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

U.S. NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REGULATORY RESEARCH

November 2015
Division 1

DRAFT REGULATORY GUIDE
Technical Lead
Eric Bowman

DRAFT REGULATORY GUIDE DG-1301
(Proposed New Regulatory Guide 1.226)

FLEXIBLE MITIGATION STRATEGIES FOR
BEYOND-DESIGN-BASIS EVENTS
A. INTRODUCTION
Purpose
This regulatory guide (RG) identifies methods and procedures the staff of the U.S. Nuclear
Regulatory Commission (NRC) considers acceptable for nuclear power reactor applicants and licensees to
demonstrate compliance with NRC regulations covering integrated planning and preparedness for
beyond-design-basis events as required by U.S. Code of Federal Regulations, Title 10, “Energy,” Part 50,
“Domestic Licensing of Production and Utilization Facilities,” (10 CFR 50) (Ref. 1), Section 50.155,
“Mitigation of Beyond-Design-Basis Events,” (10 CFR 50.155).
This RG endorses, with clarifications, the methods and procedures promulgated by the Nuclear
Energy Institute (NEI) in technical document NEI 12-06, “Diverse and Flexible Coping Strategies
(FLEX) Implementation Guide,” Revision 1A (NEI 12-06, Rev. 1A) dated October, 2015 (Ref. 2) as a
process the NRC considers acceptable for meeting, in part, the regulations in 10 CFR 50.155.
Additionally, this RG provides guidance in areas that are not covered in NEI 12-06, for meeting the
regulations in 10 CFR 50.155.
Applicable Orders and Regulations
•

NRC Order EA-12-049, “Order Modifying Licenses with Regard to Requirements for Mitigation
Strategies for Beyond-Design-Basis External Events,” was issued March 12, 2012 (Ref. 3). This
order requires nuclear power reactor licensees and construction permit holders to develop,
implement, and maintain strategies to maintain or restore core cooling, spent fuel pool (SFP)
cooling, and containment capabilities following a beyond-design-basis external event (BDBEE).

This regulatory guide is being issued in draft form to involve the public in the early stages of the development of a staff guidance
position in this area. It has not received final NRC review or approval and does not represent an official NRC final position.
Public comments are being solicited on this draft guide (including any implementation schedule) and its associated regulatory
analysis or value/impact statement. Comments should be accompanied by appropriate supporting data. Written comments may be
submitted to the Rules, Announcements, and Directives Branch, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; submitted through the NRC’s interactive rulemaking Web page at http://www.nrc.gov; or faxed to
(301) 492-3446. Copies of comments received may be examined at the NRC’s Public Document Room, 11555 Rockville Pike,
Rockville, MD. Comments would be most helpful if received within 90 days after publication of the request for comment in the
Federal Register.
Electronic copies of this draft regulatory guide are available through the NRC’s interactive rulemaking Web page (see above); the
NRC’s public Web site under Draft Regulatory Guides in the Regulatory Guides document collection of the NRC Library at
http://www.nrc.gov/reading-rm/doc-collections/; and the NRC’s Agencywide Documents Access and Management System
(ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML13168A031. The regulatory analysis may be
found in ADAMS under Accession No. ML15049A212.

•

10 CFR 50.155, “Mitigation of Beyond-Design-Basis Events,” requires nuclear power reactor
licensees to develop, implement, and maintain an integrated response capability that includes
strategies and guidelines to mitigate a BDBEE.

Related Guidance
•

JLD-ISG-2012-01, “Compliance with Order EA-12-049, Order Modifying Licenses with Regard
to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events,” Rev. 0,
was issued August 29, 2012 (Ref. 4). This interim staff guidance (ISG) endorses, with
clarifications, the methodologies described in NEI 12-06, Rev. 0 (Ref. 5) as one acceptable
method of demonstrating compliance with NRC Order EA-12-049. JLD-ISG-2012-01 is
superseded and replaced by this RG.

Purpose of Regulatory Guides
The NRC issues RGs to describe to the public methods that the NRC considers acceptable for use
in implementing specific parts of the agency’s regulations, to explain techniques that the NRC uses in
evaluating specific problems or postulated accidents, and to provide guidance to applicants. Regulatory
guides are not substitutes for regulations and compliance with them is not required. Methods and
solutions that differ from those set forth in RGs will be deemed acceptable if they provide a basis for the
findings required for the issuance or continuance of a permit or license by the Commission.
Paperwork Reduction Act
This RG contains information collection requirements covered by 10 CFR 50 and 10 CFR Part
52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” (Ref. 6) that the Office of
Management and Budget (OMB) approved under OMB control numbers 3150-0011 and 3150-151
respectively. The NRC may neither conduct nor sponsor, and a person is not required to respond to, an
information collection request or requirement unless the requesting document displays a currently valid
OMB control number.
List of Abbreviations
The following abbreviations are used in this RG
ac
ADAMS
AFW
AMS
ASCE
BDBEE
CFR
DG
EFW
ELAP
EOP
EPRI
ESBWR
ESEP

alternating current
NRC Agencywide Documents Access and Management System
auxiliary feedwater
alternate mitigating strategy
American Society of Civil Engineers
beyond-design-basis external event
U.S. Code of Federal Regulations
draft regulatory guide
emergency feedwater
extended loss of alternating current power
emergency operating procedure
Electric Power Research Institute
economic simplified boiling-water reactor
expedited seismic evaluation process
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FLEX
FR
FSAR
FSG
GL
GMRS
HF
HPCI
IAEA
IHS
IPEEE
ISG
LLNL
LUHS
NEA
NEI
NRC
NTTF
OECD
PGA
RCIC
RG
RLE
RLGM
SAMGs
SEI
SEL
SFP
SMA
SPID
SRM
SPRA
SSC
THMS

diverse and flexible coping strategies
federal register
final safety analysis report
FLEX support guidelines
generic letter
ground motion response spectrum
high frequency
high pressure core injection
International Atomic Energy Agency
IPEEE high-confidence-of-low-probability-of-failure spectrum
individual plant examination of external events
interim staff guidance
Lawrence Livermore National Laboratory
loss of normal access to the ultimate heat sink
Nuclear Energy Agency
Nuclear Energy Institute
U.S. Nuclear Regulatory Commission
Near-Term Task Force
Organization for Economic Co-operation and Development
peak ground acceleration
reactor core isolation cooling
regulatory guide
review level earthquake
review level ground motion
severe accident management guidelines
Structural Engineering Institute
seismic equipment list
spent fuel pool
seismic margin assessment
screening, prioritization, and implementation details
staff requirement memorandum
seismic probabilistic risk assessment
structure, system, and component
targeted hazard mitigating strategy

B. DISCUSSION
Reason for Issuance
One of the primary lessons learned from the events at Fukushima Dai-ichi was the significance of
the challenge presented by a loss of safety-related systems following the occurrence of a BDBEE. In the
case of Fukushima Dai-ichi, the extended loss of alternating current power (ELAP) led to loss of core
cooling and core damage including a loss of containment integrity. The design basis for U.S. nuclear
plants includes bounding analyses with margin for external events expected at each site. Extreme external

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events (e.g., seismic events, external flooding, etc.) beyond those accounted for in the design basis are
highly unlikely but could present challenges to nuclear power plants.
As one method of addressing these challenges, this RG endorses, with clarifications as detailed in
this RG, the principles and processes in NEI 12-06, Rev. 1A, as acceptable for use by applicants and
licensees to define and deploy strategies that will enhance their ability to cope with conditions resulting
from BDBEEs.
Background
Following the March 11, 2011 events at the Fukushima Dai-ichi nuclear power plant, the NRC
established a senior-level agency task force referred to as the Near-Term Task Force (NTTF). The NTTF
conducted a systematic and methodical review of the NRC regulations and processes and determine if the
agency should make additional improvements in NRC regulations or processes in light of the events at
Fukushima Dai-ichi. As a result of this review, the NTTF developed a comprehensive set of
recommendations, documented in SECY-11-0093, “Near-Term Report and Recommendations for Agency
Actions Following the Events in Japan,” dated July 12, 2011 (Ref. 7). The Commission then directed the
NRC staff in staff requirement memorandum (SRM) SRM-SECY-11-0093 (Ref. 8) to identify any actions
that could, and in the staff’s judgment should, be taken in the near term given consideration to the wide
range of regulatory tools available. The staff’s response to this Commission direction is contained in
SECY-11-0124, “Recommended Actions to be Taken without Delay from the Near-Term Task Force
Report,” dated September 9, 2011(Ref. 9). In SRM-SECY-11-0093, the Commission further directed that
all the regulatory actions in the report should be prioritized, and SECY-11-0137, “Prioritization of
Recommended Actions to be Taken in Response to Fukushima Lessons Learned,” dated October 3, 2012
(Ref. 10) provides the staff’s response to this direction.
After receiving the Commission’s direction in SRM-SECY-11-0124 (Ref. 11) and SRM-SECY11-0137 (Ref. 12), the NRC conducted public meetings to discuss enhanced mitigation strategies intended
to maintain or restore core cooling, containment, and SFP cooling capabilities following a BDBEE. At
these meetings, the industry described its proposal for a diverse and flexible mitigation capability
(FLEX), as documented in NEI’s letter, dated December 16, 2011 (Ref. 13). FLEX was proposed as a
strategy to fulfill the key safety functions of core cooling, containment integrity, and spent fuel cooling.
Stakeholder input influenced the NRC to pursue a performance-based approach to improve the safety of
operating power reactors different than envisioned in NTTF Recommendation 4.2, SECY-11-0124, and
SECY-11-0137.
On February 17, 2012, the NRC staff provided SECY-12-0025, “Proposed Orders and Requests
for Information in Response to Lessons Learned from Japan’s March 11, 2011, Great Tohoku Earthquake
and Tsunami” (Ref. 14) to the Commission, including the proposed order to implement the enhanced
mitigation strategies. As directed by SRM-SECY-12-0025 (Ref. 15), the NRC issued Order EA-12-049.
On March 30, 2012, the Commission issued Memorandum and Order CLI-12-09 (Ref. 16), which
included the requirements for mitigation strategies as a license condition for Virgil C. Summer Nuclear
Station, Units 2 and 3. These requirements were subsequently included as license condition 2.D.(13) on
both combined license NPF-93 and combined license NPF-94 for those units.
On May 4, 2012, NEI submitted NEI 12-06, Rev. B (Ref. 17), to provide specifications for an
industry developed method for the development, implementation, and maintenance of guidance and
strategies in response to the Mitigating Strategies Order. On May 13, 2012, NEI submitted NEI 12-06,
Rev. B1 (Ref. 18). The strategies and guidance described in NEI 12-06, Rev. 1A expand on those
developed and implemented by the nuclear industry to address the limited set of BDBEEs involving the
DG-1301, Page 4

