[Federal Register Volume 80, Number 219 (Friday, November 13, 2015)]
[Proposed Rules]
[Pages 70610-70647]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-28589]



[[Page 70609]]

Vol. 80

Friday,

No. 219

November 13, 2015

Part III





Nuclear Regulatory Commission





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10 CFR Parts 50 and 52





Mitigation of Beyond-Design-Basis Events; Proposed Rule

  Federal Register / Vol. 80 , No. 219 / Friday, November 13, 2015 / 
Proposed Rules  

[[Page 70610]]


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NUCLEAR REGULATORY COMMISSION

10 CFR Parts 50 and 52

[Docket Nos. PRM-50-97 and PRM-50-98; NRC-2011-0189 and NRC-2014-0240]
RIN 3150-AJ49


Mitigation of Beyond-Design-Basis Events

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to 
amend its regulations that establish regulatory requirements for 
nuclear power reactor applicants and licensees to mitigate beyond-
design-basis events. The NRC is proposing to make generically 
applicable requirements in Commission orders for mitigation of beyond-
design-basis events and for reliable spent fuel pool instrumentation. 
This proposed rule would establish regulatory requirements for an 
integrated response capability, including supporting requirements for 
command and control, drills, training and change control. This proposed 
rule also would establish requirements for enhanced onsite emergency 
response capabilities. Finally, this proposed rule would address a 
number of petitions for rulemaking (PRMs) submitted to the NRC 
following the March 2011 Fukushima Dai-ichi event. This rulemaking is 
applicable to power reactor licensees, power reactor license 
applicants, and decommissioning power reactor licensees. This 
rulemaking combines two NRC activities for which documents have been 
published in the Federal Register--Onsite Emergency Response 
Capabilities (RIN 3150-AJ11; NRC-2012-0031) and Station Blackout 
Mitigation Strategies (RIN 3150-AJ08; NRC-2011-0299). The new 
identification numbers for this consolidated rulemaking are RIN 3150-
AJ49 and NRC-2014-0240.

DATES: Submit comments by February 11, 2016. Comments received after 
this date will be considered if it is practical to do so, but the 
Commission is able to ensure consideration only for comments received 
before this date. A public meeting will be held during the public 
comment period; refer to the NRC's public meeting schedule on the NRC 
Web site at http://meetings.nrc.gov/pmns/mtg.

ADDRESSES: You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0240. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions contact 
the individuals listed in the FOR FURTHER INFORMATION CONTACT section 
of this document.
     Email comments to: [email protected]. If you do 
not receive an automatic email reply confirming receipt, then contact 
us at 301-415-1677.
     Fax comments to: Secretary, U.S. Nuclear Regulatory 
Commission at 301-415-1101.
     Mail comments to: Secretary, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, ATTN: Rulemakings and 
Adjudications Staff.
     Hand deliver comments to: 11555 Rockville Pike, Rockville, 
Maryland 20852, between 7:30 a.m. and 4:15 p.m. (Eastern Time) Federal 
workdays; telephone: 301-415-1677.
    You may submit comments on the guidance documents and the 
information collections by the methods indicated in the ``Availability 
of Guidance'' and ``Paperwork Reduction Act'' sections of this 
document.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Timothy Reed, Office of Nuclear 
Reactor Regulation, telephone: 301-415-1462, email: 
[email protected]; or Eric Bowman, Office of Nuclear Reactor 
Regulation, telephone: 301-415-2963, email: [email protected]. Both 
are staff of the U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001.

SUPPLEMENTARY INFORMATION:

Executive Summary

A. Need for the Regulatory Action

    The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend 
its regulations to establish regulatory requirements for nuclear power 
reactor applicants and licensees to mitigate beyond-design-basis 
events. This proposed rule would make Commission Order EA-12-049 and 
Order EA-12-051 generically applicable; establish regulatory 
requirements for an integrated response capability, including 
supporting requirements for command and control, drills, training and 
change control; include requirements for enhanced onsite emergency 
response capabilities; and address a number of petitions for rulemaking 
submitted to the NRC following the March 2011 Fukushima Dai-ichi event. 
This rulemaking would be applicable to operating power reactor 
licensees, power reactor license applicants, and decommissioning power 
reactor licensees. The NRC is conducting this rulemaking to amend the 
regulations to reflect requirements imposed on current licensees by 
order and to reflect the lessons learned from the Fukushima accident.

B. Major Provisions

    Major provisions of this proposed rule include amendments or 
additions to parts 50 and 52 of title 10 of the Code of Federal 
Regulations (10 CFR) that would:
     Revise the 10 CFR parts 50 and 52 ``Content of 
application'' requirements to reflect the additional information that 
would be required for applications.
     Add proposed Sec.  50.155, which contains beyond-design-
basis mitigation requirements that would make Orders EA-12-049 and EA-
12-051 generically applicable; requires an integrated response 
capability for beyond-design-basis events that includes the integration 
of two guideline sets with the existing emergency operating procedures; 
training requirements; drills or exercise requirements; and change 
control requirements.
     Revise 10 CFR part 50, appendix E, to include enhanced 
capabilities for assessing the impact and release of radioactive 
materials for multi-unit events; to remove references to specific 
technology for each licensee's emergency response data system; to 
include enhanced capabilities for onsite and offsite communications; 
and to add staffing analysis requirements to address multi-unit events.

C. Costs and Benefits

    The NRC prepared a draft regulatory analysis to determine the 
expected costs and benefits of the proposed rule. The draft analysis 
demonstrates that the proposed rule is justified. The draft analysis 
examines the benefits and costs of the proposed rule requirements 
relative to the baseline (i.e., no action alternative). Additionally, 
the draft analysis estimates the historical costs incurred as a result 
of implementation of Order EA-12-049, Order EA-12-051, and related 
industry initiatives. The proposed rule costs are associated with the 
proposed provisions that make generically-applicable Order EA-12-049 
and Order EA-12-051, as well as related industry initiatives and the 
NRC's rulemaking-related costs. Because the NRC uses a no action 
baseline to estimate incremental costs, the total cost

[[Page 70611]]

of the proposed rule is estimated to be approximately $7.2 million for 
the industry ($111,000 per site) to review the rule against the 
previous implementation of Orders EA-12-049 and EA-12-051 and make any 
additional changes to plant programs and procedures. This small impact 
stems from the fact that the proposed requirements are expected to be 
implemented prior to the effective date of the rule. However, this 
regulatory analysis does not estimate the impacts that may occur as a 
result of licensees needing to make changes to mitigation strategies 
including potential plant modifications as a result of the need to 
address the seismic and flooding reevaluated hazards for reasonable 
protection of the FLEX equipment. As part of the proposed rule, the NRC 
is seeking external stakeholder feedback to enable these impacts to be 
estimated.
    The proposed rule would result in a total one-time cost to the NRC 
of $880,000 to complete the rulemaking (i.e., complete the proposed 
rule, analyze public comments, hold public meeting(s), and develop the 
final rule and regulatory guidance).
    Based on the NRC's assessment of the costs and benefits of the 
proposed rule, the NRC has concluded that the proposed rule is 
justified. For more information, please see the draft regulatory 
analysis (Accession No. ML15265A610 in the NRC's Agencywide Documents 
Access and Management System).

Table of Contents

I. Obtaining Information and Submitting Comments
    A. Obtaining Information
    B. Submitting Comments
II. Background
    A. Fukushima Dai-ichi
    B. NRC Near-Term Task Force
    C. Implementation of the NTTF Recommendations
    D. Consolidation of Regulatory Efforts
    E. Public Involvement
III. Petitions for Rulemaking
IV. Discussion
    A. Rulemaking Objectives
    B. Rulemaking Scope
    C. Proposed Rule Organization
    D. Proposed Rule Regulatory Bases
V. Section-by-Section Analysis
VI. Specific Requests for Comments
VII. Regulatory Flexibility Certification
VIII. Availability of Regulatory Analysis
IX. Availability of Guidance
X. Backfitting and Issue Finality
XI. Cumulative Effects of Regulation
XII. Plain Writing
XIII. Environmental Assessment and Proposed Finding of No 
Significant Environmental Impact
XIV. Paperwork Reduction Act
XV. Criminal Penalties
XVI. Coordination With NRC Agreement States
XVII. Compatibility of Agreement State Regulations
XVIII. Voluntary Consensus Standards
XIX. Public Meeting
XX. Availability of Documents

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2014-0240 when contacting the U.S. 
Nuclear Regulatory Commission (NRC) about the availability of 
information for this action. You may obtain publicly-available 
information related to this action by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0240.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. For 
the convenience of the reader, instructions about obtaining materials 
referenced in this document are provided in the ``Availability of 
Documents'' section.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2014-0240 in your comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

II. Background

A. Fukushima Dai-ichi

    At 2:46 p.m. Japan standard time on March 11, 2011, the Great East 
Japan Earthquake, rated a magnitude 9.0, occurred at a depth of 
approximately 25 kilometers, 130 kilometers east of Sendai and 372 
kilometers northeast of Tokyo off the coast of Honshu Island. This 
earthquake resulted in the automatic shutdown of 11 nuclear power 
plants (NPPs) at four sites along the northeast coast of Japan 
including three of six reactors at the Fukushima Dai-ichi NPP (the 
three remaining plants were in outages). The earthquake precipitated a 
large tsunami that is estimated to have exceeded 14 meters in height at 
the Fukushima Dai-ichi NPP. The earthquake and tsunami produced 
widespread devastation across northeastern Japan, resulting in 
approximately 25,000 people dead or missing, displacing many tens of 
thousands of people, and significantly impacting the infrastructure and 
industry in the northeastern coastal areas of Japan.
    The earthquake and tsunami disabled the majority of the external 
and internal electrical power systems at the Fukushima Dai-ichi NPP, 
leaving it with only a few hours' worth of battery power. Since an NPP 
licensee typically relies on electrical power to keep its reactor core 
and spent fuel pool (SFP) cool, this loss of internal and external 
power was a significant challenge to operators at Fukushima Dai-ichi. 
In addition, the combination of severe events challenged the 
implementation of emergency plans and procedures.

B. NRC Near-Term Task Force

    The NRC Chairman's tasking memorandum, COMGBJ-11-0002, ``NRC 
Actions Following the Events in Japan,'' established a senior-level 
task force referred to as the ``Near-Term Task Force'' (NTTF) to 
conduct a systematic and methodical review of NRC regulations and 
processes to determine if the agency should make safety improvements in 
light of the events in Japan. On July 12, 2011, the NRC staff provided 
the Commission with the report of the NTTF (NTTF Report) as an 
enclosure to SECY-11-0093, ``Near-Term Report and Recommendations for 
Agency Actions Following the Events in Japan.'' The NTTF concluded that 
continued U.S. plant operation and NRC licensing activities present no 
imminent risk to public health and safety. While

[[Page 70612]]

the NTTF also concluded that the current regulatory system has served 
the NRC and the public well, it found that enhancements to safety and 
emergency preparedness are warranted and made a dozen general 
recommendations for Commission consideration. In examining the 
Fukushima Dai-ichi accident for insights for reactors in the United 
States, the NTTF addressed protecting against accidents resulting from 
natural phenomena, mitigating the consequences of such accidents, and 
ensuring emergency preparedness. The NTTF found that the Commission's 
longstanding defense-in-depth philosophy, supported and modified as 
necessary by state-of-the-art probabilistic risk assessment techniques, 
should continue to serve as the primary organizing principle of its 
regulatory framework. The NTTF concluded that the application of the 
defense-in-depth philosophy could be strengthened by including explicit 
requirements for beyond-design-basis events.
    In response to the NTTF Report, the Commission directed the NRC 
staff to engage with stakeholders to review and assess the NTTF 
recommendations in a comprehensive and holistic manner and to provide 
the Commission with fully-informed options and recommendations. The 
Commission's Staff Requirements Memorandum (SRM)-SECY-11-0093 provided 
that direction and specifically directed the NRC staff to pursue 
recommendation 1 of the NTTF Report independent of the activities 
associated with the review of the remaining recommendations. The NTTF's 
recommendation 1 was to establish a logical, systematic, and coherent 
regulatory framework for adequate protection that appropriately 
balances defense-in-depth and risk considerations. This recommendation 
included steps for the establishment of a Commission policy statement 
for a risk-informed defense-in-depth framework including extended 
design-basis requirements and the initiation of rulemaking to implement 
that framework. The results of the NRC staff work on NTTF 
recommendation 1 were provided to the Commission in SECY-13-0132, 
``Plan for Updating the U.S. Nuclear Regulatory Commission's Cost 
Benefit Guidance,'' and dispositioned by the Commission in SRM-SECY-13-
0132, which specifically disapproved the establishment of a design-
basis extension category of events and associated regulatory 
requirements and changes to the NRC's approach to defense-in-depth, but 
allowed for reevaluation, as appropriate, in the context of the 
Commission direction on the proposed policy statement for a long-term 
Risk Management Regulatory Framework. That work is outside of the scope 
of this rulemaking. The Commission has closed NTTF recommendation 1.

C. Implementation of the NTTF Recommendations

    Following the issuance of the NTTF Report, the NRC staff provided 
the Commission with recommendations for near-term action in SECY-11-
0124, ``Recommended Actions to be Taken Without Delay from the Near-
Term Task Force Report,'' dated September 9, 2011. The suggested near-
term actions addressed several NTTF recommendations associated with 
this rulemaking, including NTTF recommendations 4, 8, and 9.3. In SRM-
SECY-11-0124, dated October 18, 2011, the Commission directed the NRC 
staff to, among other things: initiate a rulemaking to address NTTF 
recommendation 4, Station Blackout (SBO) regulatory actions, as an 
Advance Notice of Proposed Rulemaking (ANPR); designate the SBO 
rulemaking associated with NTTF recommendation 4 as a high priority 
rulemaking; craft recommendations that continue to realize the 
strengths of a performance-based system as a guiding principle; and 
consider approaches that are flexible and able to accommodate a diverse 
range of circumstances and conditions. As discussed more fully in later 
portions of this proposed rule, the regulatory actions associated with 
NTTF recommendation 4 evolved substantially from this early Commission 
direction, and included issuance of Order EA-12-049 that, as 
implemented, ultimately addressed all of NTTF recommendation 4 as well 
as other recommendations.
    In SECY-11-0137, ``Prioritization of Recommended Actions To Be 
Taken in Response to Fukushima Lessons Learned,'' dated October 3, 
2011, the NRC staff, based on its assessment of the NTTF 
recommendations, proposed to the Commission a three-tiered 
prioritization for implementing regulatory actions stemming from the 
NTTF recommendations. The Tier 1 recommendations were those actions 
having the greatest safety benefit that could be implemented without 
unnecessary delay. The Tier 2 recommendations were those actions that 
needed further technical assessment or critical skill sets to 
implement, and the Tier 3 recommendations were longer-term actions that 
depended on the completion of a shorter-term action or needed 
additional study to support a regulatory action. On December 15, 2011, 
the Commission approved the staff's recommended prioritization in SRM-
SECY-11-0137.
    The NTTF recommendations that form the basis of this rulemaking 
activity are:
     NTTF recommendation 4: Strengthen SBO mitigation 
capability at all operating and new reactors for design-basis and 
beyond-design-basis external events;
     NTTF recommendation 7: Enhance spent fuel pool makeup 
capability and instrumentation for the spent fuel pool;
     NTTF recommendation 8: Strengthen and integrate onsite 
emergency response capabilities such as emergency operating procedures 
(EOPs), Severe Accident Management Guidelines (SAMGs), and extensive 
damage mitigation guidelines (EDMGs);
     NTTF recommendation 9: Require that facility emergency 
plans address staffing, dose assessment capability, communications, 
training and exercises, and equipment and facilities for prolonged 
station blackout, multi-unit events, or both;
     NTTF recommendation 10: Pursue additional emergency 
protection topics related to multi-unit events and prolonged station 
blackout, including command and control structure and the 
qualifications of decision makers; and
     NTTF recommendation 11: Pursue emergency management topics 
related to decision making, radiation monitoring, and public education, 
including the ability to deliver equipment to the site with degraded 
offsite infrastructure.
    In response to input received from stakeholders, the NRC 
accelerated the schedule originally proposed in SECY-11-0137. On 
February 17, 2012, the NRC staff recommended in SECY-12-0025, 
``Proposed Orders and Requests for Information in Response to Lessons 
Learned From Japan's March 11, 2011, Great T[omacr]hoku Earthquake and 
Tsunami,'' that the Commission issue orders and requests for 
information.
    To address Tier 1 NTTF recommendation 4, the NRC issued Order EA-
12-049 on March 12, 2012, requiring all U.S. nuclear power plant 
licensees to implement strategies that would allow them to cope without 
their permanent electrical power sources for an indefinite period of 
time. These strategies would provide additional capability to maintain 
or restore reactor core and spent fuel cooling, as well as protect the 
reactor containment. This order also addressed: portions of NTTF 
recommendation 9 to require that facility emergency plans address 
prolonged station blackouts and multi-

[[Page 70613]]

unit events; portions of NTTF recommendation 10 to pursue additional 
emergency protection topics related to multi-unit events and prolonged 
station blackout; and portions of NTTF recommendation 11 to pursue 
emergency procedure topics related to decision-making, radiation 
monitoring, and public education.
    To address Tier 1 NTTF recommendation 7, the NRC issued Order EA-
12-051 on March 12, 2012, requiring all U.S. nuclear power plant 
licensees to have a reliable indication of the water level in 
associated spent fuel storage pools.
    To address Tier 1 NTTF recommendation 8, the NRC issued an ANPR on 
April 18, 2012 (77 FR 23161), to engage stakeholders in rulemaking 
activities associated with the methodology for integration of onsite 
emergency response processes, procedures, training and exercises.

D. Consolidation of Regulatory Efforts

    While developing the NTTF rulemakings, the NRC staff recognized 
that efficiencies could be gained by consolidating the rulemaking 
efforts due to the inter-relationships among the proposed changes. The 
NRC staff recommended to the Commission in COMSECY-13-0002, 
``Consolidation of Japan Lessons Learned Near-Term Task Force 
Recommendations 4 and 7 Regulatory Activities,'' COMSECY-13-0010, 
``Schedule and Plans for Tier 2 Order on Emergency Preparedness for 
Japan Lessons Learned,'' and SECY-14-0046, ``Fifth 6-Month Status 
Update on Response to Lessons Learned From Japan's March 11, 2011, 
Great Tohoku Earthquake and Subsequent Tsunami,'' the consolidation of 
rulemaking activities that address NTTF recommendations 4, 7, 8, 
portions of 9, 10.2, and 11.1. Section II.B of this document contains a 
more complete discussion of the scope of NTTF recommendations addressed 
by this proposed rule. The Commission approved these consolidations in 
the associated SRMs. These consolidations were intended to:
    1. Align the proposed regulatory framework with ongoing industry 
implementation efforts to produce a more coherent and understandable 
regulatory framework. Given the complexity of these requirements and 
their associated implementation, the NRC concluded that this is an 
important objective for the regulatory framework.
    2. Reduce the potential for inconsistencies and complexities 
between the related rulemaking actions that could occur if the efforts 
remained as separate rulemakings.
    3. Facilitate better understanding of the proposed requirements for 
both internal and external stakeholders, and thereby lessen the impact 
on internal and external stakeholders who would otherwise need to 
review and comment on multiple rulemakings while cross-referencing both 
proposed rules and sets of guidance documents.

E. Public Involvement

    This proposed rule consolidates two previous rulemaking efforts: 
The Station Blackout Mitigation Strategies rulemaking, directed by SRM-
COMSECY-13-0002, and the Onsite Emergency Response Capabilities 
rulemaking, which implemented NTTF recommendation 8. Both regulatory 
efforts offered extensive external stakeholder involvement 
opportunities, including public meetings, ANPRs issued for public 
comment, and draft regulatory basis documents issued for public 
comment. The major opportunities for stakeholder involvement were:
    1. Station Blackout ANPR (77 FR 16175; March 20, 2012);
    2. Onsite Emergency Response Capabilities ANPR (77 FR 23161; April 
18, 2012);
    3. Station Blackout Mitigation Strategies draft regulatory basis 
and draft rule concepts (78 FR 21275; April 10, 2013). The final 
Station Blackout Mitigation Strategies regulatory basis was 
subsequently issued on July 23, 2013 (78 FR 44035); and
    4. Onsite Emergency Response Capabilities draft regulatory basis 
(78 FR 1154; January 8, 2013). The final Onsite Emergency Response 
Capabilities regulatory basis, with preliminary proposed rule language, 
was subsequently issued on October 25, 2013 (78 FR 63901).
    The NRC described in each final regulatory basis document how it 
considered stakeholder feedback in developing the respective final 
regulatory basis, including consideration of ANPR comments and draft 
regulatory basis document comments. Section 5 of the Station Blackout 
Mitigation Strategies regulatory basis document includes a discussion 
of stakeholder feedback used to develop the final regulatory basis. 
Appendix B to the Onsite Emergency Response Capabilities regulatory 
basis includes a discussion of stakeholder feedback used to develop 
that final regulatory basis.
    The public has had multiple opportunities to engage in these 
regulatory efforts. Most noteworthy were the following:
    1. Preliminary proposed rule language for Onsite Emergency Response 
Capabilities made available to the public on November 15, 2013 (78 FR 
68774).
    2. Consolidated rulemaking proof of concept language made available 
to the public on February 21, 2014.
    3. Preliminary proposed rule language for Mitigation of Beyond-
Design-Basis Events rulemaking made available to the public on August 
15, 2014.
    4. Preliminary proposed rule language for Mitigation of Beyond-
Design-Basis Events rulemaking made available to the public on November 
13, 2014, and December 8, 2014, to support public discussion with the 
Advisory Committee on Reactor Safeguards (ACRS).
    The NRC staff has had numerous interactions with the ACRS, and in 
all cases these were public meetings, including the following:
    1. The ACRS Plant Operations and Fire Protection subcommittee met 
on February 6, 2013, to discuss the Onsite Emergency Response 
Capabilities regulatory basis.
    2. The ACRS Regulatory Policies and Practices subcommittee met on 
December 5, 2013, and April 23, 2013, to discuss the Station Blackout 
Mitigation Strategies regulatory basis.
    3. The ACRS full committee met on June 5, 2013, to discuss the 
Station Blackout Mitigation Strategies regulatory basis.
    4. The ACRS Fukushima subcommittee met on June 23, 2014, to discuss 
consolidation of Station Blackout Mitigation Strategies and Onsite 
Emergency Response Capabilities rulemakings.
    5. The ACRS full committee met on July 10, 2014, to discuss 
consolidation of Station Blackout Mitigation Strategies and Onsite 
Emergency Response Capabilities rulemakings.
    6. The ACRS Fukushima subcommittee met on November 21, 2014, to 
discuss preliminary proposed Mitigation of Beyond-Design-Basis Events 
rulemaking language.
    7. The ACRS Fukushima full committee met on December 4, 2014, to 
discuss preliminary proposed Mitigation of Beyond-Design-Basis Events 
rulemaking language.
    The NRC held many additional public meetings that have supported 
the development of this proposed rule. Notwithstanding these efforts to 
engage the public during the preparation of this proposed rule, the 
Commission is committed to the rigors of the notice-and-comment process 
enacted by the Administrative Procedures Act, and is providing members 
of the public a 90-

[[Page 70614]]

day comment period on the requirements NRC is proposing today.

III. Petitions for Rulemaking

    During development of this proposed rule, the NRC gave 
consideration to the issues raised in six petitions for rulemaking 
(PRMs) submitted to the NRC, five from the Natural Resources Defense 
Council Inc. (NRDC) (PRM-50-97, PRM-50-98, PRM-50-100, PRM-50-101, and 
PRM-50-102), and one submitted by Mr. Thomas Popik (PRM-50-96). The 
petitions filed by the NRDC use the NTTF Report as the sole basis for 
the PRMs. The NTTF recommendations that the NRDC PRMs rely upon are: 
4.1, 7.5, 8.4, 9.1, and 9.2. This proposed rule addresses each of these 
recommendations, and therefore it would resolve the issues raised by 
the NRDC PRMs. The NRDC petitions were dated July 26, 2011, and 
docketed by the NRC on July 28, 2011. The NRC published a notice of 
receipt in the Federal Register on September 20, 2011 (76 FR 58165), 
and did not ask for public comment at that time.
    In PRM-50-97 (NRC-2011-0189), the NRDC requested emergency 
preparedness enhancements for prolonged station blackouts in the areas 
of communications ability, Emergency Response Data System (ERDS) 
capability, training and exercises and equipment and facilities (NTTF 
recommendation 9.2). The NRC determined that the issues raised in this 
PRM should be considered in the NRC's rulemaking process. The NRC's 
consideration of the issues raised in PRM-50-97 are reflected in the 
proposed provisions in Sec.  50.155(d) and (e), and the proposed 
amendments to appendix E in both section VI and in new section VII, 
``Communications and Staffing Requirements for the Mitigation of Beyond 
Design Basis Events.'' The NRC concludes that consideration of the PRM 
issues, as discussed herein, would address PRM-50-97. The NRC is 
closing the docket for this petition and intends to take final action 
on this petition in the Federal Register notice the NRC issues for the 
final Mitigation of Beyond-Design-Basis Events rule.
    In PRM-50-98 (NRC-2011-0189), the NRDC requested emergency 
preparedness enhancements for multi-unit events in the areas of 
personnel staffing, dose assessment capability, training and exercises, 
and equipment and facilities (NTTF recommendation 9.1). The NRC 
determined that the issues raised in this PRM should be considered in 
the NRC's rulemaking process. The NRC's consideration of the issues 
raised in PRM-50-98 are reflected in the proposed provisions in Sec.  
50.155(b)(4), (d), and (e); and the proposed amendment to appendix E in 
section IV as well as the addition of a new section VII. The NRC 
concludes that consideration of the PRM issues, as discussed herein, 
would address PRM-50-98. The NRC is closing the docket for this 
petition and intends to take final action on this petition in the 
Federal Register notice the NRC issues for the final Mitigation of 
Beyond-Design-Basis Events rule.
    In PRM-50-100, the NRDC requested enhancement of spent fuel pool 
makeup capability and instrumentation for the spent fuel pool (NTTF 
recommendation 7.5). The NRC determined that the issues raised in this 
PRM should be considered in the NRC's rulemaking process, and the NRC 
published a document in the Federal Register with this determination on 
July 23, 2013 (78 FR 44034). The NRC's consideration of the issues 
raised in PRM-50-100 are reflected in the proposed provisions in Sec.  
50.155(b)(1) and (c)(4). This proposed rule would make generically 
applicable NRC's Order EA-12-051, ``Spent Fuel Pool Instrumentation.'' 
The NRC concludes that consideration of the PRM issues, as discussed 
herein, would address PRM-50-100. The NRC has already closed the docket 
for this petition and intends to take final action on this petition in 
the Federal Register notice the NRC issues for the final Mitigation of 
Beyond-Design-Basis Events rule.
    In PRM-50-101, the NRDC requested that Sec.  50.63, ``Loss of all 
alternating current power,'' be revised to establish a minimum coping 
time of 8 hours for a loss of all alternating current (ac) power, 
establish the equipment, procedures, and training necessary to 
implement an extended loss of ac power (72 hours) for core and spent 
fuel pool cooling and for reactor coolant system and primary 
containment integrity as needed, and preplan/prestage offsite resources 
to support uninterrupted core and spent fuel pool cooling and reactor 
coolant system and containment integrity as needed (NTTF recommendation 
4.1). The NRC determined that the issues raised in this PRM should be 
considered in the NRC's rulemaking process, and the NRC published a 
document in the Federal Register with this determination on March 21, 
2012 (77 FR 16483). The NRC's consideration of the issues raised in 
PRM-50-101 is reflected in the proposed provisions in Sec.  
50.155(b)(1), (c), (d), (e), and (f). The NRC concludes that 
consideration of the PRM issues, as discussed herein, would address 
PRM-50-101. The NRC has already closed the docket for this petition and 
intends to take final action on this petition in the Federal Register 
notice the NRC issues for the final Mitigation of Beyond-Design-Basis 
Events rule.
    In PRM-50-102, the NRDC requested more realistic, hands-on training 
and exercises on SAMGs and EDMGs for licensee staff expected to 
implement those guideline sets and make decisions during emergencies 
(NTTF recommendation 8.4). The NRC determined that the issues raised in 
this PRM should be considered in the NRC's rulemaking process, and the 
NRC published a document in the Federal Register with this 
determination on April 27, 2012 (77 FR 25104). The NRC's consideration 
of the issues raised in PRM-50-102 are reflected in the proposed 
provisions in Sec.  50.155(d) and (e). The NRC concludes that 
consideration of the PRM issues, as discussed herein, would address 
PRM-50-102. The NRC has already closed the docket for this petition and 
intends to take final action on this petition in the Federal Register 
notice the NRC issues for the final Mitigation of Beyond-Design-Basis 
Events rule.
    In PRM-50-96, Mr. Thomas Popik requested that the NRC amend its 
regulations to require facilities licensed by the NRC to assure long-
term cooling and unattended water makeup of spent fuel pools in the 
event of geomagnetic storms caused by solar storms resulting in long-
term losses of power. The NRC determined that the issues raised in this 
PRM should be considered in the NRC's rulemaking process and the NRC 
published a document in the Federal Register with this determination on 
December 18, 2012 (77 FR 74788). In that Federal Register document, the 
NRC also closed the docket for this petition. Specifically, the NRC 
indicated that it would monitor the progress of the mitigation 
strategies rulemaking to determine whether the requirements established 
would address, in whole or in part, the issues raised in the PRM. In 
this context, the proposed requirements in Sec.  50.155(b)(1) and (c) 
and the associated draft regulatory guidance should address, in part, 
the issues raised because these actions would establish offsite 
assistance to support maintenance of the key functions (including both 
reactor and spent fuel pool cooling) following an extended loss of ac 
power that has been postulated for geomagnetic events. Additional 
consideration of these issues will result from NRC's participation in 
the interagency task force developing a National Space Weather Strategy 
and the associated action plan. Both the strategy and action plan are 
expected to be completed in 2015. When the

[[Page 70615]]

National plans are completed, the NRC will reevaluate the need for 
additional actions to address the impact of geomagnetic storms on 
nuclear power plants within the overall context of the National Space 
Weather Strategy and action plan.

