T H E N E A R -T E R M TA S K F O R C E
R E VIE W OF I NS I GHT S FROM T HE
FUKUSHIMA DAI-ICHI ACCIDENT

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U.S. Nuclear Regulatory Commission

Enhancing Reactor Safety in the 21st Century

T H E N E A R -T E R M TA S K F O R C E
RE VIE W OF INS I GHT S FROM T HE
FUKUSHIMA DAI-ICHI ACCIDENTJULY 12, 2011

Dr. Charles Miller
Amy Cubbage
Daniel Dorman
Jack Grobe
Gary Holahan
Nathan Sanfilippo

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Enhancing Reactor Safety in the 21st Century

DEDICATIONThis report is dedicated to the people of Japan and especially to those who have responded
heroically to the nuclear accident at Fukushima. It is the fervent hope of the Near-Term Task
Force that their hardships and losses may never be repeated.
Throughout its tenure, the Near-Term Task Force has been inspired by the strength andresilience of the Japanese people in the face of the inconceivable losses of family andproperty inflicted by the Great East Japan Earthquake and tsunami of 2011 and exacerbatedby the ongoing radioactive releases from the Fukushima Dai-ichi Nuclear Power Plant. The
heroes of Fukushima shouldered the emotional impacts of the devastation around themand labored on in the dark, through the rubble, with increasing levels of radiation andcontamination. They undertook great efforts to obtain power and cooling to prevent theunthinkable from occurring. The outcome—no fatalities and the expectation of no significant
radiological health effects—is a tribute to their efforts, their valor, and their resolve. It is ourstrong desire and our goal to take the necessary steps to assure that the result of our labors
will help prevent the need for a repetition of theirs.

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ACKNOWLEDGMENT
Over the past 4 months, the Near-Term Task Force has devoted many long hours and an
extensive amount of thought, deliberation, and collaboration to produce this report. The
recommendations presented herein will likely require the same by those who will makedecisions as to whether they should be endorsed or not. That is as it should be. The Task
Force members comprise over 135 years of regulatory experience as a collective body. I havevalued greatly their wisdom and dedication in completing this task. It has been my privilege
to serve as their leader. So to:
Amy Cubbage
Cynthia Davidson
Dan Dorman
Jack Grobe
Gary Holahan
Nathan Sanfilippo
I say thank you,
Dr. Charles L. Miller

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CONTENTS
1.

Introduction......................................................................................................................... 1

2.

Summary of Events at Fukushima Dai-ichi....................................................................... 7

3.

Regulatory Framework for the 21st Century.................................................................... 15

4.

Safety Through Defense-in-Depth.................................................................................... 25
4.1 Ensuring Protection from External Events................................................................ 25
4.2 Mitigation.................................................................................................................... 32
4.3 Emergency Preparedness.......................................................................................... 50

5.

Implications for NRC Programs....................................................................................... 63
5.1 NRC Inspection Program........................................................................................... 63
5.2 Management of NRC Records and Information........................................................ 64
5.3 International Cooperation and Coordination............................................................. 66

6.

Summary of Overarching Recommendations.................................................................. 69

7.

Applicability and Implementation Strategy for New Reactors........................................ 71

Appendix A – Summary of Detailed Recommendations by Implementation Strategy.......... 73
Appendix B – Tasking Memorandum....................................................................................... 77
Appendix C – Memo Transmitting Charter for the Nuclear Regulatory Commission Task
Force to Conduct a Near-Term Evaluation of the Need for Agency Actions
Following the Events Iin Japan.......................................................................... 79

All photographs and graphic images used in this document are either in the publicdomain, have been licensed for public use, or have no known copyright restrictions.

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EXECUTIVE SUMMARYThe Near-Term Task Force was established in response to Commission direction to conduct
a systematic and methodical review of U.S. Nuclear Regulatory Commission processes andregulations to determine whether the agency should make additional improvements to its
regulatory system and to make recommendations to the Commission for its policy direction,in light of the accident at the Fukushima Dai-ichi Nuclear Power Plant. The Task Forceappreciates that an accident involving core damage and uncontrolled release of radioactive
material to the environment, even one without significant health consequences, is inherently
unacceptable. The Task Force also recognizes that there likely will be more than 100 nuclear
power plants operating throughout the United States for decades to come. The Task Forcedeveloped its recommendations in full recognition of this environment.
In examining the Fukushima Dai-ichi accident for insights for reactors in the United States,
the Task Force addressed protecting against accidents resulting from natural phenomena,
mitigating the consequences of such accidents, and ensuring emergency preparedness.
The accident in Japan was caused by a natural event (i.e., tsunami) which was far more
severe than the design basis for the Fukushima Dai-ichi Nuclear Power Plant. As part of its
undertaking, the Task Force studied the manner in which the NRC has historically required
protection from natural phenomena and how the NRC has addressed events that exceed thecurrent design basis for plants in the United States.
In general, the Task Force found that the current NRC regulatory approach includes:• requirements for design-basis events with protection and mitigation features controlled
through specific regulations or the general design criteria (Title 10 of the Code of Federal
Regulations (10 CFR) Part 50, “Domestic Licensing of Production and Utilization Facilities,”
Appendix A, “General Design Criteria for Nuclear Power Plants”)• requirements for some “beyond-design-basis” events through specific regulations (e.g.,
station blackout, large fires, and explosions)• voluntary industry initiatives to address severe accident features, strategies, andguidelines for operating reactors
This regulatory approach, established and supplemented piece-by-piece over the decades,
has addressed many safety concerns and issues, using the best information and techniques
available at the time. The result is a patchwork of regulatory requirements and othersafety initiatives, all important, but not all given equivalent consideration and treatment
by licensees or during NRC technical review and inspection. Consistent with the NRC’s
organizational value of excellence, the Task Force believes that improving the NRC’s
regulatory framework is an appropriate, realistic, and achievable goal.
The current regulatory approach, and more importantly, the resultant plant capabilities
allow the Task Force to conclude that a sequence of events like the Fukushima accident isunlikely to occur in the United States and some appropriate mitigation measures have been
implemented, reducing the likelihood of core damage and radiological releases. Therefore,
continued operation and continued licensing activities do not pose an imminent risk to publichealth and safety.
However, the Task Force also concludes that a more balanced application of theCommission’s defense-in-depth philosophy using risk insights would provide an enhanced
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regulatory framework that is logical, systematic, coherent, and better understood. Sucha framework would support appropriate requirements for increased capability to address
events of low likelihood and high consequence, thus significantly enhancing safety.
Excellence in regulation demands that the Task Force provide the Commission with its best
insights and vision for an improved regulatory framework.
The Task Force finds 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 Task Force concludes that the application of the defense-in-depth
philosophy can be strengthened by including explicit requirements for beyond-designbasis events.
Many of the elements of such a regulatory framework already exist in the form of rules
regarding station blackout, anticipated transient without scram, maintenance, combustible
gas control, aircraft impact assessment, beyond-design-basis fires and explosions, andalternative treatment. Other elements, such as severe accident management guidelines,
exist in voluntary industry initiatives. The Task Force has concluded that a collection ofsuch “extended design-basis” requirements, with an appropriate set of quality or special
treatment standards, should be established.
The Task Force further sees this approach, if implemented, as a more comprehensive andsystematic application of defense-in-depth to NRC requirements for providing “adequate
protection” of public health and safety. Implementation of this concept would require strong
Commission support for a clear policy statement, rule changes, and revised staff guidance.
The Task Force notes that, after the attacks of September 11, 2001, the Commission
established new security requirements on the basis of adequate protection. These newrequirements did not result from any immediate or imminent threat to NRC-licensed
facilities, but rather from new insights regarding potential security events. The Task Forceconcluded that the Fukushima Dai-ichi accident similarly provides new insights regarding
low-likelihood, high-consequence events that warrant enhancements to defense-in-depthon the basis of redefining the level of protection that is regarded as adequate. The Task
Force recommendation for an enhanced regulatory framework is intended to establish
a coherent and transparent basis for treatment of the Fukushima insights. It is also
intended to provide lasting direction to the staff regarding a consistent decisionmaking
framework for future issues.
The Task Force has considered industry initiatives in this framework and sees that these
could play a useful and valuable role. The Task Force believes that voluntary industry
initiatives should not serve as a substitute for regulatory requirements but as a mechanism
for facilitating and standardizing implementation of such requirements.
The Task Force applied this conceptual framework during its deliberations. The result is aset of recommendations that take a balanced approach to defense-in-depth as applied to
low-likelihood, high-consequence events such as prolonged station blackout resulting fromsevere natural phenomena. These recommendations, taken together, are intended to clarify
and strengthen the regulatory framework for protection against natural disasters, mitigation,and emergency preparedness, and to improve the effectiveness of the NRC’s programs. The
Task Force’s overarching recommendations are:viii | U.S. Nuclear Regulatory Commission

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Clarifying the Regulatory Framework
1.