loss of a large area of the plant due to explosions and fire required pursuant to paragraph (hh)(2) of 10
CFR 50.54, “Conditions of licenses.”
On May 31, 2012, the NRC issued a draft version of an interim staff guidance, JLD-ISG-2012-01,
and published a notice of its availability for public comment in the Federal Register (77 FR 33779), with
the 30 day comment period running through July 7, 2012. The NRC received seven comments during this
time, with the NRC addressing the comments as documented in “NRC Response to Public Comments,
JLD-ISG-2012-01 (Docket ID NRC-2012-0068)” (Ref. 19).
On July 3, 2012, NEI submitted Rev. C to NEI 12-06 (Ref. 20), incorporating many of the
exceptions and clarifications included in the draft version of JLD-ISG-2012-01. On August 3, 2012, NEI
submitted Draft Rev. 0 to NEI 12-06 incorporating many of the remaining exceptions and clarifications.
On August 21, 2012, NEI submitted Rev. 0 to NEI 12-06, making various editorial corrections. The NRC
reviewed the August 21, 2012 submittal of Rev. 0 of NEI 12-06 dated August 2012 and endorsed it in
JLD-ISG-2012-01 as a process the NRC considers acceptable for meeting the regulatory requirements
with noted clarifications.
On August 25, 2015, NEI submitted Rev. 1 to NEI 12-06 (Ref. 21), incorporating lessons learned
in the implementation of Order EA-12-049 and alternative approaches taken by licensees for compliance
to that order. Following a public webinar discussion of potential exceptions and clarifications on
September 21, 2015, NEI submitted Rev. 1A to NEI 12-06.
The NRC is issuing 10 CFR 50.155 to, among other things, make the requirements of Order EA12-049 generically applicable, taking into account lessons learned during the implementation of the orders
and input from stakeholders. This RG endorses, with clarifications, NEI 12-06, Rev. 1A as an acceptable
method for applicants and licensees to demonstrate compliance, in part, with the regulatory requirements.
NEI 12-06, Rev. 1A was developed by NEI to incorporate lessons learned and additional alternative
approaches to meet the requirements of Order EA-12-049. The guidelines in NEI 12-06, Rev. 1A
recommend a three-phase approach for mitigating BDBEEs. The initial phase makes use of installed
equipment and resources to maintain or restore key safety functions including core cooling, containment,
and SFP cooling. The transition phase includes providing sufficient, portable, onsite equipment and
consumables to maintain or restore these functions until they can be accomplished with resources brought
from offsite. The final phase includes obtaining sufficient offsite resources to sustain these functions
indefinitely.
External Documents Endorsed in This Guide
This RG endorses, in part, the use of one or more codes, standards, or guidance documents
developed by external organizations. These codes, standards, and third party guidance documents may
contain references to other codes, standards, or third party guidance documents (“secondary references”).
If a secondary reference has itself been incorporated by reference into NRC regulations as a requirement,
then licensees and applicants must comply with that standard as set forth in the regulation. If the
secondary reference has been endorsed in an RG as an acceptable approach for meeting an NRC
requirement, then the standard constitutes a method acceptable to the NRC for meeting that regulatory
requirement as described in the specific RG. If the secondary reference has neither been incorporated into
NRC regulations nor endorsed in an RG, the secondary reference is neither a legally-binding requirement
nor a “generic” NRC approved acceptable approach for meeting an NRC requirement. However, licensees
and applicants may consider and use the information in the secondary reference, if appropriately justified,
consistent with current regulatory practice, and consistent with applicable NRC requirements.

DG-1301, Page 5

Harmonization with International Standards
The International Atomic Energy Agency (IAEA) has established a series of technical reports,
safety guides and standards constituting a high level of safety for protecting people and the environment.
IAEA guides present international good practices and identify best practices to help users striving to
achieve high levels of safety. This RG and the NEI technical document endorsed by it contain guidance
about BDBEE mitigation similar to guidance under revision by the IAEA.

C. STAFF REGULATORY GUIDANCE
This RG endorses, with clarifications, the methods described in NEI 12-06, Rev. 1A, dated
October 2015. The NRC staff has determined that the methods described in the NEI 12-06, Rev. 1A
constitute procedures and processes generally acceptable to the NRC for demonstrating compliance with
the regulatory requirements in 10 CFR 50.155 subject to the following clarifications.
1.

Development and Implementation Process
10 CFR 50.155(b) and (b)(1) require that applicants or licensees develop and implement an
integrated response capability that includes strategies and guidelines to mitigate beyond-designbasis external events from natural phenomena that result in an ELAP concurrent with either a loss
of normal access to the ultimate heat sink (LUHS) or, for nuclear power plants with passive
reactor designs, a loss of normal access to the normal heat sink. The strategies and guidelines
developed and implemented under those sections must be capable of being implemented site-wide
and must include maintaining or restoring core cooling, containment, and spent fuel pool cooling
capabilities; and the acquisition and use of offsite assistance and resources to support those
functions.

1.1.

Establishment of Baseline Coping Capability
Section 1.3 of NEI 12-06, Rev. 1A, discusses the objectives and guiding principles of the FLEX
program that are responsive to 10 CFR 50.155(b)(1). These principles retain the three-phase
approach that had been required under Order EA-12-049 and provide that plant-specific analyses
will determine the duration of each phase.
Section 2 of NEI 12-06, Rev. 1A, provides a high-level discussion of the site-specific nature of
the actions required by each licensee to properly implement the performance-based requirements
in the regulations. Sections 2.1 through 2.5 of NEI 12-06, Rev. 1A discuss the coping capacities,
types of external hazards, strategies, and controls each licensee should implement to meet the
requirements in the regulations.
Section 3 of NEI 12-06, Rev. 1A provides performance attributes, general criteria and baseline
assumptions for use in the development and implementation of the strategies and guidelines under
10 CFR 50.155(b)(1). NEI 12-06, Rev. 1A further provides that licensees should use these criteria
and assumptions for analyses used to establish a baseline coping capability. The assumptions
include the initial conditions listed in section 3.2.1.3 that include a loss of offsite power affecting
all units at a plant site and the specification that “[a]ll design basis installed sources of emergency
on-site ac power and SBO alternate ac power sources [as defined in 10 CFR 50.2] are assumed to
be not available and not imminently recoverable.”

DG-1301, Page 6

NEI 12-06, Rev. 1A specifies in section 3.2.1.7 that “[s]trategies that have a time constraint to be
successful should be identified and a basis provided that the time can reasonably be met.” NEI
12-06, Rev. 1A specifies in section 11.4.3 that FLEX support guidelines (FSGs) will be
developed to provide guidance that can be employed for a variety of conditions and that the FSGs
will be reviewed and validated to ensure they are feasible. NEI 12-06, Rev. 1A, Appendix E
provides a method for validation of the FSGs.
NEI 12-06, Rev. 1A, Section 3.2.1.13 specifies that best-estimate analyses are appropriate for the
purpose of establishing the baseline coping capabilities.
Staff Position: Sections 1, 2 and 3 and Appendix E of NEI 12-06, Rev. 1A provide an acceptable
method for licensees to follow to develop a baseline coping capability for mitigating an ELAP
concurrent with either an LUHS or, for nuclear power plants with passive reactor designs, a loss
of normal access to the normal heat sink with the following clarifications:
a)

It should be noted that the initial and boundary conditions described do not accurately
reflect a loss of all ac power condition due to the limitation of initial conditions (1) and
(2) of NEI 12-06, Rev. 1A, Section 3.2.1.3. The additional contingencies described in
section 1.2 of this document are necessary for compliance with the requirement to
mitigate a loss of all ac power.

b)