IV. Discussion

A. Rulemaking Objectives

    The regulatory objectives of this rulemaking are to: (1) Make the 
requirements in Order EA-12-049 and Order EA-12-051 generically 
applicable, giving consideration to lessons learned from implementation 
of the orders; (2) establish new requirements for an integrated 
response capability; (3) establish new requirements for actions that 
are related to onsite emergency response; and (4) address issues raised 
by PRMs that were submitted to the NRC following the March 2011 
Fukushima Dai-ichi event.
    1. Make the requirements in Order EA-12-049 and Order EA-12-051 
generically applicable, giving consideration to lessons learned from 
implementation of the orders.
    An objective of this rulemaking is to place the requirements in 
Order EA-12-049 and Order EA-12-051 into the NRC's regulations so that 
they apply to all current and future power reactor applicants, and to 
provide regulatory clarity and stability to power reactor licensees. In 
making the requirements of Order EA-12-049 generically-applicable, this 
proposed rule would also consider the reevaluated hazard information 
developed in response to the March 12, 2012, NRC letter issued under 
Sec.  50.54(f) as part of providing reasonable protection for 
mitigation strategies equipment against external flooding or seismic 
hazards. Because these orders were issued to current licensees, the 
requirements of these orders would not apply to future licensees. In 
the absence of this proposed rule, these requirements would need to be 
implemented for new reactor applicants or licensees through additional 
orders or license conditions (as was done for the Vogtle Electric 
Generating Plant, Units 3 and 4, Virgil C. Summer Nuclear Station, 
Units 2 and 3, and Enrico Fermi Nuclear Plant, Unit 3, combined 
licenses (COLs), respectively). As part of the rulemaking, the NRC 
considered stakeholder feedback and lessons-learned from the 
implementation of the orders, including any challenges or unintended 
consequences associated with implementation. The NRC reflected this 
stakeholder input in the draft regulatory guidance for this proposed 
rule.
    2. Establish new requirements for an integrated response 
capability.
    An objective of this rulemaking is to establish requirements for an 
integrated response capability for beyond-design-basis events that 
would integrate existing strategies and guidelines (implemented through 
guideline sets) with the existing EOPs. This would include guideline 
sets that implement the requirements of current Sec.  50.54(hh)(2) and 
Order EA-12-049. This proposed rule would require sufficient staffing, 
command and control, training, drills, and change control to support 
the integrated response capability.
    3. Establish new requirements for actions that are related to 
onsite emergency response.
    An objective of this rulemaking is to establish requirements for 
onsite emergency response capabilities being implemented in conjunction 
with the implementation of Order EA-12-049. This proposed rule contains 
new requirements for staffing and communications assessment, and 
clarifies requirements for multiple source term dose assessment.
    4. Address a number of PRMs submitted to the NRC following the 
March 2011 Fukushima Dai-ichi event.
    An objective of this rulemaking is to address the five PRMs filed 
by the NRDC that raise issues that pertain to the technical objectives 
of this rulemaking. The petitions rely solely on the NTTF Report, and 
request that the NRC undertake rulemaking in a number of areas that 
would be addressed by this proposed rule. This proposed rule would also 
address, in part, the PRM submitted by Mr. Thomas Popik.

B. Rulemaking Scope

    The scope of this rulemaking, described in terms of the 
relationship to various NTTF recommendations that provided the 
regulatory impetus for this proposed rule, includes:
    1. All the requirements that were within the scope of Station 
Blackout Mitigation Strategies rulemaking. These requirements address 
NTTF recommendations 4 and 7. This aspect of the proposed rule would 
also address NTTF recommendation 11.1 regarding onsite emergency 
resources to support multi-unit events with station blackout, including 
the need to deliver equipment to the site despite degraded offsite 
infrastructure. This provision currently is being implemented through 
Order EA-12-049.
    2. All the requirements that were within the scope of the Onsite 
Emergency Response Capabilities rulemaking. These requirements address 
NTTF recommendation 8, as directed by SRM-SECY-11-0137. This aspect of 
this proposed rule also would address command and control issues in 
NTTF recommendation 10.2.
    3. Numerous requirements regarding onsite emergency response 
actions being implemented by Order EA-12-049; in addition, NRC staff 
has developed draft guidance to support the emergency response aspect 
of this proposed rule. The specific regulatory actions related to 
emergency response in this proposed rule and the associated NTTF 
recommendations are:
    a. Staffing and communications requirements: would address NTTF 
recommendation 9.3; also discussed in NTTF recommendations 9.1 and 9.2. 
These regulatory issues currently are being implemented through Order 
EA-12-049. The proposed requirements also address supporting facilities 
and equipment, as discussed in the same NTTF recommendations.
    b. Multiple source term dose assessment requirements: would address 
NTTF recommendation 9.3; also discussed in NTTF recommendation 9.1. 
This regulatory issue is being implemented voluntarily by industry.
    c. Training and exercise requirements: would address NTTF 
recommendation 9.3; also discussed in NTTF recommendations 9.1 and 9.2. 
These regulatory issues currently are being implemented through Order 
EA-12-049.
    Accordingly, this rulemaking would address all the justifiable 
recommendations in NTTF recommendations 4, 7, 8, 9.1, 9.2, 9.3 (with 
one exception--ERDS modernization is addressed, but maintenance of ERDS 
capability throughout the accident is not addressed), 10.2, and 11.1.
    This rulemaking also would address NTTF recommendation, 9.4: 
modernize ERDS. This action differs from the other regulatory actions 
because ERDS is not an essential component of a licensee's capability 
to mitigate a beyond-design-basis external event. However, ERDS is an 
important form of communication between the licensee and the NRC. 
Modernization of ERDS has been completed voluntarily by industry; 
therefore, NRC has included amendments to remove the technology-
specific references in 10 CFR part 50, appendix E, section VI, 
``Emergency Response Data System,'' in this proposed rule.
SAMG Implementation
    Unlike the requirements for the mitigation of beyond-design-basis 
external events imposed by Order EA-

[[Page 70616]]

12-049, and requirements that address the loss of large areas of the 
plant due to explosions and fire in current Sec.  50.54(hh)(2) (NRC is 
proposing in this rule to move these requirements to a new section), 
SAMGs are not an NRC requirement imposed on licensees. Nevertheless, 
SAMGs are well established guidance documents that have been developed 
by the nuclear power industry with substantial NRC involvement, have 
been implemented by every operating nuclear power reactor licensee for 
decades, and are the subject of a license condition for combined 
licenses. Following the Three Mile Island (TMI) accident in 1979, the 
nuclear power industry revised its emergency response procedures to be 
symptom-based, and as a result, developed EOPs. In the mid-to-late 
1980s, the NRC and the nuclear power industry identified a need to 
consider plant conditions that could lead to a severe accident. These 
efforts led to the nuclear industry voluntarily initiating a 
coordinated program on severe accident management in 1990. Section 5 of 
Nuclear Energy Institute (NEI) 91-04 (formerly Nuclear Management and 
Resources Council (NUMARC) 91-04), Revision 1, ``Severe Accident 
Closure Guidelines,'' describes the elements of the industry's severe 
accident management closure actions. The program involves the 
development of: (1) A structured method by which utilities could 
systematically evaluate and enhance their ability to deal with 
potential severe accidents, (2) vendor-specific SAMGs for use by 
licensees in developing plant-specific SAMGs, and (3) guidance and 
material to support utility activities related to training for severe 
accidents. In 1992, the Electric Power Research Institute (EPRI) 
developed the SAMG Technical Basis Report (TBR). Volume one of this 
report covers general actions that could be taken to manage a severe 
accident (referred to as SAMG candidate high level actions) and their 
effects, and volume two is a detailed report on the physics of accident 
progression. By letter dated June 20, 1994, the NRC accepted the 
industry's approach for mitigating the consequences of severe 
accidents, including licensee regulatory commitments to implement 
plant-specific SAMGs, using the guidance developed in section 5 of NEI 
91-04, Revision 1, by December 31, 1998.
    The NRC assessed the ongoing implementation of SAMGs at a select 
number of plants during the 1997-1998 time frame as discussed in SECY-
97-132, ``Status of the Integration Plan for Closure of Severe Accident 
Issues and the Status of Severe Accident Research,'' and SECY-98-131, 
``Status of the Integration Plan for Closure of Severe Accident Issues 
and the Status of Severe Accident Research,'' and concluded that the 
results of the voluntary initiative achieved the NRC's overall 
objectives established for accident management in SECY-89-012, ``Staff 
Plans for Accident Management Regulatory and Research Programs.'' In 
2012, EPRI revised the TBR to account for the initial lessons learned 
from the Fukushima Dai-ichi accidents, as well as enhanced 
understanding of severe accident behavior gained from additional 
research and analyses performed since the original report was 
published.
    Following the events at Fukushima Dai-ichi, the NRC again inspected 
the implementation, ongoing training, and maintenance of licensee SAMGs 
at all power reactor sites, except those that had permanently ceased 
operation, through performance of Temporary Instruction (TI)-2515/184, 
``Availability and Readiness Inspection of Severe Accident Management 
Guidelines (SAMGs).'' The NRC found that some licensees had not 
maintained the SAMGs in accordance with the latest revisions of the 
applicable industry generic technical guidelines nor conducted training 
in a consistent and systematic manner. The NRC inspectors attributed 
the inconsistent implementation and training on SAMGs to the voluntary 
nature of this initiative.
    Based in part on the findings of the inspections previously 
described, the NTTF recommended that the NRC require licensees to 
integrate onsite emergency response capabilities, including SAMGs. 
Unlike the Mitigating Strategies Order requirements, which were 
justified as necessary for adequate protection under Sec.  50.109, 
SAMGs do not involve adequate protection. Because the imposition of 
SAMGs also would not be necessary to bring licensees into compliance 
with an existing NRC requirement, a SAMGs requirement would have to be 
justified under Sec.  50.109 as a cost-justified, substantial increase 
in protection of the public health and safety or common defense and 
security.
    In the regulatory analysis where the NRC considered an option to 
require SAMGs (i.e., option 2 of the regulatory analysis including the 
supporting proposed backfit justification), the NRC used available 
quantified risk information that might provide risk insights to inform 
the justification. In this regard, the NRC looked at its recent 
technical analysis \1\ performed in support of the Containment 
Protection and Release Reduction (CPRR) rulemaking regulatory basis.\2\ 
This analysis is relevant because it examined regulatory alternatives 
that would be implemented after core damage to determine whether any of 
the contemplated approaches can be justified under the NRC's 
backfitting provisions. In this respect, the risk insights stemming 
from this work might have relevance to NRC's consideration of SAMG 
requirements where the safety benefits would occur after core damage. 
The NRC also considered other post-Fukushima regulatory efforts (e.g., 
the safety benefits due to implementation of Order EA-12-049 mitigation 
strategies, which result in a reduction in core damage frequency) 
within this technical analysis. The NRC acknowledges that the work to 
support the CPRR rulemaking was not conducted to provide a complete 
quantitative measure of the possible safety benefits of SAMG 
requirements, particularly with regard to how SAMGs might benefit 
maintenance of containment integrity or support more informed 
protective action recommendations by the emergency response 
organization following core damage. However, this technical analysis 
work does provide valuable risk insights that the NRC concluded were 
important to fully inform the decision on this matter, and that 
additionally influenced the NRC's development of the SAMG framework 
considered in the regulatory analysis.
---------------------------------------------------------------------------

    \1\ The technical risk insights were presented to the ACRS 
Reliability and PRA, and Fukushima subcommittees on August 22, 2014, 
and to the ACRS Reliability and PRA subcommittee on November 19, 
2014. This footnote is informational only; it does not imply 
advisory committee endorsement of the technical analysis.
    \2\ Refer to the draft regulatory basis for Containment 
Protection and Release Reduction.
---------------------------------------------------------------------------

    The CPRR technical analysis includes a screening for a conservative 
high estimate of frequency-weighted individual latent cancer fatality 
risk. This screening analysis combined the highest ELAP frequency among 
all boiling water reactors (BWRs) with Mark I or Mark II containments, 
a success probability in the FLEX equipment \3\ of 0.6 per demand 
following core melt, the highest conditional individual latent cancer 
fatality (ILCF) risk among all BWRs with Mark I or Mark II 
containments, and a worst case re-habitability assumption. This yields 
a conservative high estimate of frequency-weighted individual latent

[[Page 70617]]

cancer fatality risk of approximately 7 x 10 -8 per reactor 
year. This combination of assumptions does not exist at any BWR with a 
Mark I or Mark II containment. This conservative estimate of the risk 
can be viewed as the maximum possible risk that could be removed or 
reduced through regulatory action (i.e., the CPRR technical analysis 
examines a range of post-core damage regulatory actions for BWRs with 
Mark I or Mark II containments to identify whether any of these 
proposals might result in a safety benefit large enough to be justified 
under the Commission's backfitting requirements). This estimate is 
compared against the quantitative health objective, which is a 
quantitative measure that equates to \1/10\ of 1 percent of the ILCF 
risk and relates to the Commission's Safety Goal Policy. This 
quantitative metric for the individual latent cancer fatality risk is 
approximately 2 x 10-6 per reactor year. This technical work 
shows that the risk is well below a level that equates to \1/10\ of 1 
percent of the surrounding population's latent cancer fatality risk. 
This result also means, that, from a quantitative standpoint, achieving 
risk reductions that might satisfy backfitting requirements is very 
unlikely. More refined risk estimates from the same work (i.e., which 
remove the worst case assumptions and instead use assumptions specific 
to each power reactor), push this potential risk benefit significantly 
lower, by approximately two orders of magnitude. This result 
demonstrates the benefits of the NRC's regulations to both effectively 
keep the frequency of core damage very low at BWRs with Mark I and II 
containments, and to ensure through emergency preparedness requirements 
that the surrounding population is adequately protected. Those general 
attributes of the NRC's regulations that result in this risk insight 
(i.e., requirements that resulted in reduced core damage frequencies 
and effective emergency preparedness requirements) apply to all power 
reactor designs. The NRC has not performed a comprehensive quantitative 
analysis of the potential safety benefits of SAMG requirements for all 
types of reactors. However, the general risk insights obtained from the 
CPRR work align well with NUREG-1935, ``State-of-the-Art Reactor 
Consequence Analyses (SOARCA) Report,'' (November 2012), which shows 
very low levels of risk (e.g., individual early fatality risk is 
essentially zero, ILCF risk is thousands of times lower than the NRC 
Safety Goal, and millions of times lower than the general cancer 
fatality risk in the United States from all causes). As such, the 
available risk insights point to the likely outcome that a 
comprehensive quantitative analysis, where the proposed regulatory 
action is intended to provide its safety benefit in the post-core 
damage environment (as is the case for use of SAMGs), would not 
demonstrate a substantial safety benefit. In addition, for the specific 
case of the consideration of SAMG requirements in this proposed rule, 
the proposed regulatory action's benefit must also recognize that 
imposing SAMG requirements must be compared with the current regulatory 
state, (i.e., SAMGs) exist and are voluntarily in use under an industry 
initiative.
---------------------------------------------------------------------------

    \3\ Refer to NEI 12-06, Revision 0, ``Diverse and Flexible 
Coping Strategies (FLEX) Implementation Guide,'' for a description 
of industry-developed guidance on FLEX strategies and equipment.
---------------------------------------------------------------------------

    Along with its quantitative analysis, the Commission considered a 
proposed SAMG backfit analysis that relied on qualitative factors, 
relating SAMGs to defense-in-depth. The Commission concluded that the 
imposition of SAMG requirements was not warranted as it did not meet 
the substantial additional protection criteria under 10 CFR 
50.109(a)(3), and consequently SAMGs will continue to be implemented 
and maintained through a voluntary industry initiative. The Commission 
notes that the industry indicated it would strengthen its voluntary 
initiative for SAMGs in its letter dated May 11, 2015.
Scope of Procedure and Guideline Integration
    This rulemaking limits the scope of the integrated response 
capability to two guideline sets. This proposed rule includes these new 
provisions:
    1. Sec.  50.155(b)(1), resulting from Order EA-12-049, and 
addressing beyond-design-basis external events; these requirements are 
those that the NRC termed in previous regulatory basis interactions as 
``Station Blackout Mitigation Strategies.'' The nuclear industry refers 
to these as ``FLEX Support Guidelines'' (FSGs).
    2. Sec.  50.155(b)(2) (current Sec.  50.54(hh)(2)). These 
requirements are defined in NEI 06-12, Revision 2, ``B.5.b Phase 2 & 3 
Submittal Guideline,'' as a subset of the strategies and guidelines for 
addressing the loss of large areas of the plant due to explosions and 
fires and are termed ``Extensive Damage Mitigation Guidelines.'' The 
NRC proposes to expand the scope of the generic term ``EDMGs'' to 
include all of the strategies and guidelines used to implement Sec.  
50.54(hh)(2).
    The NRC is proposing this integrated response capability structure 
to avoid unnecessarily revisiting the existing symptom-based EOPs that 
were developed following the TMI accident. The NRC has determined that 
current regulations addressing EOPs, which include the quality 
assurance requirements of criterion V, ``Instructions, Procedures, and 
Drawings,'' and criterion VI, ``Document Control,'' in appendix B to 10 
CFR part 50, and the administrative controls section of the technical 
specifications for each plant as well as the guidance provided in 
regulatory guides and technical reports (e.g., NUREG-0660, ``NRC Action 
Plan Developed as a Result of the TMI-2 Accident,'' issued May 1980; 
NUREG-0737, ``Clarification of TMI Action Plan Requirements,'' issued 
November 1980; and NUREG-0711, ``Human Factors Engineering Program 
Review Model,'' issued November 2012) provide sufficient regulation and 
control of the EOPs to provide reasonable assurance of adequate 
protection of public health and safety. In addition, the EOPs are the 
subject of a national consensus standard (American National Standards 
Institute/American Nuclear Society 3.2 1994, ``Administrative Controls 
and Quality Assurance for the Operational Phase of Nuclear Power 
Plants''). In order to avoid the unnecessary regulatory burden that 
would result by restructuring the EOPs, proposed Sec.  50.155(b)(3) 
would require that the FSGs, and EDMGs be integrated with the EOPs, 
rather than moving the requirements for EOPs to Sec.  50.155.
Guideline Sets Excluded From This Proposed Rule
    During the development of this proposed rule, other guideline sets 
were considered for inclusion within the integrated response 
capability. The guideline sets considered included fire response 
procedures, alarm response procedures (ARPs), and abnormal operating 
procedures (AOPs).
    Similar to the EOPs, ARPs and AOPs are subject to existing NRC 
regulations (e.g., 10 CFR part 50, appendix B, criteria V and VI) that 
adequately ensure integration with other procedure sets in use at power 
reactors. These procedures have been used by operating power reactor 
licensees in actual and simulated events for many years; any further 
integration effort to address potential issues would likely have 
already been identified and corrected by existing processes (or will be 
identified and corrected under the quality assurance program).
    The issue of whether to include fire response procedures in the 
scope of proposed Sec.  50.155(b) was initially raised as 
recommendation 1.g. by the ACRS in its letter to the then-Chairman 
Jaczko dated October 13, 2011, ``Initial ACRS Review of: (1) The NRC 
Near-Term Task

[[Page 70618]]

Force Report on Fukushima and (2) Staff's Recommended Actions to be 
Taken Without Delay.'' That letter expressed the ACRS view that:

    [The] efforts to integrate the onsite emergency response 
capabilities should be expanded to include the plant fire response 
procedures. These procedures provide operator guidance for coping 
with fires that are beyond a plant's original design basis. Some 
plant-specific fire response procedures instruct operators to 
manually de-energize major electrical buses and realign fluid 
systems in configurations that may not be consistent with the 
guidance or expectations in the EOPs. Experience from actual fire 
events has shown that parallel execution of fire procedures, 
Abnormal Operating Procedures (AOPs), and EOPs can be difficult and 
can introduce operational complexity. Therefore, these procedures 
should also be included in the comprehensive efforts to better 
coordinate and integrate operator responses during challenging plant 
conditions.

    This recommendation was reiterated in the ACRS letter of November 
8, 2011, ``ACRS Review of Staff's Prioritization of Recommended Actions 
to Be Taken in Response to Fukushima Lessons Learned (SECY-11-0137).''
    In SECY-12-0025, enclosure 3, the NRC documented the formal process 
used in evaluating additional recommendations that were made by the 
ACRS as follows:

    The staff developed a process to disposition all additional 
issues, including recommendations by the ACRS. All issues are 
reviewed by a panel of senior-level advisors from different NRC 
program offices. The panel determines whether each issue represents 
a valid safety concern, and whether there is a clear nexus to the 
Fukushima Dai-ichi accident. If neither criterion is met, or only 
one criterion is met, the panel chooses to either disposition the 
issue with no action, or direct it to one of the NRC's existing 
regulatory processes (e.g., generic issue process). If both criteria 
are met, the issue is forwarded for further consideration by the 
cognizant technical staff in the appropriate NRC line organization. 
Should the issue go forward, the cognizant technical staff is tasked 
with developing a proposal for Steering Committee (SC) disposition. 
The SC may elect to take no further action, disposition the issue 
using an existing NRC process, or prioritize the issue as a Tier 1, 
2, or 3 item under the Japan Lessons-Learned Program.

    By letter dated February 27, 2012, the NRC responded to the ACRS 
recommendations of October 13, 2011, and November 8, 2011, discussing 
the disposition of ACRS recommendation 1.g. as follows:

    The NRC staff evaluated how to appropriately integrate the fire 
response procedure into a licensee's onsite emergency response 
capabilities and determined that the fire response procedures would 
be best considered with the agency's Tier 3 actions associated with 
NTTF Recommendation 3.

    This disposition of the ACRS recommendation also was documented in 
SECY-12-0025. In its letter of March 13, 2012, the ACRS acknowledged 
that the formal screening process used by the NRC for additional 
recommendations was acceptable, but nevertheless expressed the view 
that integration of the fire response procedures presents similar 
challenges to those associated with the integration of other guideline 
sets such as the EDMGs with the EOPs. Accordingly, the ACRS recommended 
that the integration effort should address fire response procedures as 
part of NTTF recommendation 8 rather than as a seismic-induced-fire 
issue under NTTF recommendation 3.
    Recognizing the continued ACRS interest in the integration of fire 
response procedures with onsite emergency actions and the existence of 
an additional program of work to be taken up on the ACRS 
recommendation, the NRC has concluded that the reasoning underlying the 
initial prioritization of ACRS recommendation 1.g was sound and it 
would be inappropriate to include fire response procedure integration 
within this rulemaking effort. The NRC offers the following reasons for 
the exclusion of firefighting strategies and procedures from the scope 
of integration in this rulemaking:
    1. The NRC-required fire protection program is designed to function 
autonomously from other ongoing activities and is implemented by a fire 
brigade that is manned in all modes of operation and is well-trained. 
Firefighting activities are led by personnel knowledgeable of overall 
plant operations, including the equipment necessary for safe shutdown 
of the plant. These personnel communicate with the main control room in 
order to prioritize and deconflict activities.
    2. Comprehensive firefighting strategies and implementing 
procedures have been developed for each area of the plant and fire 
brigade qualified individuals participate in drills on a quarterly 
basis to demonstrate proficiency with the use of these strategies and 
procedures in the context of concurrent use of other, non-integrated 
procedures throughout the plant.
    3. The EOPs, EDMGs, and FSGs account for equipment lost due to 
concurrent fires during events by providing alternate methods to 
accomplish the functions the equipment was to have performed.

C. Proposed Rule Organization

    To accomplish the NRC's rulemaking objectives in a manner 
consistent with the described scope, this proposed rule has been based 
on these precepts:
    1. The central requirement would be an integrated response 
capability that includes currently existing procedures and guideline 
sets. Additional requirements would support this integrated response 
capability.
    The mitigation strategies under Order EA-12-049 established the 
basic framework for broader capability to mitigate beyond-design-basis 
external events that impact an entire reactor site. This framework 
includes: Supporting drills, training, change control, staffing, 
communications capability, multiple source term dose assessment 
capability, and command and control. As a result, the proposed new 
Sec.  50.155 is structured to have:
    1. Integrated response requirements in paragraph (b).
    2. Supporting equipment requirements in paragraph (c) that include 
equipment required by both Order EA-12-049 and Order EA-12-051.
    3. External hazard equipment protection requirements in paragraph 
(c) that reflect the hazard information developed under the Sec.  
50.54(f) letter of March 12, 2012.
    4. Supporting training, drills, and change control requirements in 
paragraphs (d), (e), and (f).
    5. Implementation requirements that establish compliance deadlines 
in paragraph (g).
    In addition to proposed Sec.  50.155, this proposed rulemaking is 
structured to have (1) supporting power reactor operating license 
application requirements (under either 10 CFR parts 50 or 52 processes) 
in the appropriate content of applications portions, and (2) 
requirements that relate to enhanced onsite emergency response 
capabilities located in appendix E to 10 CFR part 50, to include a new 
section VII.
    The proposed requirements previously described would apply to both 
current licensees and new applicants (under either 10 CFR parts 50 or 
52) as established by proposed paragraph Sec.  50.155 (a). Finally, 
this proposed rule contains provisions to facilitate power reactor 
decommissioning.