The Task Force recommends establishing a logical, systematic, and coherent regulatory
framework for adequate protection that appropriately balances defense-in-depth andrisk considerations. (Section 3)

Ensuring Protection
2.

The Task Force recommends that the NRC require licensees to reevaluate and upgrade
as necessary the design-basis seismic and flooding protection of structures, systems, andcomponents for each operating reactor. (Section 4.1.1)

3.

The Task Force recommends, as part of the longer term review, that the NRC evaluate
potential enhancements to the capability to prevent or mitigate seismically induced fires
and floods. (Section 4.1.2)

Enhancing Mitigation
4.

The Task Force recommends that the NRC strengthen station blackout mitigation
capability at all operating and new reactors for design-basis and beyond-design-basis
external events. (Section 4.2.1)

5.

The Task Force recommends requiring reliable hardened vent designs in boiling water
reactor facilities with Mark I and Mark II containments. (Section 4.2.2)

6.

The Task Force recommends, as part of the longer term review, that the NRC identify
insights about hydrogen control and mitigation inside containment or in other buildings
as additional information is revealed through further study of the Fukushima Dai-ichi
accident. (Section 4.2.3)

7.

The Task Force recommends enhancing spent fuel pool makeup capability andinstrumentation for the spent fuel pool. (Section 4.2.4)

8.

The Task Force recommends strengthening and integrating onsite emergency response
capabilities such as emergency operating procedures, severe accident management
guidelines, and extensive damage mitigation guidelines . (Section 4.2.5)

Strengthening Emergency Preparedness
9.

The Task Force recommends that the NRC require that facility emergency plans address
prolonged station blackout and multiunit events. (Section 4.3.1)

10. The Task Force recommends, as part of the longer term review, that the NRC pursue
additional emergency preparedness topics related to multiunit events and prolonged
station blackout. (Section 4.3.1)
11. The Task Force recommends, as part of the longer term review, that the NRC
should pursue emergency preparedness topics related to decisionmaking, radiation
monitoring, and public education. (Section 4.3.2)

Improving the Efficiency of NRC Programs
12. The Task Force recommends that the NRC strengthen regulatory oversight of licensee
safety performance (i.e., the Reactor Oversight Process) by focusing more attention on
defense-in-depth requirements consistent with the recommended defense-in-depth
framework. (Section 5.1)

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The Task Force presents further details on its recommendations in this report andan implementation strategy in Appendix A. The strategy includes several rulemaking
activities to establish new requirements. Recognizing that rulemaking and subsequent
implementation typically take several years to accomplish, the Task Force recommends
interim actions to enhance protection, mitigation, and preparedness while the rulemaking
activities are conducted.
These recommendations are based on the best available information regarding the
Fukushima Dai-ichi accident and a review of relevant NRC requirements and programs. The
Task Force concludes that these are a reasonable set of actions to enhance U.S. reactor
safety in the 21st century.

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Introduction

1.

INTRODUCTION

In the days following the Fukushima Dai-ichi nuclear accident in Japan, the U.S. Nuclear
Regulatory Commission (NRC) directed the staff to establish a senior-level agency task force(the Task Force) to conduct a methodical and systematic review of the NRC’s processes andregulations to determine whether the agency should make additional improvements to its
regulatory system and to make recommendations to the Commission for its policy direction.
The Commission direction was provided in a tasking memorandum dated March 23, 2011,
from Chairman Gregory B. Jaczko to the Executive Director for Operations. The tasking
included objectives for a near-term and a longer term review. Appendix B provides a copy of
the tasking memorandum.
In response to the Commission’s direction, the Executive Director for Operations establishedan agency Task Force to conduct a near-term evaluation of the need for agency actions. The
Task Force’s charter, dated March 30, 2011, called for a methodical and systematic review of
relevant NRC regulatory requirements, programs, and processes, and their implementation,and to recommend whether the agency should make near-term improvements to its
regulatory system. The charter also directed the Task Force to identify topics for review andassessment for the longer term effort. Appendix C presents a copy of the Task Force charter.
The Task Force structured its review activities to reflect insights from past lessons-learned
efforts. For example, after the 1979 accident at Three Mile Island (TMI), the NRC had notyet developed much of the decisionmaking framework that is in place today. In addition,
the post-TMI review considered a number of actions that were proposed for general safetyenhancement rather than being directed at specific safety weaknesses revealed by the TMI
accident. As a result, some of the actions taken by the NRC after TMI were not subjected
to a structured review and were subsequently not found to be of substantial safety benefit
and were removed.
In establishing a systematic and methodical process for its review, the Task Forcedetermined that it would focus its efforts on areas that had a nexus to the Fukushima Daiichi accident regarding the initiating event, the response of equipment and personnel, andthe progression of the accident as the Task Force understands it from available information.
The Task Force recognized that detailed information in each of these areas was, in manycases, unavailable, unreliable, or ambiguous because of damage to equipment at the siteand because the Japanese response continues to focus on actions to stop the ongoingradioactive release and to achieve long-term core and spent fuel pool cooling. Even without
a detailed understanding of all aspects of the Fukushima accident, the Task Force identified
those key areas most relevant to the safety of U.S. reactors, such as external events that
could damage large areas of the plant, protection against and mitigation of a prolonged
station blackout, and management of severe accidents. Section 2 of this report presents thesequence of events at Fukushima, as relevant to the Task Force’s deliberations. The Task
Force did not address insights from the NRC’s incident response activities related to the
Fukushima accident as these are being addressed by the NRC’s Office of Nuclear Security
and Incident Response.
Consistent with its tasking and charter, the Task Force remained independent of industry
efforts, while obtaining a broad range of inputs. The Task Force had full access to the NRC
staff to obtain information on existing programs, received briefings from staff experts in

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Introduction
the Headquarters offices, and solicited inputs from all four NRC regional offices. The Task
Force also obtained valuable insights from the members of the NRC site team in Japan.
The Task Force requested information on the status of licensees’ implementation of severe
accident management guidelines (SAMGs); the regional offices collected this information
through Temporary Instruction (TI) 2515/184, “Availability and Readiness Inspection of Severe
Accident Management Guidelines (SAMGs),” dated April 29, 2011, and the Office of Nuclear
Reactor Regulation compiled the information for the Task Force. During the implementationof this TI, members of the Task Force accompanied the inspectors at two nuclear power plantsites to gain an independent perspective and additional insights into licensee decisionmaking
during severe accidents. While maintaining its independence of industry efforts, membersof the Task Force met with representatives of the Institute of Nuclear Power Operations togather information on the industry’s post-Fukushima actions. The Task Force also met withrepresentatives of the Federal Emergency Management Agency (FEMA) to discuss offsite
emergency preparedness and to obtain insights on the U.S. National Response Framework.
Finally, the Task Force appropriately screened and considered information and suggestions
received from internal and external stakeholders. The Task Force monitored, directly or
indirectly, related international activities of the International Atomic Energy Agency (IAEA),
Nuclear Energy Agency (NEA), and other organizations.
In developing recommendations, the Task Force considered the existing approach forregulatory decisionmaking. This approach includes the existing technical requirements
related to the licensing, operation, and maintenance of commercial nuclear power plants.
It is informed by the Commission’s Policy Statement on Safety Goals for the Operations of
Nuclear Power Plants, which appeared in the Federal Register in August 1986 (51 FR 30028).
The approach includes the agency’s historical commitment to a defense-in-depth philosophy
that ensures that the design basis includes multiple layers of defense. In developing
its recommendations, the Task Force was also guided by the NRC’s Principles of Good
Regulation, particularly as related to ensuring a clear nexus between the Commission’s
requirements and its goals and objectives. In addition, the Task Force was appropriately
informed by the “Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission”(NUREG/BR-0058, Revision 4, issued September 2004), which provide a framework forevaluating potential new requirements.
The Policy Statement on Safety Goals sets forth two qualitative safety goals, which are supported
by two quantitative supporting objectives. The following are the qualitative safety goals:
Individual members of the public should be provided a level of protection fromthe consequences of nuclear power plant operation such that individuals bear no
significant additional risk to life and health.
Societal risks to life and health from nuclear power plant operation should becomparable to or less than the risks of generating electricity by viable competing
technologies and should not be a significant addition to other societal risks.
The quantitative supporting objectives are as follows:
The risk to an average individual in the vicinity of a nuclear power plant of prompt
fatalities that might result from reactor accidents should not exceed one-tenth of one
percent (0.1 percent) of the sum of prompt fatality risks resulting from other accidents
to which members of the U.S. population are generally exposed.