An element of a set of strategies to maintain or restore core and SFP cooling and
containment functions includes knowledge of the time a licensee or applicant can
withstand challenges to these key safety functions using installed equipment during a
BDBEE. This knowledge provides an input to the choice of storage locations and
conditions of readiness of the equipment required for the follow-on phase. This duration
is related to, but distinct from the specified duration for the requirements of 10 CFR
50.63, “Loss of All Alternating Current Power,” paragraph (a), because it represents the
current capabilities of the licensee or applicant rather than a required capability and
licensees and applicants should 1) account for the SFP cooling function, which is not
addressed by 10 CFR 50.63(a), and 2) assume the non-availability of alternate ac sources,
which may be included in meeting the specified durations of 10 CFR 50.63(a). This is
implicit in the NEI 12-06, Rev. 1A principles described in Section 3.2.1.7, Paragraph (6)
and Section 3.2.2, Paragraph (1). However, maintenance of the guidance and strategies
requires that the estimate of capability be kept current to reflect plant conditions
following facility changes such as modifications or equipment outages. Changes in the
facility can impact the duration for which the initial response phase can be accomplished,
the required initiation times for the transition phase, and the required delivery and
initiating times for the final phase.

c)

The use of best-estimate analyses for establishing the baseline coping capabilities is
appropriate in the context of the beyond-design-basis external events for 10 CFR
50.155(b)(1). This includes the use of normal fluid levels for tanks that are maintained by
procedure or administrative controls rather than the minimum levels allowed by
Technical Specifications.

d)

Consistent with the goal of mitigation strategies for BDBEE, the validation method
documented in Appendix E of NEI 12-06, Rev. 1A is endorsed as a method to (1) assess
whether it is feasible, considering design basis, or reevaluated hazard conditions
determined under the § 50.54(f) request for information of March 12, 2012 (as
applicable), to execute tasks, manual actions and decisions (i.e., human actions) required
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by the mitigation strategies described in NEI 12-06, Rev. 1A and (2) support a conclusion
that the strategies mitigate, to the extent practical, the adverse effects of BDBEEs on the
ability of personnel to perform the required human actions. NEI 12006, Rev. 1A,
Appendix E neither proposes nor is endorsed as a method to assess whether required
human actions are reliable.

1.1.1.

1.

The use of Level C validation methods should be limited to those tasks, manual
actions and decisions that do not have a time constraint for the strategy to be
successful. This is because the Level C validation methods do not result in an
estimate of the time necessary to perform the tasks, manual action, or decision
and cannot provide a basis that a time constraint can reasonably be met. Tasks,
manual actions, or decisions that have time constraints may be validated using a
Level A or Level B method that results in an estimate of the time required to
complete the task or manual action or to make and communicate the decision in
order to confirm that the time constraint can reasonably be met as specified in
NEI 12-06, Rev. 1A, Section 3.2.1.7, principle 6, which states that “[s]trategies
that have a time constraint to be successful should be identified and a basis
provided that the time can reasonably be met.”

2.

Consistent with NEI 12-01, “Guideline for Assessing Beyond Design Basis
Accident Response Staffing and Communications Capabilities” (Ref. 22) as
endorsed by NRC letter dated May 15, 2012 (Ref. 23), additional staff and
resources may be assumed to be available commencing at the 6 hour point (or
less, with justification). As a result, the use of Level B validation techniques,
which are less stringent, may typically be substituted for the use of Level A
validation techniques at the 6 hour point. However, prior to substituting Level B
validation, licensees should confirm that staff augmentation will improve the
capability of plant personnel to complete tasks, manual actions, and decisions.
Level B validation should not be substituted for tasks, manual actions, and
decisions that have time constraints and that cannot performed more effectively
or efficiently through addition of personnel. In addition, the results of the
integrated review should be used to assess the need to adjust and revalidate tasks,
manual actions, or decisions for which the validation did not provide reasonable
confidence in the ability of plant personnel to execute a required task, manual
action, or decision.

Phased Approach

The regulations in 10 CFR 50.155 do not contain specific requirements for a multiple phase
approach to mitigating and recovering from a BDBEE as had been the case under Order
EA-12-049. NEI 12-06, Rev. 1A, carries the definitions of the phases from that order forward as a
conceptual framework for the development of the FLEX strategies. Maintenance of core and SFP
cooling and containment functions requires overlap between the initiating times for the phases
with the duration for which each licensee can perform the prior phases. The NRC recognizes that
for certain BDBEEs, the damage state could prevent maintenance of key safety functions using
the equipment intended for particular phases. Under such circumstances, prompt initiation of the
follow-on phases to restore core and SFP cooling and containment functions is appropriate.
Staff Position: NEI 12-06, Rev. 1A provides an acceptable method for developing an approach to
mitigate and cope with BDBEEs.
DG-1301, Page 8

1.1.1.1. Initial Response Phase
The initial response phase will be accomplished using installed equipment. Licensees and
applicants should establish and maintain current estimates of their capabilities to maintain core
and SFP cooling and containment functions assuming a loss of all alternating current (ac) electric
power to the essential and nonessential switchgear buses except for those fed by station batteries
through inverters. These estimates provide the time period in which the licensee should be able to
initiate the transition phase and maintain or restore the key safety functions using portable on-site
equipment. These estimates should be considered in selecting the storage locations for that
equipment and the prioritization of resources to initiate their use.
Staff Position: NEI 12-06, Rev. 1A, Section 3.0, provides an acceptable method for determining
the baseline coping capabilities for the initial response phase.
1.1.1.2. Transition Phase
The transition phase will be accomplished by supplementing the use of installed equipment with
portable equipment stored on-site. The strategies for this phase must be capable of maintaining
core cooling, containment, and SFP cooling capabilities (following their restoration, if applicable)
from the time they are implemented until they can be supplemented by offsite resources in the
final phase. The duration of the transition phase should provide sufficient overlap with both the
initial and final phases to account for the time it takes to install equipment and for uncertainties.
Staff Position: NEI 12-06, Rev. 1A, Section 3.0, provides an acceptable method for determining
the baseline coping capabilities for the transition phase.
1.1.1.3. Final Phase
The final phase will be accomplished using the portable equipment stored on-site augmented with
additional equipment and consumables obtained from off-site until power, water, and coolant
injection systems are restored or commissioned.
Staff Position: NEI 12-06, Rev. 1A, Section 3.0, provides an acceptable method for determining
the baseline coping capabilities for the final phase. NEI 12-06, Rev. 1A, Section 12.2, provides an
acceptable method for establishing the capability to obtain equipment and consumables from offsite until power, water, and coolant injection systems are restored or commissioned.
1.2.

Contingencies for Loss of All Alternating Current Power
NEI 12-06, Rev. 1A, Section 3.2.2 provides 17 guidelines for use in the development of the
guidance and strategies under 10 CFR 50.155(b)(1). Guideline (2) of this sections states:
“Plant procedures/guidance should recognize the importance of
AFW/HPCI/RCIC/IC during the early stages of the event and direct the operators
to invest appropriate attention to assuring its initiation and continued, reliable
operation throughout the transient since this ensures decay heat removal.”
The risk of core damage due to ELAP can be significantly reduced by assuring the
availability of auxiliary feedwater (AFW) (emergency feedwater (EFW) at some plants),
high pressure core injection (HPCI), reactor core isolation cooling (RCIC), or isolation
condensers (IC), particularly in the first 30 minutes to one hour of the event. Assuring
DG-1301, Page 9

that one of these systems has been initiated to provide early core heat removal, even if
local initiation and control is required is an important initial action. A substantial portion
of the decay and sensible reactor heat can be removed during this period. The availability
of AFW/HPCI/RCIC/IC can be improved by providing a reliable supply of water,
monitoring turbine conditions (particularly lubricating oil flow and temperature),
bypassing automatic trips, and maintaining nuclear boiler/steam generator water levels.
These actions help ensure that the core remains adequately covered and cooled during an
extended loss of ac power event.
Appendices C and D of NEI 12-06, Rev. 1A contain summaries of performance attributes for
boiling-water and pressurized-water reactors respectively, address guideline (2) of NEI 12-06,
Rev. 1A, Section 3.2.2 by specifying that procedures/guidance will include local manual initiation
of AFW/EFW/HPCI/RCIC/IC.
NEI 12-06, Rev. 1A, Section 5.3.3, describes interface considerations for seismic events, expands
on this contingency to specify that the strategies and guidelines should include:
“…a reference source for the plant operators that provides approaches to
obtaining necessary instrument readings to support the implementation of the
coping strategy. Such a resource could be provided as an attachment to the plant
procedures/guidance. Guidance should include critical actions to perform until
alternate indications can be connected and on how to control critical equipment
without associated control power.”
“This reference source should include control room and non-control room
readouts and should also provide guidance on how and where to measure key
instrument readings using a portable instrument (e.g., a Fluke meter) at a location
that does not rely on the functioning of intervening electrical equipment (e.g. I/E
convertors, analog to digital converters, relays, etc.) that could be adversely
affected by BDB seismic events. An instrument reading should be obtained at the
closest accessible termination point to the containment penetration or parameter
of measurement, as practical.”
Staff Position: NEI 12-06, Rev. 1A, Section 3.2.2, guideline (2) and the provisions in NEI 12-06,
Rev. 1A, Appendices C and D, for manual initiation of AFW/EFW/HPCI/RCIC/IC coupled with
the NEI 12-06, Rev. 1A, Section 5.3.3 provisions for the development of guidance on obtaining
instrument readings and controlling critical equipment without the associated power provide an
acceptable method for licensees to develop the contingencies for the loss of all ac power that are
necessary to comply with the 10 CFR 50.155(b)(1) requirement to mitigate an extended loss of all
ac power. The need for the NEI 12-06, Rev. 1A, Section 5.3.3 contingencies to show compliance
with the § 50.155(b)(1) condition of loss of all ac power is not limited to seismically-induced
events; it is a necessary element of compliance for that requirement regardless of the initiating
event. Because NEI 12-06, Rev. 1A, Section 5 is applicable to all power reactor licensees,
conformance to NEI 12-06, Rev. 1A, Section 5.3.3 can provide the capabilities necessary to meet
that element regardless of the initiating event.
2.