D. Proposed Rule Regulatory Bases

Applicability
    This proposed rule would apply, in whole or in part, to applicants 
for and holders of an operating license for a nuclear power reactor 
under 10 CFR

[[Page 70619]]

part 50, or combined license under 10 CFR part 52.
    This proposed rule would not apply to applicants for, or holders 
of, an operating license for a non-power reactor under 10 CFR part 50. 
Non-power reactor licensees would not be subject to this proposed rule 
because non-power reactors pose lower radiological risks to the public 
from accidents than do power reactors because: (1) The core 
radionuclide inventories in non-power reactors are lower than in power 
reactors as a result of their lower power levels and often shorter 
operating cycle lengths; and (2) non-power reactors have lower decay 
heat associated with a lower risk of core melt and fission product 
release in a loss-of-coolant accident than power reactors.
    A holder of a general or specific 10 CFR part 72 independent spent 
fuel storage installation (ISFSI) license for dry cask storage would 
not be subject to this proposed rule for the ISFSI, because the decay 
heat load of the irradiated fuel would be sufficiently low prior to 
movement to dry cask storage that it could be air-cooled. This would 
meet the proposed sunsetting criteria (discussed later in this section 
of this document).
    The GE Morris facility in Illinois, which is the only spent fuel 
pool licensed under 10 CFR part 72 as an ISFSI would not need to comply 
with this proposed rule because it is excluded by the rule 
applicability described in proposed Sec.  50.155(a). The NRC considered 
including the GE Morris facility within the scope of this proposed rule 
but found that the age (and corresponding low decay heat load) of the 
fuel in the facility made it unnecessary. The GE Morris facility also 
would meet this proposed rule's sunsetting criteria. While this 
proposed rule would leave in force the requirements of the current 
Sec.  50.54(hh)(2), those requirements are not applicable to GE Morris 
due to its status as a non-10 CFR part 50 licensee. In the course of 
the development and implementation of the guidance and strategies 
required by the current Sec.  50.54(hh)(2), the NRC evaluated whether 
additional mitigation strategies were warranted at GE Morris and 
concluded that no mitigating strategies were warranted beyond existing 
measures, due to the extended decay time since the last criticality of 
the fuel stored there, the resulting low decay heat levels, and the 
assessment that a gravity drain of the GE Morris SFP is not possible 
due to the low permeability of the surrounding rock and the high level 
of upper strata groundwater.
    This proposed rule would establish a ``sunsetting'' or phased 
removal of requirements for licensees of decommissioning power 
reactors. Licensees would not need to meet requirements that relate to 
the reactor source term and associated fission product barriers once 
all fuel has been permanently removed from the reactor vessel and 
placed in the spent fuel pool. This proposed rule would require 
secondary containment for reactor designs that employ this feature as a 
fission product barrier for the spent fuel pool source term.
    Once the NRC has docketed a licensee's Sec.  50.82(a)(1) or Sec.  
52.110(a) certification of permanent removal of fuel from the reactor 
vessel and certification of permanent cessation of operations, that 
licensee would not be subject to requirements to have mitigation 
strategies and guidelines for maintaining or restoring core cooling and 
containment capabilities. As discussed previously, these proposed 
requirements are based on Order EA-12-049. The licensees for the 
Kewaunee Power Station, Crystal River Unit 3 Nuclear Generating Plant, 
San Onofre Nuclear Generating Station, Units 2 and 3, and Vermont 
Yankee Nuclear Power Station, submitted Sec.  50.82(a)(1) 
certifications after issuance of Order EA-12-049; the NRC has rescinded 
Order EA-12-049 to this group of NPP licensees (Shutdown NPP Group). 
These rescissions were based on the NRC's conclusion that the lack of 
fuel in the licensee's reactor core and the absence of challenges to 
the containment rendered unnecessary the development of guidance and 
strategies to maintain or restore core cooling and containment 
capabilities. Consistent with these rescissions, the NRC proposes to 
relieve licensees in decommissioning from the requirement to comply 
with proposed requirements to have mitigation strategies and guidelines 
to maintain or restore core cooling and containment capabilities. 
Moreover, these licensees would not need to comply with any of the 
other requirements in this proposed rule that support compliance with 
the proposed requirement to have mitigation strategies and guidelines 
for maintaining or restoring core cooling and containment capabilities.
    This proposed rule treats the EDMG requirements in a manner similar 
to the requirements for FSGs. For a licensee who has Sec.  50.82(a)(1) 
or Sec.  52.110(a) certifications docketed at the NRC, the lack of fuel 
in their reactor core and the absence of challenges to the containment 
would render unnecessary EDMGs for core cooling and containment 
capabilities. This licensee would not need to comply with any 
requirements in this proposed rule associated with core cooling or 
containment capabilities; rather, the licensee would be required to 
comply with the proposed requirement to have EDMGs as based on the 
presence of fuel in the spent fuel pool.
    Once the NRC has docketed a licensee's Sec.  50.82(a)(1) or Sec.  
52.110(a) certifications, that licensee would not need to comply with 
the requirement proposed by this rule that the equipment relied on for 
the mitigation strategies include reliable means to remotely monitor 
wide-range spent fuel pool levels to support effective prioritization 
of event mitigation and recovery actions. This proposed requirement is 
based on the requirements in Order EA-12-051. This order requires a 
reliable means of remotely monitoring wide-range SFP levels to support 
effective prioritization of event mitigation and recovery actions in 
the event of a beyond-design-basis external event with the potential to 
challenge both the reactor and SFP.
    The NRC has also rescinded Order EA-12-051 for the Shutdown NPP 
Group mentioned previously. These rescissions were based, in part, on 
the NRC's conclusions that once a licensee certifies the permanent 
removal of the fuel from its reactor vessel, the safety of the fuel in 
the SFP becomes the primary safety function for site personnel. In the 
event of a challenge to the safety of fuel stored in the SFP, decision-
makers would not have to prioritize actions and the focus of the staff 
would be the SFP condition. Therefore, once fuel is permanently removed 
from the reactor vessel, the basis for the Order EA-12-051 would no 
longer apply. Consistent with the NRC order rescissions, the NRC 
proposes to no longer require licensees in decommissioning to have a 
reliable means to remotely monitor wide-range spent fuel pool levels to 
support effective prioritization of event mitigation and recovery 
actions in the event of a beyond-design-basis external event with the 
potential to challenge both the reactor and SFP.
    Once the NRC has docketed a licensee's Sec.  50.82(a)(1) or Sec.  
52.110(a) certifications, that licensee would not need to comply with 
the requirements in proposed Section VII, ``Communications and Staffing 
Requirements for the Mitigation of Beyond Design Basis Events,'' in 10 
CFR part 50, appendix E. These proposed requirements are based on the 
March 12, 2012, Sec.  50.54(f) letters that requested operating power 
reactor licensees to perform, among other things, emergency 
preparedness communication and

[[Page 70620]]

staffing evaluations for prolonged loss of power events consistent with 
NTTF recommendation 9.3. Once the licensees for the Shutdown NPP Group 
were no longer operating power reactors, they informed the NRC that 
they would no longer proceed with implementing recommendation 9.3. In 
response to the filings, the NRC determined that, for beyond-design-
basis external events challenging the safety of the spent fuel at the 
Shutdown NPP Group:

recovery and mitigation actions could be completed over a long 
period of time due to the slow progression of any accident as a 
result of the very low decay heat levels present in the pool within 
a few months following permanent shutdown of the reactor. Thus, 
spent fuel pool beyond design basis accident scenarios at 
decommissioning reactor sites do not require the enhanced 
communication and staffing that may be necessary for the reactor-
centered events the 50.54(f) letter addresses.\4\
---------------------------------------------------------------------------

    \4\ See the ``Availability of Documents'' section of this 
document for the NRC letters to the licensees for Kewaunee Power 
Station, Crystal River Unit 3 Nuclear Generating Plant, San Onofre 
Nuclear Generating Station, Units 2 and 3, and Vermont Yankee 
Nuclear Power Station.

    Order EA-12-049 also required power reactor licensees to have 
certain spent fuel pool cooling capabilities. In the rescission letters 
to the licensees for the Shutdown NPP Group, the NRC determined that, 
due to the passage of time, the fuel's low decay heat and the long time 
to boil off the water inventory in the spent fuel pool obviated the 
need for the Shutdown NPP Group licensees to have guidance and 
strategies necessary for compliance with Order EA-12-049. The 
rescission of Order EA-12-049 for those licensees eliminated the 
requirement for them to comply with the Order's requirements concerning 
beyond-design-basis event strategies and guidelines for spent fuel pool 
cooling capabilities. Consistent with the basis for the Order 
rescissions, licensees in decommissioning could be relieved from the 
proposed requirements concerning beyond-design-basis event strategies 
and guidelines for spent fuel pool cooling capabilities and any related 
requirements. These licensees would have to perform and retain an 
analysis demonstrating that sufficient time has passed since the fuel 
within the spent fuel pool was last irradiated such that the fuel's low 
decay heat and boil-off period provide sufficient time for the licensee 
to obtain offsite resources to sustain the spent fuel pool cooling 
function indefinitely. Licensees could make use of the equipment in 
place for EDMGs should that equipment be available, recognizing that 
the protection for that equipment is against the hazards posed by 
events that result in losses of large areas of the plant due to fires 
or explosions rather than beyond-design-basis external events resulting 
from natural phenomena. If the EDMG equipment is not available, the 
offsite resources would be used by the licensee for only onsite 
emergency response (i.e., spent fuel pool cooling). This proposed 
amendment would not impact any commitments licensees have made 
regarding exemptions from offsite emergency planning requirements, 
which consider a beyond-design-basis event that could result in a 
zirconium cladding fire due to a loss of SFP inventory and do not 
consider offsite resources in mitigation strategies.
    The NRC proposes to maintain the EDMGs requirement, because an 
event for which EDMGs would be required is not based on the condition 
of the fuel, but may instead result from aircraft impact and a beyond-
design-basis security event which could introduce kinetic energy into 
the spent fuel pool independent from the decay heat of the fuel. These 
types of events and their potential consequences were considered as a 
part of the rulemaking dated March 7, 2009, on Power Reactor Security 
Requirements (74 FR 13926). In the course of that rulemaking, the NRC 
took into account stakeholder input and determined that it would be 
inappropriate to apply the EDMG requirements to permanently shutdown 
and defueled reactors where the fuel was removed from the site or moved 
to an ISFSI. However the resulting rule was written to remove the EDMG 
requirements once the certifications of permanent cessation of 
operations and removal of fuel from the reactor vessel were submitted 
rather than upon removal of fuel from the SFP. The NRC proposes to 
correct this error from the 2009 final rule in this proposed rule as 
explained in the ``EDMGs'' portion of this section.
    The NRC proposes to exclude from proposed Sec.  50.155, the 
licensee for Millstone Power Station Unit 1, Dominion Nuclear 
Connecticut, Inc. Dominion Nuclear Connecticut, Inc. is also the 
licensee for Millstone Power Station Units 2 and 3, but this exclusion 
would apply to Dominion Nuclear Connecticut, Inc. in its capacity as 
licensee for only Unit 1, which is not operating but has irradiated 
fuel in its spent fuel pool and satisfies the proposed criteria for not 
having to comply with this proposed rule except for the EDMG 
requirements. In the course of the development and implementation of 
the guidance and strategies required by current Sec.  50.54(hh)(2), the 
NRC evaluated whether additional mitigation strategies were warranted 
at Millstone Power Station Unit 1 and concluded that no mitigating 
strategies were warranted beyond existing measures, principally due to 
the extended decay time since the last criticality there on November 4, 
1995, and the resulting low decay heat levels allowing sufficient time 
for the use of existing strategies augmented by mitigation strategies 
existing in 2005. The exclusion for Millstone Power Station Unit 1 in 
this proposed rule is based upon that conclusion, recognizing that 
additional mitigating capabilities will be present due to the 
implementation of the Sec.  50.54(hh)(2) strategies at the collocated 
Millstone Power Station Units 2 and 3.
    In contrast to Millstone Power Station Unit 1, the Shutdown NPP 
Group licensees were issued license conditions for the mitigating 
strategies corresponding to the Sec.  50.54(hh)(2) strategies. These 
license conditions are condition 2.C.(10) to Renewed Operating License 
No. DPR-43 for Kewaunee Power Station, condition 2.C.(14) to Facility 
Operating License No. DPR-72 for Crystal River Unit 3 Nuclear 
Generating Plant, condition 2.C.(26) to Facility Operating License NPF-
10 for San Onofre Nuclear Generating Station Unit 2, condition 2.C.(27) 
to Facility Operating License NPF-15 for San Onofre Nuclear Generating 
Station Unit 3, and condition 3.N to Renewed Operating License No. DPR-
28 for Vermont Yankee Nuclear Power Station. Those licensees and future 
power reactor licensees that enter decommissioning would have the 
burden to show that operation in a decommissioning status with 
irradiated fuel in the spent fuel pool without the EDMG license 
condition or the proposed requirement to comply with the proposed EDMG 
requirement would provide adequate protection of public health and 
safety.
Integrated Response Capability
    Each applicant or licensee subject to the proposed requirements 
would be required to develop, implement, and maintain an integrated 
response capability that includes FSGs, EDMGs, EOPs, sufficient 
staffing, and a supporting organizational structure with defined roles, 
responsibilities, and authorities for directing and performing these 
strategies, guidelines, and procedures.
    As discussed in the NTTF Report, EOPs have long been part of the 
NRC's safety requirements. The NRC regulations address them through the 
quality assurance requirements of

[[Page 70621]]

criterion V and criterion VI in appendix B to 10 CFR part 50, and in 
the administrative controls section of the technical specifications for 
each plant. Following the accident at TMI Unit 2, EOPs were upgraded to 
address human factors considerations in order to improve human 
reliability including the operator's ability to mitigate the 
consequences of a broad range of initiating events and subsequent 
multiple failures without the need to diagnose specific events. In 
other words, EOPs were modified from their previous event-driven nature 
to be symptom-based. Numerous subsequent regulatory guides (RGs) and 
technical reports (e.g., NUREG-0660, NUREG-0737, and NUREG-0711) also 
address EOPs. In addition, the EOPs are the subject of a national 
consensus standard (American National Standards Institute/American 
Nuclear Society 3.2-2012, ``Administrative Controls and Quality 
Assurance for the Operational Phase of Nuclear Power Plants''). The 
subject matter for the initial and requalification training, written 
exam, and operating test for reactor operators and senior reactor 
operators also includes the EOPs. While implementing EOPs, the event 
command and control functions remain in the control room under the 
direction of the senior licensed operator on shift.
    The nuclear industry developed EDMGs following the terrorist events 
of September 11, 2001, in response to security advisories, orders, and 
license conditions issued by the NRC that required licensees to develop 
and implement guidance and strategies intended to maintain or restore 
core cooling and containment and spent fuel pool cooling capabilities 
under the circumstances associated with the loss of large areas of the 
plant due to fire or explosion. The EDMGs further extend the range of 
initiating events and plant damage states for which strategies and 
guidelines are available for use by operators to include the loss of 
large areas of the plant and a subsequent impairment of the operability 
and functionality of structures, systems and components that are within 
that area. NEI 06-12, ``B.5.b Phase 2&3 Submittal Guideline,'' Revision 
2, December 2006 (the NRC-endorsed guidance for the requirements 
associated with EDMGs) provides appropriate coordination of the EDMGs 
with the voluntarily maintained SAMGs through its guidance that the 
EDMGs ``must be interfaced with existing SAMGs so that potential 
competing considerations associated with implementing these and other 
strategies are appropriately addressed.''
    Based upon these considerations, the NTTF recommended that the NRC 
require licensees to further integrate EOPs, SAMGs and EDMGs, including 
a clarification of transition points, command and control, decision 
making, and rigorous training that includes conditions that are as 
close to real accident conditions as feasible.
    Subsequent to issuance of the NTTF Report, the range of initiating 
events and plant damage states for which strategies and guidelines are 
available for use by operators was further extended through the 
development of mitigating strategies for beyond-design-basis external 
events in response to Order EA-12-049. The development and 
implementation of this set of strategies and guidelines was 
accomplished with the knowledge of the existence of the other NTTF 
recommendations and took them into account to the extent practical. In 
order to provide better integration with the EOPs, the resulting 
strategies and guidelines (FSGs) leave the designation of command and 
control and decision-making functions within the EOPs or SAMGs, as 
maintained under the voluntary industry initiative, as appropriate. As 
recommended in the NTTF Report, this proposed rule would require that 
EDMGs and FSGs be integrated with EOPs, consistent with the expectation 
that EOPs remain the central element of a licensee's initial response 
capability.
    In establishing a requirement for a response capability that 
encompasses the use of EOPs, EDMGs, and FSGs, the NRC considered the 
fact that these strategies, guidelines and procedures were, and are 
currently being, developed at separate times over a period of several 
decades and that the associated efforts have been focused on responding 
to different types of initiating events and plant damage states. As a 
result, these strategies, guidelines and procedures may not properly 
reflect consideration of the interfaces (e.g., procedure transitions), 
dependencies (e.g., reliance on common systems or resources) and 
interactions (e.g., alignment of response strategies) among strategies, 
guidelines and procedures that may be used in combination, either 
consecutively or concurrently, to mitigate a design-basis or beyond-
design-basis event.
    Additionally, the NRC considered that these strategies, guidelines 
and procedures are not used by a single licensee organizational unit 
but will often require coordination and transfer of responsibilities 
amongst licensee organizational units. For example, the EDMGs may be 
implemented under conditions of loss of the main control room and 
therefore initiated and directed by knowledgeable and available site 
personnel until coordination and augmentation efforts enable transition 
to a more stable command and control structure. The mitigation 
strategies for extreme external events, though initiated by the main 
control room complement of licensed operators, may require coordination 
with and augmentation by offsite organizations. Further, and as noted 
previously, there are potential accident scenarios in which a licensee 
might employ strategies from more than one of these strategies, 
guidelines and procedures during its response to an accident. One 
plausible sequence is for an initial response to be under the EOPs, 
supplemented by actions under the FSGs, and ultimately transition to 
actions under the SAMGs, which are implemented under a voluntary 
initiative. Such an accident progression would engage and require the 
coordination of multiple licensee organizational units.
    In light of the preceding considerations, this proposed rule would 
require that the mitigating strategies, guidelines and procedures, 
staffing, and supporting organizational structure be developed, 
implemented, and maintained such that they function as an 
``integrated'' response capability. The intent is to ensure that 
applicants and licensees establish and maintain a functional capability 
to produce a coordinated and logical response under a wide range of 
accident conditions. The intent is not to require physical integration 
(e.g., organizations need not be merged and strategies, guidelines and 
procedures need not be combined), but rather to require a functional 
integration of the elements of the response capability. To achieve this 
functional integration, the NRC expects that applicants and licensees 
would have addressed the interfaces, dependencies, and interactions 
among the elements of their response capability such that elements work 
together to support effective performance under the full range of 
accident conditions. For example, functional integration of the 
strategies, guidelines and procedures would ensure that transition 
points are explicitly identified and conflicts between strategies are 
eliminated to the extent practical. Functional integration of response 
organizations would ensure that organizations working together to use 
these strategies, guidelines, and procedures (e.g., to coordinate 
actions or provide support) have clearly defined lines of communication 
between the

[[Page 70622]]

organizations, as well as clearly defined authorities and 
responsibilities relative to each other, such that there are no gaps or 
conflicts.
    The proposed requirements for FSGs would make generically-
applicable requirements previously imposed on licensees by Order EA-12-
049, for Virgil C. Summer Nuclear Station Units 2 and 3 by license 
condition as described in Memorandum and Order CLI-12-09,\5\ and for 
Enrico Fermi Nuclear Plant Unit 3, License No. NPF-95, by license 
condition 2.D.(12)(g). These proposed requirements would provide 
additional defense-in-depth measures that increase the capability of 
nuclear power plant licensees to mitigate consequences of beyond-
design-basis external events. Consistent with Order EA-12-049 and 
associated license conditions, these proposed provisions would be made 
generically-applicable in recognition that beyond-design-basis events 
have an associated significant uncertainty, and that the NRC concluded 
additional measures were warranted in light of this uncertainty.
---------------------------------------------------------------------------

    \5\ Summer, CLI-12-09, 75 NRC at 440, and the V.C. Summer Unit 2 
license, License No. NPF-93, Condition 2.D.(13) and V.C. Summer Unit 
3 license, License No. NPF-94, Condition 2.D.(13).
---------------------------------------------------------------------------

    The proposed FSG strategies and guideline requirements are intended 
to mitigate consequences of beyond-design-basis external events from 
natural phenomenon that result in an ELAP concurrent with either a loss 
of normal access to the ultimate heat sink, or for passive reactor 
designs, a loss of normal access to the normal heat sink. Recognizing 
that beyond-design-basis external events are fundamentally unbounded, 
and that these events can result in a multitude of damage states and 
associated accident conditions, a significant regulatory challenge is 
developing bounded requirements that meaningfully address the 
regulatory issue. From a practical standpoint, development of 
mitigation strategies requires that there be some definition (or 
boundary conditions established) for an onsite damage state for which 
the strategies would then address and thereby provide an additional 
capability to mitigate beyond-design-basis external event conditions 
that might occur. The damage state should ideally be representative of 
a large number of potential damage states that might occur as a result 
of extreme external events, and it should present an immediate 
challenge to the key safety functions, so that the resultant strategies 
actually improve safety. The assumed damage state for this proposed 
rule is the same as that assumed to implement the requirements of EA-
12-049, attachment 2 for currently operating power reactors: An ELAP 
condition concurrent with loss of normal access to the ultimate heat 
sink (LUHS). This assumed damage state is effective at immediately 
challenging the key safety functions following a beyond-design-basis 
external event (i.e., core cooling, containment and spent fuel pool 
cooling). Requiring strategies to maintain or restore these key 
functions under such circumstances would result in an additional 
mitigation capability consistent with the Commission's objective when 
it issued Order EA-12-049.
    This proposed rule would not be prescriptive in terms of the 
specific set of initial and boundary conditions assumed for the ELAP 
and LUHS condition, recognizing that the damage state for current 
operating reactors, defined in more detail in draft regulatory guidance 
for this proposed rule (DG)-1301, ``Flexible Mitigation Strategies for 
Beyond-Design-Basis Events,'' reflects current operating power reactor 
designs and the reliance of those designs on ac power, while the 
assumed damage state for a future design may be different depending 
upon the design features. Specifically, this damage state was 
implemented through the assumption of the ELAP to the onsite emergency 
ac buses, but did allow for ac power from the inverters to be assumed 
available in order to establish event sequence and the associated times 
for when mitigation actions would be assumed to be required. To address 
the Order EA-12-049 requirement for an actual loss of all ac power, 
including ac power from the batteries (through inverters), 
contingencies are included in the mitigation strategies to enable 
actions to be taken under those circumstances (e.g., sending operators 
to immediately take manual control over a non ac-powered core cooling 
pump). As such, this proposed provision is meant to make generically-
applicable the current implementation under EA-12-049 (i.e., there is 
no intent to either relax or impose new requirements), and be 
performance-based to allow some flexibility for future designs. As an 
example, some reactor designs (e.g., Westinghouse AP1000 and General 
Electric Economic Simplified Boiling Water Reactor (ESBWR)) use passive 
safety systems to meet NRC requirements for maintaining key safety 
functions. The inherent design of those passive safety systems makes 
certain assumptions, such as loss of access to the ultimate heat sink, 
not credible. Accordingly, the assumed condition for the FSG 
requirements for passive reactors is the loss of normal access to the 
normal heat sink, discussed further in this section. Nevertheless, in 
this proposed rule the NRC is requiring that the strategies and 
guidelines be capable of implementation during a loss of all ac power.
    Regarding the assumed LUHS for combined licenses or applications 
referencing the AP1000 or the ESBWR designs, the assumption was 
modified to be a loss of normal access to the normal heat sink (see 
attachment 3 to Order EA-12-049, Summer, CLI-12-09, 75 NRC at 440, the 
V.C. Summer Unit 2 license, License No. NPF-93, Condition 2.D.(13), the 
V.C. Summer Unit 3 license, License No. NPF-94, Condition 2.D.(13) and 
Enrico Fermi Nuclear Plant Unit 3 License, License No. NPF-95, 
Condition 2.D.(12)(g)). This modified language reflects the passive 
design features of the AP1000 and the ESBWR that provide core cooling, 
containment, and spent fuel cooling capabilities for 72 hours without 
reliance on ac power. These features do not rely on access to any 
external water sources for the first 72 hours because the containment 
vessel and the passive containment cooling system serve as the safety-
related ultimate heat sink for the AP1000 design and the isolation 
condenser system serves as the safety-related ultimate heat sink for 
the ESBWR design.
    As discussed previously, the range of beyond-design-basis external 
events is unbounded. These proposed provisions are not intended, and 
should not be understood to mean, that the mitigation strategies can 
adequately address all postulated beyond-design-basis external events. 
It is always possible to postulate a more severe event that causes 
greater damage and for which the mitigation strategies may not be able 
to maintain or restore the functional capabilities (e.g., meteorite 
impact). Instead, the proposed requirements provide additional 
mitigation capability in light of uncertainties associated with 
external events, consistent with the NRC's regulatory objective when it 
issued Order EA-12-049.
    This proposed rule would require that the FSGs be capable of being 
implemented site-wide. This recognizes that severe external events are 
likely to impact the entire reactor site, and for multi-unit sites, 
damage all the power reactor units on the site. This requirement means 
that there needs to be sufficient equipment and supporting staff to 
enable the core cooling, containment, and spent fuel pool

[[Page 70623]]

cooling functions to be maintained or restored for all the power 
reactor units on the site. This is a distinguishing characteristic of 
this set of mitigating strategies from those that currently exist for 
Sec.  50.54(hh)(2), for which the damage state was a more limited, 
albeit large area of a single plant, reflecting the hazards for which 
that set of strategies was developed.
    The NRC gave consideration to whether there should be changes made 
to Sec.  50.63 to link those requirements with this proposed rule. This 
consideration stemmed from recommendation 4.1 of the NTTF Report to 
``initiate rulemaking to revise 10 CFR 50.63'' and the understanding 
that this proposed rule could result in an increased station blackout 
coping capability, in addition to the regulatory objective of the 
proposed provisions, which is to provide additional beyond-design-basis 
external event mitigation. Because of the substantive differences 
between the requirements of Sec.  50.63 for licensees to be able to 
withstand and recover from a station blackout and the proposed 
requirements, the NRC determined that such a linkage was not necessary 
and could lead to regulatory confusion.
    The principal regulatory objective of Sec.  50.63 was to establish 
station blackout coping durations for a specific scenario (i.e., loss-
of-offsite power coincident with a failure of both trains of emergency 
onsite ac power, typically, the failure of multiple emergency diesel 
generators). In meeting this regulatory objective, the NRC recognized 
that there would be safety benefits accrued through the provision of an 
alternate ac source diverse from the emergency diesel generators and 
therefore defined such a source in Sec.  50.2. In furtherance of this 
alternative means to comply with Sec.  50.63, the NRC also defined the 
event a licensee must withstand and recover from as a station blackout 
rather than a loss of all ac power. A station blackout allows for 
continued availability of ac power to buses fed by station batteries 
through inverters or by alternate ac sources. This proposed rule would 
provide an additional capability to mitigate beyond-design-basis 
external events. Because the condition assumed for the mitigation 
strategies to establish the additional mitigation capability includes 
an ELAP, which is more conservative than a station blackout as defined 
in Sec.  50.2, there can be a direct relationship between the two 
different sets of requirements with regard to the actual implementation 
at the facility. Specifically, implementation of the proposed 
mitigation strategies links into the station blackout procedures (e.g., 
the applicable strategies would be implemented to maintain or restore 
the key safety functions when the EOPs reach a ``response not 
obtained'' juncture).\6\
---------------------------------------------------------------------------

    \6\ One of the formats for symptom-based EOPs that are used in 
the operating power reactors has the operators take an action and 
verify that the system responds to the action in a manner that 
confirms that the action was effective. For example, a step in an 
EOP could be to open a valve in order to allow cooling water flow 
and the verification would be obtained by confirming there are 
indications that flow has commenced such as lowering temperature of 
the system being cooled. If those indications are not obtained, the 
procedure would provide instructions on the next step to accomplish 
in a separate column labeled ``response not obtained.''
---------------------------------------------------------------------------