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Introduction
The risk to the population in the area near a nuclear power plant of cancer fatalities thatmight result from nuclear power plant operation should not exceed one-tenth of one
percent (0.1 percent) of the sum of cancer fatality risks resulting from all other causes.
In the Policy Statement on Safety Goals, the Commission emphasized the importance offeatures such as containment, siting, and emergency planning as “integral parts of the
defense-in-depth concept associated with its accident prevention and mitigation philosophy.”
A cursory review of documents discussing the agency’s approach to defense-in-depth
provides a range of explanations and applications.
The Commission’s policy on probabilistic risk assessment (PRA) (“Use of Probabilistic Risk
Assessment Methods in Nuclear Regulatory Activities,” dated August 16, 1995), states thefollowing:
Defense-in-depth is a philosophy used by the NRC to provide redundancy for facilities
with “active” safety systems, e.g. a commercial nuclear power [plant], as well as thephilosophy of a multiple-barrier approach against fission product releases.
An instructive discussion of the defense-in-depth philosophy also appears in director’s
decisions relating to a petition on Davis-Besse (FirstEnergy Nuclear Operating Company(Davis-Besse Nuclear Power Station, Unit 1), DD-03-3, 58 NRC 151, 163 (2003)).
The decision described defense-in-depth as encompassing the following requirements:(1) require the application of conservative codes and standards to establish substantial
safety margins in the design of nuclear plants;(2) require high quality in the design, construction, and operation of nuclear plants to
reduce the likelihood of malfunctions, and promote the use of automatic safety system
actuation features;(3) recognize that equipment can fail and operators can make mistakes and, therefore,
require redundancy in safety systems and components to reduce the chance thatmalfunctions or mistakes will lead to accidents that release fission products from thefuel;(4) recognize that, in spite of these precautions, serious fuel-damage accidents may
not be completely prevented and, therefore, require containment structures and safety
features to prevent the release of fission products; and(5) further require that comprehensive emergency plans be prepared and periodically
exercised to ensure that actions can and will be taken to notify and protect citizens in
the vicinity of a nuclear facility.
The Task Force has found that the defense-in-depth philosophy is a useful and broadly
applied concept. It is not, however, susceptible to a rigid definition because it is a philosophy.
For the purposes of its review, the Task Force focused on the following application of the
defense-in-depth concept:• protection from external events that could lead to fuel damage• mitigation of the consequences of such accidents should they occur, with a focus on
preventing core and spent fuel damage and uncontrolled releases of radioactive material
to the environment
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Introduction• emergency preparedness (EP) to mitigate the effects of radiological releases to the public
and the environment, should they occur
These levels of defense-in-depth are appropriate for significant external challenges to afacility. In applying these defense-in-depth features, the Task Force sought to ensure that
the Commission’s regulatory requirements, processes, and programs effectively addresseach layer of protection while maintaining appropriate balance among them. The Task Forcenotes that this approach is also consistent with Levels of Defense 3, 4, and 5 in IAEA Draft
Safety Standard DS 414, “Safety of Nuclear Power Plants: Design,” dated January 2010.
The framework of the Regulatory Analysis Guidelines (NUREG/BR-0058) informed the Task
Force as it developed recommendations and evaluated potential new requirements. These
guidelines provide a range of potential actions depending on the nature of the underlying
safety issue and the means of addressing it. These potential actions include the following:• actions necessary to bring a facility into compliance with existing requirements (Title 10 of
the Code of Federal Regulations (10 CFR) 50.109(a)(4)(i))• actions necessary to ensure adequate protection (10 CFR 50.109(a)(4)(ii))• actions defining or redefining what level of protection of the public health and safety orcommon defense and security should be regarded as adequate (10 CFR 50.109(a)(4)(iii))• actions that provide substantial additional protection and for which the direct and indirect
costs are justified by the increased protection afforded (10 CFR 50.109(a)(3))
In developing its recommendations, the Task Force considered all of the above-mentioned
guidance (defense-in-depth, the Policy Statement on Safety Goals, Regulatory Analysis
Guidelines, and the backfit rule as codified in 10 CFR 50.109, “Backfitting”). In the Task
Force’s deliberations, it became apparent that the existing guidance does not present acompletely clear and consistent framework for decisionmaking. The Policy Statement on
Safety Goals and Regulatory Analysis Guidelines address only limited aspects of defensein-depth. The safety goals policy addresses defense-in-depth by setting a “subsidiary
numerical objective” (a third-level goal) for core damage frequency, in addition to the higher
level health effects goals. In the Regulatory Analysis Guidelines, which provide guidance on
implementation of the backfit rule, the phrase “defense-in-depth” occurs only twice in 46
pages of guidance; those references are limited to dealing with core damage frequency andcontainment performance in the screening criteria. The defense-in-depth concept is notably
absent from the Regulatory Analysis Guidelines as part of the decision rationale.
Extensive efforts have been made in the past to address safety goals and their relationshipto defense-in-depth. The most extensive attempt occurred as part of the technologyneutral framework effort described in NUREG-1860, “Feasibility Study for a Risk-Informed
and Performance-Based Regulatory Structure for Future Plant Licensing,” issued
December 2007. The technology-neutral framework effort attempted to reconcile theseissues in a single process in which defense-in-depth plays a fundamental role in addressing
and compensating for uncertainties.
Ultimately, the Task Force chose a decision rationale built around the defense-in-depth
concept in which each level of defense-in-depth (namely protection, mitigation, and EP) is
critically evaluated for its completeness and effectiveness in performing its safety function.
The Task Force has therefore developed a comprehensive set of recommendations toincrease safety and redefine what level of protection of the public health is regarded as
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Introduction
adequate. The Task Force’s logic is that (1) these recommendations are intended to makeeach level of defense-in-depth complete and effective and (2) ensuring the completenessand effectiveness of each level of defense-in-depth is an essential element in the overall
approach to ensuring safety, called “adequate protection.” With this in mind, the Task Forcehas recommended what it concludes are significant reinforcements to NRC requirements
and programs.
Finally, the Task Force considered its recommendations and the Commission’s requirements
for nuclear power plants in the context of the NRC’s Principles of Good Regulation. In
particular, the principle of “clarity” states: “Regulations should be coherent, logical,
and practical. There should be a clear nexus between regulations and agency goals andobjectives, whether explicitly or implicitly stated. Agency positions should be readily
understood and easily applied.” Also, the principle of “efficiency” states, in part, that
“Regulatory activities should be consistent with the degree of risk reduction they achieve.”
The principles of “independence” and “openness” focus on the importance of obtaining
inputs from the full range of stakeholders, including consideration of many and possibly
conflicting public interests, and open channels of communication. The duration and scopeof the Task Force’s effort have necessarily limited the degree of stakeholder interaction thatwas possible. The implementation of Task Force recommendations will require additional
effort by the NRC staff to conduct stakeholder outreach through its normal processes (e.g.,rulemaking, licensing, public meetings, and workshops).

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Summary of Events at Fukushima Dai-ichi

2.

SUMMARY OF EVENTS AT FUKUSHIMA DAI-ICHI

Source: GRS, 2011

At 14:46 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 (15 miles),
130 kilometers (81 miles) east of Sendai and 372 kilometers (231 miles) northeast of Tokyo
off the coast of Honshu Island. This earthquake resulted in the automatic shutdown of 11
nuclear power plants at four sites along the northeast coast of Japan (Onagawa 1, 2, and 3;
Fukushima Dai-ichi 1, 2, and 3; Fukushima Dai-ni 1, 2, 3, and 4; and Tokai 2). The earthquake
precipitated a large tsunami that is estimated to have exceeded 14 meters (45 feet) in heightat the Fukushima Dai-ichi Nuclear Power Plant site. 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 Fukushima Daiichi Nuclear Power
Plant is located on thenortheastern coast ofJapan in the Fukushima
Prefecture.

Figure 1: Nuclear Power Plants in Japan

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Summary of Events at Fukushima Dai-ichi
Fukushima Dai-ichi Units 1 through 4 are located in the southern part of the station and areoriented such that Unit 1 is the northernmost and Unit 4 is the southernmost. Fukushima
Dai-ichi Units 5 and 6 are located farther north and at a somewhat higher elevation than the
Unit 1–4 cluster, and Unit 6 is located to the north of Unit 5.

6
5

1
2
3

4

Unit

Power

Status1

460 MWe

Operating2

784 MWe

Operating3

784 MWe

Operating

4784 MWe

Outage5784 MWe

Outage6

1,100 MWe

OutageFigure 2: Fukushima Dai-ichi Nuclear Power Plant (status before earthquake)

The Fukushima Dai-ichi site includes six boiling water reactors (BWRs).