Equipment Capacity and Capability
10 CFR 50.155(c)(1) requires that the equipment relied upon for the mitigation strategies required
by § 50.155(b)(1) have sufficient capacity and capability to simultaneously maintain or restore
DG-1301, Page 10

core cooling, containment, and SFP cooling capabilities for all the power reactor units within the
site boundary.
NEI 12-06, Rev. 1A, Section 3.2.2, Guideline (16), provides guidance for the minimum number
of sets of equipment a licensee should provide in order to achieve reasonable assurance that the
equipment will be available in sufficient quantity to have the capacity and capability necessary to
comply with § 50.155(c)(1). This includes guidance for the provision of spare hoses and cables
in a quantity that is either (1) equivalent to 10% of the total length of each type of hose or cable
necessary; or (2) of sufficient length and sizing to replace the single longest run needed to support
any single strategy.
NEI 12-06, Rev. 1A, Sections 11.1 and 11.2 provide guidance on the quality attributes and
equipment design a licensee may use to achieve reasonable assurance that the individual pieces of
equipment have the capability to perform the functions they are intended for in the FLEX
strategies.
Staff Position: NEI 12-06, Rev. 1A, Section 3.2.2, Guideline (16) and Sections 11.1-2, provide an
acceptable method to demonstrate compliance with § 50.155(c)(1).
3.

Reasonable Protection
10 CFR 50.155(c)(2) requires that the equipment relied upon for the mitigation strategies required
by § 50.155(b)(1) be reasonably protected from the effects of natural phenomena.
NEI 12-06, Rev. 1A, Appendix A, defines reasonable protection as… “Storing on-site FLEX
equipment in configurations such that no one external event can reasonably fail the site FLEX
capability (N) when the required FLEX equipment is available.”
Staff Position: NEI 12-06, Rev. 1A, provides an acceptable approach for reasonably protecting
equipment from the effects of natural phenomena. This approach includes the following:
•

Identification of the natural phenomena for which reasonable protection is necessary,

•

Determination of the method of protection to be used,

•

Establishment of controls on unavailability of the equipment, and

•

Provision of a method of transporting the portable equipment from its storage location to
the site in which it will be used.

Individual elements of reasonable protection are discussed below.
3.1.

Evaluation of External Hazards
Section 4 of NEI 12-06, Rev. 1A, discusses the overall methodology for identifying external
hazards and evaluating their impact. Appendix B of NEI 12-06, Rev. 1A discusses the
identification of external hazards for which licensees should provide reasonable protection. NEI
12-06, Rev. 1A, Sections 5 through 9, discuss the evaluation of the effects of natural phenomena
to meet the baseline coping capability.

DG-1301, Page 11

Staff Position: Sections 5 through 9 and Appendix B of NEI 12-06, Rev. 1A, provide an
acceptable method for the evaluation and equipment considerations to address the effects of
external hazards in order to satisfy that element of reasonable protection.
3.2.

Protection from External Hazards
Sections 5 through 9 of NEI 12-06, Rev. 1A discuss methodologies for the protection of the
equipment. The methods of protection comprise: 1) physical protection of the equipment; 2)
protection by relocation of the equipment from a position in which a licensee may have indication
of an impending hazard; and 3) provision of multiple, redundant pieces of equipment or methods
to accomplish a function, stored in diverse locations in order to provide assurance that at least one
method of accomplishing that function will survive an event of a localized nature such as a
tornado missile impact.
Staff Position: Sections 5 through 9 and Appendix B of NEI 12-06, Rev. 1A provide an
acceptable method for protecting the equipment from the effects of external hazards in order to
satisfy that element of reasonable protection.

3.3.

Deployment of Equipment
Sections 5 through 9 of NEI 12-06, Rev. 1A discuss methods for transporting the equipment from
the location in which it is stored to the location in which it would be used. These sections
additionally discuss the connection of the equipment to structures, systems, and components
(SSCs) necessary for completion of the deployment of the equipment from storage to a state in
which it can supplement the functions of the installed SSCs.
Staff Position: Sections 5 through 9 and Appendix B of NEI 12-06, Rev. 1A provide an
acceptable method for deployment of the equipment in order to satisfy that element of reasonable
protection.

3.4.

Programmatic Controls for Unavailability
Section 11.5.3 of NEI 12-06, Rev. 1A discusses the programmatic controls for equipment and
connections between that equipment and permanently installed SSCs. These controls include
limited time periods in which the equipment and connection points may be unavailable for any
reason, with the duration of the acceptable time period being based on the ability of the licensee
to accomplish the intended function of the equipment by other means.
When a licensee is unable to accomplish the intended function of the equipment by other means,
unavailability durations are limited to periods comparable to those allowed by Technical
Specifications for safety-related SSCs with similar functions. (See, e.g., the completion times
allowed for restoration of turbine-driven auxiliary feedwater trains in limiting condition for
operation 3.7.5, “Auxiliary Feedwater (AFW) System,” of NUREG-1431, “Standard Technical
Specifications – Westinghouse Plants,” Rev. 4.0, Volume 1, “Specifications,” which range from
24 hours to 7 days. [Ref. 24])
When a licensee is able to accomplish the intended function of the equipment by other means
(i.e., the equipment is spare equipment beyond the minimum necessary to accomplish the
intended function), unavailability of the equipment is limited to 90 days based on a normal plant
work cycle of 12 weeks in order to avoid displacing maintenance actions for other safetysignificant equipment or SSCs.
DG-1301, Page 12

When a licensee is able to accomplish the intended function of the equipment by other means, but
that means is not protected from all possible effects of natural phenomena, unavailability of the
equipment is limited to 45 days based on a short-cycle work period of 6 weeks in order to avoid
displacing maintenance actions for other safety-significant equipment or SSCs.
Similar controls are applied to connection points for the equipment to installed SSCs.
Staff Position: Section 11.5.3 of NEI 12-06, Rev. 1A provides an acceptable method for
controlling unavailability of the equipment in order to satisfy that element of reasonable
protection.
4.

Equipment Maintenance
10 CFR 50.155(c)(3) requires that the equipment relied on for the mitigation strategies under
§ 50.155(b)(1) receive adequate maintenance such that it is capable of fulfilling its intended
function.
Section 11.5 of NEI 12-06, Rev. 1A discusses the maintenance and testing of the equipment.
Section 3.2.1.13 discusses the Electric Power Research Institute (EPRI) program developed for
maintenance of the equipment, which is documented in the EPRI technical report 3002000623,
“Applications Center: Preventive Maintenance Basis for FLEX Equipment – Project Overview
Report” (Ref. 25). The EPRI technical report 3002000623 was endorsed by NRC letter dated
October 7, 2013 (Ref. 26).
Staff Position: Sections 11.5 and 3.2.1.13 of NEI 12-06, Rev. 1A, provide an acceptable method
for maintaining the equipment relied on for the mitigation strategies under § 50.155(b)(1).

5.

Configuration Control
10 CFR 50.155(b) and (b)(1) require that applicants or licensees maintain an integrated response
capability that includes strategies and guidelines to mitigate beyond-design-basis external events
from natural phenomena that result in an ELAP concurrent with either an LUHS or, for nuclear
power plants with passive reactor designs, a loss of normal access to the normal heat sink.
10 CFR 50.155(f) allows licensees to make changes to the implementation of the requirements of
10 CFR 50.155 without NRC approval provided that the licensee performs an evaluation
demonstrating that 10 CFR 50.155 continues to be met prior to making the change.
Section 11.8 of NEI 12-06, Rev 1A discusses the configuration control of the strategies and
guidelines as well as the maintenance of an overall program document and record of changes.
Staff Position: Section 11.8 of NEI 12-06, Rev. 1A provides an acceptable method for
maintaining configuration control of an integrated response capability under 10 CFR 50.155(b)(1)
and (f).

6.

Treatment of Reevaluated Hazards under the Requests for Information of March 12, 2012
10 CFR 50.155(c)(2)(i) requires that each licensee that received the March 12, 2012, NRC letter
issued under § 50.54(f) concerning reevaluations of seismic and flooding hazard levels provide
DG-1301, Page 13

reasonable protection against that reevaluated seismic or flooding hazard(s) if it exceeds the
design basis of its facility.
6.1.

Treatment of Reevaluated Seismic Hazards
The following guidance establishes the necessary considerations to evaluate the equipment used
in the guidance and strategies required by 10 CFR 50.155(b)(1) with respect to the reevaluated
hazard as required under 10 CFR 50.155(c)(2)(i). Detailed implementation methods will be
developed as necessary.
Staff Position: Licensees with reevaluated seismic hazards that exceed the design basis of the
facility should demonstrate reasonable protection for the reevaluated seismic hazard of the
equipment used in the guidance and strategies required by 10 CFR 50.155(b)(1) as required under
10 CFR 50.155(c)(2)(i). Past seismic reevaluations, to the extent they meet the standards of
reasonable protection described here, can be used to meet the requirements.
a.