    Step-by-step procedures are not necessary for many aspects of the 
proposed mitigating strategies and guidelines. Rather, the strategies 
and guidelines should be flexible, and therefore enable plant personnel 
to adapt them to the conditions that result from the beyond-design-
basis external event. The proposed provisions typically would result in 
strategies and guidelines that use both installed and portable 
equipment, instead of only relying on installed ac power sources (with 
the exception of protected battery power) to maintain or restore core 
cooling, containment, and spent fuel pool cooling capabilities. By 
using equipment that is separate from the normal installed ac-powered 
equipment, the strategies and guidelines have a diverse attribute. By 
having available multiple sets of portable equipment that can be 
deployed and used in multiple ways depending on the circumstances of 
the event, operators are able to implement strategies and guidelines 
that are flexible and adaptable.
    The proposed mitigation strategies requirements are both 
performance-based and functionally-based. The proposed performance-
based requirements recognize that the new requirements would provide 
most benefit to future reactors whose designs could differ 
significantly from current power reactor designs and as such, use of 
more prescriptive requirements could be problematic and create 
unnecessary regulatory impact and need for exemptions. Use of 
functionally-based requirements results from the need to have 
requirements that can address a wide range of damage states that might 
exist following beyond-design-basis external events. Maintaining or 
restoring three key functions (core cooling, containment' and spent 
fuel pool cooling) supports maintenance of the fission product barriers 
(i.e., fuel clad, reactor coolant pressure boundary, and containment) 
and results in an effective means to mitigate these events, while 
remaining flexible such that the strategies and guidelines can be 
adapted to the damage state that occurs. Functionally-based 
requirements also result in strategies that align well with the 
symptom-based procedures used by power reactors to respond to 
accidents. Accordingly, Order EA-12-049 contained requirements for a 
three-phased approach for current operating reactors. This proposed 
rule does not specify a number of phases; instead, the NRC is proposing 
higher level, performance-based requirements consistent with this 
discussion.
    The NRC gave consideration to incorporating into this proposed rule 
a requirement that licensees be capable of implementing the strategies 
and guidelines ``whenever there is irradiated fuel in the reactor 
vessel or spent fuel pool.'' This provision would have been a means of 
making generically-applicable the requirement from Order EA-12-049 that 
licensees be capable of implementing the strategies and guidelines ``in 
all modes.'' The NRC considers the terminology ``whenever there is 
irradiated fuel in the reactor vessel or spent fuel pool'' would be a 
better means to address the Order requirement since the phrase does not 
use technical specification type language (i.e., modes), which would 
not be in effect when a licensee completely offloads the fuel from the 
reactor vessel into the spent fuel pool during an outage. The NRC 
concluded that the use of the phrases ``whenever there is irradiated 
fuel in the reactor vessel or spent fuel pool'' or ``in all modes'' is 
not necessary because the proposed applicability provisions would 
ensure that licensees would be required to have mitigation strategies 
for beyond-design-basis external events for the various configurations 
that can exist for the reactor and spent fuel pools throughout the 
operational, refueling and decommissioning phases.
    The mitigation strategies and guidelines implemented under NRC 
Order EA-12-049 assume a demanding condition that maximizes decay heat 
that would need to be removed from the reactor core and spent fuel pool 
source terms on site. This implementation results in a more restrictive 
timeline (i.e., mitigation actions required earlier following the event 
to take action to maintain or restore cooling to these source terms) 
and a greater resulting additional capability. These assumed at-power 
conditions are 100 days at 100 percent power prior to the event for the 
reactor core as was used for Sec.  50.63. This assumption establishes a 
conservative decay heat for the reactor source term. The assumed spent 
fuel

[[Page 70624]]

pool conditions include the design basis heat load for the spent fuel 
pool, typically a full core offload following a refueling outage. This 
establishes a conservative heat load for the spent fuel pool. The NRC 
recognizes that, as a practical reality, these conditions would not 
exist simultaneously. The NRC considers the development of timelines 
for the proposed mitigating strategies using the maximum heat load for 
either the reactor core or the spent fuel pool to be appropriate. While 
establishing the capability to mitigate the maximum heat load for both 
simultaneously would be compliant with the proposed requirements, it 
would not be necessary.
    The NRC recognizes the difficulty of developing engineered 
strategies for the extraordinarily large number of possible plant and 
equipment configurations that might exist under shutdown conditions 
(i.e., at shutdown when equipment may be removed from service, when 
there is ongoing maintenance and repairs or refueling operations, or 
modifications are being implemented). The proposed requirements mean 
that licensees should be cognizant of such configurations, equipment 
availability, and decay heat states that could present greater 
challenges under these conditions, and design mitigation strategies 
that can be implemented under such circumstances.
    The NRC considered requiring the strategies to be developed 
considering the need to plan for delays in the receipt of offsite 
resources as a result of damage to the transportation infrastructure. 
While severe events could damage local infrastructure, and could create 
challenges with regard to the delivery of offsite resources, the NRC 
concluded that having this level of specificity in the proposed 
provisions would not be necessary. Instead, this proposed rule contains 
provisions that are more performance-based, requiring continued 
maintenance or restoration of the functional capabilities until 
acquisition of offsite assistance and resources. Potential delays and 
other challenges presented by extreme events that affect acquisition 
and use of offsite resources would be addressed by licensee programs 
that implement the proposed provisions.
    Order EA-12-049 included a requirement that licensees develop 
guidance and strategies to obtain ``sufficient offsite resources to 
sustain [the functions of core cooling, containment, and spent fuel 
pool cooling] indefinitely.'' The NRC considered using this language in 
this proposed rule, but concluded that this would be better phrased as 
``indefinitely, or until sufficient site functional capabilities can be 
maintained without the need for the mitigation strategies.'' The NRC 
concluded that this phrase better communicates the existence of a 
transition from the use of the mitigating strategies to recovery 
operations.
    The NRC recognizes that the use of the proposed mitigating 
strategies would potentially require departure from a license condition 
or a technical specification (contained in a license issued under 10 
CFR part 50 or 52) and could be considered a proceduralization of the 
allowance provided under Sec.  50.54(x). Given that the initiation of 
the use of these strategies may be included in emergency operating 
procedures or other procedures, which might be considered procedures 
described in the final safety analysis report (as updated), there is an 
interaction with the provisions of Sec.  50.59(c)(1) regarding the need 
to obtain a license amendment in order to make the necessary change to 
those procedures. The NRC considered including provisions in this 
proposed rule specifically to allow departures from license conditions 
or technical specifications in order to clarify this situation, but 
found these provisions unnecessary. For holders of operating licenses 
under 10 CFR part 50 and combined licenses under 10 CFR part 52 that 
were subject to Order EA-12-049, the provisions of that Order provided 
more specific criteria for making the necessary changes than Sec.  
50.59, making that section inapplicable as set forth in Sec.  
50.59(c)(4). Those criteria included the provision of submitting an 
overall integrated plan to the NRC for review. Similar criteria were 
included in license conditions for the combined licenses for Virgil C. 
Summer Nuclear Station, Units 2 and 3, and Enrico Fermi Nuclear Plant 
Unit 3.
EDMGs
    The NRC proposes to move the EDMGs requirement currently in Sec.  
50.54(hh)(2) to a new mitigation of beyond-design-basis events section 
of 10 CFR part 50. In addition to moving the text, the NRC proposes to 
make a few editorial changes. The wording used to describe these 
requirements has evolved from ``guidance and strategies,'' in Interim 
Compensatory Measures Order EA-02-026, dated February 25, 2002, to 
``strategies,'' in the corresponding license conditions, to ``guidance 
and strategies,'' in Sec.  50.54(hh)(2), to its proposed form 
``strategies and guidelines.'' The word ``guidelines'' was chosen 
rather than ``guidance'' to better reflect the nature of the 
instructions that could be developed as appropriate by a licensee and 
to avoid confusion with the term ``regulatory guidance.'' The word 
``strategies'' is used in this proposed rule to reflect its meaning, 
``plans of action.'' The resulting plans of action could include plant 
procedures, methods, or other guideline documents, as deemed 
appropriate by the licensee during the development of these strategies. 
These plans of action would also include the arrangements made with 
offsite responders for support during an actual event. No substantive 
change to the requirements is intended by this proposed change in the 
wording.
    Applicability of the requirements of Sec.  50.54(hh)(2) is 
currently governed by Sec.  50.54(hh)(3), which makes these 
requirements inapplicable following the submittal of the certifications 
required under Sec.  50.82(a) or Sec.  52.110(a)(1). As discussed in 
the statement of considerations for the Power Reactor Security 
Rulemaking (74 FR 13926), the NRC believes that it would be 
inappropriate for the requirements for EDMGs to apply to a permanently 
shutdown, defueled reactor, where the fuel was removed from the site or 
moved to an ISFSI. The NRC proposes to require EDMGs for a licensee 
with permanently shutdown defueled reactors, but with irradiated fuel 
still in its spent fuel pool, because the licensee must be able to 
implement effective mitigation measures for large fires and explosions 
that could impact the spent fuel pool while it contains irradiated 
fuel. The difference between this proposed rule and Sec.  50.54(hh)(3) 
would correct the wording of the latter provision to implement the 
sunsetting of the associated requirement as was intended by the 
Commission in 2009. This change would not constitute backfitting for 
currently operating reactors because the proposed change concerns 
decommissioning reactors. The proposed change would not constitute 
backfitting for currently decommissioning reactors because the EDMGs 
are also required by the licensees' license conditions that were made 
generically applicable through the Power Reactor Security Rulemaking 
and remain in effect.
Integration With EOPs
    In developing a proposed requirement for the integration of FSGs 
and EDMGs with the EOPs, the NRC considered their differences in 
content and the standards for usage applied to them. The EOPs are a 
specific and prescribed set of instructions implemented in accordance 
with exacting standards for usage and adherence (e.g., step-by-step 
sequential performance, concurrent execution of multiple sections) that

[[Page 70625]]

operators and plant staff are required to follow when performing a 
specific task or addressing plant conditions. When implementing 
procedures, each step is to be performed as prescribed, with rare 
exceptions. The strategies and guidelines that would be required differ 
from EOPs primarily in terms of the level of detail to which they are 
written and expectations regarding usage. These strategies and 
guidelines may be a less prescriptive set of instructions not subject 
to the same constraints imposed by standards of usage for procedure 
implementation (e.g., may not be followed in a step-by-step manner). 
This is because of: (1) The large number of possible event initiators, 
plant configurations, and sequences; and (2) the high degree of 
uncertainties in event progression and consequences. The strategies and 
guidelines can take the form of high level plans that identify and 
describe potential, previously evaluated, success paths for addressing 
specific conditions such as loss of core cooling. As a result, 
strategies and guidelines provide operators and plant staff the 
information and latitude to respond as necessary to unpredictable and 
dynamic situations, allowing them to adapt to the actual conditions and 
damage states without the burden of detailed procedures and the 
challenge of determining which procedure may be applicable and 
effective under the uncertain conditions of a beyond design basis 
accident.
    Given these differences in content and standards for usage, the 
intent of this proposed rule is not to require conformance of the 
strategies and guidelines to the level of detail and standards of usage 
for EOPs, or consolidation of the strategies, guidelines and procedures 
into a single set of instructions, but rather, as previously described, 
to require functional integration of strategies and guidelines with the 
EOPs. The objective is for the strategies, procedures, and guidelines 
to retain or employ the characteristics that support their effective 
use under the range of conditions to which they are each intended to 
apply while ensuring that the strategies and guidelines, in conjunction 
with the EOPs, constitute a useable and cohesive set of instructions 
for mitigating the consequences of a wide range of initiating events 
and plant damage states. To achieve this functional integration, the 
NRC expects that applicants and licensees would have addressed the 
interfaces, dependencies, and interactions among the strategies and 
guidelines that would be required under this proposed rule and the 
EOPs, such that they can be implemented in concert with each other, as 
necessary, to effectively use available plant resources and direct a 
logical and coordinated response to a wide range of accident 
conditions.
    In keeping with the basis for a functional integration of the 
strategies and guidelines with EOPs, this proposed rule would require 
that the FSGs and EDMGs be integrated ``with the Emergency Operating 
Procedures (EOPs).'' This proposed language is intended to communicate 
the NRC's expectation that the EOPs retain their role as the primary 
means of directing emergency operations and that the strategies and 
guidelines that would be required under this proposed rule would be 
integrated with EOPs to support their implementation or augment them 
where their implementation is not successful in preventing significant 
fuel damage.
    The NRC considered establishing specific criteria for the 
integration of the strategies and guidelines with EOPs but opted to 
specify only a high level requirement to allow applicants and licensees 
flexibility in the means by which they achieve the functional 
integration described previously. Approaches for achieving functional 
integration could include the following:
    1. Strategies, guidelines, and procedures have clearly defined 
transitions (e.g., entry and exit conditions with distinct pointers) 
from one strategy, guideline, or procedure to another.
    2. Individuals are cued by the document or trained to know when 
transitions between the strategies, guidelines, and procedures result 
in corresponding changes in the associated standards for usage (e.g., 
when transitioning from EOPs to the voluntarily maintained SAMGs, the 
operator is able to recognize the transition from a step-by-step 
procedure to a flexible guideline set where it is permissible to 
deviate from the order or method of accomplishing the steps).
    3. Licensees establish expectations (e.g., through standards for 
usage) pertaining to the parallel use of strategies, guidelines, and 
procedures. Plant personnel using different strategies, guidelines, and 
procedures concurrently understand which is the controlling procedure 
and therefore which actions take precedence.
    4. Licensees identify and resolve conflicts between the strategies, 
guidelines and procedures.
    5. Licensees identify competing considerations when using the 
strategies, guidelines and procedures and eliminate or address them in 
guidance.
    6. Licensees control the development and maintenance of their 
content and format in accordance with human factors standards and 
guidelines (e.g., writer's guides) that recognize and address the 
interfaces between them in order to achieve compatibility of the 
strategies, guidelines, and procedures.
Staffing
    The NRC proposes to require licensees to provide the staffing 
necessary for having an integrated response capability to support 
implementation of the FSGs and EDMGs. To be effective, staffing for an 
expanded response capability should include the trained and qualified 
individuals who would be relied upon to analyze, recommend, authorize, 
and implement the mitigating strategies. The staffing must directly 
support the assessment and implementation of a range of mitigation 
strategies intended to maintain or restore the functions of core 
cooling, containment, and spent fuel pool cooling.
    The staffing analyses required by proposed appendix E, section VII, 
should determine when personnel performing expanded response functions 
should report to the site, within a timeframe sufficient to support 
implementation of the strategies that are not assigned to the on-shift 
staff. This would ensure that the functions of core cooling, 
containment, and spent fuel pool cooling are continuously maintained or 
are promptly restored.
    The NRC has endorsed the industry guidance for conducting staffing 
analyses, NEI 10-05, ``Assessment of On-Shift Emergency Response 
Organization Staffing and Capabilities,'' Revision 0, and NEI 12-01, 
``Guideline for Assessing Beyond Design Basis Accident Response 
Staffing and Communications Capabilities,'' Revision 0, and the NRC has 
issued Interim Staff Guidance (ISG), NSIR/DPR-ISG-01, ``Emergency 
Planning for Nuclear Power Plants,'' that provides the requisite 
details for determining the staffing levels and for which positions, as 
well as which beyond design basis external events, the applicants and 
licensees should evaluate.
    The recommended minimum positions and staffing levels for emergency 
plans were initially provided in 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.'' 
Following the September 11, 2001, events, the NRC issued Enhancements

[[Page 70626]]

to Emergency Preparedness Regulations (EP final rule) (76 FR 72560) to 
amend 10 CFR part 50, appendix E, to address, in part, concerns about 
the assignment of tasks or responsibilities to on-shift emergency 
response organization (ERO) personnel that would potentially overburden 
them and prevent the timely performance of their functions under the 
emergency plan. Licensees must have enough on-shift staff to perform 
specified tasks in various functional areas of emergency response 24 
hours a day, 7 days a week. This proposed rule would address the 
staffing requirements for the expanded response capabilities for on-
shift response and the ERO.
    This proposed rule would require adequate staffing to implement the 
FSGs and EDMGs with the EOPs without requiring further analysis to 
supplement analyses that were completed as a result of post-Fukushima 
orders or the EP final rule. Staffing levels should be established to 
ensure that if strategies are executed there would be no delays in 
completing them caused by the lack of qualified personnel. The NRC 
expects that the use of drills, existing training analyses and other 
methods would verify sufficient staffing levels.
Command and Control
    The NRC proposes to require licensees to have a supporting 
organizational structure with defined roles, responsibilities, and 
authorities for directing and performing the FSGs and EDMGs. The 
objective is to ensure that licensees address the organizational 
implications of: (1) Implementing the FSGs; and (2) integrating the 
FSGs and EDMGs with the EOPs such that organizational units responsible 
for on-site accident mitigation (e.g., main control room, emergency 
operations facility, and technical support center staff) can support a 
coordinated implementation of these procedures and guidelines under the 
challenging conditions presented by beyond-design-basis events.
    Additional requirements currently exist in 10 CFR part 50, appendix 
E, section IV.A, for the inclusion within the emergency plan of a 
description of the organization for coping with radiological 
emergencies, including definition of authorities, responsibilities, and 
duties of individuals assigned to the licensee's emergency organization 
and the means for notification of such individuals in the event of an 
emergency. These requirements provide the command and control structure 
for use in the execution of the emergency plan. The current 10 CFR part 
50, appendix E, sections IV.A.2.a. and IV.A.5., further require that 
the emergency plan include: (1) A detailed description of the 
authorities, responsibilities, and duties of the individual(s) who will 
take charge during an emergency; (2) plant staff emergency assignments, 
authorities, responsibilities, and duties of an onsite emergency 
coordinator who shall be in charge of the exchange of information with 
offsite authorities responsible for coordinating and implementing 
offsite emergency measures; and (3) the identification, by position and 
function to be performed, of other employees of the licensee with 
special qualifications for coping with emergency conditions that may 
arise.
    The need for defined command and control structures and 
responsibilities for use in beyond-design-basis conditions was 
recognized in the course of the development of the guidance and 
strategies for the current Sec.  50.54(hh)(2). As stated in the 
industry's guidance document for that set of requirements, NEI 06-12, 
``B.5.b Phase 2 & 3 Submittal Guideline,'' Revision 2, ``Experience 
with large scale incidents has shown that command and control execution 
can be a key factor to mitigation success.'' The guidance and 
strategies developed for that effort include an EDMG for initial 
response to provide a bridge between normal operational command and 
control and the command and control that is provided by the ERO in the 
event that the normal command and control structure is disabled. The 
NRC considers that the actions taken in the development of the EDMG for 
initial response for the guidance and strategies for the current Sec.  
50.54(hh)(2) would continue to be adequate for compliance with this 
proposed rule for EDMGs following the proposed movement of those 
requirements.
    The endorsed industry guidance in NEI 12-06, Revision 0, ``Diverse 
and Flexible Coping Strategies (FLEX) Implementation Guide,'' for the 
guidance and strategies required by Order EA-12-049, specifies that the 
existing command and control structure will be used for transition to 
the voluntarily maintained SAMGs
    All previous requirements did not specify a command and control 
structure for a multi-unit event that includes the potential need for 
acquisition of offsite assistance to support onsite event mitigation. 
Additionally, these requirements were not understood to require such a 
response since they preceded the Fukushima event and the regulatory 
actions that stemmed from that event. As a practical matter, the 
current command and control structures, including any changes that 
resulted from the implementation of Order EA-12-049 requirements, are 
expected to be sufficient to ensure that the functional objectives of 
this proposed rule are achieved. Accordingly, the NRC recognizes that 
this new requirement may not be necessary and is requesting stakeholder 
feedback on this issue (refer to section VI of this notice).
Equipment
    The NRC proposes to have requirements for licensee equipment, 
including instrumentation, that is relied upon for use in proposed 
mitigation strategies and guidelines. This rulemaking does not propose 
to modify the regulatory treatment of equipment relied upon for the 
EDMGs currently required by Sec.  50.54(hh)(2). The regulatory 
treatment of that equipment will remain as it is described in the 
endorsed guidance document for those strategies and guidelines.
    This proposed rule would make generically applicable requirement 
(2) of Order EA-12-049, attachments 2 and 3, which reads as follows: 
``These strategies must . . . have adequate capacity to address 
challenges to core cooling, containment, and SFP cooling capabilities 
at all units on a site subject to this Order.''
    The industry guidance of NEI 12-06, as endorsed by NRC interim 
staff 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,'' included 
specifications for licensee provision of a spare capability in order to 
assure the reliability and availability of the equipment required to 
provide the capacity and capability requirements of the Order. This 
spare capability was also referred to within the guidance as an ``N+1'' 
capability, where ``N'' is the number of power reactor units on a site. 
The NRC considered including requirements similar to the spare 
capability specification of NEI 12-06 in this proposed rule but 
determined that such an inclusion would be too prescriptive and could 
result in the need to grant exemptions for alternate approaches that 
provide an effective and efficient means to provide the required 
capability of the Order. One example of this is in the area of flexible 
hoses, for which a strict application of the sparing guidance could 
necessitate provision of spare hose or cable lengths sufficient to 
replace the longest run of hoses when significant operating experience 
with similar hoses for fire protection does not show a failure rate 
that would support this as a need.

[[Page 70627]]

    The development of the mitigating strategies in response to Order 
EA-12-049 relied upon a variety of initial and boundary conditions that 
were provided in the regulatory guidance of JLD-ISG-2012-01, Revision 
0, and NEI 12-06, Revision 0. These initial and boundary conditions 
followed the philosophy of the basis for imposition of the requirements 
of Order EA-12-049, which was to require additional defense-in-depth 
measures to provide continued reasonable assurance of adequate 
protection of public health and safety. As a result, the industry 
response to Order EA-12-049 includes diverse and flexible means of 
accomplishing safety functions rather than providing an additional 
further hardened train of safety equipment. These requirements and 
conditions included the acknowledgement that, due to the fact that 
initiation of an event requiring use of the strategies would include 
multiple failures of safety-related structures, systems, and components 
(SSCs), it is inappropriate to postulate further failures that are not 
consequential to the initiating event. As a result, the NRC has 
determined that the conditions to which the instrumentation relied on 
for the mitigating strategies would be exposed do not include 
conditions stemming from fuel damage, but instead are limited as 
described previously. The NRC has determined that it should not be 
necessary for the instrumentation to be designed specifically for use 
in the mitigating strategies and guidelines, but instead it would be 
necessary that the design and associated functional performance be 
sufficient to meet the demands of those strategies.
    The underlying proposed requirements are for events that are not 
included in the design basis events as that term is used in the Sec.  
50.2 definition of safety-related SSCs. Because of this, reliance on 
equipment for use in the related strategies would not result in the 
applicability of 10 CFR part 50, appendix A, General Design Criterion 
(GDC)-2, ``Design bases for protection against natural phenomena,'' or 
the principal design criterion (PDC) applicable to a plant's operating 
license if issued prior to GDC-2. This proposed rule would require 
reasonable protection for the equipment relied on for the mitigation 
strategies to a hazard level as severe as that originally determined 
for the facility under GDC-2 or the applicable PDC unless the 
reevaluated hazards stemming from the March 12, 2012, NRC letter issued 
under Sec.  50.54(f), as assessed by the NRC show that increased 
protection is necessary. The March 12, 2012, NRC letter requested 
information on licensees' seismic and flooding hazards; licensees and 
the NRC are currently scheduled to complete most of the work on the 
flooding reevaluations prior to the anticipated effective date of this 
proposed rule. The NRC notes that there are some licensees whose 
licensing bases include requirements for protection from natural 
phenomena beyond those established at the original licensing (e.g., 
North Anna Power Station for the seismic hazard), but anticipates that 
these different hazard levels would be captured in the reevaluation of 
external hazards under the March 12, 2012, NRC letter.
    As discussed in COMSECY-14-0037, ``Integration of Mitigating 
Strategies for Beyond-Design-Basis External Events and The Reevaluation 
of Flooding Hazards,'' and its associated SRM, the requirements of 
Order EA-12-049 were imposed in parallel with the agency's March 12, 
2012, requests for information on the reevaluation of external hazards. 
As a result, Order EA-12-049 included a requirement in both attachment 
2 and 3 for licensees to provide reasonable protection for equipment 
associated with the required mitigating strategies from external events 
without specific reference to the necessary level of protection. The 
appropriate level of protection from external hazards, particularly 
flooding, was the subject of discussion in the course of NRC-held 
public meetings leading up to the issuance of JLD-ISG-2012-01 and its 
endorsement of the industry guidance for Order EA-12-049, NEI 12-06. 
Section 6.2.3.1 of NEI 12-06 specifies that the level of protection for 
flooding should be ``the flood elevation from the most recent site 
flood analysis. The evaluation to determine the elevation for storage 
should be informed by flood analysis applicable to the site from early 
site permits, combined license applications, and/or contiguous licensed 
sites.'' The choice of this hazard level was driven by the recognition 
that, while the flooding hazard reevaluations by holders of operating 
licenses and construction permits may not be complete in advance of the 
development and implementation of the mitigating strategies, 
information available from flood analyses for nearby sites could be 
taken into account in choosing the appropriate level in order to avoid 
the need for rework or modification of the strategies. Many licensees 
took the former approach, using their best estimates of potential 
hazard levels and providing additional margin to the current licensing 
basis. (See, e.g., the description of the flooding strategies for Fort 
Calhoun Station on page B-43 et seq., of Omaha Public Power District's 
Overall Integrated Plan (Redacted) in Response to March 12, 2012, Order 
EA-12-049.)
    In COMSECY-14-0037, the NRC staff requested that the Commission 
affirm that: (1) Licensees for operating nuclear power plants need to 
address the reevaluated flooding hazards within their mitigating 
strategies for beyond-design-basis external events; (2) licensees for 
operating nuclear power plants may need to address some specific 
flooding scenarios that could significantly damage the power plant site 
by developing targeted or scenario-specific mitigating strategies, 
possibly including unconventional measures, to prevent fuel damage in 
reactor cores or spent fuel pools; and (3) the NRC staff should revise 
the flooding assessments and integrate the decision-making into the 
development and implementation of mitigating strategies in accordance 
with Order EA-12-049 and this rulemaking. These principles reflect the 
NEI 12-06 reference to the ``most recent flood analysis'' previously 
discussed and the documentation by licensees in their overall 
integrated plans for the mitigating strategies that, at the time of 
their submittals, ``flood and seismic reevaluations pursuant to the 
Sec.  50.54(f) letter of March 12, 2012, are not completed and 
therefore not assumed in this submittal. As the reevaluations are 
completed, appropriate issues would be entered into the corrective 
action system and addressed on a schedule commensurate with other 
licensing bases changes.'' In SRM-COMSECY-14-0037, the Commission 
approved the first two items recommended by the NRC staff, regarding 
the need for operating nuclear power plant licensees to address the 
reevaluated flood hazards within the mitigating strategies and the 
potential for using targeted or scenario specific mitigating 
strategies. The Commission did not approve the third recommendation, 
but that recommendation is outside the scope of this rulemaking effort. 
The NRC drafted the proposed rule to reflect this direction and in 
recognition of the fact that the wording of Order EA-12-049 and its 
associated guidance did not make clear that the mitigating strategies 
equipment would require protection to the reevaluated hazard levels 
resulting from the Sec.  50.54(f) request for information of March 12, 
2012.
    Because the events for which the proposed mitigating strategies are 
to be used are outside the scope of the design basis events considered 
in establishing the basis for the design of the facility, equipment 
that is relied upon for those

