generating Station 

Wolf Creek Nuclear Operating Corp. 

Docket No. 50-482 

License No. NPF-42 

Mr. Matthew W. Sunseri 

President and Chief Executive Officer 

Wolf Creek Nuclear Operating Corporation 

P. O. Box 411 

Burlington, KS 66839 
- 6 ­

If you have any questions on this matter, please contact your NRC licensing Project Manager. 

Sincerely, 

Ira! 

Eric J. Leeds, Director 

Office of Nuclear Reactor Regulation 

Michael R. Johnson, Director 

Office of New Reactors 

Enclosures: 

1. Recommendation 2.1: Seismic 

2. Recommendation 2.1: Flooding 

3. Recommendation 2.3: Seismic 

4. Recommendation 2.3: Flooding 

5. Recommendation 9.3: EP 

6. Licensees and Holders of Construction Permits 

cc: Listserv 

Distribution: See next page 

ADAMS Accession Nos.: 

OFFICE PM: NRR/JLD 

NAME GEMilier SRohrer 

DATE 03/121 03/12/2012 

BC:N OD: NRO 

DSkeen MS eneer MJohnson ELeeds 

03/11/2012 03/11/2012 03/12/2012 03/12/2012 

OFFICIAL RECORD COPY