The demonstration of reasonable protection should address all phases of mitigation and
consider FLEX equipment as well as installed equipment or structures, systems, and
components of the facility relied upon in mitigating strategies under
10 CFR 50.155(b)(1).

b.

Licensees should evaluate the equipment and SSCs within the scope of the demonstration
considering all pertinent failure modes (both the individual and system level failure
modes) that could prevent the functional performance needed for the mitigating strategies
as discussed in provisions c through k below. This should include consideration of
seismic interactions, evaluation of soil related failure modes, and consequential failures.

c.

Licensees should evaluate structures containing cooling and makeup water, fuel, and
equipment relied on for the mitigating strategies under NEI 12-06, Rev. 1A, Sections
3.2.1.3.3 through 3.2.1.3.7, 3.2.2.5, and footnote 4 to Section 3.2.3 for robustness as
defined in NEI 12-06, Rev. 1A, Appendix A using the reevaluated seismic hazard rather
than the design basis seismic hazard. Deformation of the structures is acceptable so long
as they will remain functional (i.e., retain the fluids and allow access for deployment and
use of the equipment as well as support equipment functionality).

d.

Licensees should evaluate delivery systems for cooling and makeup water and fuel relied
on for the mitigating strategies under NEI 12-06, Rev. 1A, Sections 3.2.1.3.4, 3.2.1.3.10,
3.2.2.5 and 3.2.2.13 for robustness as defined in NEI 12-06, Rev. 1A, Appendix A using
the reevaluated seismic hazard rather than the design basis seismic hazard. Deformation
of the systems is acceptable so long as they will remain functional (i.e., retain the fluids
and allow their flow).

e.

Licensees should evaluate portable equipment and the means to move that equipment that
is stored as described in NEI 12-06, Rev. 1A, Section 5.3.1.1.a to confirm that securing
and protection from seismic interactions at the new seismic hazard level remains
acceptable. The structure housing the equipment should be evaluated to confirm it has
adequate seismic margin to protect the equipment at the reevaluated seismic hazard.

f.

Licensees should evaluate portable equipment and the means to move that equipment that
is stored as described in NEI 12-06, Rev. 1A, Section 5.3.1.1.b to confirm that securing
and protection from seismic interactions at the new seismic hazard level remains
DG-1301, Page 14

acceptable. The structure housing the equipment should be evaluated to confirm that,
while deformation of the structure is possible, it will protect the equipment at the
reevaluated seismic hazard so as to allow deployment of the equipment.

6.1.1.

g.

Licensees should evaluate portable equipment and the means to move that equipment that
is stored as described in NEI 12-06, Rev. 1A, Section 5.3.1.1.c to confirm that securing
and protection from seismic interactions at the reevaluated seismic hazard level remains
acceptable.

h.

Licensees should review routes for the deployment of equipment from a storage location
to its usage location to ensure no adverse impact is created by a seismic event at the
reevaluated hazard level, such as potential soil liquefaction.

i.

Licensees should evaluate locations for connection points described in NEI 12-06, Rev.
1A, Section 5.3.2.2 to confirm they only require access through structures that are
seismically robust for the reevaluated seismic hazard.

j.

Licensees should evaluate large internal flooding sources considered under NEI 12-06,
Rev 1A, Section 5.3.3.2 for robustness as defined in NEI 12-06, Rev 1A, Appendix A
using the reevaluated seismic hazard rather than the design basis seismic hazard.

k.

Licensees may consider equipment from off-site in diverse, redundant locations to be
protected from the reevaluated seismic hazard.

Case 1: Demonstration of Reasonable Protection for Exceedances Limited to High
Frequency
Staff Position: If the ground motion response spectrum (GMRS) for a licensee’s reevaluated
seismic hazard is fully bounded by the licensee’s design basis seismic hazard from 1 Hz to 10 Hz,
but exceeds the design basis seismic hazard above 10 Hz, the licensee should evaluate high
frequency (HF) sensitive in-plant SSCs relied upon for execution of the mitigating strategies
using the methodologies of the EPRI Report 3002004396, “High Frequency Program:
Application Guidance for Functional Confirmation and Fragility Evaluation,” (Ref. 27) consistent
with its endorsement by letter dated September 17, 2015, (Ref. 28).

6.1.2.

Alternate Mitigating Strategies
Staff Position: Development of an Alternate Mitigating Strategy (AMS) that provides a
capability to mitigate the BDBEE by mitigating or preventing an ELAP that would occur as a
result of the BDBEE through exhaustion of fuel for operating emergency power sources is an
acceptable method of compliance with 10 CFR 50.155(b)(1) when the hazard level for the AMS
is identified.
a.

Reevaluated seismic hazard should be used in place of the safe shutdown earthquake
(SSE) as described in Section 6.1 of this RG.

b.

The initial condition of NEI 12-06, Rev. 1A, Section 3.2.1.3.1 should be modified to
reflect the reevaluated seismic hazard as the initiating event for the associated AMS.
Timing of the loss of off-site power should reflect impact of the flooding mechanism on
the delivery of off-site power to the facility.
DG-1301, Page 15

c.

The initial condition of NEI 12-06, Rev. 1A, Section 3.2.1.3.2 should be modified to
reflect the availability of emergency on-site ac power sources and station blackout
alternate ac power sources unless and until they are rendered unavailable by the
reevaluated seismic hazard.

6.1.2.1. Case 1: Demonstration of Reasonable Protection Based upon the Individual Plant
Examination of External Events
Licensees completed the Individual Plant Examination of External Events (IPEEE) in the 1990s
under Generic Letter (GL) 88-20 Supplements 4 (Ref. 29) and 5 (Ref. 30) using the guidance of
NUREG-1407, “Procedural and Submittal Guidance for the Individual Plant Examination of
External Events (IPEEE) for Severe Accident Vulnerabilities,” (Ref. 31). Acceptable approaches
to perform the IPEEE included the NRC seismic margin assessment (SMA) method, the EPRI
SMA method described in EPRI NP-6041-SL, Rev. 1, “A Methodology for Assessment of
Nuclear Plant Seismic Margin,” (Ref. 32) or a seismic probabilistic risk assessment (SPRA). For
each approach, a seismic equipment list (SEL) was developed that included multiple redundant
safe shutdown success paths and/or accident sequences. Under NUREG-1407, plants performed
the seismic portion of the IPEEE in three categories, reduced scope, focused scope and full scope.
The seismic IPEEEs were generally performed using input motions based on the following:
a.

Median-centered response spectrum using the shape from NUREG/CR-0098,
“Development of Criteria for Seismic Review of Selected Nuclear Power Plants,”
(Ref. 33) anchored to 0.3g peak ground acceleration (PGA).

b.

For SPRAs, plants generally used the mean uniform hazard response spectra and hazard
curves developed by Lawrence Livermore National Laboratory (LLNL) in NUREG1488, “Revised Livermore Seismic Hazard Estimates for Sixty-Nine Nuclear Power Plant
Sites East of the Rocky Mountains,” (Ref. 34), and/or the EPRI in the EPRI NP-6395-D,
“Probabilistic Seismic Hazard Evaluations at Nuclear Plant Sites in the Central and
Eastern US: Resolution of the Charleston Earthquake Issue” (Ref. 35).

c.

In some cases, past SPRAs were submitted for IPEEE closure that used input motions
and hazard curves that preceded the LLNL and EPRI hazard curves of NUREG-1488 and
EPRI NP-6395-D respectively.

Consistent with the input spectrum shape used in an IPEEE, a licensee can develop an IPEEE
high-confidence-of-low-probability-of-failure spectrum (IHS).
Staff Position: The evaluation of redundant safe shutdown success paths under the IPEEE
demonstrates the reasonable protection of equipment necessary to maintain or restore core
cooling and containment capabilities for licensees provided that:
a.

The IHS envelopes the GMRS for the reevaluated seismic hazard from 1 to 10 Hz, with
the exception of small narrowband exceedances that meet the criteria of the EPRI
1025287, “Seismic Evaluation Guidance: Screening, Prioritization and Implementation
Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation
2.1: Seismic” (Ref. 36); and

b.

The previous seismic evaluation was accepted by the NRC in the letter dated May 9,
2014, (Ref. 37), or is subsequently accepted by the NRC to screen out of conducting a
seismic risk evaluation based on the IHS; and
DG-1301, Page 16

c.

If the licensee performed an EPRI SMA, a plant-specific evaluation shows that SSCs that
limit the coping duration to 72 hours are available for an indefinite period to support
continued maintenance of the safe shutdown conditions.

Licensees relying on the IHS to demonstrate reasonable protection of equipment necessary to
maintain or restore core cooling and containment capabilities should:
a.

Evaluate the seismic capacity of equipment necessary to maintain or restore SFP cooling
capabilities to the GMRS for the reevaluated seismic hazard; and

b.

Evaluate HF sensitive in-plant SSCs relied upon to maintain or restore core cooling,
containment and SFP cooling capabilities using the methods of the EPRI Report
3002004396, “High Frequency Program: Application Guidance for Functional
Confirmation and Fragility Evaluation,” if the GMRS from the reevaluated seismic
hazard exceeds the design basis seismic hazard above 10 Hz.