REACTORS AT THE FUKUSHIMA DAI-ICHI NUCLEAR POWER PLANT
Unit

Net MWe*

Reactor, Containment, and Cooling Systems**1

460

BWR-3, Mark I, IC, HPCI2784

BWR-4, Mark I, RCIC, HPCI3784

BWR-4, Mark I, RCIC, HPCI4784

BWR-4, Mark I, RCIC, HPCI5784

BWR-4, Mark I, RCIC, HPCI

6

1,100

BWR-5, Mark II, RCIC, HPCS

* MWe—megawatts electric**	IC—isolation condenser, HPCI—high-pressure coolant injection system, RCIC—reactor coreisolation cooling system, HPCS—high-pressure core spray system

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Summary of Events at Fukushima Dai-ichi
On March 11, 2011, Units 1, 2, and 3 were in operation, and Units 4, 5, and 6, were shut down forroutine refueling and maintenance activities; the Unit 4 reactor fuel was offloaded to the Unit 4
spent fuel pool.
As a result of the earthquake, all of the operating units appeared to experience a normal reactor
trip within the capability of the safety design of the plants. The three operating units at Fukushima
Dai-ichi automatically shut down, apparently inserting all control rods into the reactor. As a resultof the earthquake, offsite power was lost to the entire facility. The emergency diesel generators
started at all six units providing alternating current (ac) electrical power to critical systems at each
unit, and the facility response to the seismic event appears to have been normal.
Approximately 40 minutes following the earthquake and shutdown of the operating units, thefirst large tsunami wave inundated the site followed by multiple additional waves. The estimatedheight of the tsunami exceeded the site design protection from tsunamis by approximately
8 meters (27 feet). The tsunami resulted in extensive damage to site facilities and a complete lossof ac electrical power at Units 1 through 5, a condition known as station blackout (SBO). Unit 6
retained the function of one of the diesel generators.
The operators were faced with a catastrophic, unprecedented emergency situation. They had
to work in nearly total darkness with very limited instrumentation and control systems. The
operators were able to successfully cross-tie the single operating Unit 6 air-cooled diesel
generator to provide sufficient ac electrical power for Units 5 and 6 to place and maintain thoseunits in a safe shutdown condition, eventually achieving and maintaining cold shutdown.
Despite the actions of the operators following the earthquake and tsunami, cooling was lost to
the fuel in the Unit 1 reactor after several hours, the Unit 2 reactor after about 71 hours, and theUnit 3 reactor after about 36 hours, resulting in damage to the nuclear fuel shortly after the lossof cooling. Without ac power, the plants were likely relying on batteries and turbine-driven anddiesel-driven pumps. The operators were likely implementing their severe accident management
program to maintain core cooling functions well beyond the normal capacity of the station
batteries. Without the response of offsite assistance, which appears to have been hampered by
the devastation in the area, among other factors, each unit eventually lost the capability to further
extend cooling of the reactor cores.
The current condition of the Unit 1, 2, and 3 reactors is relatively static, but those units have yet toachieve a stable, cold shutdown condition. Units 1, 2, 3, and 4 also experienced explosions further
damaging the facilities and primary and secondary containment structures. The Unit 1, 2, and3 explosions were caused by the buildup of hydrogen gas within primary containment produced
during fuel damage in the reactor and subsequent movement of that hydrogen gas from thedrywell into the secondary containment. The source of the explosive gases causing the Unit 4
explosion remains unclear. In addition, the operators were unable to monitor the condition of andrestore normal cooling flow to the Unit 1, 2, 3, and 4 spent fuel pools.
Below is a sequence of events early in the accident for the six Fukushima Dai-ichi reactors.
Only information from the Japanese utility and official Japanese Government sources,
including the “Report of the Japanese Government to the IAEA Ministerial Conference on
Nuclear Safety,” is included. More detail is expected to emerge as recovery from the accident
continues and access to various locations, facilities, data, and staff improves. This sequenceof events provides only the level of detail necessary for the near-term assessment of insights
and the recommendation of actions for consideration at U.S. nuclear facilities. When available,
times indicated are in Japan standard time.
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Summary of Events at Fukushima Dai-ichi
Unit 1 Sequence of Events
March 11
14:47

Earthquake, loss of offsite ac power, and plant trip

14:52

Isolation condenser operated to cool reactor

15:03

Isolation condenser stopped operating

15:37

Tsunami and total loss of ac power—SBO

15:37

Loss of ability to inject water to the reactor~17:00

Water level below top of fuel--:--Partial core damage (several hours after tsunami)

14:30

Vent primary containment

15:36

Explosion results in severe damage to the reactor building (secondary
containment)

March 12Figure 3: Damage to Unit 1

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Summary of Events at Fukushima Dai-ichi
Unit 2 Sequence of Events
March 11
14:47

Earthquake, loss of offsite ac power, and plant trip

~14:50RCIC manually operated to inject water to reactor

15:41

Tsunami and total loss of ac power at site—SBO

March 13--:--

RCIC continued to be used to cool reactor~11:00Vent primary containment

March 14
13:25

RCIC stopped operating~18:00

Water level below top of fuel--:--Partial core damage (approximately 3 days after tsunami)--:--

Blowout panel open on side of reactor building~06:00Explosion; suppression chamber pressure decreased indicating the possibility that primary containment was damaged

March 15

Figure 4: Damage to Unit 2

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Unit 3 Sequence of Events
March 11
14:47

Earthquake, loss of offsite ac power, and plant trip

15:05

RCIC manually started to inject water into reactor

15:41

Tsunami and total loss of ac power at site—SBO

11:36

RCIC stopped operating

12:35

HPCI automatically started injecting water into reactor

02:42

HPCI stopped operating~08:00

Water level below top of fuel--:--Partial core damage (approximately 2 days after tsunami)

05:20

Vent primary containment

11:01

Explosion results in severe damage to the reactor building (secondary
containment)

March 12

March 13

March 14

Figure 5: Damage to Unit 3

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Summary of Events at Fukushima Dai-ichi
Unit 4 Sequence of Events (Unit 4 reactor was defueled)
March 11
14:46

Earthquake and loss of offsite ac power

15:38

Tsunami and total loss of ac power at site—SBO~06:00Explosion in reactor building

March 15

Figure 6: Damage to Unit 4

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Summary of Events at Fukushima Dai-ichi
Unit 5 & 6 Sequence of Events (Both units were shut down for periodic inspection)
March 11
14:46

Earthquake and loss of offsite ac power

15:41

Tsunami and total loss of ac power at site—SBO

14:30

Unit 5 enters cold shutdown

19:27

Unit 6 enters cold shutdown

March 20

Protective Action Recommendations at Fukushima Dai-ichi
March 11
Evacuation of residents within 3 kilometers (1.9 miles) and shelter-in-place forresidents within 10 kilometers (6.2 miles)
March 12
Evacuation of residents within 20 kilometers (12.4 miles)
March 15
Evacuation of residents within 30 kilometers (18.6 miles)
April 11
“Planned Evacuation Areas” and “Evacuation Prepared Area” established in theareas beyond 20 kilometers (12.4 miles)
April 21
Restricted area within 20 kilometers (12.4 miles) established to allow temporary accessand exclusion area of 3 kilometers (1.9 miles) established for members of the public

As directed in the Chairman’s tasking memorandum, the Task Force emphasizes the needfor the staff to maintain awareness and develop further insights during the long-term
followup of the Fukushima event. Additional information is expected to evolve and become
better understood regarding (1) the condition of structures, systems, and components(SSCs) in Units 1 through 6, including the effects of the seismic and flooding events andthe explosions, (2) the timing and success of the actions taken by facility staff, including theadequacy of procedures and SAMGs, and (3) the progression of the accidents in the Unit 1, 2,
and 3 reactors and the Unit 1, 2, 3, and 4 spent fuel pools.

Figure 7: Fukushima Dai-ichi Units 1–4 following explosions

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3.