[[Page 70628]]

mitigating strategies may not fall within the scope of Sec.  50.65, 
``Requirements for monitoring the effectiveness of maintenance at 
nuclear power plants.'' Nevertheless, the NRC proposes that such 
equipment should receive adequate maintenance in order to assure that 
it is capable of fulfilling its intended function when called upon.
    The NRC proposes to require licensees to have a means to remotely 
monitor wide-range SFP level as a part of the equipment relied upon to 
support the FSGs. This provision would make generically-applicable the 
requirements imposed by Order EA-12-051. The NRC considered including 
the detailed requirements from Order EA-12-051 within this proposed 
rule, but determined that the more performance-based approach taken 
with this proposed rule would better enable an applicant for a new 
reactor license or design certification to provide innovative solutions 
to address the need to effectively prioritize event mitigation and 
recovery actions between the source term contained in the reactor 
vessel and that contained within the spent fuel pool.
Training
    The NRC anticipates that mitigation of the effects of beyond-
design-basis events using the proposed strategies and guidelines would 
be principally accomplished through manual actions rather than 
automated plant responses. Additionally, the instructions provided for 
event mitigation may be largely provided as high level strategies and 
guidelines rather than step-by-step procedures. The use of strategies 
and guidelines supports the ability to adapt the mitigation measures to 
the specific plant damage and operational conditions presented by the 
event. However, effective use of this flexibility would depend upon the 
knowledge and abilities of personnel to select appropriate strategies 
or guidelines from a range of options and implement mitigation measures 
using equipment or methods that may differ from those employed for 
normal operation or design-basis event response. As a result, the NRC 
considers personnel training and qualification necessary to ensure that 
individuals would be capable of effectively performing their roles and 
responsibilities in accordance with the strategies and guidelines that 
would be required by this proposed rule.
    The NRC acknowledges that licensee training programs, such as those 
required for licensed operators under 10 CFR part 55, ``Operators' 
Licenses,'' the programs for plant personnel specified under Sec.  
50.120, ``Training and Qualification of Nuclear Power Plant 
Personnel,'' and the training for emergency response personnel required 
by 10 CFR part 50, appendix E, section IV.F, ``Training,'' would likely 
provide for many of the knowledge and abilities required for performing 
activities in accordance with the strategies and guidelines that would 
be required by this proposed rule. Nevertheless, as noted previously, 
the NRC anticipates that these strategies and guidelines may use new 
methods or equipment that require knowledge and abilities not currently 
addressed under existing training programs and, as a result, there may 
be gaps in these training programs that must be addressed to support 
effective use of the strategies and guidelines. Accordingly, this 
proposed rule would further require that licensees provide for the 
training of personnel using a systems approach to training as defined 
in Sec.  55.4 (the Systems Approach to Training (SAT) process), except 
for elements already covered under other NRC regulations.\7\ The SAT 
process, which is acceptable for meeting training requirements under 10 
CFR part 55 and Sec.  50.120, would also be appropriate for licensee 
identification and resolution of any current gaps or future 
modifications to personnel training that may be necessary to provide 
for the training of personnel performing activities in accordance with 
the mitigating strategies and guidelines that would be required by this 
proposed rule. The NRC recognizes that there are other training 
programs that are currently acceptable for meeting other regulatory 
required training (e.g., 10 CFR part 50, appendix E, section IV.F) that 
do not use the SAT process. In light of the existence of these training 
programs, which have been found acceptable for more frequently 
occurring design-basis events, the NRC has determined that these 
training programs can meet the needs for common elements with beyond-
design-basis event mitigation. Therefore, the NRC would not require 
licensees to revise these training programs to use the SAT process to 
meet the proposed requirements. Licensees would be required to use the 
SAT process for newly identified training requirements supporting the 
effective use of the strategies and guidelines that would be required 
by this proposed rule.
---------------------------------------------------------------------------

    \7\ This definition of a systems approach to training (SAT), is 
a training program that includes the following five elements: (1) 
Systematic analysis of the jobs to be performed; (2) learning 
objectives derived from the analysis which describe desired 
performance after training; (3) training design and implementation 
based on the learning objectives; (4) evaluation of trainee mastery 
of the objectives during training; and (5) evaluation and revision 
of the training based on the performance of trained personnel in the 
job setting.
---------------------------------------------------------------------------

    By using the SAT process, licensees would identify and train on any 
additional tasks that would be necessary to implement the strategies 
and guidelines for the mitigation of beyond-design-basis events as 
defined in this proposed rule. The additional tasks identified would be 
incorporated into the training program to ensure appropriate training 
would be administered for each qualified individual designated to 
implement the strategies and guidelines required by this proposed rule.
Change Control
    The proposed requirements address beyond-design-basis events, and 
as such, currently existing change control processes do not address all 
aspects of a contemplated change, including most notably Sec.  50.59. 
As such, the proposed change control provision is intended to 
supplement the existing change control processes and focus on the 
beyond-design-basis aspects of the proposed change.
    This proposed rule would not contain criteria typically included in 
other change control processes that are used as a threshold for 
determining when a licensee needs to seek NRC review and approval prior 
to implementing the proposed change. Instead, the proposed provisions 
would require that the evaluations of the proposed change reach a 
conclusion that all new requirements continue to be met and that this 
evaluation is documented and maintained to support NRC inspection.
    Proposed changes that remain consistent with regulatory guidance 
would be acceptable, since such changes would ensure continued 
compliance with the proposed provisions in this rulemaking. The NRC 
recognizes that the proposed change control provisions may result in 
licensees seeking NRC review and approval of proposed changes that do 
not follow current regulatory guidance for this proposed rulemaking 
potentially through a license amendment or through NRC review of new or 
revised regulatory guidance. Accordingly, the NRC is requesting 
stakeholder feedback on this issue to determine whether there is a 
better regulatory approach for change control (refer to the ``Specific 
Requests for Comments'' section of this document).
    During public discussions before issuance of this proposed rule, 
there was a suggestion that the NRC should consider a provision to 
allow a licensee to request NRC review of a proposed change, and that 
if the NRC did not act

[[Page 70629]]

upon the request for a suggested time period (e.g., 180 days) that the 
request be considered ``acceptable.'' The NRC did not include this 
``negative consent'' type of approval process in this proposed rule and 
instead the proposed change control process places the responsibility 
on the licensees to ensure that proposed changes result in continued 
compliance with the proposed rule provisions, or are otherwise 
submitted to the NRC following the Sec.  50.12 exemption process. The 
NRC expects to obtain stakeholder feedback on this issue and will 
consider that feedback when developing the final rule provisions.
    A licensee may intend to change its facility, procedures, or 
guideline sets to revise some aspect of beyond-design-basis mitigation 
(i.e., governed by the proposed provisions of this rulemaking), and the 
same change can impact multiple aspects of the facility (i.e., impact 
``design basis'' aspects of the facility and be subject to other 
regulations and change control processes). As previously discussed, the 
NRC anticipates that a licensee would ensure that a proposed change is 
consistent with endorsed guidance to ensure continued compliance with 
the proposed provisions. This same change could also impact safety-
related structures, systems, and components, either directly (e.g., a 
proposed change that impacts a physical connection of mitigation 
strategies equipment to a safety-related component or system) or 
indirectly (e.g., a proposed change that involves the physical location 
of mitigation equipment in the vicinity of safety-related equipment 
that presents a potential for adverse physical/spatial interactions 
with safety-related components). As such, Sec.  50.59 would need to be 
applied to evaluate the proposed change for any potential impacts to 
safety-related SSCs.
    Additionally, proposed changes can impact numerous aspects of the 
facility beyond the safety-related impacts, including implementation of 
fire protection requirements, security requirements, emergency 
preparedness requirements, or safety/security interface requirements. 
Accordingly, it would be necessary for a licensee to ensure that all 
applicable change control provisions are used to judge the 
acceptability of facility changes including, for example, change 
control requirements for fire protection, security, and emergency 
preparedness. Additionally, recognizing the nature of mitigation 
strategies and the reliance on human actions, it is also necessary to 
ensure that the proposed changes satisfy the safety/security interface 
requirements of Sec.  73.58. It is the obligation of the licensee to 
comply with all applicable requirements, and as such, the proposed 
change control provisions could be viewed as unnecessary. However 
recognizing the potential complexity of proposed facility changes and 
the complexity of existing regulatory requirements that govern change 
control, the NRC concluded that adding the proposed change control 
provision, for the purposes of regulatory clarity, was warranted.
Implementation
    The NRC proposes a compliance schedule of 2 years following the 
effective date of the rule. This proposed rule does not include any 
special provision for a holder of a COL as of the effective date of the 
rule for which the Commission has not made the finding required under 
Sec.  52.103(g) (i.e., a COL holder still in the construction phase). 
The NRC considers the duration of 2 years prior to compliance with the 
requirements of this proposed rule to be acceptable because the 
majority of these requirements have been previously implemented under 
Orders EA-12-049 and Order EA-12-051 or Sec.  50.54(hh)(2), or are in 
response to the Sec.  50.54(f) requests for information issued March 
12, 2012.
Regulatory Basis for New Emergency Response Capability Requirements
    A significant objective of this rulemaking is to make the 
requirements that were previously imposed under Order EA-12-049 
generically applicable. As an implicit part of the implementation of 
Order EA-12-049, additional emergency response capabilities were 
included to address a beyond-design-basis external event that impacts 
multiple power reactor units, and potentially multiple source terms, on 
the site. In all cases, these additional proposed revisions are 
considered to be necessary to effectively mitigate such an event, 
consistent with the NRC's intent in issuing Order EA-12-049. These 
proposed requirements were not explicitly addressed in the previous 
regulatory basis documents issued for the two rulemakings that were 
consolidated into this rulemaking. This section discusses the basis for 
these proposed emergency response capability provisions.
    The March 12, 2012, Sec.  50.54(f) letters (i.e., Request for 
Information Pursuant to title 10 of the Code of Federal Regulations 
50.54(f)) requested information from the licensees that, in part, was 
intended to verify the adequacy of emergency planning to address what 
was then termed prolonged SBO \8\ and multi-unit events. The accident 
at Fukushima highlighted the need to determine and implement the 
required staff to fill all necessary positions responding to multi-unit 
events. Additionally, NRC recognizes that the communication equipment 
relied upon to coordinate the event response during an ELAP should be 
powered and maintained.
---------------------------------------------------------------------------

    \8\ While the letter made use of the term ``prolonged SBO,'' the 
request for information was for a loss of all alternating current 
power, which was subsequently termed an ELAP. The phrase ``prolonged 
SBO'' is retained here to accurately reflect the wording used in the 
letter.
---------------------------------------------------------------------------

1. Onsite and Offsite Communications Capability
    This proposed rule would require additional communications 
capabilities for events that result in extended loss of ac power 
onsite, or potential destruction of offsite communications 
infrastructure. Because of the destruction to communications capability 
that occurred at Fukushima, the NRC would propose requirements for 
licensees to provide a greater capability to communicate with onsite 
staff to support mitigation of the event, and to support offsite 
communications to gain any additional support or to perform emergency 
preparedness functions. The proposed requirements would support 
effective implementation of the FSGs and were included as part of the 
implementation of Order EA-12-049.
2. Staffing Assessment
    This proposed rule would require an assessment that is considered 
essential for effective implementation of the FSGs. This assessment 
matches the one that was conducted under the March 12, 2012, request 
for information that was developed to align with the requirements 
included in Order EA-12-049 (i.e., the staffing analysis specifically 
considered the staffing needs for implementing Order EA-12-049); 
licensees would not be required to repeat the staffing analysis. A 
lesson-learned from the Fukushima event is that there are increased 
staffing demands following a beyond-design-basis external event, and 
this coupled with the subsequent NRC requirements issued in Order EA-
12-049 required the staffing analysis to provide a level of assurance 
that the FSGs can be implemented. This provision would then support the 
proposed requirements of the rule to have sufficient staffing to 
implement the FSGs and EDMGs in conjunction with the EOPs.

[[Page 70630]]

3. Change Control
    The NRC would not require a power reactor applicant or licensee to 
address or implement the proposed communications and staffing analysis 
requirements through the licensee's or applicant's emergency plan or 
maintain the capabilities as a part of the emergency preparedness 
program. This approach would allow for site-specific flexibility in 
implementation. Therefore, the requirements of maintaining the 
communications and staffing analysis in an effective emergency plan and 
controlling changes to it under Sec.  50.54(q) would not apply when 
implementation of the requirements is not in the emergency plan, but in 
all cases, the change control process of this proposed rule would 
apply. However, if an applicant or a licensee incorporates the 
communications and staffing analysis into the emergency preparedness 
program through the emergency plan or emergency plan implementing 
procedures, the requirements of Sec.  50.54(q) would apply.
4. Multiple Source Dose Assessment Capability
    This proposed rule would require licensees to have a means for 
determining the magnitude of, and for continually assessing the impact 
of, the release of radioactive materials, including from all reactor 
core and spent fuel pool sources. A lesson learned from the Fukushima 
Dai-ichi event is that there is a potential for a beyond-design-basis 
external event to result in multiple source terms from multiple release 
points, and under such a situation, additional capabilities are 
necessary to support development of appropriate protective action 
recommendations. In COMSECY-13-0010, ``Schedule and Plans for Tier 2 
Order on Emergency Preparedness for Japan Lessons Learned,'' dated 
March 27, 2013, the NRC staff informed the Commission that licensees 
would provide information about their current multiple source term dose 
assessment capability, or a schedule for implementing such a 
capability, and that associated implementation would occur by the end 
of calendar year 2014. Licensee implementation of the multiple source 
term dose assessment capability would be verified by inspection under 
TI-2515/191, ``Inspection of the Licensee's Responses to Mitigation 
Strategies Order EA-12-049, Spent Fuel Pool Instrumentation Order EA-
12-051 and Emergency Preparedness Information Requested in NRC March 
12, 2012.'' The NRC has been working with the industry and stakeholders 
through public meetings to review and provide feedback on NEI 13-06, 
``Enhancements to Emergency Response Capabilities for Beyond Design 
Basis Accidents and Events,'' Revision 0, which, in part, would provide 
licensees with guidance on implementing a multiple source term dose 
assessment capability.
    The capability should be available to support responses during 
events both within and beyond the plant design basis. Also, the 
licensee should discuss the site's multi-unit and multiple source term 
dose assessment capability with the offsite response organizations, 
particularly, with the agencies that are responsible for making 
decisions on public protective action recommendations. Agreement on the 
methods and results would avoid unnecessary delays during the event in 
making the public protective action decisions, public notification, and 
the implementation of protective actions.
5. Technology-Neutral Emergency Response Data System
    The proposed requirements of 10 CFR part 50, appendix E, section 
VI, for the Emergency Response Data System (ERDS) would reflect the use 
of up-to-date technologies and remain technology-neutral so that the 
equipment supplied by NRC would continue to be replaced as needed, 
without the need for future rulemaking because equipment becomes 
obsolete. In 2005, the NRC initiated a comprehensive, multi-year effort 
to modernize all aspects of the ERDS, including the hardware and 
software that constitute the ERDS infrastructure at NRC headquarters, 
as well as the technology used to transmit data from licensed power 
reactor facilities. As described in NRC Regulatory Issue Summary 2009-
13, ``Emergency Response Data System Upgrade From Modem to Virtual 
Private Network Appliance,'' the NRC engaged licensees in a program 
that replaced the existing modems used to transmit ERDS data with 
Virtual Private Network (VPN) devices. The licensees now have less 
burdensome testing requirements, faster data transmission rates, and 
increased system security.

V. Section-by-Section Analysis

Proposed Sec.  50.8 Information Collection Requirements: OMB Approval

    This section, which lists all information collections in 10 CFR 
part 50 that have been approved by the Office of Management and Budget 
(OMB), is revised by adding a reference to Sec.  50.155, the mitigation 
of beyond-design-basis events rule. As discussed in the ``Paperwork 
Reduction Act Statement'' section of this document, the OMB has 
approved the information collection and reporting requirements in the 
final mitigation of beyond-design-basis events rule. No specific 
requirement or prohibition is imposed on applicants or licensees in 
this section.

Proposed Sec.  50.34 Contents of Applications; Technical Information

    Section 50.34 identifies the technical information that must be 
provided in applications for construction permits and operating 
licenses. Paragraphs (a) and (b) of this section identify the 
information to be submitted as part of the preliminary or final safety 
analysis report, respectively. New paragraph (i) of this section would 
identify information to be submitted as part of an operating license 
application, but not necessarily included in the final safety analysis 
report.
    The NRC is proposing an administrative change to Sec.  50.34(a)(13) 
and (b)(12) to remove the word ``stationary'' from the requirement for 
power reactor applicants who apply for a construction permit or 
operating license, respectively. Section 50.34(a)(13) and 50.34(b)(12) 
were added to the regulations in 2009 to reflect the requirements of 
Sec.  50.150(b) regarding the inclusion of information within the 
preliminary or final safety analysis reports for applicants subject to 
Sec.  50.150. Section 50.34(a)(13) and (b)(12) were inadvertently 
limited to ``stationary power reactors,'' matching the wording of Sec.  
50.34(a)(1), (a)(12), (b)(10), and (b)(11), which pertain to seismic 
risk hazards for stationary power reactors. The NRC does not intend to 
change the meaning of this requirement by removing the word 
``stationary'' from these requirements. This change is intended to 
ensure consistency in describing the types of applications to which the 
requirements apply.
    Proposed Sec.  50.34(i) would require each application for an 
operating license to include the applicant's plans for implementing the 
requirements of proposed Sec.  50.155 and 10 CFR part 50, appendix E, 
section VII, including a schedule for achieving full compliance with 
these requirements. This paragraph would also require the application 
to include a description of: (1) The integrated response capability 
that would be required by proposed Sec.  50.155(b); (2) the equipment 
upon which the strategies and guidelines that would be required by 
proposed Sec.  50.155(b)(1) rely, including the

[[Page 70631]]

planned locations of the equipment and how the equipment and SSCs would 
meet the design requirements of proposed Sec.  50.155(c); and (3) the 
strategies and guidelines that would be required by proposed Sec.  
50.155(b)(2).

Proposed Sec.  50.54 Conditions of Licenses

    Applicability of the requirements of Sec.  50.54(hh) is currently 
governed by Sec.  50.54(hh)(3), which makes these requirements 
inapplicable to a nuclear power plant for which the certifications 
required under Sec.  50.82(a) or Sec.  52.110(a)(1) have been 
submitted. This rulemaking proposes to renumber Sec.  50.54(hh)(3) to 
reflect the proposed movement of the requirements currently within 
Sec.  50.54(hh)(2) to proposed Sec.  50.155(b)(2). The proposed Sec.  
50.54(hh)(2) includes editorial changes to reflect that the 
applicability is to the licensee rather than the facility and to 
correct the section numbers for the required certifications. 
Additionally, proposed Sec.  50.54(hh)(2) clarifies that the 
inapplicability is dependent upon the NRC docketing of the 
certifications rather than licensee submittal because Sec.  50.82(a)(2) 
and Sec.  52.110(b) set the docketing of the certifications as the 
point at which operation of the reactor is no longer authorized and 
fuel cannot be placed in the reactor vessel.

Proposed Sec.  50.155(a), ``Applicability''

    Proposed Sec.  50.155(a) would describe which entities would be 
subject to this proposed rule. Proposed Sec.  50.155(a)(1) would 
provide that each holder of an operating license for a nuclear power 
reactor under part 50 and each holder of a combined license under part 
52 after the Commission has made the finding under Sec.  52.103(g) that 
the acceptance criteria have been met, would be required to comply with 
the requirements of this proposed rule until the time when the NRC has 
docketed the certifications described in Sec.  50.82(a)(1) or Sec.  
52.110(a). These certifications inform the NRC that the licensee has 
permanently ceased to operate the reactor and permanently removed all 
fuel from the reactor vessel. Upon the docketing of the certifications, 
by operation of law under Sec.  50.82(a)(2) or Sec.  52.110(b), the 
licensee's part 50 or 52 license, respectively, no longer authorizes 
operation of the reactor or emplacement or retention of fuel in the 
reactor vessel. At this point, many portions of this proposed rule 
would not apply to the licensee because the removal of fuel from the 
reactor vessel would eliminate the risk of a reactor-based beyond-
design-basis event and the need to prepare to mitigate those events. 
Proposed Sec.  50.155(a)(3) would set forth the requirements that would 
apply to the licensee with Sec.  50.82(a)(2) or Sec.  52.110(b) 
certification.
    Proposed Sec.  50.155(a)(2) would provide that each applicant for 
an operating license for a nuclear power reactor under part 50 and each 
holder of a combined license before the Commission makes the finding 
under Sec.  52.103(g) would be required to comply with the requirements 
of this proposed rule no later than the date on which the Commission 
issues the operating license under Sec.  50.57 or makes the finding 
under Sec.  52.103(g), respectively. Under this regulation, operating 
license applicants and COL holders would be in compliance with this 
proposed rule before they begin operating their reactors, thereby 
providing additional defense-in-depth capabilities at the inception of 
power operations.
    Proposed Sec.  50.155(a)(3) would address power reactor licensees 
that permanently stop operating and defuel their reactors and begin 
decommissioning the reactors. The proposed paragraph would provide that 
when an entity subject to the requirements of proposed Sec.  50.155 
submits to the NRC the certifications described in Sec.  50.82(a)(1) or 
Sec.  52.110(a), and the NRC dockets those certifications, then that 
licensee would be required to comply with the requirements of proposed 
Sec.  50.155(b) through (e) associated with maintaining or restoring 
secondary containment, if applicable, and spent fuel pool cooling 
capabilities for the reactor described in the Sec.  50.82(a)(1) or 
Sec.  52.110(a) certifications, except for the requirements in proposed 
Sec.  50.155(c)(4) and proposed in 10 CFR part 50, appendix E, section 
VII. In other words, the licensee could discontinue compliance with the 
requirements in proposed Sec.  50.155 associated with maintaining or 
restoring core cooling or the primary reactor containment functional 
capability for the reactor described in the Sec.  50.82(a)(1) or Sec.  
52.110(a) certifications. Compliance with the requirements of proposed 
Sec.  50.155(b) through (e) associated with maintaining or restoring 
secondary containment, if applicable, and spent fuel pool cooling 
capabilities would continue as long as spent fuel remains in the spent 
fuel pool(s) associated with the reactor described in the Sec.  
50.82(a)(1) or Sec.  52.110(a) certifications.
    Proposed Sec.  50.155(a)(3)(i) would discontinue the requirement to 
comply with proposed Sec.  50.155(b)(1) requirements concerning beyond-
design-basis event strategies and guidelines for spent fuel pool 
cooling capabilities, and any requirements based on compliance with 
proposed Sec.  50.155(b)(1), for certain licensees in decommissioning. 
These licensees would have to perform and retain an analysis 
demonstrating that sufficient time has passed since the fuel within the 
spent fuel pool was last irradiated such that the fuel's low decay heat 
and boil-off period provide sufficient time in an emergency for the 
licensee to obtain off-site resources to sustain the spent fuel pool 
cooling function indefinitely and therefore obviate the need to comply 
with proposed Sec.  50.155(b)(1) using installed or on-site portable 
equipment.
    Proposed Sec.  50.155(a)(3)(i) also would discontinue the 
requirement to comply with the remaining provisions of proposed Sec.  
50.155 except proposed Sec.  50.155(b)(2) when the fuel in the spent 
fuel pool reaches the point where beyond-design-basis event strategies 
and guidelines for spent fuel cooling capabilities would no longer be 
needed.
    Proposed Sec.  50.155(a)(3)(ii) would exempt the licensee for 
Millstone Power Station Unit 1, Dominion Nuclear Connecticut, Inc. from 
the requirements of proposed Sec.  50.155.
    Under proposed Sec.  50.155(a)(3), once a power reactor licensee 
has permanently stopped operating and defueled its reactor and has 
removed all irradiated fuel from the spent fuel pool(s) associated with 
the reactor described in the Sec.  50.82(a)(1) or Sec.  52.110(a) 
certifications, the licensee could cease compliance with all 
requirements in proposed Sec.  50.155 for the unit(s) described in the 
Sec.  50.82(a)(1) or Sec.  52.110(a) certifications.

Proposed Sec.  50.155(b), ``Integrated Response Capability''

    Proposed paragraph (b) would require that each applicant or 
licensee develop, implement, and maintain an integrated response 
capability that includes: (1) Mitigation strategies for beyond-design-
basis external events, (2) extensive damage mitigation guidelines, (3) 
integration of these strategies and guidelines with emergency operating 
procedures, (4) sufficient staffing to support implementation of the 
guidelines in conjunction with the EOPs, and (5) a supporting 
organizational structure with defined roles, responsibilities, and 
authorities for directing and performing these strategies, guidelines, 
and procedures. The intent is to require that the operating and 
combined license holders described in Sec.  50.155(a) be able to 
mitigate the consequences of a wide range of initiating events and 
plant

[[Page 70632]]

damage states that can challenge public health and safety.
    The specification of strategies, guidelines and procedures for the 
response capability not only defines the required scope of the 
capability but sets forth the expectation that the response capability 
must include planned methods for responding that are documented in some 
form of written instruction. To serve their function, these strategies, 
guidelines and procedures must be acted upon by individuals capable of 
understanding their appropriate application and implementing them. 
Accordingly, proposed Sec.  50.155(b)(4), in conjunction with proposed 
Sec.  50.155(d), would require that the response capability include an 
adequate number of personnel with the knowledge and skills to implement 
the strategies, guidelines and procedures and that the mitigation 
activities of these individuals be coordinated in accordance with a 
defined command and control structure as would be required by proposed 
Sec.  50.155(b)(5).
    Proposed Sec.  50.155(b) would specify that the integrated response 
capability be ``developed, implemented, and maintained.'' This language 
reflects NRC consideration that whereas certain elements of the 
integrated response capability have been developed and are currently in 
place (e.g., the EDMGs), other elements (e.g., guidelines to mitigate 
beyond-design-basis external events) may require additional efforts to 
complete and integrate. The term ``implement'' is used in proposed 
Sec.  50.155(b) to mean that the integrated response capability is 
established and available to respond, if needed (e.g., the licensee has 
approved the strategies, guidelines, and procedures for use). The term 
``maintain'' as used in proposed Sec.  50.155(b) reflects the NRC's 
intent that licensees ensure that the integrated response capability, 
once established, be preserved consistent with the change control 
provisions of proposed Sec.  50.155(g).
    Proposed Sec.  50.155(b)(1) would establish requirements for 
applicants and licensees to develop, implement and maintain strategies 
and guidelines to mitigate beyond-design-basis external events from 
natural phenomenon that result in an extended loss of ac power 
concurrent with either a loss of normal access to the ultimate heat 
sink or, for passive reactor designs, a loss of normal access to the 
normal heat sink. These provisions would require that the strategies 
and guidelines be capable of being implemented site-wide and include:
    i. Maintaining or restoring core cooling, containment, and spent 
fuel pool cooling capabilities; and
    ii. Enabling the use and receipt of offsite assistance and 
resources to support the continued maintenance of the functional 
capabilities for core cooling, containment, and spent fuel pool cooling 
indefinitely, or until sufficient site functional capabilities can be 
maintained without the need for the mitigation strategies.
    New reactors may establish different approaches from operating 
reactors in developing strategies to mitigate beyond-design-basis 
events. For example, new reactors may use installed plant equipment for 
both the initial and long-term response to an ELAP with less reliance 
on portable equipment and offsite resources than currently operating 
nuclear power plants. The NRC would consider the specific plant 
approach when evaluating the SSCs relied on as part of the mitigating 
strategies for beyond-design-basis events. Additional information on 
these strategies is provided in DG-1301, which would endorse an updated 
version of the industry guidance, for use by applicants and licensees, 
that incorporates lessons learned and feedback stemming from the 
implementation of Order EA-12-049, consistent with Commission 
direction.
    The proposed Sec.  50.155(b)(1) would limit the requirements for 
mitigation strategies to addressing ``external events from natural 
phenomena.'' This proposed language is meant to differentiate these 
requirements from those that currently exist within Sec.  50.54(hh)(2), 
which address beyond-design-basis external events leading to loss of 
large areas of the plant due to explosions and fire. This proposed 
provision also results in the need to have mitigation equipment be 
reasonably protected from the effects of external natural phenomena as 
discussed in later portions of this proposed notice.
    The proposed requirements to enable ``the acquisition and use of 
offsite assistance and resources to support the functions required by 
(b)(1)(i) of this section indefinitely, or until sufficient site 
functional capabilities can be maintained without the need for the 
mitigation strategies'' means that licensees would need to plan for 
obtaining sufficient resources (e.g., fuel for generators and pumps, 
cooling and makeup water) to continue removing decay heat from the 
irradiated fuel in the reactor vessel and spent fuel pool as well as to 
remove heat from containment as necessary until an alternate means of 
removing heat is established. The alternate means of removing heat 
could be achieved through repairs to existing SSCs, commissioning of 
new SSCs, or reduction of decay heat levels through the passage of time 
sufficient to allow heat removal through losses to the ambient 
environment. More detailed planning for offsite assistance and 
resources would be necessary for the initial period following the 
event; less detailed planning would be necessary as the event 
progresses and the licensee can mobilize additional support for 
recovery.
    Proposed Sec.  50.155(b)(2) would move requirements for EDMGs that 
currently exist in Sec.  50.54(hh)(2) to proposed Sec.  50.155(b)(2). 
This move would consolidate the requirements for beyond-design-basis 
strategies and guidance into a single section to promote efficiency in 
their consideration and allow for better integration. Although the 
wording of proposed Sec.  50.155(b)(2) differs from that of Sec.  
50.54(hh)(2), no substantive change in the requirements is intended.
    The preamble to Sec.  50.155(b)(2) that is contained in Sec.  
50.155(b) is worded so that it would require that licensees ``develop, 
implement, and maintain'' the strategies and guidance required in Sec.  
50.155(b)(2) rather than using the wording of Sec.  50.54(hh)(2) to 
require that licensees ``develop and implement'' the described guidance 
and strategies. The addition of the word ``maintain'' was proposed in 
order to correct an inconsistency with the wording of Sec.  
50.54(hh)(1), which was promulgated along with Sec.  50.54(hh)(2) in 
the Power Reactor Security Rulemaking, issued on March 27, 2009 (74 FR 
13926), and to clarify that the NRC considers the plain language 
meaning of the transitive verb ``to implement,'' ``to put into 
effect,'' as it was used in the context of Sec.  50.54(hh)(2) as 
including maintenance of the resulting guidance and strategies. The 
requirement as it was originally issued in the Interim Compensatory 
Measures Order, EA-02-026, dated February 25, 2002, was worded to 
require licensees to ``develop'' specific guidance, while the 
corresponding license conditions imposed by the conforming license 
amendment was worded to require each affected licensee to ``develop and 
maintain'' strategies. The NRC believes that the phrase ``develop, 
implement, and maintain'' would provide better clarity of what is 
necessary for compliance with the requirements without substantively 
changing the requirements.
    Proposed Sec.  50.155(b)(3) would establish requirements for 
licensees to integrate the strategies and guidelines in

[[Page 70633]]

(b)(1) and (2) with EOPs. The Commission's intent regarding integration 
of strategies, guidelines, and procedures was introduced in the 
section-by-section analysis of the proposed Sec.  50.155(b) requirement 
for an integrated response capability and is described further under 
``Integration with EOPs'' of Section IV.D, Proposed Rule Regulatory 
Bases.
    Proposed Sec.  50.155(b)(4) would establish requirements for 
licensees to provide the staffing necessary for having an integrated 
response capability to support implementation of the strategies and 
guidelines in proposed (b)(1) and (2). The number and composition of 
the response staff should be sufficient to implement mitigation 
strategies intended to maintain or restore the functions of core 
cooling, containment, and spent fuel pool cooling for all affected 
units. The word ``sufficient'' is used in the proposed paragraph to 
reflect its meaning ``adequate.''
    Proposed Sec.  50.155(b)(5) would establish requirements for 
licensees to have a supporting organizational structure with defined 
roles, responsibilities, and authorities for directing and performing 
the guidelines in (b)(1) and (2).