6.1.2.2. Case 2: Demonstration of Reasonable Protection Based upon the Expedited Seismic
Evaluation Process
The EPRI report 3002000704, “Seismic Evaluation Guidance: Augmented Approach for the
Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic,” (Ref. 38),
described an expedited seismic evaluation process (ESEP) to evaluate the seismic ruggedness of
SSCs for a review level ground motion (RLGM) derived by linearly scaling the design basis
seismic hazard by the maximum ratio of it to the GMRS for the reevaluated seismic hazard in the
1 to 10 Hz range, with this ratio limited to a maximum of two times the design basis seismic
hazard. Alternatively, licensees conducted the ESEP using the GMRS itself. The ESEP
evaluated seismic adequacy of components in a single success path for core cooling, reactor
coolant system makeup and containment capabilities for the RLGM or the GMRS resulting from
the reevaluated seismic hazard. This process was endorsed by the NRC by letter dated May 7,
2013 (Ref. 39).
Staff Position: The ESEP demonstrates reasonable protection of evaluated SSCs necessary for the
maintenance or restoration of core cooling and containment capabilities for those licensees having
reevaluated seismic hazards less than twice the design basis seismic hazard.
1.

SSCs not within the scope of the ESEP should be evaluated for reasonable protection as
follows:
a.

Qualitatively based on Seismic Experience. EPRI NP-6041-SL and EPRI TR104871, “Generic Seismic Technical Evaluations of Replacement Items for
Nuclear Power Plants,” (Ref. 40) provide guidance on rugged SSCs. Such
equipment that was not included within the ESEP review and that have high
seismic capacities would require no further actions to demonstrate reasonable
assurance to withstand the new seismic hazard. These SSCs include:
1.

Piping, cabling, conduit, and their supports

2.

Manual valves, check valves, and rupture disks

DG-1301, Page 17

b.

c.

3.

Power operated valves not required to change state as part of the FLEX
mitigation strategies

4.

Nuclear steam supply system components (e.g. reactor pressure vessel
and internals, control rod drive mechanisms (CRDMs), fuel rods, reactor
coolant pumps and seals, etc.)

5.

Portable FLEX equipment (tie downs and seismic interactions should be
addressed using the approach 3 below)

6.

Safety-related buildings

Quantitatively as described in item c below for SSCs and seismic interactions
that were not included in the ESEP review and cannot be justified to be
inherently rugged for seismic accelerations and displacements. Examples of
these SSCs and seismic interactions include:
1.

Haul Path – including liquefaction, slope stability and interactions

2.

FLEX Equipment Storage Building and Non-Seismic Category 1
Structures

3.

Operator Pathways – interaction pathway review, use the beyond-designbasis seismic evaluation criteria described in § 6.1.2.2.1.c, below, if
calculation is required

4.

Tie down of FLEX portable equipment that are required to be restrained
during the earthquake

Beyond-Design-Basis Seismic Evaluation Criteria
In order to demonstrate reasonable protection of equipment that was not included
in the ESEP review and is not inherently rugged for seismic accelerations, a
licensee should demonstrate that the GMRS level of seismic hazard at the site
results in an acceptably low probability of failure. Licensees may rely on the
guidance in the American Society of Civil Engineers/Structural Engineering
Institute (ASCE/SEI) Standard 43-05, “Seismic Design Criteria for Structures,
Systems and Components in Nuclear Facilities” (Ref. 41) related to beyonddesign-basis seismic evaluation for purposes of defining an appropriately low
probability of failure. ASCE/SEI 43-05 defines a 10% probability of
unacceptable performance (C10%) which is reviewed against the beyond-designbasis seismic event (150% of the design-basis event (DBE) ground motion for the
ASCE/SEI 43-05 case).
The process for calculating the C10% values is defined in this section. Table 1
provides recommended values for βC, βR, βU, and the ratio of the median capacity
C50% to the C1% capacity taken from the SPID determined in the EPRI 1025287.
The recommended βC values are based on Kennedy’s recommendations in
“Overview of Methods for Seismic PRA and Margin Analysis Including Recent
Innovations,” Proceedings of the Organization for Economic Co-operation and
Development (OECD)-Nuclear Energy Agency (NEA) Workshop on Seismic
DG-1301, Page 18

Risk, Tokyo, Japan, August, 1999 (Ref. 42), and on average area biased slightly
conservative (i.e., slightly low βC on average). Because random variability βR is
primarily due to ground motion variability, a constant βR value of 0.24 is
recommended regardless of the SSC being considered. The recommended
uncertainty βU values are back-computed from the recommended composite βC
and βR values. The β values for Table 1 apply to fragilities tied to ground motion
parameters (e.g., PGA or Peak Spectral Acceleration at 5 Hz). The ratios of the
10% failure probability capacity C10% to the C1% capacity have been calculated
and are shown in the last column of Table 1. The method for demonstrating the
adequate seismic ruggedness for mitigation systems would follow the approach
for an SMA wherein a defined capacity is shown to exceed the defined demand.
In the case of an SMA the demand for the assessment is referred to as the review
level earthquake (RLE). The following steps would be undertaken for SSCs
within the mitigation systems that undertake the C10% review:
•

The GMRS will be the RLE for the beyond-design-basis seismic review of
the mitigation strategies

•

The seismic capacity aligned with reasonable protection will be the C10%
value. The C10% can be calculated by:

•

o

Calculate the C1% capacity using the methods documented in past SPRA
and seismic margin documentation and as summarized in the SPID
defined in EPRI 1025287.

o

Multiply the C1% capacity by the C10%/C1% ratio from Table 1 based on
the type of SSC being evaluated

Verify that the C10% capacity exceeds the RLE demand

Table 1: βC, βR, βU, and C50%/C1% Values for Hybrid Method for Various Types of SSCs
Type SSC

Composite
βC

Random
βR

Uncertainty
βU

C50%/C1%

C10%/C1%

Structures & Major Passive
Mechanical Components
Mounted on Ground or at
Low Elevation Within
Structures

0.35

0.24

0.26

2.26

1.44

Active Components
Mounted at High Elevation
in Structures

0.45

0.24

0.38

2.85

1.60

Other SSCs

0.40

0.24

0.32

2.54

1.52

DG-1301, Page 19

2.

6.2.

Licensees relying on the ESEP to demonstrate reasonable protection of equipment
necessary to maintain or restore core cooling and containment capabilities should:
a.

Evaluate the seismic capacity of equipment necessary to maintain or restore SFP
cooling capabilities to the GMRS for the reevaluated seismic hazard; and

b.

Evaluate HF sensitive in-plant SSCs relied upon to maintain or restore core
cooling, containment and SFP cooling capabilities using the methods in the EPRI
Report 3002004396, “High Frequency Program: Application Guidance for
Functional Confirmation and Fragility Evaluation,” ADAMS Accession No.
ML15223A095, consistent with its endorsement by letter dated September 17,
2015, ADAMS Accession No. ML15218A569, if the GMRS from the
reevaluated seismic hazard exceeds the design basis seismic hazard above 10 Hz.

Treatment of Reevaluated Flooding Hazards
Appendix G of NEI 12-06, Rev. 1A discusses a method to assess the results of the flooding
hazard reevaluations with respect to the guidance and strategies required by 10 CFR 50.155(b)(1).
6.2.1. (Modified) Mitigating Strategies
Sections G.4.1 and G.4.2 of Appendix G of NEI 12-06, Rev. 1A discuss a method to assess or
modify the mitigating strategies to show they comply with the requirements of 10 CFR
50.155(b)(1) and 10 CFR 50.155(c)(2)(i) by using the new flooding hazard information, referred
to as mitigating strategies flood hazard information.
Staff Position: Sections G.4.1 and G.4.2 of Appendix G of NEI 12-06, Rev. 1A provide
acceptable methods to show that the existing strategies and guidelines comply with the
requirements of 10 CFR 50.155(b)(1) and 10 CFR 50.155(c)(2)(i) for the new flooding hazard
information or for developing modified strategies and guidelines.
6.2.2. Alternate Mitigating Strategies
Section G.4.3 of Appendix G of NEI 12-06, Rev. 1A discusses a method to develop AMS to meet
the requirements of 10 CFR 50.155(b)(1) and 10 CFR 50.155(c)(2)(i).
Staff Position: Development of an AMS that provides a capability to mitigate the BDBEE by
mitigating or preventing an ELAP that would occur as a result of the BDBEE through exhaustion
of fuel for operating emergency power sources is an acceptable method of compliance with
10 CFR 50.155(b)(1) when the hazard level for the AMS is identified. Section G.4.3 of Appendix
G of NEI 12-06, Rev. 1A provides an acceptable method to develop AMS to meet the
requirements of 10 CFR 50.155(b)(1) and 10 CFR 50.155(c)(2)(i).
6.2.3. Targeted Hazard Mitigating Strategies
Section G.4.4 of Appendix G of NEI 12-06, Rev. 1A discusses a method to develop targeted
hazard mitigating strategies (THMS) to meet the requirements of 10 CFR 50.155(b)(1) and 10
CFR 50.155(c)(2)(i).
Staff Position: Development of a THMS that provides a capability to mitigate the BDBEE by
mitigating or preventing an ELAP that would occur as a result of the BDBEE through exhaustion
DG-1301, Page 20

of fuel for operating emergency power sources is an acceptable method of compliance with 10
CFR 50.155(b)(1) when the hazard level for the THMS is identified. Section G.4.4 of Appendix
G of NEI 12-06, Rev. 1A provides an acceptable method to develop THMS to meet the
requirements of 10 CFR 50.155(b)(1) and 10 CFR 50.155(c)(2)(i).
7.