REGULATORY FRAMEWORK FOR THE 21 ST CENTURYBACKGROUND
The combination of the massive earthquake and devastating tsunami at Fukushima were
well in excess of external events considered in the plant design. The Fukushima accident
also challenged the plant’s mitigation capabilities and EP.
With that in mind, this section addresses the elements of the NRC regulatory frameworkthat play a part in providing protection from design-basis events, as well as events as severe
and complex as the Fukushima accident. Those elements include protection against seismicand flooding events (characterized as design-basis events), protection for loss of all ac
power (characterized as a beyond-design-basis event), and mitigation of severe accidents(addressing beyond-design-basis topics of core damage and subsequent containment
performance), as well as EP. The Fukushima accident therefore highlights the full spectrumof considerations necessary for a comprehensive and coherent regulatory framework.
Similar issues were raised by the TMI accident, and many beyond-design-basis
requirements, programs, and practices were derived from that experience and fromthe concurrent development of PRA as a practical tool. Other sections of this reportaddress specific elements of protection, mitigation, and preparedness and evaluate
their current capabilities, limitations, and potential enhancements. This section will
evaluate the overall NRC regulatory approach for ensuring safety through a defense-indepth philosophy that includes design-basis requirements and additional risk reduction
requirements and programs. The goal of this section is to identify possible ways tobetter integrate the fundamental design-basis concept with the beyond-design-basis
considerations, including EP.
Design-basis events became a central element of the safety approach almost 50 years ago
when the U.S. Atomic Energy Commission (AEC) formulated the idea of requiring safetysystems to address a prescribed set of anticipated operational occurrences and postulated
accidents. In addition, the design-basis requirements for nuclear power plants included a
set of external challenges including seismic activity and flooding from various sources. That
approach and its related concepts of design-basis events and design bases were used inlicensing the current generation of nuclear plants in the 1960s and 1970s.
Frequently, the concept of design-basis events has been equated to adequate protection,and the concept of beyond-design-basis events has been equated to beyond adequate
protection (i.e., safety enhancements). This vision of adequate protection has typically only
led to requirements addressing beyond-design-basis concerns when they were found to be
associated with a substantial enhancement in safety and justified in terms of cost.
Starting in the 1980s and continuing to the present, the NRC has maintained the designbasis approach and expanded it to address issues of concern. The NRC added requirements
to address each new issue as it arose but did not change the fundamental concept of designbasis events or the list of those events; nor did the NRC typically assign the concept ofadequate protection to these changes. The following paragraphs include discussion of thehistorical development of requirements to address issues beyond the design basis, including
the potential loss of all ac power (i.e., SBO) and other issues. In reading these paragraphs, it
is helpful to keep in mind that actions to address issues beyond the design basis were largely
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considered safety enhancements. They were considered to be beyond what was required foradequate protection.
In the early 1970s, the AEC began questioning applicants and licensees about the potential
consequences of an anticipated operational occurrence (a design-basis event) with thebeyond-design-basis failure of the shutdown system (a sequence called anticipatedtransient without scram (ATWS)). This issue was clearly outside the design basis of theplants, yet it remained a concern for many years, eventually resulting in a regulation(10 CFR 50.62, “Requirements for Reducing the Risk from Anticipated Transients without
Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants”). That regulation
established several requirements to address ATWS concerns but did not expand the list ofdesign-basis events to include ATWS.
In the mid-1980s, the Commission established the SBO rule (10 CFR 50.63, “Loss of All
Alternating Current Power”) to address concerns related to loss of all ac power. Again
the rule established new requirements but did not alter the design-basis concept or list ofdesign-basis events. SBO is not a design-basis event, nor is SBO-related equipment subject
to the NRC’s quality assurance requirements in Appendix B, “Quality Assurance Criteria
for Nuclear Power Plants and Fuel Reprocessing Plants,” to 10 CFR Part 50, “Domestic
Licensing of Production and Utilization Facilities.”
In addition to the established regulatory approach, the AEC commissioned and the NRC
later published the first PRA of nuclear power plants in WASH-1400, “Reactor Safety
Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants,” issued
October 1975. That study addressed possible reactor accidents based on the frequency
of initiating events and their likely consequences. The study proceeded without regard toclassification of events as “anticipated operational occurrences” or “postulated accidents”
and therefore without regard to event categorization as “design-basis events.” The treatmentof events based solely on frequency and consequences remains a characteristic of PRA andis one of its strengths in complementing the traditional regulatory approach.
Following the TMI accident, numerous lessons-learned efforts were commissioned. Oneof those studies (NUREG/CR-1250, “Three Mile Island; A Report to the Commissioners andto the Public,” issued in 1980 and generally referred to as the Rogovin Report) evaluatedthe then-existing NRC regulatory approach (characterized in the report as “the so-called‘design basis accident’ concept”). The report concluded that “More rigorous and quantitative
methods of risk analysis have been developed and should be employed to assess the safetyof design and operation,” and “The best way to improve the existing design review process
is by relying in a major way upon quantitative risk analysis.” The report’s recommendations
to the Commission included “Expand the spectrum of design basis accidents…” and “On aselective basis, determine whether some design features to mitigate the effects of someClass Nine accidents [i.e., severe accidents in current terminology] should be required.”On the basis of all the recommendations developed by the NRC following the TMI accident,
the Commission established numerous new requirements (approximately 120 actions
per plant), some within and some beyond the design bases. In addition, the Commission
considered action to address core damage scenarios, again beyond the design basis.
Ultimately, the Commission encouraged licensees to use the newly developed PRA
methodology to search for vulnerabilities (in the Individual Plant Examination (IPE) programand Individual Plant Examination for External Events (IPEEE) program) and requested
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information on their findings. The Commission also encouraged the development of SAMGs
based on PRA insights and severe accident research. However, the Commission did not takeaction to require the IPEs, IPEEEs, or SAMGs. While the Commission has been partially
responsive to recommendations calling for requirements to address beyond-design-basis
accidents, the NRC has not made fundamental changes to the regulatory approach forbeyond-design-basis events and severe accidents for operating reactors.
Currently, risk-informed regulation (i.e., regulation using PRAs) serves the limited rolesof maintenance rule implementation, Regulatory Analysis Guidelines, the search forvulnerabilities (e.g., through the IPE and IPEEE programs), the Reactor Oversight Process(ROP) and its significance determination process, and voluntary license amendment
applications (e.g., risk-informed inservice inspection).
In contrast, for new reactors, the Commission has moved further from a largely designbasis accident concept, requiring applicants for design certifications and combined licenses(COLs) under 10 CFR Part 52, “Licenses, Certification, and Approvals for Nuclear Power
Plants,” to perform a PRA and provide a description and analysis of design features for theprevention and mitigation of severe accidents (10 CFR 52.47(23) and 10 CFR 52.79(48)). Eachdesign certification rule (10 CFR Part 52, Appendix A, “Design Certification Rule for the U.S.
Advanced Boiling Water Reactor,” and other Part 52 appendices) then codifies the severe
accident features of each approved standard design.
Following the terrorist events of September 11, 2001, the NRC issued security advisories,
orders, license conditions, and ultimately a new regulation (10 CFR 50.54(hh)) to require
licensees to develop and implement guidance and strategies to maintain or restore
capabilities for core cooling and containment and spent fuel pool cooling under thecircumstances associated with the loss of large areas of the plant due to a fire or explosion.
These requirements have led to the development of extensive damage mitigation guidelines(EDMGs) at all U.S. nuclear power plants. The NRC has inspected the guidelines andstrategies that licensees have implemented to meet the requirements of 10 CFR 50.54(hh)(2).
However, there are no specific quality requirements associated with these requirements, andthe quality assurance requirements of 10 CFR Part 50, Appendix B, do not apply. The EDMGs
are requirements for addressing events well beyond those historically considered to be thedesign basis and were implemented as adequate protection backfits. In order to address
the changing security threat environment, the Commission effectively redefined what levelof protection should be regarded as adequate. This is a normal and reasonable, albeit
infrequent, exercise of the NRC’s responsibilities of protecting public health and safety.
All of the above indicate the Commission’s desire and commitment to act either throughregulatory requirements or voluntary industry initiatives to address concerns related to thedesign basis or beyond the design basis where appropriate.