Proposed Sec.  50.155(c) Equipment Requirements

    Proposed Sec.  50.155(c)(1) would require that equipment relied on 
for the mitigation strategies of proposed paragraph (b)(1) have 
sufficient capacity and capability to simultaneously maintain or 
restore core cooling, containment, and spent fuel pool capabilities for 
all the power reactor units and spent fuel pools within the licensee's 
site boundary.
    The phrase sufficient ``capacity and capability'' in proposed Sec.  
50.155(c)(1) means that the equipment, and the instrumentation relied 
on to support the decision making necessary to accomplish the 
associated mitigating strategies of Sec.  50.155(b)(1), should have the 
design specifications necessary to assure that it would function and 
provide the requisite plant information when subjected to the 
conditions it is expected to be exposed to in the course of the 
execution of those mitigating strategies. These design specifications 
would include appropriate consideration of environmental conditions 
that are predicted in the thermal-hydraulic and room heat up analyses 
used in the development of the mitigating strategies responsive to 
Sec.  50.155(b)(1).
    Proposed Sec.  50.155(c)(2) would require reasonable protection of 
the Sec.  50.155(b)(1) equipment rather than the treatment of SSCs 
important to safety under GDC-2, which requires that those SSCs be 
designed to withstand the effects of natural phenomena without loss of 
capability to perform their safety functions. The phrase ``reasonable 
protection'' was initially proposed in recommendation 4.2 of the NTTF 
Report in the context of a proposed NRC Order to licensees to require 
``reasonable protection'' of equipment required by Sec.  50.54(hh)(2) 
from the effects of design-basis external events along with providing 
additional sets of equipment as an interim measure during a subsequent 
rulemaking on prolonged SBO. The NTTF based this recommendation on the 
potential usefulness of the EDMGs in circumstances that do not involve 
loss of a large area of the plant and explained that reasonable 
protection from external events as used in the NTTF Report meant that 
the equipment must ``be stored in existing locations that are 
reasonably protected from significant floods and involve robust 
structures with enhanced protection from seismic and wind-related 
events.''
    The NRC carried forward the use of the phrase ``reasonable 
protection'' in Order EA-12-049 with regard to the protection required 
for equipment associated with the mitigation strategies. That Order did 
not, however, define ``reasonable protection.'' The NRC guidance in 
JLD-ISG-2012-01 discussed ``reasonable protection'' as follows:

    Storage locations chosen for the equipment must provide 
protection from external events as necessary to allow the equipment 
to perform its function without loss of capability. In addition, the 
licensee must provide a means to bring the equipment to the 
connection point under those conditions in time to initiate the 
strategy prior to expiration of the estimated capability to maintain 
core and spent fuel pool cooling and containment functions in the 
initial response phase.

    In JLD-ISG-2012-01, the NRC endorsed NEI 12-06, Revision 0, as 
providing an acceptable method to provide reasonable protection, 
storage, and deployment of the equipment associated with Order EA-12-
049. The NEI 12-06, Revision 0, also omitted a definition for the 
phrase ``reasonable protection,'' but did provide guidelines for use by 
licensees for protecting the equipment from the hazards that would be 
commonly applicable: (1) Seismic hazards; (2) flooding hazards; (3) 
severe storms with high winds; (4) snow, ice and extreme cold; and(5) 
high temperatures. These guidelines included the use of structures 
designed to or evaluated equivalent to American Society for Civil 
Engineers (ASCE) Standard 7-10, ``Minimum Design Loads for Buildings 
and Other Structures,'' for the seismic and high winds hazards, rather 
than requiring the use of a structure that meets the plant's design 
basis for the Safe Shutdown Earthquake or high winds hazards including 
missiles. The NEI 12-06 guidelines also allow storage of the equipment 
above the flood elevation from the most recent site flood analysis, 
storage within a structure designed to protect the equipment from the 
flood, or storage below the flood level if sufficient time would be 
available and plant procedures would address the need to relocate the 
equipment above the flood level based on the timing of the limiting 
flood scenario(s). The NEI 12-06 guidelines further provide that 
multiple sets of equipment may be stored in diverse locations in order 
to provide assurance that sufficient equipment would remain deployable 
to assure the success of the strategies following an initiating event. 
The NRC-endorsed guidelines in NEI 12-06 do not consider concurrent, 
unrelated beyond-design-basis external events to be within the scope of 
the initiating events for the mitigating strategies. There is an 
assumption of a beyond-design-basis external event that establishes the 
event conditions for reasonable protection, and then it is assumed that 
the event leads to an ELAP and LUHS. But, for example, there is not an 
assumption of multiple beyond-design-basis external events occurring at 
the same time. As a result, reasonable protection for the purposes of 
compliance with Order EA-12-049 would allow the provision of specific 
sets of equipment for specific hazards with the required protection for 
those sets of equipment being against the hazard for which the 
equipment is intended to be used.
    The NRC proposes to continue the use of the phrase ``reasonable 
protection'' in proposed Sec.  50.155(c)(2) in order to distinguish the 
character of the required protection of GDC-2, which requires that SSCs 
important to safety be designed to withstand the effects of natural 
phenomena, from that of proposed Sec.  50.155(c)(2), which would allow 
damage to or loss of specific pieces of equipment so long as the 
capability to use some of the equipment to accomplish its intended 
purpose is retained. ``Reasonable protection'' would also allow for 
protection of the equipment using structures that could deform as a 
result of natural phenomena so long as the equipment could be

[[Page 70634]]

deployed from the structure to its place of use.
    The remaining portion of proposed Sec.  50.155(c)(2) would set the 
hazard level for which ``reasonable protection'' of the equipment must 
be provided. The hazard level would be the level determined for the 
design basis for the facility for protection of safety-related SSCs 
from the effects of natural phenomena, or, for the seismic or flooding 
hazards, the greater of the hazard level determined for the design 
basis for the facility and the licensee's reevaluated hazards, stemming 
from the March 12, 2012, NRC letter issued under Sec.  50.54(f). The 
timing for the proposed requirement for reasonable protection against 
the reevaluated hazards is set by Sec.  50.155(g) at 2 years following 
the effective date of this proposed rule. Operating power reactor 
licensees that were requested to reevaluate their seismic and flooding 
hazard levels by the NRC by letter dated March 12, 2012, under 10 CFR 
50.54(f) are currently on a submittal and NRC review schedule to have 
confirmation of the reevaluated hazard levels by December 2015. Given 
that the rulemaking schedule for this proposed rule is to provide the 
final rule to the Commission in December 2016, the anticipated 
effective date of the final rule would be mid-to-late 2017. Requiring 
compliance within 2 years following the effective date of the final 
rule would allow licensees with a new hazard level the opportunity to 
take measurements to support any necessary plant modifications during 
the first refueling outage following NRC confirmation of those levels 
and the opportunity to implement those modifications in a subsequent 
refueling outage after the effective date of the rule. The NRC is 
requesting feedback on this proposed implementation schedule in section 
VI of this notice.
    Proposed paragraph (c)(3) would require that licensees perform 
adequate maintenance on the equipment relied on for the mitigation 
strategies responsive to proposed paragraph (b)(1) to assure that the 
equipment is capable of fulfilling its intended function following a 
beyond-design-basis external event. The phrase ``adequate maintenance'' 
means sufficient routine maintenance and testing are performed, 
reflecting the storage and readiness conditions of the equipment, for a 
licensee to conclude that the equipment is capable of performing its 
function to a degree that would support the successful execution of the 
mitigation strategies of paragraph (b)(1). Provision of ``adequate 
maintenance'' also entails the establishment of a system of 
programmatic controls for the equipment to limit the quantity of 
equipment taken out of service for maintenance and testing in order to 
limit the unavailability of that equipment appropriately and to provide 
assurance that sufficient equipment would remain available to satisfy 
proposed paragraph (c)(1).
    Proposed paragraph (c)(4) would make generically applicable the 
requirements of Order EA-12-051 by requiring that licensees include a 
reliable means to remotely monitor wide-range spent fuel pool levels to 
support effective prioritization of event mitigation and recovery 
actions.

Proposed Sec.  50.155(d) Training Requirements

    Proposed Sec.  50.155(d) would require that each licensee specified 
in Sec.  50.155(a) provide for the training and qualification of 
personnel that perform activities in accordance with the strategies and 
guidelines identified in Sec.  50.155(b)(1) and (2).

Proposed Sec.  50.155(e) Drills and Exercises

    Proposed Sec.  50.155(e) would require that each licensee and 
applicant specified in Sec.  50.155(a) conduct drills and exercises for 
personnel that would perform activities in accordance with the 
strategies and guidelines identified in Sec.  50.155(b)(1) and (2). The 
use of drills and exercises allows demonstration and evaluation of the 
licensee's capability to execute the integrated response capability 
required by Sec.  50.155(b) mitigation strategies and guidelines in 
light of the specific plant damage and operational conditions presented 
by an initiating event. ``Integrated'' is used to describe the 
licensee's or applicant's approach to using all tools, spaces, 
qualified personnel and resources during a performance enhancing 
experience to the furthest extent practical given a set of initiating 
conditions and within the bounds of a drill or exercise scenario. When 
two or more strategies or guidelines in Sec.  50.155(b)(1) and (2) are 
potentially useful, ``integrated'' is meant that transitions to and 
from one set of strategies or guidelines in Sec.  50.155(b)(1) and (2) 
to another are coordinated.
    This proposed rule uses the words ``drill'' and ``exercise'' as 
they are defined in NUREG-0654/FEMA-REP-1, Revision 1,\9\ meaning an 
evaluated performance-enhancing experience that reasonably simulates 
the interactions between appropriate centers, work groups, strike 
teams, or individuals that would be expected to occur during the event. 
For the initial drill or exercise, the licensee would be required to 
demonstrate its capability to transition to and use one or more of the 
strategies that would be required by Sec.  50.155(b)(1) and (2) from 
the AOPs or EOPs, whichever would govern for the initiating event and 
plant degraded conditions, using the equipment and communication 
systems used for the EOPs and guidelines.
---------------------------------------------------------------------------

    \9\ Planning Standards N.1 Exercise and N.2 Drills.
---------------------------------------------------------------------------

    Proposed Sec.  50.155(e)(1) would require the initial drill or 
exercise to be conducted within 12 months prior to the issuance of the 
first operating license (OL) for the unit described in the application. 
This would allow the license applicant to implement any improvements or 
corrective actions identified during the drill or exercise, and allow 
the Commission to consider the results of any drill or exercise actions 
in the decision on whether to authorize the OL. Because Sec.  
50.155(e)(1) applies only to applicants for operating licenses, it 
would not apply to holders of operating licenses under 10 CFR part 50, 
who are subject to proposed Sec.  50.155(e)(4), or holders of combined 
licenses under 10 CFR part 52, who are subject to proposed Sec.  
50.155(e)(2) through (4). Following issuance of the operating license, 
the applicant, as a licensee, would be subject to proposed Sec.  
50.155(e)(3).
    Proposed Sec.  50.155(e)(2) would require the licensee to conduct 
an initial drill or exercise that demonstrates the capability to 
transition from the AOPs or EOPs, use one or more of the strategies and 
guidelines in paragraphs (b)(1) and (2) of this section, and use 
communications equipment required in 10 CFR part 50, appendix E, 
section VII, no more than 12 months before the date specified for 
completion of the last inspections, tests, and analyses in the 
inspections, tests, analyses, and acceptance criteria (ITAAC) 
completion schedule as required by Sec.  52.99(a) for the unit 
described in the combined license.
    This proposed rule would set the completion date for the initial 
drill or exercise at ``no more than 12 months before the date specified 
for completion of the last inspections, tests, and analyses in the 
ITAAC completion schedule required by Sec.  52.99(a) for the unit 
described in the combined license'' in order to allow the licensee to 
implement any improvements or corrective actions identified during the 
drill or exercise, and allow the Commission to consider the results of 
any drill or exercise actions.
    The proposed Sec.  50.155(e)(2) requirement for initial drills or 
exercises is limited to holders of combined

[[Page 70635]]

licenses under 10 CFR part 52 before the Commission has made the 
finding under Sec.  52.103(g). A combined license holder for whom the 
Commission has already made the finding under Sec.  52.103(g) as of the 
effective date of the rule would not be subject to proposed Sec.  
50.155(e)(2), but would instead be subject to Sec.  50.155(e)(4) for 
the proposed initial drill requirements.
    Proposed Sec.  50.155(e)(3) would require holders of operating 
power reactor licenses issued under 10 CFR part 50 subsequent to the 
effective date of this rule, and holders of combine licenses issued 
under 10 CFR part 52 for whom the Commission has made the finding under 
Sec.  52.103(g) subsequent to the effective date of this rule, to 
conduct subsequent drills, exercises, or both that collectively 
demonstrate a capability to use at least one of the strategies and 
guidelines in each of proposed Sec.  50.155(b)(1) and (2) in succeeding 
8-year intervals. This would require that the drills and exercises 
performed to demonstrate this capability include transitions from other 
procedures and guidelines, as applicable, and the use of communications 
equipment that would be required by proposed 10 CFR part 50, appendix 
E, section VII. This proposed requirement differs from the proposed 
Sec.  50.155(e)(1) and (2) initial demonstration requirement, in that 
it would require licensees to demonstrate a continuing capability, and 
as such, it is structured to require licensees to demonstrate at least 
one of the strategies and guidelines from each of the guidelines during 
the 8-year interval.
    Proposed Sec.  50.155(e)(4) would require holders of operating 
licenses or combined licenses for which the Commission has made the 
finding under Sec.  52.103(g) to conduct an initial drill or exercise 
that demonstrates the capability to transition to and use one or more 
of the strategies and guidelines in proposed Sec.  50.155(b)(1) and (2) 
and use communications equipment required in 10 CFR part 50, appendix 
E, section VII. Proposed Sec.  50.155(e)(4) would be equivalent to 
proposed Sec.  50.155(e)(1) and (2) for initial drills or exercises, 
but would apply to current licensees. Following this initial drill or 
exercise, the licensee would be required to conduct subsequent drills, 
exercises, or both that collectively demonstrate a capability to use at 
least one of the strategies and guidelines in each of proposed Sec.  
50.155(b)(1) and (2) in succeeding 8-year intervals. Proposed Sec.  
50.155(e)(4) would be equivalent to proposed Sec.  50.155(e)(3) for 
subsequent drills or exercises, but would apply to current licensees 
under 10 CFR part 50 and those under 10 CFR part 52 for whom the 
Commission has made the finding under Sec.  52.103(g) as of the 
effective date of the rule.

Proposed Sec.  50.155(f) Change Control

    Proposed Sec.  50.155(f) would establish requirements that govern 
changes in the implementation of the requirements of proposed Sec.  
50.155 and 10 CFR part 50, appendix E, section VII. Prior to 
implementing a proposed change, proposed Sec.  50.155(f)(1) would 
require the licensee to perform an evaluation to ensure that the 
provisions of proposed Sec.  50.155 and 10 CFR part 50, appendix E, 
section VII, continue to be met. Proposed Sec.  50.155(f)(2) would 
require that licensees maintain documentation of the paragraph (f)(1) 
evaluations until the requirements of this proposed Sec.  50.155 and 10 
CFR part 50, appendix E, section VII, no longer apply. Finally, 
proposed Sec.  50.155(f)(3) would inform licensees that proposed 
changes must continue to be subject to all other applicable change 
control processes.

Proposed Sec.  50.155(g) Implementation

    Proposed Sec.  50.155(g) would set schedules for compliance for 
different classes of licensees depending on the circumstances unique to 
each class. Paragraphs (g)(1) and (2) would require licensees of 
operating reactors to comply with all requirements within 2 years of 
the effective date of the rule.

Proposed 10 CFR Part 50, Appendix E, Section I, Introduction

    The NRC proposes adding the sentence, ``Section VII of this 
appendix also provides for `Communications and Staffing Requirements 
for the Mitigation of Beyond-Design-Basis Events' that do not need to 
be contained within a licensee's emergency plan'' to the end of 
paragraph I.2. The NRC is not proposing to require an applicant or 
licensee to address or implement the proposed requirements in Section 
VII of Appendix E through the applicant's or licensee's emergency plan 
or to maintain the capabilities as a part of the emergency preparedness 
program. This would allow for site-specific flexibility in 
implementation.

Proposed 10 CFR Part 50, Appendix E, Section IV.B, Assessment Actions

    The NRC proposes adding the phrase, ``including from all reactor 
core and spent fuel pool sources,'' into paragraph B.1 following 
``determining the magnitude of, and for continually assessing the 
impact of, the releases of radioactive materials.'' This proposed rule 
would require all licensees to establish the capability to perform 
offsite dose assessments during an event involving concurrent 
radiological releases from all on-site units and spent fuel pools, and 
for multiple release points. The capability would quantify the total 
releases from the site and estimate the offsite dose consequences.

Proposed 10 CFR Part 50, Appendix E, Section IV.E, Emergency Facilities 
and Equipment

    The NRC proposes adding the phrase, ``including from all reactor 
core and spent fuel pool sources,'' into paragraph E.2 following 
``equipment for determining the magnitude of, and for continuously 
assessing the impact of, the release of radioactive materials to the 
environment.'' This proposed rule would require that equipment used for 
multi-unit dose assessment be maintained in a ready state.

Proposed 10 CFR Part 50, Appendix E, Section IV, Training

    This proposed rule would move the Sec.  50.54(hh)(2) exercise 
requirement from 10 CFR part 50, appendix E, section IV.F.2.j, to Sec.  
50.155(e). This move would change the exercise requirement to a drill 
requirement, aligning the requirement with the mitigation strategies 
drill requirements described in Sec.  50.155(e).
    This proposed rule would also require that periodic opportunities 
for a performance-enhancing experience should be provided to personnel 
responsible for performing multiple source term dose assessment and 
assessing the results in accordance with the site's emergency plan and 
implementing procedures.

Proposed 10 CFR Part 50, Appendix E, Section VI, Emergency Response 
Data Systems

    The NRC proposes to change its Emergency Response Data Systems 
regulations to require the use of technology-neutral equipment. The NRC 
proposes to restate the requirements in paragraph 3.c to replace the 
phrase ``onsite modem'' with ``equipment'' and removing references to a 
specific ``unit'' or equipment use.

Proposed 10 CFR Part 50, Appendix E, Section VII, Communications and 
Staffing Requirements for the Mitigation of Beyond-Design-Basis Events

    Proposed section VII would require power reactor applicants and 
licensees to conduct a detailed analysis to provide the basis for the 
staffing necessary for responding to a beyond-design-basis external 
event as described in Sec.  50.155(b)(1) during an extended loss of ac 
power (ELAP), and while access to the plant and normal access to the

[[Page 70636]]

ultimate or normal heat sink are lost. Additionally, the proposed 
section VII would require power reactor applicants and licensees to 
maintain at least one onsite and one offsite communications system 
functional during an ELAP and a loss of the local communication 
infrastructure.
    The current rule in 10 CFR part 50, appendix E, section IV.E.9, 
requires, ``At least one onsite and one offsite communication system; 
each system shall have a backup power source.'' However, the current 
rule doesn't address an interruption in the offsite communication 
services. This proposed rule would require the power reactor applicants 
and licensees to maintain the communication capabilities of 
communication amongst onsite staff and between onsite staff and offsite 
personnel in light of the lessons learned at Fukushima Dai-ichi. 
Furthermore, this proposed rule would require the power reactor 
applicants and licensees to submit the staffing analysis, results and 
implementation plans to meet the requirements, and the submissions 
would afford the NRC the opportunity to identify any common industry 
implementation problems and address them in guidance.
    This proposed rule would require an applicant for an operating 
license to complete a detailed staffing analysis at least 2 years 
before the issuance of the first operating license for full power (one 
authorizing operation above 5 percent of rated thermal power). The time 
frame allows the applicant to implement any improvements or corrective 
actions identified during the analysis, and the results of any analysis 
to inform the Commission's decision in authorizing the operating 
license.
    This proposed rule would require that an applicant for a combined 
license conduct a detailed staffing analysis and submit the analysis 
and results to the NRC 2 years before the date specified for completion 
of the last inspections, tests, and analyses in the ITAAC completion 
schedule required by Sec.  52.99(a) for the unit described in the 
combined license. The time frame allows the applicant to implement any 
staffing and communications system improvements and corrective actions 
identified during the analysis.
    This proposed rule would provide that when the NRC has docketed the 
certifications described in Sec.  50.82(a)(1) or Sec.  52.110(a) for a 
power reactor licensee, then that licensee would no longer be subject 
to section VII of appendix E to 10 CFR part 50 for the unit described 
in the Sec.  50.82(a)(1) or Sec.  52.110(a) certifications.

Proposed Sec.  52.80 Contents of Applications; Additional Technical 
Information

    Section 52.80 identifies the required additional technical 
information to be included in an application for a combined license. 
Proposed paragraph (d) would be amended to require a combined license 
applicant to include the applicant's plans for implementing the 
requirements of proposed Sec.  50.155 and 10 CFR part 50, appendix E, 
section VII, including a schedule for achieving full compliance with 
these requirements. This paragraph would also require the application 
to include a description of: (1) The integrated response capability 
that would be required by proposed Sec.  50.155(b); (2) the equipment 
upon which the strategies and guidelines that would be required by 
proposed Sec.  50.155(b)(1) rely, including the planned locations of 
the equipment and how the equipment and SSCs would meet the design 
requirements of proposed Sec.  50.155(c); and (3) the strategies and 
guidelines that would be required by proposed Sec.  50.155(b)(2).

VI. Specific Requests for Comments

    The NRC is seeking advice and recommendations from the public on 
this proposed rule. We are particularly interested in comments and 
supporting rationale from the public on the following:
    1. Change Control. The provisions governing change control in 
proposed Sec.  50.155(f) do not contain a criterion or a set of 
criteria that would establish a threshold beyond which prior NRC review 
and approval would be necessary to support a proposed change to the 
facility impacting the beyond-design-basis aspects of this proposed 
rulemaking and its supporting implementation guidance. For example, a 
set of criteria that asks whether a proposed facility change adversely 
impacts the capability to maintain and restore core cooling, 
containment, and spent fuel pool cooling capabilities, in conjunction 
with a criterion that asks whether the proposed facility change 
adversely impacts the supporting equipment requirements in proposed 
paragraph (c) might be sufficient for judging whether changes to the 
facility that impact the implementation of the mitigation strategies of 
proposed (b)(1) require prior NRC review and approval. What are 
stakeholders' views on this proposed change control structure, and what 
do stakeholders suggest for revising the change control process to 
contain criteria for determining the need for prior NRC review and 
approval?
    2. Application of Other Change Control Processes. Proposed Sec.  
50.155(f)(3) contains a requirement for licensees to use all applicable 
change control processes for facility changes, and not simply apply 
proposed paragraph (f) (i.e., the proposed change control process of 
paragraph (f) is only applicable to facility changes with respect to 
their beyond-design-basis aspects and to the extent that such changes 
impact implementation of the requirements of proposed Sec.  50.155 or 
the proposed 10 CFR part 50, appendix E, section VII) to the exclusion 
of other change control processes. This recognizes that facility 
changes can impact multiple aspects of the plant having different 
applicable requirements, and being subject to different change control 
requirements. For example, a licensee may want to make a facility 
change (e.g., a physical connection device) to support implementation 
of the beyond-design-basis external event mitigation strategies, and 
this change might impact safety-related SSCs. In addition to applying 
the new change control provision to ensure beyond-design-basis aspects 
of the proposed change result in continued compliance with the new 
requirements of this proposed rule, the licensee would also need to 
apply 10 CFR 50.59 to ensure that the facility change does not, due to 
its impact on safety-related SSCs, require prior NRC approval. The NRC 
requests feedback on the need for this proposed provision, or 
suggestions on how it might be improved.
    3. Reasonable Protection. This proposed rule contains a requirement 
in proposed Sec.  50.155(c)(2) that equipment supporting the proposed 
mitigation requirements of paragraph (b)(1) be ``reasonably protected'' 
from the effects of natural phenomenon including both those in the 
current plant design basis as well as the reevaluated hazards under the 
March 12, 2012, Sec.  50.54(f) request concerning flooding and seismic 
hazards. As a practical matter, implementation of Order EA-12-049 began 
before the reevaluated hazard information was available. The NRC 
recognizes that licensees were mindful of the hazard information, and 
attempted to address it during implementation. The NRC requests 
feedback concerning any costs and impacts that licensees would expect 
to occur as a result of this proposed requirement to include such 
things as rework or changes to previously implemented mitigation 
strategies.
    4. Mitigation of Beyond-Design-Basis Events Staffing Analysis. 
Proposed 10 CFR part 50, appendix E, section VII,