Coordination with Severe Accident Management Guidelines
In SRM-COMSECY-15-0065, (Ref. 43) the Commission directed the NRC staff to “…ensure
that any NRC-endorsed guidance for the proposed rule will provide for appropriate coordination
of the FLEX support guidelines, extreme damage mitigating guidelines, and voluntarily
maintained SAMGs with the existing emergency operating procedures (EOPs) at each plant….”
Section 3.2.1.10 of NEI 12-06, Rev. 1A provides criteria for the selection of parameters to be
monitored as part of the minimum set of parameters necessary to support strategy
implementation. These criteria include the ability to demonstrate the success of the strategies at
maintaining the key safety functions as well as indicating imminent or actual core damage to
facilitate a decision to manage the response to the event within the EOPs and FSGs or within the
Severe Accident Management Guidelines (SAMGs).
Section 11.4 of NEI 12-06, Rev. 1A provides that FSGs will be used to supplement (not replace)
the existing procedure that establishes the command and control for the event. This section
further provides that the existing command and control procedure structure will be used to
transition to SAMGs if FLEX mitigation strategies are not successful.
Staff Position: Sections 3.2.1.10 and 11.4 provide appropriate coordination between the FSGs
and voluntarily maintained SAMGs, retaining command and control direction as defined within
the EOPs unless and until a licensee transitions to the use of SAMGs.

8.

Guidance for AP-1000 Design
Appendix F of NEI 12-06, Rev. 1A provides specific guidance for licensees with reactors of the
AP-1000 design on how to satisfy provisions of the aforementioned regulations for sufficient
offsite resources to sustain functions indefinitely.
Staff Position: The guidance of NEI 12-06, Rev. 1A, Appendix F, provides an acceptable means
to meet the requirements of the regulations or license conditions imposing similar requirements.

D. IMPLEMENTATION.
The purpose of this section is to provide information on how applicants and licensees1 may use
this guide and information regarding the NRC’s plans for using this RG. In addition, it describes how the
NRC complies with the Backfit Rule found in 10 CFR 50.109(a)(1) or any applicable finality provisions
in 10 CFR Part 52.

1

In this section, “licensees” refers to holders of, and “applicants” refers to applicants for, licenses for nuclear power
plants under 10 CFR Parts 50 and 52.

DG-1301, Page 21

Use by Applicants and Licensees
Applicants and licensees may voluntarily2 use the guidance in this document to demonstrate
compliance with the underlying NRC regulations. Methods or solutions that differ from those described in
this RG may be deemed acceptable if they provide sufficient basis and information for the NRC to verify
that the proposed alternative demonstrates compliance with the appropriate NRC regulations. Current
licensees may continue to use guidance the NRC found acceptable for complying with the identified
requirements as long as their current licensing basis remains unchanged.
Licensees may use the information in this RG for actions that do not require NRC review and
approval. Licensees may use the information in this RG or applicable parts to resolve regulatory or
inspection issues.
Use by NRC
The NRC does not intend or approve any imposition or backfitting of the guidance in this RG.
The NRC does not expect any existing licensee to use or commit to using the guidance in this RG, unless
the licensee makes a change to its licensing basis. The NRC does not expect or plan to request licensees to
voluntarily adopt this RG to resolve a generic regulatory issue. The NRC does not expect or plan to
initiate NRC regulatory action that would require the use of this RG. Examples of such unplanned NRC
regulatory actions include issuance of an order requiring the use of the RG, generic communication, or
promulgation of a rule requiring the use of this RG without further backfit consideration.
During regulatory discussions on plant specific operational issues, the NRC staff may discuss
with licensees various actions consistent with NRC positions in this RG, as one acceptable means of
meeting the underlying NRC regulatory requirement. Such discussions would not ordinarily be
considered backfitting. However, unless this RG is part of the licensing basis for a facility, the NRC may
not represent to the licensee that the licensee’s failure to comply with the positions in this RG constitutes
a violation.
If an existing licensee voluntarily seeks a license amendment or change and (1) the NRC’s
consideration of the request involves a regulatory issue directly relevant to this RG and (2) the specific
subject matter of this RG is an essential consideration in the NRC’s determination of the acceptability of
the licensee’s request, then the NRC may request that the licensee either follow the guidance in this RG or
provide an equivalent alternative process that demonstrates compliance with the underlying NRC
regulatory requirements. This is not considered backfitting as defined in 10 CFR 50.109(a)(1) or a
violation of any applicable finality provisions in 10 CFR Part 52.
If a licensee believes that the NRC is either using this RG or requesting or requiring the licensee
to implement the methods or processes in this RG in a manner inconsistent with the discussion in this
Implementation section, then the licensee may file a backfit appeal with the NRC in accordance with the
guidance in NUREG-1409, “Backfitting Guidelines,” (Ref. 44) and the NRC Management Directive 8.4,
“Management of Facility-Specific Backfitting and Information Collection” (Ref. 45).

2

In this section, “voluntary” and “voluntarily” means that the licensee is seeking the action of its own accord, without
the force of a legally binding requirement or an NRC representation of further licensing or enforcement action.

DG-1301, Page 22

REFERENCES3
1.

U.S. Code of Federal Regulations, Title 10, “Energy,” Chapter 1, “Nuclear Regulatory
Commission,” Part 50, “Domestic Licensing of Production and Utilization Facilities.”

2.

Nuclear Energy Institute (NEI) document NEI 12-06, “Diverse and Flexible Coping Strategies
(FLEX) Implementation Guide,” Revision 1A, dated October 2015, Washington, DC. (ADAMS
Accession No. ML15279A426)4

3.

NRC, Order EA-12-049, “Order Modifying Licenses with Regard to Requirements for Mitigation
Strategies for Beyond-Design-Basis External Events,” dated March 12, 2012, Washington, DC.
(ADAMS Accession No. ML12054A736).

4.

NRC, JLD-ISG-2012-01, “Compliance with Order EA-12-049, Order Modifying Licenses with
Regard to Requirements for Mitigating Strategies for Beyond-Design-Basis External Events,”
Revision 0, Issued August 2012, NRC, Washington, DC. (ADAMS Accession No.
ML12229A174)

5.

NEI, document NEI 12-06, “Diverse and Flexible Coping Strategies (FLEX) Implementation
Guide,” Revision 0, dated August 21, 2012, Washington, DC. (ADAMS Accession No.
ML12242A378)

6.

CFR, Title 10, Energy, Part 52, “Licenses, Certifications, and Approvals for Nuclear Power
Plants.”

7.

NRC, SECY-11-0093, “Near-Term Report and Recommendations for Agency Actions Following
the Events in Japan,” dated July 12, 2011, Washington, DC. (ADAMS Accession No.
ML11186A950)

8.

NRC, SRM-SECY-11-0093, “Staff Requirements – SECY-11-0093 – Near-Term Report and
Recommendations for Agency Actions Following the Events in Japan,” dated August 19, 2011,
Washington, D.C. (ADAMS Accession No. ML112310021)

9.

NRC, SECY-11-0124, “Recommended Actions to be Taken without Delay from the Near-Term
Task Force Report,” dated September 9, 2011, Washington, DC. (ADAMS Accession No.
ML11245A158)

10.

NRC, SECY-11-0137, “Prioritization of Recommended Actions to be Taken in Response to
Fukushima Lessons Learned,” dated October 3, 2011, Washington, DC. (ADAMS Accession No.
ML11272A111)

3

Publicly available documents from the U.S. Nuclear Regulatory Commission (NRC) are available electronically
through the NRC Library on the NRC’s public Web site at http://www.nrc.gov/reading-rm/doc-collections/. The
documents can also be viewed on-line for free or printed for a fee in the NRC’s Public Document Room (PDR) at
11555 Rockville Pike, Rockville, MD; the mailing address is USNRC PDR, Washington, DC 20555; telephone (301)
415-4737 or (800) 397-4209; fax (301) 415 3548; and e-mail [email protected].

4

Publications from the Nuclear Energy Institute (NEI) are available at their Web site: http://www.nei.org/ or by
contacting the headquarters at Nuclear Energy Institute, 1776 I Street NW, Washington DC 20006-3708, Phone: 202739-800, Fax 202-785-4019.

DG-1301, Page 23

11.

NRC, SRM-SECY-11-0124, “Staff Requirements - SECY-11-0124 - Recommended Actions to
be Taken without Delay from the Near-Term Task Force Report,” dated October 18, 2011,
Washington, DC. (ADAMS Accession No. ML112911571)

12.

NRC, SRM-SECY-11-0137, “Staff Requirements - SECY-11-0137 - Prioritization of
Recommended Actions to be Taken in Response to Fukushima Lessons Learned,” dated
December 15, 2011, Washington, DC. (ADAMS Accession No. ML113490055)

13.