TASK FORCE EVALUATIONAs presented above, the current NRC regulatory approach includes (1) requirements fordesign-basis events with features controlled through specific regulations or the generaldesign criteria (GDC) (10 CFR Part 50, Appendix A, “General Design Criteria for Nuclear
Power Plants”) and the quality requirements of 10 CFR Part 50, Appendix B, (2) beyonddesign-basis requirements through specific rules (e.g., the SBO rule) with specified
quality requirements, (3) voluntary industry initiatives to address severe accident features,
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strategies, and guidelines for operating reactors, and (4) specific requirements to address
damage from fires and explosions and their mitigation.
The Task Force presents the following observations on the NRC regulatory approach:• Although complex, the current regulatory approach has served the Commission and the
public well and allows the Task Force to conclude that a sequence of events like thoseoccurring in the Fukushima accident is unlikely to occur in the United States and could be
mitigated, reducing the likelihood of core damage and radiological releases.
• Therefore, in light of the low likelihood of an event beyond the design basis of a U.S.
nuclear power plant and the current mitigation capabilities at those facilities, the Task
Force concludes that continued operation and continued licensing activities do not posean imminent risk to the public health and safety and are not inimical to the common
defense and security. Nonetheless, the Task Force is recommending building on thesafety foundation laid in the 1960s and 1970s, and the safety improvements added fromthe 1980s to the present, to produce a regulatory structure well suited to licensing andoverseeing the operation of nuclear power plants for decades to come. The Task Forcesees these recommendations, not as a rejection of the past, but more as a fulfillment of
past intentions.
• Adequate protection has been, and should continue to be, an evolving safety standard
supported by new scientific information, technologies, methods, and operating
experience. This was the case when new information about the security environment
was revealed through the events of September 11, 2001. Licensing or operating a
nuclear power plant with no emergency core cooling system or without robust security
protections, while done in the past, would not occur under the current regulations. As
new information and new analytical techniques are developed, safety standards need
to be reviewed, evaluated, and changed, as necessary, to insure that they continue toaddress the NRC’s requirements to provide reasonable assurance of adequate protectionof public health and safety. The Task Force believes, based on its review of the information
currently available from Japan and the current regulations, that the time has come forsuch change.
• In response to the Fukushima accident and the insights it brings to light, the Task Forceis recommending actions, some general, some specific, that it believes would be areasonable, well-formulated set of actions to increase the level of safety associated withadequate protection of the public health and safety.
• The Commission has come to rely on design-basis requirements and a patchworkof beyond-design-basis requirements and voluntary initiatives for maintaining
safety. Design-basis requirements include consideration of anticipated operational
occurrences and postulated accidents such as loss-of-coolant accidents. Beyonddesign-basis considerations such as ATWS and SBO are discussed below. Voluntary
initiatives have addressed some severe accident considerations (through the IPE and
IPEEE programs), shutdown risk issues, containment vents for BWR Mark I designs,
and SAMGs.
• The concept of beyond-design-basis requirements applies, for example, to ATWS,
SBO, aircraft impact assessment (AIA), combustible gas control, and EDMGs. Since
fire protection is not based on a design-basis fire, it too can be considered beyond
design basis. Although the phrase “beyond design basis” appears only once in the NRC
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regulations (i.e., in 10 CFR 50.150, “Aircraft Impact Assessment,” known as the AIArule), regulators and industry use it often. Unfortunately, the phrase “beyond designbasis” is vague, sometimes misused, and often misunderstood. Several elements ofthe phrase contribute to these misunderstandings. First, some beyond-design-basis
considerations have been incorporated into the requirements and therefore directly affectreactor designs. The phrase is therefore inconsistent with the normal meaning of thewords. In addition, there are many other beyond-design-basis considerations that are notrequirements. The phrase therefore fails to convey the importance of the requirements towhich it refers.
• The Task Force has noted that other international regulatory systems also addressconsiderations beyond the design basis. For example, while the NRC addresses
regulatory requirements in five categories—three design basis (normal operation,
anticipated operational occurrences, and postulated accidents) and two beyond designbasis (one required and one voluntary)—Finland addresses regulatory requirements insix categories—four design basis (normal operation, anticipated operational occurrences,and two postulated accident categories), one “design extension condition,” plus severe
accidents. France also addresses both design-basis requirements and additional
requirements in categories called “Risk Reduction Category 1” and “Risk Reduction
Category 2.” In addition, the phrase used in the IAEA Draft Safety Standard DS 414
addresses considerations beyond the design basis, referring to them as thoseaddressing “design extension conditions.” In this report, the Task Force will refer topast considerations beyond the design basis using that phrase (e.g., “beyond-designbasis events”). In the context of the Task Force recommendation for a new regulatory
framework for the future, the Task Force will refer to such considerations as “extended
design basis” requirements.
• The primary responsibility for safety rests with licensees, and the NRC holds licensees
accountable for meeting regulatory requirements. In addition, voluntary safety initiatives
by licensees can enhance safety if implemented and maintained effectively, but should
not take the place of needed regulatory requirements. The NRC inspection and licensing
programs give less attention to beyond-design-basis requirements and little attention
to industry voluntary initiatives since there are no requirements to inspect against.
Because of this, the NRC gives much more attention to design-basis events than to severe
accidents.
• With the exception of a few special cases, licensees of operating reactors are not required
to develop or maintain a PRA, although all licensees currently have a PRA. These PRAs
are of varying scope and are generally not required to meet NRC-endorsed quality
standards. New reactor applications must include a description of a design-specific PRA
and its results and must address severe accident protection and mitigation features.
• The Commission has expressed its intent with respect to industry initiatives in theRegulatory Analysis Guidelines (NUREG/BR-0058, Revision 4). That document states, “Itmust be clear to the public that substituting industry initiatives for NRC regulatory action
can provide effective and efficient resolution of issues, will in no way compromise plantsafety, and does not represent a reduction in the NRC’s commitment to safety and soundregulation.”• Lastly, the Task Force observes that for new reactor designs, the Commission’s
expectations that beyond-design-basis and severe accident concerns be addressed
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and resolved at the design stage are largely expressed in policy statements and staff
requirements memoranda, only reaching the level of rulemaking when each design iscodified through design certification rulemaking.
In summary, the major elements of the NRC regulatory approach relevant to the Fukushima
accident, or a similar accident in the United States, are seismic and flooding protection (wellestablished in the design-basis requirements); SBO protection (required, but beyond thedesign-basis requirements); and severe accident mitigation (expected but neither the severe
accident mitigation features nor the SAMGs are required). In addition, U.S. facilities could
employ EDMGs as further mitigation capability. The Task Force observes that this collectionof approaches is largely the product of history; it was developed for the purpose of reactor
licensing in the 1960s and 1970s and supplemented as necessary to address significant
events or new issues. This evolution has resulted in a patchwork regulatory approach.
The Fukushima accident clearly demonstrates the importance of defense-in-depth. Whether
through extraordinary circumstances or through limited knowledge of the possibilities,
plants can be challenged beyond their established design bases protection. In suchcircumstances, the next layer of defense-in-depth, mitigation, is an essential element ofadequate protection of public health and safety. Mitigation is provided for beyond-designbasis events and severe accidents, both of which involve external challenges or multiple
failures beyond the design basis. This beyond-design-basis layer of defense-in-depth isbroadly consistent with the IAEA concept of “design extension conditions” (presented in Draft
Safety Standard DS 414).
The Task Force concludes that the NRC’s safety approach is incomplete without a strong
program for dealing with the unexpected, including severe accidents. Continued reliance on
industry initiatives for a fundamental level of defense-in-depth similarly would leave gaps
in the NRC regulatory approach. The Commission has clearly established such defense-indepth severe accident requirements for new reactors (in 10 CFR 52.47(23), 10 CFR 52.79(38),and each design certification rule), thus bringing unity and completeness to the defense-indepth concept. Taking a similar action, within reasonable and practical bounds appropriate
to operating plants, would do the same for operating reactors.
The Task Force therefore concludes that the future regulatory framework should be basedon the defense-in-depth philosophy, supported and modified as necessary by state-of-theart PRA techniques. The Task Force also concludes that the application of defense-in-depth
should be strengthened by formally establishing, in the regulations, an appropriate levelof defense-in-depth to address requirements for ”extended” design-basis events. Many ofthe elements of such regulations already exist in the form of the SBO rule (10 CFR 50.63),
ATWS rule (10 CFR 50.62), maintenance rule (10 CFR 50.65), AIA rule (10 CFR 50.150), therequirements for protection for beyond-design-basis fires and explosions (10 CFR 50.54(hh)),and the alternative treatment requirements (10 CFR 50.69) and new reactor policy
regarding regulatory treatment of nonsafety systems as described in SECY-94-084, “Policy
and Technical Issues Associated with the Regulatory Treatment of Nonsafety Systems
in Passive Plant Designs,” dated March 28, 1994. Other elements such as SAMGs exist
in voluntary industry initiatives. The Task Force envisions this collection of beyonddesign-basis requirements as a coherent whole in a separate section of 10 CFR Part 50(e.g., 10 CFR 50.200, 10 CFR 50.201) or as a dedicated appendix to 10 CFR Part 50. This
separate section would have an appropriate set of quality standards, analogous toAppendix B to 10 CFR Part 50, plus a change process similar to the 10 CFR Part 52 “50.5920 | U.S. Nuclear Regulatory Commission

