[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
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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
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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
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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-
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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-
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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\
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\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\
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\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.
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\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.
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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.
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\9\ Planning Standards N.1 Exercise and N.2 Drills.
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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
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ACRS Transcript--Fukushima Subcommittee, Discuss ML14337A671
Preliminary Mitigation of Beyond-Design-Basis
Events Rulemaking Language, November 21, 2014.
ACRS Transcript--Full Committee, Discuss ML14223A631
Consolidation of Station Blackout Mitigation
Strategies and Onsite Emergency Response
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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
Subcommittees, Discuss CPRR Technical Analysis,
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ACRS Transcript--Plant Operations and Fire ML13063A403
Protection Subcommittee, Discuss the Onsite
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CLI-12-09, South Carolina Electric & Gas Co. and ML12090A531
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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
Lessons Learned,'' dated March 27, 2013.
COMSECY-14-0037, ``Integration of Mitigating ML14309A256
Strategies for Beyond-Design-Basis External
Events and The Reevaluation of Flooding
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Federal Register Notice--Onsite Emergency 78 FR 63901
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Federal Register Notice--Onsite Emergency 77FR 23161
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Federal Register Notice--Onsite Emergency 78 FR 1154
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Federal Register Notice--Onsite Emergency 78 FR 68774
Response.
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
for Rulemaking Submitted by the Natural
Resources Defense Council, Inc., March 21, 2012.
Federal Register Notice--PRM-50-102, Petition 77 FR 25104
for Rulemaking; Submitted by the Natural
Resources Defense Council, Inc., April 27, 2012.
Federal Register Notice--PRM-50-96, Long-Term 77 FR 74788
Cooling and Unattended Water Makeup of Spent
Fuel Pools, Consideration in the Rulemaking
Process, December 18, 2012.
Federal Register Notice--PRM-50-97, PRM-50-98,.. 76 FR 58165
PRM-50-99, PRM-50-100, PRM-50-101, PRM-50-102,
Petitions for Rulemaking Submitted by the
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Federal Register Notice--Statement of Principles 62 FR 46517
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Federal Register Notice--Station Blackout 78 FR 21275
Mitigation Strategies, Draft Regulatory Basis
and Draft Rule Concepts, April 10, 2013.
Federal Register Notice--Station Blackout 78 FR 44035
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23, 2013.
Federal Register Notice--Station Blackout, 77 FR 16175
Advance Notice of Proposed Rulemaking, March
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JLD-ISG-2012-01, ``Compliance with Order EA-12- ML12229A166
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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 Chairman Jaczko, ``Initial ML11284A136
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Delay,'' October 13, 2011.
Letter from ACRS to Mr. R. W. Borchardt, ML12072A197
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Related ACRS Recommendations In Letters Dated
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Letter from R.W. Borchardt to J. Sam Amijo, ML12030A198
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Review Of (1) The U.S. Nuclear Regulatory
Commission Near-Term Task Force Report On
Fukushima, (2) Staff's Recommended Actions To
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Staff's Prioritization Of Recommended Actions
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Learned,'' February 27, 2012.
Letter from ACRS to Chairman Stephen G. Burns, ML15111A271
``Draft SECY Paper Proposed Rulemaking:
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3150-AJ49),'' April 22, 2015.
Letter from Mark Satorius to John Stetkar, ML15125A485
``Draft SECY Paper Proposed Rulemaking:
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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
Response Organization Staffing and
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NEI 12-01, ``Guideline for Assessing Beyond ML12125A412
Design Basis Accident Response Staffing and
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NEI 12-06, ``Diverse and Flexible Coping ML15279A426
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NEI 13-06, ``Enhancements to Emergency Response ML14269A230
Capabilities for Beyond Design Basis Accidents
and Events,'' Revision 0, September 2014.
NEI 14-01, ``Emergency Response Procedures and ML14269A236
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Severe Accidents,'' Revision 0, September 2014.
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
Preparation and Evaluation of Radiological
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Support of Nuclear Power Plants,'' Revision 1,
November 1980.
NUREG-0660, Volume1 and 2, ``NRC Action Plan ML072470526 and
Developed as a Result of the TMI-2 Accident,'' ML072470524
May 1980.
NUREG-0711, ``Human Factors Engineering Program ML12324A013
Review Model,'' Revision 3, November 2012.
NUREG-0737, ``Clarification of TMI Action Plan ML102560051
Requirements,'' November 1980.
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,''
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Omaha Public Power District's Overall Integrated ML13116A208
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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
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Order EA-12-051, ``Order Modifying Licenses with ML12056A044
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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
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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