Letter from Adrian Heymer (Nuclear Energy Institute) to David L. Skeen (NRC), “An Integrated,
Safety-Focused Approach to Expediting Implementation of Fukushima Dai-ichi Lessons
Learned,” dated December 16, 2011, NEI, Washington, DC. (ADAMS Accession No.
ML11353A008)

14.

NRC, SECY-12-0025, “Proposed Orders and Requests for Information in Response to Lessons
Learned from Japan’s March 11, 2011, Great Tohoku Earthquake and Tsunami,” dated February
17, 2012, Washington, DC. (ADAMS Accession No. ML12039A103)

15.

NRC, SRM-SECY-12-0025, “Staff Requirements - SECY-12-0025 - Proposed Orders and
Requests for Information in Response to Lessons Learned from Japan’s March 11, 2011, Great
Tohoku Earthquake and Tsunami,” dated March 9, 2012, Washington, DC. (ADAMS Accession
No. ML120690347)

16.

NRC, Commission Memorandum and Order, CLI-12-09, dated March 30, 2112, Washington,
DC. (ADAMS Accession No. ML12090A531).

17.

NEI, document NEI 12-06, “Diverse and Flexible Coping Strategies (FLEX) Implementation
Guide,” Revision B, dated May 4, 2012, Washington, DC. (ADAMS Accession No.
ML12144A419)

18.

NEI, document NEI 12-06, “Diverse and Flexible Coping Strategies (FLEX) Implementation
Guide,” Revision B1, May 13, 2012, Washington, DC. (ADAMS Accession No. ML12143A232)

19.

NRC, Response to Public Comments on JLD-ISG-2012-01 (Docket ID NRC-2012-0068), dated
August 29, 2012, Washington DC. (ADAMS Accession No. ML12229A253)

20.

NEI, document 12-06, “Diverse and Flexible Coping Strategies (FLEX) Implementation Guide,”
Revision C, July 3, 2012, Washington, DC. (ADAMS Accession No. ML15027A542)

21.

NEI, document NEI 12-06, “Diverse and Flexible Coping Strategies (FLEX) Implementation
Guide,” Revision 1, dated October, 2015, Washington, DC. (ADAMS Accession No.
ML15244B006)

22.

NEI, document NEI 12-01, “Guideline for Assessing Beyond Design Basis Accident Response
Staffing and Communication Capabilities,” Revision 0, dated May 2012. Washington, DC.
(ADAMS Accession No. ML12125A412)

23.

Letter from David L. Skeen (NRC) to Susan Perkins-Grew (NEI), Subject: “U.S. Nuclear
Regulatory Commission Review of NEI 12-01, ‘Guidelines for Assessing Beyond Design Basis
Accident Response Staffing and Communication Capabilities,’ Revision 0, dated May 2013,”
dated May 15, 2012, Washington, DC. (ADAMS Accession No. ML12131A043)
DG-1301, Page 24

24.

NRC, NUREG-1431, “Standard Technical Specifications – Westinghouse Plants,” Revision 4.0,
Volume 1, “Specifications,” Washington, DC. (ADAMS Accession No. ML12100A222)

25.

Electric Power Research Institute (EPRI) technical report 3002000623, “Applications Center:
Preventive Maintenance Basis for FLEX Equipment – Project Overview Report,” Palo Alto, CA5
(ADAMS Accession Number ML13276A573)

26.

Letter from Jack R. Davis (NRC) to Joseph E. Pollock (NEI), no subject line, dated October 7,
2013, Washington, DC. (ADAMS Accession Number ML13276A224)

27.

EPRI, Technical Report 3002004396, High Frequency Program: Application Guidance for
Functional Confirmation and Fragility Evaluation,” dated July 2015, Palo Alto, CA. (ADAMS
Accession No. ML15223A102)

28.

Letter from Jack R. Davis (NRC) to Andrew Mauer (EPRI), Subject: “Endorsement of Electric
Power Research Institute Final Draft Report 3002004396, ‘High Frequency Program:
Application Guidance for Functional Confirmation and Fragility,” dated September 17, 2015,
Washington, DC. (ADAMS Accession No. ML15218A569)

29.

NRC, Generic Letter 88-20, Supplement 4, Subject: “Individual Plant Examination of External
Events (IPEEE) for Severe Accident Vulnerabilities - 10CFR 50.54(f)” dated June 28, 1991.
http://www.nrc.gov/reading-rm/doc-collections/gen-comm/gen-letters/1988/gl88020s4.html

30.

NRC, Generic Letter 88-20, Supplement 5, Subject: “Individual Plant Examination of External
Events for Severe Accident Vulnerabilities,” dated September 8, 1995.
http://www.nrc.gov/reading-rm/doc-collections/gen-comm/gen-letters/1988/gl88020s5.html/

31.

NRC, NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of
External Events (IPEEE) for Severe Accident Vulnerabilities, dated June 1991, Washington, DC.
(ADAMS Accession No. ML063550238)

32.

EPRI, Report NP-6041-SL, Revision 1, “A Methodology for Assessment of Nuclear Plant
Seismic Margin,” dated August 1, 1991, Palo-Alto, CA.
http://www.epri.com/abstracts/Pages/ProductAbstract.aspx?ProductId=NP-6041-SLR1

33.

NRC, NUREG/CR-0098, “Development of Criteria for Seismic Review of Selected Nuclear
Power Plants,” dated May 1978, Washington, DC. (ADAMS Accession No. ML061880412)

34.

NRC, NUREG-1488, “Revised Livermore Seismic Hazard Estimates for Sixty-Nine Nuclear
Power Plant Sites East of the Rocky Mountains,” dated April 1994, Washington, DC. (ADAMS
Accession No. ML052640591)

35.

EPRI, Report NP-6395-D, “Probabilistic Seismic Hazard Evaluations at Nuclear Plant Sites in the
Central and Eastern US: Resolution of the Charleston Earthquake Issue,” dated May 1, 1989, Palo
Alto, CA. http://www.epri.com/abstracts/Pages/ProductAbstract.aspx?ProductId=NP-6395-D

5

Copies of Electric Power Research Institute (EPRI) documents may be obtained by contacting the Electric Power
Research Institute, 3420 Hillview Avenue, Palo Alto, CA 94304, Telephone: 650-855-2000 or on-line at
http://my.epri.com/portal/server.pt.

DG-1301, Page 25

36.

EPRI, Report 1025287, “Seismic Evaluation Guidance: Screening, Prioritization and
Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force
Recommendation 2.1: Seismic,” dated February 28, 2013, Palo Alto, CA. (ADAMS Accession
No. ML12333A170)

37.

Letter from Eric J. Leeds (NRC) to All Power Reactor Licensees and Holders of Construction
Permits in Active or Deferred Status on the Enclosed List, “Screening and Prioritization Results
Regarding Information pursuant to Title 10 of the Code of Federal Regulations 50.54(F)
Regarding Seismic Hazard Re-evaluations for Recommendation 2.1 of the Near-Term Task Force
Review of Insights from the Fukushima Dai-Ichi Accident,” dated May 9, 2014, (ADAMS
Accession No. ML14111A147)

38.

EPRI technical report 3002000704, Seismic Evaluation Guidance: Augmented Approach for the
Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic,” April 2013,
Palo Alto, CA (ADAMS Accession No. ML13102A142)

39.

Letter from Eric J. Leeds (NRC) to Joseph E. Pollock (NEI), Subject: “Electric Power Research
Institute Final Draft Report XXXXXX, ‘Seismic Evaluation Guidance: Augmented Approach For
The Resolution Of Fukushima Near-Term Task Force Recommendation 2.1: Seismic,’ as an
Acceptable Alternative to the March 12, 2012, Information Request for Seismic Reevaluations,”
dated May 7, 2013 (ADAMS Accession No. ML13106A331)

40.

EPRI technical report TR-104871, “Generic Seismic Technical Evaluations of Replacement Items
for Nuclear Power Plants,” May 1, 1995, Palo Alto, CA
http://www.epri.com/abstracts/Pages/ProductAbstract.aspx?ProductId=TR-104871

41.

American Society of Civil Engineers (ASCE)/Structural Engineering Institute (SEI) standard 4305, “Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities,”
May 2005, Reston, VA6

42.

Kennedy, R. P., “Overview of Methods for Seismic PRA and Margin Analysis Including Recent
Innovations”, Proceedings of the Organization for Economic Co-operation and Development
(OECD)-Nuclear Energy Agency (NEA) Workshop on Seismic Risk, Tokyo, Japan, August,
1999. https://www.oecd-nea.org/nsd/docs/1999/csni-r99-28.pdf

43.

NRC, SRM, SECY-15-0065, “Proposed Rulemaking: Mitigation of Beyond-Design-Basis Events
(RIN 3150-AJ49),” August 27, 2015, Washington, DC (ADAMS Accession No. ML15239A767)

44.

NRC, NUREG-1409, "Backfitting Guidelines," July 1990, Washington, DC. (ADAMS Accession
No. ML032230247)

45.

NRC, Management Directive 8.4, "Management of Facility specific Backfitting and Information
Collection," Washington, DC.

6

Copies of American Society of Civil Engineers (ASCE)/Structural Engineering Institute (SEI) standards are available
through their Web site (www.asce.org), or by contacting them at American Society of Civil Engineers, 1801 Alexander
Bell Drive, Reston, VA 20191, Phone 800-548-2723.

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