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like” process codified in the rule for each certified design (e.g., 10 CFR Part 52, Appendix A,
Section VIII, “Processes for Changes and Departures”).
The Task Force envisions a framework in which the current design-basis requirements(i.e., for anticipated operational occurrences and postulated accidents) would remain largely
unchanged and the current beyond-design-basis requirements (e.g., for ATWS and SBO)
would be complemented with new requirements to establish a more balanced and effective
application of defense-in-depth.
This framework, by itself, would not create new requirements nor eliminate any current
requirements. It would provide a more coherent structure within the regulations to facilitate
Commission decisions relating to what issues should be subject to NRC requirements andwhat those requirements ought to be. The Task Force envisions that implementation ofa new regulatory framework would result in the addition of some currently unregulated
issues (i.e., those important to defense-in-depth) to a category of “extended design-basis”
requirements. The framework could also support shifting issues currently addressed as
design-basis requirements to the “extended design-basis” category of requirements. Suchchanges would establish a more logical, systematic, and coherent set of requirements
addressing defense-in-depth.
As discussed earlier, the Task Force believes that voluntary industry initiatives could play auseful and valuable role in the suggested framework. Voluntary industry initiatives would
not serve as substitutes for regulatory requirements but as a mechanism for facilitating
and standardizing implementation of such requirements. The development of symptombased emergency operating procedures (EOPs) in the 1980s and development of the EDMGs
following the events of September 11, 2001, are just two examples of notable industry
contributions to effective implementation of regulatory initiatives.
A newly organized part of the regulations would not only strengthen mitigation of accidents
as severe as the Fukushima accident, but could also help to resolve some longstanding
regulatory issues. For example, it could provide a natural location for requirements suchas reactor coolant system breaks beyond the transition break size (being considered in therevision to 10 CFR 50.46, “Acceptance Criteria for Emergency Core Cooling Systems forLight-Water Nuclear Power Reactors”). It would also be a logical location for a requirement
for PRAs and IPE and IPEEE generic safety insights for operating reactors, should theCommission desire such a requirement.
In a new regulatory framework, risk assessment and defense-in-depth would be combined
more formally. PRA would help ensure that the design-basis requirements address events ofa specific frequency with strict quality standards (i.e., 10 CFR Part 50, Appendix B) and thatbeyond-design-basis requirements address less frequent, but nonetheless important, events
through appropriate quality standards. The Task Force concludes that the new framework
could be implemented on the basis of full-scope Level 1 core damage assessment PRAs andLevel 2 containment performance assessment PRAs.
The current NRC approach to land contamination relies on preventing the release ofradioactive material through the first two levels of defense-in-depth, namely protection
and mitigation. Without the release of radioactive material associated with a core damage
accident, there would be no significant land contamination. The Task Force also concludes
that the NRC’s current approach to the issue of land contamination from reactor accidents
is sound. The Task Force’s objective of ensuring that protection and mitigation provide
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strong and effective levels of defense-in-depth is therefore fully consistent with reducing the
likelihood of land contamination without introducing any new safety concepts or methods.
The Task Force also considered the value of requiring a Level 3 PRA (i.e., a probabilistic
assessment of accident dose and health effects) as part of a new regulatory framework.
The Task Force concluded that for large light-water reactors, the metrics of core damage
frequency and large early release provide very effective, relatively simple, well-documented
and understood measures of safety for decisionmaking. Therefore, the Task Force has notrecommended including Level 3 PRA as a part of a regulatory framework. However, some
limited Level 3 PRA analyses, such as those done for a few selected plants and reported in
NUREG1150, “Severe Accident Risks: An Assessment of Five U.S. Nuclear Power Plants,”
issued December 1990, could confirm that the selected frequency ranges for design-basis
and beyond-design-basis requirements are consistent with the Commission’s safety goals.
The Task Force considered the role of the IPEs and IPEEEs (mentioned above) in a new
regulatory framework. If the new regulatory framework had been in place before the start of
the IPE and IPEEE programs, they would likely have been different, perhaps being required
rather than encouraged. However, the programs have been completed, the NRC reviewed
the efforts to some extent, and the licensees have taken voluntary actions, in some cases,
to address identified vulnerabilities. Given this reality, the Task Force concludes that themost appropriate step consistent with the new risk-informed, defense-in-depth regulatory
framework would be to revisit the results of those efforts to identify any significant items ofa generic nature and consider them for possible inclusion under design-basis or extended
design-basis regulations. Similarly, this effort could identify significant plant-specific items
as candidates for plant-specific regulatory requirements (e.g., license conditions).
Finally, a new and dedicated portion of the regulations would allow the Commission torecharacterize its expectations for safety features beyond design basis more clearly andmore positively as “extended design-basis” requirements. The Task Force recognizes fully
that a comprehensive reevaluation and restructuring of the regulatory framework would
be no small feat. The Task Force also recognizes that strengthening the roles of defensein-depth and risk assessment, emphasizing beyond-design-basis and severe accident
mitigation, and establishing a clear, coherent, and well-integrated regulatory framework
would be a significant accomplishment. Therefore, the Task Force concludes that additional
steps would be prudent to further enhance the NRC regulatory framework to encompass theprotections for accidents beyond the design basis.

Recommendation 1
The Task Force recommends establishing a logical, systematic, and coherent
regulatory framework for adequate protection that appropriately balances defensein-depth and risk considerations.
The Task Force recommends that the Commission direct the staff to initiate action toenhance the NRC regulatory framework to encompass beyond-design-basis events and theiroversight through the following steps:
1.1 Draft a Commission policy statement that articulates a risk-informed defense-in-depth
framework that includes extended design-basis requirements in the NRC’s regulations as
essential elements for ensuring adequate protection.

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1.2 Initiate rulemaking to implement a risk-informed, defense-in-depth framework
consistent with the above recommended Commission policy statement.
1.3 Modify the Regulatory Analysis Guidelines to more effectively implement the defense-indepth philosophy in balance with the current emphasis on risk-based guidelines.
• The Task Force believes that the Regulatory Analysis Guidelines could be modified by
implementing some of the concepts presented in the technology-neutral framework(NUREG-1860) to better integrate safety goals and defense-in-depth.
1.4 Evaluate the insights from the IPE and IPEEE efforts as summarized in NUREG-1560,“Individual Plant Examination Program: Perspectives on Reactor Safety and Plant
Performance,” issued December 1997, and NUREG-1742, “Perspectives Gained from theIndividual Plant Examination of External Events (IPEEE) Program,” issued April 2002, toidentify potential generic regulations or plant-specific regulatory requirements.
Subsequent sections of this report will evaluate issues relevant to the Fukushima
accident and will propose recommendations to strengthen associated NRC programs andrequirements. The objective of strengthening those programs and requirements is to makethem fully effective in providing defense-in-depth, thus supporting this recommended
framework.

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4.

SAFETY THROUGH DEFENSE-IN-DEPTH

The key to a defense-in-depth approach is creating multiple independent and redundantlayers of defense to compensate for potential failures and external hazards so that no single
layer is exclusively relied on to protect the public and the environment. In its application of
the defense-in-depth philosophy, the Task Force has addressed protection from design-basis
natural phenomena, mitigation of the consequences of accidents, and EP.

4.1

ENSURING PROTECTION FROM EXTERNAL EVENTSThe first level of defense-in-depth is protection. Specifically, the Task Force examined thehistorical development of facility design bases for protection from external hazards that
could cause the loss of large areas of a plant. The combined effects of the Great East
Japan Earthquake of 2011 and the ensuing tsunami at the Fukushima Nuclear Power Plant
site represent the most significant external event challenges that any commercial nuclear
reactor has ever faced. The following section discusses the importance of establishingappropriate protection from design-basis natural phenomena. In evaluating protection
from design-basis natural phenomena, the Task Force considered earthquakes, floods, highwinds (due to hurricanes or tornadoes), and external fires. The Task Force concluded thatearthquakes and flooding hazards warranted further Task Force consideration due, in part,to significant advancements in the state of knowledge and the state of analysis in theseareas in the time period since the operating plants were sited and licensed. In addition, theearthquake and subsequent tsunami highlighted the need to evaluate concurrent related
events, such as seismically induced fires and floods.

4.1.1 Protection from Design-Basis Natural Phenomena
BACKGROUND
The NRC has long recognized the importance of protection from natural phenomena as ameans to prevent core damage and to ensure containment and spent fuel pool integrity. The
NRC established several requirements addressing natural phenomena in 1971 with GDC 2,“Design Bases for Protection Against Natural Phenomena,” of Appendix A to 10 CFR Part 50.
GDC 2 requires, in part, that SSCs important to safety be designed to withstand the effects
of natural phenomena such as floods, tsunami, and seiches without loss of capability toperform their safety functions. GDC 2 also requires that design bases for these SSCs reflect(1) appropriate consideration of the most severe of the natural phenomena that have been
historically reported for the site and surrounding region, with sufficient margin for thelimited accuracy and quantity of the historical data and the period of time in which the data
have been accumulated, (2) appropriate combinations of the effects of normal and accident
conditions with the effects of the natural phenomena, and (3) the importance of the safetyfunctions to be performed.
Since the establishment of GDC 2, the NRC’s requirements and guidance for protection
from seismic events, floods, and other natural phenomena have continued to evolve. The
agency has developed new regulations, new and updated regulatory guidance, and several
regulatory programs aimed at enhancements for previously licensed reactors.