[[Page 70637]]

would require an analysis for the staffing necessary to support 
mitigation of a beyond-design-basis external event. This requirement 
would supplement the separate staffing analysis requirement that 
already exists in 10 CFR part 50, appendix E, section IV.A.9. The 
reason for the two separate staffing analysis requirements is related 
to the historical imposition of the requirements for the staffing 
analyses in the emergency preparedness rulemaking of 2011 and the March 
12, 2012, Request for Information under 10 CFR 50.54(f). The NRC is 
seeking feedback on whether it would be more efficient in practice for 
the two staffing analyses and their corresponding requirements to be 
combined, particularly for future reactor applicants. Would there be 
any unintended consequences to keeping the analyses separate or 
combining them? Is there a better way of achieving the underlying 
purpose of this requirement?
    5. Training Requirements. Section 50.155(d) of this proposed rule 
would require licensees to provide for the training and qualification 
of personnel that perform activities in accordance with the strategies 
and guidelines identified in paragraphs (b)(1) and (2) (i.e., 
mitigation strategies for beyond-design-basis external events and 
extensive damage mitigation guidelines) using the SAT process as 
defined in Sec.  55.4. The NRC notes that whereas many individuals at 
licensee facilities that would be subject to this proposed rule are 
trained under the SAT process (e.g., individuals specified under Sec.  
50.120), some individuals (e.g., firefighting and emergency 
preparedness personnel) may be currently trained under programs that 
are not required by NRC regulation to use the SAT process (e.g., 
National Fire Protection Association standards for training and 10 CFR 
part 50, appendix E). It is not the NRC's intent to extend the 
requirement for SAT-based training to the entirety of such programs. 
Rather, the intent of the proposed requirement would be to ensure that 
any training that is not currently part of existing programs but would 
be needed for performing activities in accordance with the strategies 
and guidelines identified in paragraphs proposed Sec.  50.155(b)(1) and 
(2) be identified and provided for in accordance with the SAT process. 
The NRC requests comment on potential unintended consequences of the 
proposed rule language for programs not currently required to be SAT-
based and if unintended consequences are identified, proposed 
alternative language for requiring the necessary amendments to such 
programs.
    6. Drill or Exercise Frequency. Proposed Sec.  50.155(e)(3) and (4) 
would require that following an initial drill or exercise, licensees 
would be required to conduct subsequent drills, exercises, or both, 
that collectively demonstrate a capability to use at least one of the 
strategies and guidelines in each of proposed Sec.  50.155(b)(1) and 
(2) in succeeding 8-year intervals. This would require that the drills 
or exercises performed to demonstrate this capability include 
transitions from other procedures and guidelines as applicable, and the 
use of communications equipment that would be required by proposed 10 
CFR part 50, appendix E, section VII, and that licensees shall not 
exceed 8 years between any consecutive drills or exercises. These 
requirements would be separate from the 8-year emergency preparedness 
exercise cycle requirements in 10 CFR part 50, appendix E, section 
IV.F. The NRC is seeking feedback on whether the drill or exercise 
frequency proposed by Sec.  50.155(e)(3) and (4) is appropriate.
    7. Equipment Requirements. Proposed Sec.  50.155(c)(1) would 
require the capacity and capability of the equipment relied on for the 
mitigation strategies required by proposed Sec.  50.155 (b)(1) to be 
sufficient to simultaneously maintain or restore core cooling, 
containment, and spent fuel pool cooling capabilities for all the power 
reactor units within the site boundary. Additionally, proposed Sec.  
50.155(c)(3) would require the equipment relied on for the mitigation 
strategies in proposed Sec.  50.155(b)(1) to receive adequate 
maintenance such that the equipment is capable of fulfilling its 
intended function. The intent of these two proposed provisions is to 
make elements of Order EA-12-049 generically-applicable. Order EA-12-
049 did not contain a specific maintenance requirement, but instead 
contained a performance-based requirement ``to develop, implement and 
maintain strategies,'' and failure to perform adequate maintenance 
would likely lead to a failure to meet this more general requirement, 
which is also contained in proposed Sec.  50.155(b)(1). Additionally, 
the supporting guidance for this proposed rule for proposed Sec.  
50.155(b)(1) carries forward the same approach that was used for 
implementation of Order EA-12-049, and contains a number of 
programmatic controls that in an analogous fashion to the maintenance 
provision in proposed Sec.  50.155(c)(3), if not followed, would likely 
lead to a loss of equipment capacity and capability and result in a 
failure to comply with the proposed Sec.  50.155(b)(1). Therefore, the 
NRC would like stakeholder views on the need for a separate maintenance 
provision.
    8. Equipment Protection Implementation Deadline. The NRC is 
proposing to require licensees to reasonably protect the equipment 
relied upon to implement the mitigation strategies required by proposed 
Sec.  50.155(b)(1). That equipment would need to be reasonably 
protected from the effects of natural phenomena that are, at a minimum, 
equivalent to the design basis of the facility. This proposed rule 
would require each licensee that received the March 12, 2012, NRC 
letter issued under Sec.  50.54(f) to provide reasonable protection 
against that reevaluated seismic or flooding hazard(s) by 2 years 
following the effective date of the final rule, if the reevaluated 
hazard exceeds the design basis of its facility. This is based on the 
anticipated completion dates for the licensees' hazard reevaluations 
and their confirmation by the NRC and the potential need for planning 
and implementing modifications during refueling outages. The NRC 
recognizes that certain licensees may need input into their analyses of 
reevaluated hazards from other government agencies, without any 
certainty of when that input would be provided. This reliance on 
information from other entities could remove from the licensee's 
control the ability to comply with the rule by a specific date. The NRC 
requests comments on the proposed implementation schedule, including 
suggestions for the criteria that licensees would need to satisfy to 
extend the schedule.
    9. Methodology for addressing reevaluated hazards. In SRM-COMSECY-
14-0037, the Commission affirmed that: (1) Licensees for operating 
nuclear power plants need to address the reevaluated flooding hazards 
within their mitigating strategies for beyond-design-basis external 
events; and (2) licensees for operating nuclear power plants may need 
to address some specific flooding scenarios that could significantly 
damage the power plant site by developing targeted or scenario-specific 
mitigating strategies, possibly including unconventional measures, to 
prevent fuel damage in reactor cores or spent fuel pools. The NRC is 
proposing to require licensees for operating nuclear power plants to 
address the reevaluated flooding hazard levels by reasonably protecting 
the mitigating strategies equipment to those levels if they exceed the 
design-basis flood level

[[Page 70638]]

for the facility. Alternatively, the NRC could: (1) Place this 
requirement within Sec.  50.155(b)(1) as a condition the associated 
strategies and guidelines must be capable of addressing; or (2) include 
a separate requirement for targeted or scenario-specific mitigating 
strategies as an option to address the reevaluated flooding hazards. 
The NRC seeks comment on whether the first of these options would be a 
better means to communicate the need for a licensee's strategies and 
guidelines to be capable of execution in the context of the new 
flooding hazard levels than including the requirement in Sec.  
50.155(c)(2). The NRC seeks additional comment on whether it would be 
appropriate to allow further flexibility in the licensee's strategies 
and guidelines by establishing an alternative means of compliance that 
does not include the surrogate condition of a loss of all alternating 
current power for specific beyond-design-basis conditions such as the 
reevaluated flooding hazards. For example, if a licensee could protect 
their internal power distribution system and emergency diesel 
generators from the reevaluated flooding hazard, it may not be 
necessary for the licensee to assume the loss of all alternating 
current power.
    10. Command and Control. Requirements for command and control and 
organizational structures currently exist in numerous locations, 
including 10 CFR part 50, appendix E, section IV.A, as well as within 
the typical administrative controls portions of technical 
specifications for power reactor licensees. These requirements do not 
plainly limit the scope of the roles, responsibilities and authorities 
to events within the design or licensing basis of the facility, 
although past NRC practice has been to treat these requirements in that 
manner. This proposed rule includes a further requirement on the 
subject in order to clarify the scope of what is required for 
organizational structures at power reactor licensees. Alternatively, 
the NRC is considering whether the expansion of scope of regulatory 
oversight of the organizational structures would require imposition of 
a new requirement or the expansion of scope would be better 
accomplished by communicating the understanding that the scope of the 
existing requirements covers the full spectrum of events that would be 
included in this rulemaking. The latter method of accomplishing this 
would have the potential advantage of leaving the requirements for 
command and control and organizational structures in a single 
regulation (i.e., 10 CFR part 50, appendix E, section IV.A). The NRC 
seeks stakeholder input on this subject.

VII. Regulatory Flexibility Certification

    Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the NRC 
certifies that this rule would not, if promulgated, have a significant 
economic impact on a substantial number of small entities. This 
proposed rule affects only the licensing and operation of nuclear power 
plants. The companies that own these plants do not fall within the 
scope of the definition of ``small entities'' set forth in the 
Regulatory Flexibility Act or established in 10 CFR 2.810, ``NRC size 
standards.''

VIII. Availability of Regulatory Analysis

    The NRC has prepared a draft regulatory analysis on this proposed 
regulation. The analyses examine the costs and benefits of the 
alternatives considered by the NRC. The NRC requests public comment on 
the draft regulatory analysis. The draft regulatory analysis is 
available as indicated in the ``Availability of Documents'' section of 
this document. Comments on the draft analysis may be submitted to the 
NRC as indicated in the ADDRESSES section of this document.

IX. Availability of Guidance

    The NRC is issuing for comment draft regulatory guidance (DG) to 
support the implementation of the proposed requirements in this 
rulemaking. You may access information and comment submissions related 
to the DGs by searching on http://www.regulations.gov under Docket ID 
NRC-2014-0240.
    The DG-1301, ``Flexible Mitigation Strategies for Beyond-Design-
Basis Events,'' provides licensees and applicants with an acceptable 
method of responding to an ELAP and demonstrating compliance with the 
proposed regulations requiring additional defense-in-depth measures for 
the mitigation of beyond-design-basis external events.
    The DG-1317, ``Wide-Range Spent Fuel Pool Level Instrumentation,'' 
describes one method of providing safety enhancements in the form of 
reliable spent fuel pool instrumentation for beyond-design-basis 
external events.
    The DG-1319, ``Integrated Response Capabilities for Beyond-Design-
Basis Events,'' describes one method the NRC endorses to enhance a 
site's ability to implement the on-site emergency preparedness programs 
and guidelines and better cope with conditions resulting from a beyond-
design-basis external event.
    You may submit comments on the draft regulatory guidance by the 
following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0240. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected].
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.

X. Backfitting and Issue Finality

Proposed Rule

    As required by Sec. Sec.  50.109, 52.63, 52.83, and 52.98, the 
Commission has completed a backfit and issue finality analysis for this 
proposed rule. The Commission finds that the backfit contained in this 
proposed rule, (i.e., multiple source term dose assessment), is 
considered, as part of the set of emergency preparedness (EP) 
requirements, to provide continued reasonable assurance of adequate 
protection of public health and safety under 10 CFR 50.109(a)(4)(ii), 
consistent with the regulatory basis for EP that has existed for more 
than three decades. Availability of the backfit and issue finality 
analysis is indicated in the ``Availability of Documents'' section of 
this document.

Draft Regulatory Guidance

    The NRC is issuing, for public comment, three DGs that would 
support implementation of this proposed rule: DG-1301, ``Flexible 
Mitigation Strategies for Beyond-Design-Basis Events''; DG-1317, 
``Wide-Range Spent Fuel Pool Level Instrumentation''; and DG-1319, 
``Integrated Response Capabilities for Beyond-Design-Basis Events.'' 
These DGs would provide guidance on the methods acceptable to the NRC 
for complying with this proposed rule. The DGs would apply to all 
current holders of, and applicants for operating licenses under 10 CFR 
part 50 and combined licenses under 10 CFR part 52.
    Issuance of the DGs in final form would not constitute backfitting 
under Sec.  50.109 and would not otherwise be inconsistent with the 
issue finality provisions in 10 CFR part 52. As discussed in the 
``Implementation'' section of each DG, the NRC has no current intention 
to impose the DGs, if finalized, on current holders of an operating 
license or combined license.
    Applying the DGs, if finalized, to applications for operating 
licenses or combined licenses would not constitute

[[Page 70639]]

backfitting as defined in Sec.  50.109 or be otherwise inconsistent 
with the applicable issue finality provisions in 10 CFR part 52, 
because such applicants are not within the scope of entities protected 
by Sec.  50.109 or the applicable issue finality provisions in 10 CFR 
part 52. Neither Sec.  50.109 nor the issue finality provisions under 
10 CFR part 52--with certain exceptions--were intended to apply to 
every NRC action that substantially changes the expectations of current 
and future applicants.

XI. Cumulative Effects of Regulation

    The NRC engaged extensively with external stakeholders throughout 
this rulemaking and related regulatory activities. Public involvement 
has included: (1) Issuance of two ANPRs and two draft regulatory basis 
documents that requested stakeholder feedback; (2) issuance of 
conceptual and preliminary proposed rule language in support of public 
meetings; (3) numerous public meetings with the ACRS; and (4) many more 
public meetings that supported both the development of the draft 
regulatory basis documents as well as development of the implementing 
guidance for the two orders that this rulemaking would make generically 
applicable (i.e., Orders EA-12-049 and EA-12-051). Section II.E of this 
notice provides a more detailed discussion of public involvement.
    The NRC is following its CER process with regard to the issuance of 
draft guidance with this proposed rule to support more informed 
external stakeholder feedback. The ``Availability of Guidance'' section 
of this document describes how the public can access the draft guidance 
for which the NRC seeks external stakeholder feedback.
    Finally, the NRC is requesting CER feedback on the following 
questions:
    1. In light of the current or projected CER challenges, does this 
proposed rule's compliance dates provide sufficient time to implement 
the new proposed requirements, including changes to programs, 
procedures, and the facility? Specifically, the current proposed rule 
would require each holder of an operating license or holder of a 
combined license for which the Commission made the finding specified in 
Sec.  52.103(g) to comply with all provisions of this proposed rule no 
later than 2 years following the effective date of the rule, unless 
otherwise specified in proposed 10 CFR part 50, appendix E, section 
VII. The NRC requests feedback on what this time period should be.
    2. If current or projected CER challenges exist, what should be 
done to address this situation? For example if more time is required 
for implementation of the new requirements, what period of time would 
be sufficient?
    3. Do other NRC regulatory actions, including the post-Fukushima 
actions and any other actions (e.g., generic communications, license 
amendment requests, inspection findings of a generic nature), influence 
the implementation of this proposed rule's requirements?
    4. Are there unintended consequences associated with implementation 
of these requirements, including implementing the requirements as a 
priority over other facility modifications that are currently being 
prioritized and scheduled?
    5. Please provide feedback on the NRC's supporting regulatory 
analysis for this rulemaking. Of note, the regulatory analysis 
estimates the cost of implementing both Order EA-12-049 and Order EA-
12-051. The NRC would appreciate feedback regarding those estimates.

XII. Plain Writing

    The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal 
agencies to write documents in a clear, concise, and well-organized 
manner. The NRC has written this document to be consistent with the 
Plain Writing Act as well as the Presidential Memorandum, ``Plain 
Language in Government Writing,'' published June 10, 1998 (63 FR 
31883). The NRC requests comment on this document with respect to the 
clarity and effectiveness of the language used.

XIII. Environmental Assessment and Proposed Finding of No Significant 
Environmental Impact

    The Commission has determined under the National Environmental 
Policy Act of 1969, as amended, and the Commission's regulations in 
subpart A of 10 CFR part 51, that this proposed rule, if adopted, would 
not be a major Federal action significantly affecting the quality of 
the human environment, and an environmental impact statement is not 
required. The basis of this determination reads as follows: The 
proposed action would not result in any radiological effluent impact as 
it would not change any design basis structures, systems, or components 
that function to limit the release of radiological effluents during or 
after an accident. This proposed rule does not change the standards and 
requirements for radiological releases and effluents. None of the 
revisions or additions in this proposed rule would affect current 
occupational or public radiation exposure. The proposed rule would not 
cause any significant non-radiological impacts, as it would not affect 
any historic sites or any non-radiological plant effluents. The NRC 
concludes that this proposed rule would not cause any significant 
radiological or non-radiological impacts on the human environment.
    The determination of this environmental assessment is that there 
would be no significant effect on the quality of the human environment 
from this action. Public stakeholders should note, however, that 
comments on any aspect of this environmental assessment may be 
submitted to the NRC as indicated in the Addresses section of this 
document. The environmental assessment is available as indicated under 
the ``Availability of Documents'' section.
    The NRC has sent a copy of the environmental assessment and this 
proposed rule to every State Liaison Officer and has requested 
comments.

XIV. Paperwork Reduction Act

    This proposed rule contains new or amended information collection 
requirements that are subject to the Paperwork Reduction Act of 1995 
(44 U.S.C. 3501 et seq). This proposed rule has been submitted to the 
OMB for approval of the information collection requirements.
    Type of submission, new or revision: Revision.
    The title of the information collection: Mitigation of Beyond-
Design-Basis Events Proposed Rule.
    The form number if applicable: Not applicable.
    How often the collection is required: Once.
    Who will be required or asked to report: Operating nuclear power 
reactor sites (comprised of 65 operating sites).
    An estimate of the number of annual responses: 65 (65 
recordkeepers).
    The estimated number of annual respondents: 65.
    An estimate of the total number of hours needed to complete the 
requirement or request: 6500.
    Abstract: In response to the Great East Japan Earthquake of March 
11, 2011, the NRC is seeking to: (1) Make the requirements in Order EA-
12-049 and Order EA-12-051 generically-applicable giving consideration 
to lessons learned from implementation of the orders; (2) establish new 
requirements for an integrated response capability; (3) establish new 
requirements for actions that are related to onsite emergency response; 
and (4) address a number of PRMs submitted following the March 2011 
Fukushima Dai-ichi event.

[[Page 70640]]

    The NRC is seeking public comment on the potential impact of the 
information collections contained in this proposed rule and on the 
following issues:
    1. Is the proposed information collection necessary for the proper 
performance of the functions of the NRC, including whether the 
information will have practical utility?
    2. Is the estimate of burden accurate?
    3. Is there a way to enhance the quality, utility, and clarity of 
the information to be collected?
    4. How can the burden of the information collection be minimized, 
including the use of automated collection techniques?
    A copy of the OMB clearance package and proposed rule is available 
in ADAMS under Accession No. ML15274A031 or may be viewed free of 
charge at the NRC's PDR, One White Flint North, 11555 Rockville Pike, 
Room O-1 F21, Rockville, MD 20852. You may obtain information and 
comment submissions related to the OMB clearance package by searching 
on http://www.regulations.gov under Docket ID NRC-2014-0240.
    You may submit comments on any aspect of these proposed information 
collections, including suggestions for reducing the burden and on the 
previously stated issues, by the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0059.
     Mail comments to: FOIA, Privacy, and Information 
Collections Branch, Office of Information Services, Mail Stop: T-5 F53, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 or to 
Vlad Dorjets, Desk Officer, Office of Information and Regulatory 
Affairs (3150-0011 and 3150-0151), NEOB-10202, Office of Management and 
Budget, Washington, DC 20503; telephone: 202-395-7315, email: 
[email protected].
    Submit comments by December 14, 2015. Comments received after this 
date will be considered if it is practical to do so, but the NRC staff 
is able to ensure consideration only for comments received on or before 
this date.

Public Protection Notification

    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.

XV. Criminal Penalties

    For the purposes of Section 223 of the Atomic Energy Act of 1954, 
as amended (AEA), the NRC is issuing this proposed rule that would 
amend 10 CFR parts 50 and 52 under one or more of Sections 161b, 161i, 
or 161o of the AEA. Willful violations of the rule would be subject to 
criminal enforcement. Criminal penalties as they apply to regulations 
in 10 CFR parts 50 and 52 are discussed in Sec. Sec.  50.111 and 
52.303.

XVI. Coordination with NRC Agreement States

    The Agreement States are receiving notification of the publication 
of this proposed rule.

XVII. Compatibility of Agreement State Regulations

    Under the ``Policy Statement on Adequacy and Compatibility of 
Agreement State Programs,'' approved by the Commission on June 20, 
1997, and published in the Federal Register (62 FR 46517; September 3, 
1997), this proposed rule is classified as compatibility category 
``NRC.'' Compatibility is not required for Category ``NRC'' 
regulations. The NRC program elements in this category are those that 
relate directly to areas of regulation reserved to the NRC by the AEA 
or the provisions of title 10 of the Code of Federal Regulations, and 
although an Agreement State may not adopt program elements reserved to 
the NRC, it may wish to inform its licensees of certain requirements 
via a mechanism that is consistent with a particular State's 
administrative procedure laws, but does not confer regulatory authority 
on the State.

XVIII. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, 
Public Law 104-113, requires that Federal agencies use technical 
standards that are developed or adopted by voluntary consensus 
standards bodies unless the use of such a standard is inconsistent with 
applicable law or otherwise impractical. In this proposed rule, the NRC 
would add requirements for the mitigation of beyond-design-basis 
events. This action does not constitute the establishment of a standard 
that contains generally applicable requirements.

XIX. Public Meeting

    The NRC will conduct a public meeting on this proposed rule for the 
purpose of describing the proposed rule to the public and answering 
questions from the public on the proposed rule.
    The NRC will publish a notice of the location, time, and agenda for 
the meeting on the NRC's public meeting Web site within at least 10 
calendar days before the meeting. Stakeholders should monitor the NRC's 
public meeting Web site for information about the public meeting at: 
http://www.nrc.gov/public-involve/public-meetings/index.cfm. The 
meeting notice will also be added to the Federal rulemaking Web site at 
http://www.regulations.gov under Docket ID NRC-2014-0240. See the 
``Availability of Documents'' section of this document for instructions 
on how to subscribe to a docket on the Federal rulemaking Web site.

XX. Availability of Documents

    The documents identified in the following table are available to 
interested persons through one or more of the following methods, as 
indicated.

------------------------------------------------------------------------
                                                   ADAMS accession No./
                    Document                         web link/Federal
                                                     Register citation
------------------------------------------------------------------------
                      Primary Rulemaking Documents
------------------------------------------------------------------------
Draft Regulatory Analysis and Backfit and Issue   ML15265A610
 Finality Analysis.
Environmental Assessment........................  ML15260B014
------------------------------------------------------------------------
                         Draft Regulatory Guides
------------------------------------------------------------------------
DG-1301, Flexible Mitigation Strategies for       ML13168A031
 Beyond-Design-Basis Events.
DG-1317, Wide-Range Spent Fuel Pool Level         ML14245A454
 Instrumentation.
DG-1319, Integrated Response Capabilities for     ML14265A070
 Beyond-Design-Basis Events.

[[Page 70641]]

 
                            Other References
------------------------------------------------------------------------
ACRS Transcript--Full Committee, Discuss          ML14345A387
 Preliminary Mitigation of Beyond-Design-Basis
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ACRS Transcript--Fukushima Subcommittee, Discuss  ML14337A671
 Preliminary Mitigation of Beyond-Design-Basis
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ACRS Transcript--Full Committee, Discuss          ML14223A631
 Consolidation of Station Blackout Mitigation
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ACRS Transcript--Full Committee, Discuss the      ML13175A344
 Station Blackout Mitigation Strategies
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ACRS Transcript--Joint Fukushima and PRA          ML14265A059
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ACRS Transcript--Plant Operations and Fire        ML13063A403
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ACRS Transcript--Reactor Safeguards Reliability   ML14337A651
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ACRS Transcript--Regulatory Policies and          ML13148A404
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CLI-12-09, South Carolina Electric & Gas Co. and  ML12090A531
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 Events in Japan,'' March, 21, 2011.
COMSECY-13-0002, ``Consolidation of Japan         ML13011A037
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COMSECY-13-0010, ``Schedule and Plans for Tier 2  ML12339A262
 Order on Emergency Preparedness for Japan
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COMSECY-14-0037, ``Integration of Mitigating      ML14309A256
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 Hazards,'' November 21, 2014.
Conceptual Consolidated Preliminary Proposed      ML14052A057
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Crystal River Unit 3, Final Response to March     ML13274A341
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Federal Register Notice--Enhancements to          76 FR 72560
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Federal Register Notice--Onsite Emergency         78 FR 63901
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 October 25, 2013.
Federal Register Notice--Onsite Emergency         77FR 23161
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Capabilities, Advance Notice of Proposed
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Federal Register Notice--Onsite Emergency         78 FR 1154
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Federal Register Notice--Onsite Emergency         78 FR 68774
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Capabilities, Preliminary Proposed Rule
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Federal Register Notice--Power Reactor Security   74 FR 13926
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Federal Register Notice--PRM-50-100, Petition     78 FR 44034
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Federal Register Notice--PRM-50-101, Petition     77 FR 16483
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Federal Register Notice--PRM-50-102, Petition     77 FR 25104
 for Rulemaking; Submitted by the Natural
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Federal Register Notice--PRM-50-96, Long-Term     77 FR 74788
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Federal Register Notice--PRM-50-97, PRM-50-98,..  76 FR 58165
PRM-50-99, PRM-50-100, PRM-50-101, PRM-50-102,
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Federal Register Notice--Station Blackout         78 FR 21275
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Federal Register Notice--Station Blackout         78 FR 44035
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Federal Register Notice--Station Blackout,        77 FR 16175
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Interim Staff Guidance, NSIR/DPR-ISG-01,          ML113010523
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JLD-ISG-2012-01, ``Compliance with Order EA-12-   ML12229A166
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Inspection Manual Chapter (IMC) 0308, ``Reactor   ML062890421
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Kewaunee Power Station, 60-Day Response to March  ML13123A004
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[[Page 70642]]

 
Kewaunee Power Station, Rescission of Order EA-   ML14059A411
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Kewaunee Power Station, Response to Request for   ML13322B255
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Letter from ACRS to Mr. R. W. Borchardt,          ML12072A197
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Letter from R.W. Borchardt to J. Sam Amijo,       ML12030A198
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 Learned,'' February 27, 2012.
Letter from ACRS to Chairman Stephen G. Burns,    ML15111A271
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Letter from Mark Satorius to John Stetkar,        ML15125A485
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Letter from NEI to Mark Satorious, ``Use of       ML15217A314
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NEI 06-12, ``B.5.b Phase 2&3 Submittal            ML070090060
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NEI 10-05, ``Assessment of On-Shift Emergency     ML111751698
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NEI 12-01, ``Guideline for Assessing Beyond       ML12125A412
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NEI 12-06, ``Diverse and Flexible Coping          ML15279A426
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NEI 13-06, ``Enhancements to Emergency Response   ML14269A230
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NEI 14-01, ``Emergency Response Procedures and    ML14269A236
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NEI 91-04 (formerly NUMARC 91-04), Severe         ML072850981
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Non-concurrence NCP-2015-003....................  ML15091A646
NUREG-0654/FEMA-REP-1, ``Criteria for             ML040420012
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 November 1980.
NUREG-0660, Volume1 and 2, ``NRC Action Plan      ML072470526 and
 Developed as a Result of the TMI-2 Accident,''    ML072470524
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NUREG-0711, ``Human Factors Engineering Program   ML12324A013
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NUREG-0737, ``Clarification of TMI Action Plan    ML102560051
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NUREG-0737, ``Clarification of TMI Action Plan    ML102560009
 Requirements,'' Supplement 1, November 1980.
NUREG-1935, ``State-of-the-Art Reactor            ML12332A057
 Consequence Analyses (SOARCA) Report,''
 November 2012.
Omaha Public Power District's Overall Integrated  ML13116A208
 Plan (Redacted) in Response to March 12, 2012,
 Order EA-12-049, February 28, 2013.
Order EA-02-026, ``Order for Interim Safeguards   ML020510635
 and Security Compensatory Measures,'' February
 25, 2002.
Order EA-12-049, ``Issuance of Order to Modify    ML12054A735
 Licenses With Regard to Requirements for
 Mitigation Strategies for Beyond-Design-Basis
 External Events,'' (Mitigating Strategies
 Order), March 12, 2012.
Order EA-12-051, ``Order Modifying Licenses with  ML12056A044
 Regard to Reliable Spent Fuel Pool
 Instrumentation''.
Preliminary Proposed Rule Language for            ML14336A641
 Mitigation of Beyond-Design-Basis Events
 Rulemaking made available to the public on
 November 13, 2014, and December 8, 2014, to
 support public discussion with the ACRS.
Preliminary Proposed Rule Language for            ML14218A253
 Mitigation of Beyond-Design-Basis Events
 Rulemaking, August 15, 2014.
PRM 50-102, ``NRDC's Petition For Rulemaking to   ML11216A242
 Require More Realistic Training on Severe
 Accident Mitigation Guidelines,'' July 26, 2011.
PRM 50-97, ``NRDC's Petition For Rulemaking to    ML11216A237
 Require Emergency Preparedness Enhancements for
 Prolonged Station Blackouts,'' July 26, 2011.
PRM-50-100, ``NRDC's Petition For Rulemaking to   ML11216A240
 Require Licensees to Improve Spent Nuclear Fuel
 Pool Safety,'' July 26, 2014.
PRM-50-101, ``NRDC's Petition For Rulemaking to   ML11216A241
 Revise 10 CFR Sec.   50.63,'' July 26, 2011.
PRM-50-96, ``Petition for Rulemaking Submitted    ML110750145
 by Thomas Popik on Behalf of the Foundation for
 Resilient Societies to adopt regulations that
 would require facilities licensed by the NRC
 under 10 CFR Part 50 to assure long-term
 cooling and unattended water makeup of spent
 fuel pools,'' March 14, 2011.
PRM-50-98, ``NRDC's Petition For Rulemaking to    ML11216A238
 Require Emergency Preparedness Enhancements for
 Multiunit Events,'' July 26, 2011.
Regulatory Issue Summary 2009-13, ``Emergency     ML092670124
 Response Data System Upgrade from Modem to
 Virtual Private Network Appliance,'' September
 28, 2009.
Request for Information Pursuant to Title 10 of   ML12053A340
 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, March 12,
 2012.
Severe Accident Management Guidance Technical     http://www.epri.com/
 Basis Report, Volume 1: Candidate High-Level      abstracts/Pages/
 Actions and Their Effects. EPRI, Palo Alto, CA:   ProductAbstract.aspx?
 2012. 1025295.                                    ProductId=1025295
Severe Accident Management Guidance Technical
 Basis Report, Volume 2: The Physics of Accident
 Progression. EPRI, Palo Alto, CA: 2012. 1025295.
San Onofre Nuclear Generating Station Units 2     ML14113A572
 and 3, ``Rescission of Order EA-12-049, 'Order
 Modifying Licenses with Regard to Requirements
 for Mitigation Strategies for Beyond Design
 Basis External Events','' June 30, 2014.

[[Page 70643]]

 
San Onofre Nuclear Generating Station Units 2     ML13329A826
 and 3, ``NRC Response To Southern California
 Edison's Final Response to the March 2012
 Request for Information Letter,'' January 22,
 2014.
San Onofre Nuclear Generating Station Units 2     ML13276A020
 and 3, Final Response to the March 12, 2012
 Information Request Regarding Near-Term Task
 Force Recommendations 2.1, 2.3, and 9.3 and
 Corresponding Commitments San Onofre Nuclear
 Generating Station (SONGS) Units 2 and 3,
 September 30, 2013.
San Onofre Nuclear Generating Station Units 2     ML14111A069
 and 3, ``Rescission of Order EA-12-051, `Order
 Modifying Licenses with Regard to Reliable
 Spent Fuel Pool Instrumentation','' June 30,
 2014.
SECY-11-0093, ``Near-Term Report and              ML11186A950
 Recommendations for Agency Actions Following
 the Events in Japan,'' July 12, 2011.
SECY-11-0124, ``Recommended Actions to be Taken   ML11245A127
 Without Delay from the Near-Term Task Force
 Report,'' September 9, 2011.
SECY-11-0137, ``Prioritization of Recommended     ML11272A111
 Actions to Be Taken in Response to Fukushima
 Lessons Learned,'' October 3, 2011.
SECY-12-0025, ``Proposed Orders and Requests for  ML12039A103
 Information in Response to Lessons Learned From
 Japan's March 11, 2011, Great T[omacr]hoku
 Earthquake and Tsunami,'' February 17, 2012.
SECY-13-0132, ``Plan for Updating the U.S.        ML13274A495
 Nuclear Regulatory Commission's Cost Benefit
 Guidance,'' January 2, 2014.
SECY-14-0046, ``Fifth 6-Month Status Update on    ML14064A523
 Response to Lessons Learned From Japan's March
 11, 2011, Great Tohoku Earthquake and
 Subsequent Tsunami,'' April 17, 2014.
SECY-15-0065, ``Proposed Rulemaking: Mitigation   ML15049A201
 of Beyond-Design-Basis Events (RIN 3150-
 AJ49),'' April 30, 2015.
SECY-89-012, ``Staff Plans for Accident           ML12251A414
 Management Regulatory and Research Programs,''
 January 18, 1989.
SECY-97-132, ``Status of the Integration Plan     ML992930144
 for Closure of Severe Accident Issues and the
 Status of Severe Accident Research,'' June 23,
 1997.
SECY-98-131, ``Status of the Integration Plan     ML992880008
 for Closure of Severe Accident Issues and the
 Status of Severe Accident Research,'' June 8,
 1998.
SRM-SECY-15-0065, ``Proposed Rulemaking:          ML15239A767
 Mitigation of Beyond-Design-Basis Events (RIN
 3150-AJ49)''.
SRM-COMSECY-14-0037, ``Integration of Mitigating  ML15089A236
 Strategies for Beyond-Design-Basis External
 Events and The Reevaluation of Flooding
 Hazards''.
SRM-COMSECY-13-0002, ``Consolidation of Japan     ML13063A548
 Lessons Learned Near-Term Task Force
 Recommendations 4 and 7 Regulatory Activities''.
SRM-SECY-11-0093, ``Near-Term Report and          ML112310021
 Recommendations for Agency Actions Following
 the Events in Japan,'' August 19, 2011.
SRM-SECY-11-0137, ``Prioritization of             ML113490055
 Recommended Actions to Be Taken in Response to
 Fukushima Lessons Learned,'' December 15, 2011.
SRM-SECY-13-0132, ``U.S. Nuclear Regulatory       ML14139A104
 Commission Staff Recommendation for the
 Disposition of Recommendation 1 of the Near-
 Term Task Force Report,'' May 19, 2014.
SRM-SECY-2011-0124, ``Recommended Actions to be   ML112911571
 Taken Without Delay From the Near-Term Task
 Force Report,'' October 18, 2011.
Temporary Instruction 2515/191, ``Inspection of   ML14273A444
 the Licensee's Responses to Mitigation
 Strategies Order EA-12-049, Spent Fuel Pool
 Instrumentation Order EA-12-051 and Emergency
 Preparedness Information Requested in NRC March
 12, 2012,'' March 12, 2012.
Temporary Instruction 2515/184, ``Availability    ML11115A053
 and Readiness Inspection of Severe Accident
 Management Guidelines (SAMGs),'' April 29, 2011.
Vermont Yankee Nuclear Power Station,             ML14321A685
 ``Rescission of Order EA-12-049, 'Order
 Modifying Licenses with Regard to Requirements
 for Mitigation Strategies for Beyond Design
 Basis External Events','' March 2, 2015.
Vermont Yankee Nuclear Power Station,             ML14321A696
 ``Rescission of Order EA-12-051, 'Order
 Modifying Licenses with Regard to Reliable
 Spent Fuel Pool Instrumentation','' March 2,
 2015.
------------------------------------------------------------------------

    Throughout the development of this rulemaking, the NRC may post 
documents related to this rulemaking, including public comments, on the 
Federal rulemaking Web site at http://www.regulations.gov under Docket 
ID NRC-2014-0240. The Federal rulemaking Web site allows you to receive 
alerts when changes or additions occur in a docket folder. To 
subscribe: (1) Navigate to the docket folder (NRC-2014-0240); (2) click 
the ``Sign up for Email Alerts'' link; and (3) enter your email address 
and select how frequently you would like to receive emails (daily, 
weekly, or monthly).

List of Subjects

10 CFR Part 50

    Administrative practice and procedure, Antitrust, Classified 
information, Criminal penalties, Education, Fire prevention, Fire 
protection, Incorporation by reference, Intergovernmental relations, 
Nuclear power plants and reactors, Penalties, Radiation protection, 
Reactor siting criteria, Reporting and recordkeeping requirements, 
Whistleblowing.

10 CFR Part 52

    Administrative practice and procedure, Antitrust, Backfitting, 
Combined license, Early site permit, Emergency planning, Fees, 
Incorporation by reference, Inspection, Limited work authorization, 
Nuclear power plants and reactors, Penalties, Probabilistic risk 
assessment, Prototype, Reactor siting criteria, Redress of site, 
Reporting and recordkeeping requirements, Standard design, Standard 
design certification.

    For the reasons set out in the preamble and under the authority of 
the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is proposing 
to adopt the following amendments to 10 CFR parts 50 and 52.

[[Page 70644]]

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

0
1. The authority citation for 10 CFR part 50 continues to read as 
follows:

    Authority:  Atomic Energy Act of 1954, secs. 11, 101, 102, 103, 
104, 105, 108, 122, 147, 149, 161, 181, 182, 183, 184, 185, 186, 
187, 189, 223, 234 (42 U.S.C. 2014, 2131, 2132, 2133, 2134, 2135, 
2138, 2152, 2167, 2169, 2201, 2231, 2232, 2233, 2234, 2235, 2236, 
2237, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs. 
201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste 
Policy Act of 1982, sec. 306 (42 U.S.C. 10226); National 
Environmental Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504 
note; Sec. 109, Pub. L. 96-295, 94 Stat. 783.

0
2. In Sec.  50.8, paragraph (b) is revised to read as follows:


Sec.  50.8  Information collection requirements: OMB approval.

* * * * *
    (b) The approved information collection requirements contained in 
this part appear in Sec. Sec.  50.30, 50.33, 50.34, 50.34a, 50.35, 
50.36, 50.36a, 50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 50.54, 50.55, 
50.55a, 50.59, 50.60, 50.61, 50.61a, 50.62, 50.63, 50.64, 50.65, 50.66, 
50.68, 50.69, 50.70, 50.71, 50.72, 50.74, 50.75, 50.80, 50.82, 50.90, 
50.91, 50.120, 50.150, 50.155, and appendices A, B, E, G, H, I, J, K, 
M, N, O, Q, R, and S to this part.
* * * * *
0
3. In Sec.  50.34, paragraphs (a)(13), (b)(12), and (i) are revised to 
read as follows:


Sec.  50.34  Contents of applications; technical information.

    (a) * * *
    (13) On or after July 13, 2009, power reactor applicants who apply 
for a construction permit shall submit the information required by 10 
CFR 50.150(b) as a part of their preliminary safety analysis report.
    (b) * * *
    (12) On or after July 13, 2009, power reactor applicants who apply 
for an operating license which is subject to 10 CFR 50.150(a) shall 
submit the information required by 10 CFR 50.150(b) as a part of their 
final safety analysis report.
* * * * *
    (i) Mitigation of beyond-design-basis events. Each application for 
a power reactor operating license under this part must include the 
applicant's plans for implementing the requirements of Sec.  50.155 and 
10 CFR part 50, appendix E, section VII, including a schedule for 
achieving full compliance with these requirements. The application must 
also include a description of:
    (1) The integrated response capability required by Sec.  50.155(b);
    (2) The equipment upon which the strategies and guidelines required 
by Sec.  50.155(b)(1) rely, including the planned locations of the 
equipment and how the equipment and SSCs meet the design requirements 
of Sec.  50.155(c); and
    (3) The strategies and guidelines required by Sec.  50.155(b)(2).
0
4. In Sec.  50.54 remove paragraph (hh)(2), redesignate paragraph 
(hh)(3) as (hh)(2) and revise it to read as follows:


Sec.  50.54  Conditions of licenses.

* * * * *
    (hh) * * *
    (2) This section does not apply to a licensee that has submitted 
the certifications required under Sec.  50.82(a)(1) or Sec.  52.110(a) 
of this chapter once the NRC has docketed those certifications.
* * * * *
0
5. Add Sec.  50.155 under the undesignated center heading Additional 
Standards for Lisences, Certifications, and Regulatory Approvals to 
read as follows:


Sec.  50.155  Mitigation of Beyond-Design-Basis Events.

    (a) Applicability. (1) Each holder of an operating license for a 
nuclear power reactor under this part and each holder of a combined 
license under part 52 of this chapter after the Commission has made the 
finding under Sec.  52.103(g), before the NRC's docketing of the 
license holder's certifications described in Sec.  50.82(a)(1) or Sec.  
52.110(a) of this chapter, shall comply with the requirements of this 
section and section VII of appendix E to 10 CFR part 50.
    (2) Each applicant for an operating license for a nuclear power 
reactor under this part and each holder of a combined license under 
part 52 of this chapter before the Commission has made the finding 
under Sec.  52.103(g) shall comply with the requirements of this 
section and section VII of appendix E to 10 CFR part 50 no later than 
the date on which the Commission issues the operating license under 
Sec.  50.57 or makes the finding under Sec.  52.103(g), respectively.
    (3) When the NRC has docketed the certifications described in Sec.  
50.82(a)(1) or Sec.  52.110(a) of this chapter, submitted by a licensee 
subject to the requirements of this section and section VII of appendix 
E to 10 CFR part 50, then that licensee shall comply with the 
requirements of Sec.  50.155(b) through (e) associated with maintaining 
or restoring secondary containment capabilities, if applicable, and 
spent fuel pool cooling capabilities, but need not comply with Sec.  
50.155(c)(4) and section VII of appendix E to 10 CFR part 50, for the 
unit described in the Sec.  50.82(a)(1) or Sec.  52.110(a) 
certifications until the spent fuel pool(s) is empty of all irradiated 
fuel.
    (i) Holders of operating licenses or combined licenses for which 
the NRC has docketed the certifications described in Sec.  50.82(a)(1) 
or Sec.  52.110(a) of this chapter need not meet the requirements of 
this section except for paragraph (b)(2) of this section once the decay 
heat of the fuel in the spent fuel pool can be removed solely by 
heating and boiling of water within the spent fuel pool and the boil-
off period provides sufficient time for the licensee to obtain off-site 
resources to sustain the spent fuel pool cooling function indefinitely, 
as demonstrated by an analysis performed and retained by the licensee.
    (ii) Dominion Nuclear Connecticut, Inc. (Millstone Power Station 
Unit 1) is not subject to the requirements of this section.
    (b) Integrated response capability. Each applicant or licensee 
shall develop, implement, and maintain an integrated response 
capability that includes:
    (1) Mitigation Strategies for Beyond-Design-Basis External Events. 
Strategies and guidelines to mitigate beyond-design-basis external 
events from natural phenomena that result in an extended loss of all ac 
power concurrent with either a loss of normal access to the ultimate 
heat sink or, for passive reactor designs, a loss of normal access to 
the normal heat sink. These strategies and guidelines must be capable 
of being implemented site-wide and must include:
    (i) Maintaining or restoring core cooling, containment, and spent 
fuel pool cooling capabilities; and
    (ii) The acquisition and use of offsite assistance and resources to 
support the functions required by paragraph (b)(1)(i) of this section 
indefinitely, or until sufficient site functional capabilities can be 
maintained without the need for the mitigation strategies.
    (2) Extensive Damage Mitigation Guidelines (EDMGs). Strategies and 
guidelines to maintain or restore core cooling, containment, and spent 
fuel pool cooling capabilities under the circumstances associated with 
loss of large areas of the plant due to explosions or fire, to include 
strategies and guidelines in the following areas:
    (i) Firefighting;
    (ii) Operations to mitigate fuel damage; and

[[Page 70645]]

    (iii) Actions to minimize radiological release.
    (3) Integration of strategies and guidelines in paragraphs (b)(1) 
and (2) of this section with the Emergency Operating Procedures (EOPs).
    (4) Sufficient staffing to support implementation of the strategies 
and guidelines in paragraphs (b)(1) and (2) of this section in 
conjunction with the EOPs to respond to events.
    (5) A supporting organizational structure with defined roles, 
responsibilities, and authorities for directing and performing the 
strategies and guidelines in paragraphs (b)(1) and (2) of this section.
    (c) Equipment. (1) The capacity and capability of the equipment 
relied on for the mitigation strategies required by paragraph (b)(1) of 
this section must be sufficient to simultaneously maintain or restore 
core cooling, containment, and spent fuel pool cooling capabilities for 
all the power reactor units within the site boundary.
    (2) The equipment relied on for the mitigation strategies required 
by paragraph (b)(1) of this section must be reasonably protected from 
the effects of natural phenomena that are equivalent to the design 
basis of the facility.
    (i) Each licensee that received the March 12, 2012, NRC letter 
issued under Sec.  50.54(f) concerning reevaluations of seismic and 
flooding hazard levels, shall provide reasonable protection against 
that reevaluated seismic or flooding hazard(s) if it exceeds the design 
basis of its facility.
    (3) The equipment relied on for the mitigation strategies in 
paragraph (b)(1) of this section must receive adequate maintenance such 
that the equipment is capable of fulfilling its intended function.
    (4) The equipment relied on for the mitigation strategies in 
paragraph (b)(1) of this section must include reliable means to 
remotely monitor wide-range spent fuel pool levels to support effective 
prioritization of event mitigation and recovery actions.
    (d) Training requirements. Each licensee shall provide for the 
training and qualification of personnel that perform activities in 
accordance with the strategies and guidelines identified in paragraphs 
(b)(1) and (2) of this section. The training and qualification on these 
activities must be developed using the systems approach to training as 
defined in Sec.  55.4 of this chapter except for elements already 
covered under other NRC regulations.
    (e) Drills and Exercises. (1) An applicant for an operating license 
issued under this part shall conduct an initial drill or exercise that 
demonstrates the capability to transition to and use one or more of the 
strategies and guidelines in paragraphs (b)(1) and (2) of this section 
and use the communications equipment required in 10 CFR part 50, 
appendix E, section VII, no more than 12 months before issuance of an 
operating license for the unit described in the license application.
    (2) A holder of a combined license issued under 10 CFR part 52 
before the Commission has made the finding under Sec.  52.103(g), shall 
conduct an initial drill or exercise that demonstrates the capability 
to transition to and use one or more of the strategies and guidelines 
in paragraphs (b)(1) and (2) of this section and use the communications 
equipment required in 10 CFR part 50, appendix E, section VII, no more 
than 12 months before the date specified for completion of the last 
inspections, tests, and analyses in the inspections, tests, analyses, 
and acceptance criteria (ITAAC) completion schedule required by Sec.  
52.99(a) for the unit described in the combined license.
    (3) Once the Commission issues an operating license to an entity 
described in paragraph (e)(1) of this section or makes the finding 
under Sec.  52.103(g) of this chapter for an entity described in 
paragraph (e)(2) of this section, the licensee shall conduct subsequent 
drills, exercises, or both that collectively demonstrate a capability 
to use at least one of the strategies and guidelines in each of 
paragraphs (b)(1) and (2) of this section in succeeding 8-year 
intervals. The drills and exercises performed to demonstrate this 
capability must include transitions from other procedures and 
guidelines as applicable, and the use of communications equipment 
required in 10 CFR part 50, appendix E, section VII. Each licensee 
shall not exceed 8 years between any consecutive drills or exercises.
    (4) A holder of an operating license issued under this part or a 
combined license under 10 CFR part 52 for which the Commission has made 
the finding specified in Sec.  52.103(g) as of [EFFECTIVE DATE OF THE 
FINAL RULE], shall conduct an initial drill or exercise that 
demonstrates the capability to transition to and use one or more of the 
strategies and guidelines in paragraphs (b)(1) and (2) of this section 
and use communications equipment required in 10 CFR part 50, appendix 
E, section VII, by [DATE 4 YEARS AFTER EFFECTIVE DATE OF THE FINAL 
RULE]. Following this initial drill or exercise, the licensee shall 
conduct subsequent drills, exercises, or both that collectively 
demonstrate a capability to use at least one of the strategies and 
guidelines in each of paragraphs (b)(1) and (2) of this section in 
succeeding 8-year intervals. The drills and exercises performed to 
demonstrate this capability must include transitions from other 
procedures and guidelines as applicable, and the use of communications 
equipment required in 10 CFR part 50, appendix E, section VII. Each 
licensee shall not exceed 8 years between any consecutive drills or 
exercises.
    (f) Change Control. (1) A licensee may make changes in the 
implementation of the requirements in this section and 10 CFR part 50, 
appendix E, section VII, without NRC approval, provided that before 
implementing each such change, the licensee performs an evaluation 
demonstrating that the provisions of this section and 10 CFR part 50, 
appendix E, section VII, continue to be met.
    (2) Documentation of all changes, including the evaluation required 
by paragraph (f)(1) of this section, shall be maintained until the 
requirements of this section and section VII of appendix E to 10 CFR 
part 50 no longer apply.
    (3) Changes in the implementation of requirements in this chapter 
subject to change control processes other than paragraph (f) of this 
section and resulting from changes in the implementation of the 
requirements in this section and 10 CFR part 50, appendix E, section 
VII, must be processed via their respective change control processes.
    (g) Implementation. Unless otherwise specified in this section or 
10 CFR part 50, appendix E, section VII:
    (1) Each holder of an operating license under this part on 
[EFFECTIVE DATE OF THE FINAL RULE] shall comply with all the provisions 
of this section no later than 2 years following [EFFECTIVE DATE OF THE 
FINAL RULE].
    (2) Each holder of a combined license under 10 CFR part 52 for 
which the Commission made the finding specified in Sec.  52.103(g) as 
of [EFFECTIVE DATE OF THE FINAL RULE] shall comply with all the 
provisions of this section no later than 2 years following [EFFECTIVE 
DATE OF THE FINAL RULE].
0
6. In appendix E to part 50 revise paragraphs I.2, IV.B.1, IV.E.2, 
IV.F.2.j, and VI.3.c and add section VII to read as follows:

Appendix E to Part 50--Emergency Planning and Preparedness for 
Production and Utilization Facilities

* * * * *
    I. * * *
    2. This appendix establishes minimum requirements for emergency 
plans for use in attaining an acceptable state of emergency

[[Page 70646]]

preparedness. These plans shall be described generally in the 
preliminary safety analysis report for a construction permit and 
submitted as part of the final safety analysis report for an 
operating license. These plans, or major features thereof, may be 
submitted as part of the site safety analysis report for an early 
site permit. Section VII of this appendix also provides for 
``Communications and Staffing Requirements for the Mitigation of 
Beyond-Design-Basis Events'' that do not need to be contained within 
a licensee's emergency plan.
* * * * *
    IV. * * *
    B. * * *
    1. The means to be used for determining the magnitude of, and 
for continually assessing the impact of, the release of radioactive 
materials, including from all reactor core and spent fuel pool 
sources, shall be described, including emergency action levels that 
are to be used as criteria for determining the need for notification 
and participation of local and State agencies, the Commission, and 
other Federal agencies, and the emergency action levels that are to 
be used for determining when and what type of protective measures 
should be considered within and outside the site boundary to protect 
health and safety. The emergency action levels shall be based on in-
plant conditions and instrumentation in addition to onsite and 
offsite monitoring. By June 20, 2012, for nuclear power reactor 
licensees, these action levels must include hostile action that may 
adversely affect the nuclear power plant. The initial emergency 
action levels shall be discussed and agreed on by the applicant or 
licensee and state and local governmental authorities, and approved 
by the NRC. Thereafter, emergency action levels shall be reviewed 
with the State and local governmental authorities on an annual 
basis.
* * * * *
    E. * * *
    2. Equipment for determining the magnitude of and for 
continuously assessing the impact of the release of radioactive 
materials, including from all reactor core and spent fuel pool 
sources, to the environment;
* * * * *
    F. * * *
    2. * * *
    j. The exercises conducted under paragraph 2 of this section by 
nuclear power reactor licensees must provide the opportunity for the 
ERO to demonstrate proficiency in the key skills necessary to 
implement the principal functional areas of emergency response 
identified in paragraph 2.b of this section. Each exercise must 
provide the opportunity for the ERO to demonstrate key skills 
specific to emergency response duties in the control room, TSC, OSC, 
EOF, and joint information center. Additionally, in each eight 
calendar year exercise cycle, nuclear power reactor licensees shall 
vary the content of scenarios during exercises conducted under 
paragraph 2 of this section to provide the opportunity for the ERO 
to demonstrate proficiency in the key skills necessary to respond to 
the following scenario elements: hostile action directed at the 
plant site, no radiological release or an unplanned minimal 
radiological release that does not require public protective 
actions, an initial classification of or rapid escalation to a Site 
Area Emergency or General Emergency, and integration of offsite 
resources with onsite response. The licensee shall maintain a record 
of exercises conducted during each eight year exercise cycle that 
documents the content of scenarios used to comply with the 
requirements of this paragraph. Each licensee shall conduct a 
hostile action exercise for each of its sites no later than December 
31, 2015. The first 8-year exercise cycle for a site will begin in 
the calendar year in which the first hostile action exercise is 
conducted. For a site licensed under 10 CFR part 52, the first 8-
year exercise cycle begins in the calendar year of the initial 
exercise required by section IV.F.2.a of this appendix.
* * * * *
    VI. * * *
    3. * * *
    c. In the event of a failure of NRC-supplied equipment, a 
replacement will be furnished by the NRC for licensee installation.
* * * * *

VII. Communications and Staffing Requirements for the Mitigation of 
Beyond Design Basis Events

    All changes associated with implementation of the requirements 
in this section are subject to Sec.  50.155(f). The change control 
provisions of Sec.  50.54(q) do not apply to proposed changes 
associated with implementation of the requirements in this section, 
unless the requirements in this section are implemented within the 
licensee's emergency plan.
    1. Each nuclear power reactor applicant or licensee shall 
perform a detailed analysis demonstrating that sufficient staff is 
available to implement the guidelines and strategies to respond to a 
beyond design basis external event resulting in impeded access to 
the nuclear power plant, an extended loss of ac power sources 
concurrent with either a loss of normal access to the ultimate heat 
sink or, for passive reactor designs, a loss of normal access to the 
normal heat sink, and affecting all units on-site.
    a. An applicant for a power reactor operating license under this 
part shall perform this analysis and submit it to the NRC under 
Sec.  50.4 at least 2 years before the issuance of the first 
operating license for full power (one authorizing operation above 5 
percent of rated thermal power).
    b. A holder of a combined license issued under 10 CFR part 52 
before the Commission has made the finding under Sec.  52.103(g) of 
this chapter shall perform this analysis and submit it to the NRC 
under Sec.  52.3 of this chapter at least 2 years before the date 
specified for completion of the last inspections, tests, and 
analyses in the inspections, tests, analyses, and acceptance 
criteria (ITAAC) completion schedule required by Sec.  52.99(a) of 
this chapter for the plant.
    c. Each holder of a power reactor operating license or combined 
license for which the Commission has made the finding specified in 
Sec.  52.103(g) of this chapter as of [EFFECTIVE DATE OF THE FINAL 
RULE], before the NRC's docketing of the license holder's 
certifications described in Sec.  50.82(a)(1) or Sec.  52.110(a) of 
this chapter, shall perform this analysis and submit it to the NRC 
under Sec.  50.4 no later than [DATE 365 DAYS AFTER EFFECTIVE DATE 
OF THE FINAL RULE].
    2. Each nuclear power reactor applicant or licensee shall make 
and describe adequate provisions for at least one onsite and one 
offsite communications system capable of remaining functional during 
an extended loss of alternating current power including the effects 
of the loss of the local communications infrastructure.
    a. An applicant for a power reactor operating license under this 
part shall make these provisions no later than the issuance of the 
first operating license for full power (one authorizing operation 
above 5 percent of rated thermal power).
    b. A holder of a combined license issued under 10 CFR part 52 
before the Commission has made the finding under Sec.  52.103(g) of 
this chapter shall make these provisions no later than the date 
specified for completion of the last inspections, tests, and 
analyses in the ITAAC completion schedule required by Sec.  52.99(a) 
of this chapter for the plant.
    c. Each holder of a power reactor operating license under this 
part or a combined license issued under 10 CFR part 52 for which the 
Commission has made the finding specified in Sec.  52.103(g) of this 
chapter as of [EFFECTIVE DATE OF THE FINAL RULE], before the NRC's 
docketing of the license holder's certifications described in Sec.  
50.82(a)(1) or Sec.  52.110(a) of this chapter, shall make these 
provisions no later than [DATE 365 DAYS AFTER EFFECTIVE DATE OF THE 
FINAL RULE].

PART 52--LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER 
PLANTS

0
7. The authority citation for part 52 continues to read as follows:

    Authority:  Atomic Energy Act of 1954, secs. 103, 104, 147, 149, 
161, 181, 182, 183, 185, 186, 189, 223, 234 (42 U.S.C. 2133, 2134, 
2167, 2169, 2201, 2231, 2232, 2233, 2235, 2236, 2239, 2273, 2282); 
Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 
U.S.C. 5841, 5842, 5846, 5851); 44 U.S.C. 3504 note.


[[Page 70647]]


0
8. In Sec.  52.80, revise paragraph (d) to read as follows:


Sec.  52.80  Contents of applications; additional technical 
information.

* * * * *
    (d) The applicant's plans for implementing the requirements of 
Sec.  50.155 of this chapter and 10 CFR part 50, appendix E, section 
VII, including a schedule for achieving full compliance with these 
requirements, and a description of:
    (1) The integrated response capability required by Sec.  50.155(b) 
of this chapter;
    (2) The equipment upon which the strategies and guidelines required 
by Sec.  50.155(b)(1) of this chapter rely, including the planned 
locations of the equipment and how the equipment and SSCs meet the 
design requirements of Sec.  50.155(c) of this chapter; and
    (3) The strategies and guidelines required by Sec.  50.155(b)(2) of 
this chapter.

    Dated at Rockville, Maryland, this 2nd day of November, 2015.
    For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 2015-28589 Filed 11-12-15; 8:45 am]
 BILLING CODE 7590-01-P