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Safety Through Defense-in-Depth
In 1973, Appendix A, “Seismic and Geologic Siting Criteria for Nuclear Power Plants,” to10 CFR Part 100 was established to provide detailed criteria to evaluate the suitability of
proposed sites and the suitability of the plant design basis established in consideration of theseismic and geologic characteristics of the proposed sites.
In 1977, the NRC initiated the Systematic Evaluation Program (SEP) to review the designs ofolder operating nuclear reactor plants in order to reconfirm and document their safety. The
purpose of the review was to provide (1) an assessment of the significance of differences
between then-current technical positions on safety issues and those that existed whena particular plant was licensed, (2) a basis for deciding how these differences should be
resolved in an integrated plant review, and (3) a documented evaluation of plant safety. The
plants selected for SEP review included several that were licensed before a comprehensive
set of licensing criteria (i.e., the GDC) had been developed or finalized. The SEP covered
topics including seismic events, floods, high winds, and tornadoes.
In 1980, the NRC was concerned that licensees had not conducted the seismic qualificationof electrical and mechanical equipment in some older nuclear reactor plants in accordancewith the licensing criteria for the seismic qualification of equipment acceptable at thattime. As a result, the NRC established the Unresolved Safety Issue (USI) A-46, “Seismic
Qualification of Mechanical and Electrical Equipment in Operating Nuclear Power Plants,”
program in December 1980. In February 1987, the agency issued Generic Letter (GL) 8702, “Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating
Reactors, Unresolved Safety Issue (USI) A-46,” to address this concern. The objective of USI
A-46 was to develop alternative seismic qualification methods and acceptance criteria that
could be used to assess the capability of mechanical and electrical equipment in operating
nuclear power plants to perform their intended safety functions. The scope of the review
was limited to equipment required to bring each affected plant to hot shutdown and maintainit for a minimum of 72 hours.
In 1991, the NRC issued Supplement 4 to GL 88-20, “Individual Plant Examination ofExternal Events (IPEEE) for Severe Accident Vulnerabilities, 10 CFR 50.54(f).” This GL
requested that “each licensee perform an individual plant examination of external events
to identify vulnerabilities, if any, to severe accidents and report the results together withany licensee determined improvements and corrective actions to the Commission.” Theexternal events considered in the IPEEE program include seismic events, internal fires, highwinds, and floods. The primary goal of the IPEEE program was for each licensee to identify
plant-specific vulnerabilities to severe accidents, if any, and to report the results, with any
licensee-proposed improvements and corrective actions, to the NRC.
In 1996, the NRC established two new seismic regulations for applications submitted on orafter January 10, 1997. These regulations were not applied to existing reactors. The first
regulation, 10 CFR 100.23, “Geologic and Seismic Siting Criteria,” sets forth the principal
geologic and seismic considerations that guide the Commission in its evaluation of thesuitability of a proposed site and adequacy of the design bases established in considerationof the geologic and seismic characteristics of the proposed site. The second regulation,
Appendix S, “Earthquake Engineering Criteria for Nuclear Power Plants,” to 10 CFR Part 50,
requires that nuclear power plants be designed so that certain SSCs remain functional
if the safe shutdown earthquake (SSE) ground motion occurs. These plant features arethose necessary to ensure (1) the integrity of the reactor coolant pressure boundary, (2)the capability to shut down the reactor and maintain it in a safe shutdown condition, or
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Safety Through Defense-in-Depth(3) the capability to prevent or mitigate the consequences of accidents that could result in
potential offsite exposures comparable to the guideline exposures of 10 CFR 50.34(a)(1) or
10 CFR 100.11, “Determination of Exclusion Area, Low Population Zone, and Population
Center Distance.”
In 1996, the staff also established a new requirement in 10 CFR 100.20, “Factors To Be
Considered When Evaluating Sites,” for the evaluation of the nature and proximity of manrelated hazards, such as dams, for applications submitted on or after January 10, 1997. This
regulation was not applied to existing reactors.
In 1975, the NRC published the Standard Review Plan (SRP) (NUREG/75-087, later published
as NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for
Nuclear Power Plants: LWR Edition”), which provides standardized review criteria to assist
the staff in evaluating safety analysis reports submitted by license applicants. Since its firstpublication, the SRP has undergone several revisions to incorporate new developments indesign and analysis technology. Since the last SRP update in 2007, the staff has establishedinterim staff guidance (ISG) in three areas related to protection from natural phenomena:(1) DC/COL-ISG-1, “Interim Staff Guidance on Seismic Issues of High Frequency Ground
Motion,” (2) DC/COLISG7, “Assessment of Normal and Extreme Winter Precipitation Loads
on the Roofs of Seismic Category I Structures,” and (3) DC/COL-ISG-20, “Seismic Margin
Analysis for New Reactors Based on Probabilistic Risk Assessment.” This interim guidancehas been applied only to new reactor reviews.
The staff has also published several regulatory guides (RGs) that address specific technicalissues related to protection from natural phenomena. These documents provide guidance
to licensees and applicants on implementing specific parts of the NRC’s regulations,
techniques used by the NRC staff in evaluating specific problems or postulated accidents,
and data needed by the staff in its review of applications for permits or licenses. These
guides include the following:• RG 1.29, “Seismic Design Classification,” issued in 1972 and updated in 1973, 1976, 1978,
and 2007• RG 1.59, “Design Basis Floods for Nuclear Power Plants,” issued in 1973 and updated in
1976 and 1977
• RG 1.60, “Design Response Spectra for Seismic Design of Nuclear Power Plants,” issued
in 1973• RG 1.102, “Flood Protection for Nuclear Power Plants,” issued in 1975 and updated in
1976• RG 1.125, “Physical Models for Design and Operation of Hydraulic Structures and Systems
for Nuclear Power Plants,” issued in 1977 and updated in 1978 and 2009
• RG 1.208, “A Performance-Based Approach To Define the Site-Specific Earthquake
Ground Motion,” issued in 2007
The NRC staff continually evaluates new information regarding natural phenomena,
including operational experience, and its potential impact on risk and overall plant safety.
These evaluations have led to new requirements or guidance as discussed above, updated
regulatory guidance, generic communications, and plant-specific actions to address
identified issues. Several examples are presented below.

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Safety Through Defense-in-Depth
Following the Sumatra earthquake and its accompanying tsunami in December 2004, the
NRC staff initiated a study to examine tsunami hazards at nuclear power plant sites, toreview offshore and onshore modeling of tsunami waves, to describe the effects of tsunami
waves on nuclear power plant SSCs, to develop potential approaches for screening sites
for tsunami effects, to identify the repository of historic tsunami data, and to examine ways
for an NRC reviewer to approach site safety assessment for a tsunami. The study, NUREG/
CR-6966, “Tsunami Hazard Assessment at Nuclear Power Plant Sites in the United Statesof America,” was published March 2009. The results of this study were incorporated in the2007 update of SRP Section 2.4.6, “Probable Maximum Tsunami Hazards.” As discussed in
NUREG/CR-6966, the 1977 revision to RG 1.59 (Revision 2) was expected to include guidance
for assessment of tsunamis as a flooding hazard, but that effort was not completed. The
staff is in the process of updating RG 1.59 to address tsunamis and other advances inflooding analysis. Since 1977, flood estimation techniques have significantly improved withthe availability of more accurate datasets and newer hydrologic, hydraulic, and hydrodynamic
models. It should be noted that the current fleet of reactors was sited before RG 1.59,
Revision 2, was issued.
In August 2010, the NRC initiated a proposed generic issue (GI) regarding flooding of nuclear
power plant sites following upstream dam failures. The staff evaluation of this issue is ongoing.
Lastly, the NRC is evaluating seismic hazards based on new Electric Power ResearchInstitute models used to estimate earthquake ground motion and updated models forearthquake sources in the Central and Eastern United States. The NRC is addressing thisissue through the ongoing evaluation of GI-199, “Implications of Updated Probabilistic
Seismic Hazard Estimates in Central and Eastern United States on Existing Plants,” initiatedJune 9, 2005. The results of the GI-199 safety/risk assessment stage were summarizedin Information Notice 2010-018, “Implications of Updated Probabilistic Seismic Hazard
Estimates in Central and Eastern United States on Existing Plants,” dated September 2,
2010. As discussed in Information Notice 2010-018, currently available seismic data andmodels show increased seismic hazard estimates for some operating nuclear power plantsites in the Central and Eastern United States. Determination of site-specific seismic
hazards and associated plant risk are planned for the next phase of GI-199.

TASK FORCE EVALUATION
Current NRC regulations and associated regulatory guidance provide a robust regulatory
approach for evaluation of site hazards associated with natural phenomena. However,this framework has evolved over time as new information regarding site hazards and theirpotential consequences has become available. As a result, the licensing bases, design, andlevel of protection from natural phenomena differ among the existing operating reactors in
the United States, depending on when the plant was constructed and when the plant waslicensed for operation. Over the years, the NRC has initiated several efforts to evaluate risks
and potential safety issues resulting from these differences.
The SEP, mentioned earlier, was a one-time evaluation, and integrated plant safetyassessments were published in the early 1980s for each of the plants included in the SEP.
The SEP covered several technical topics, including protection from natural phenomena(i.e., floods, seismic events, tornadoes, high winds). Even that reassessment was conducted
before satellite imaging, Doppler radar, and well-established theories of plate tectonics
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Safety Through Defense-in-Depthwere available. It is clear that our current state of knowledge far exceeds that available todecisionmakers three decades ago.
With regard to the IPEEE program, the staff performed a limited review of the IPEEE
submittals to determine whether the licensees’ IPEEE processes were capable of identifying
and addressing severe accident vulnerabilities caused by external events. The staff
published a summary of the results of the IPEEE program in NUREG-1742, “Perspectives
Gained from the Individual Plant Examination of External Events (IPEEE) Program,” in April
2002. However, the NRC reviews did not attempt to validate or verify the licensees’ IPEEE
results or the acceptability of proposed improvements. Further, the IPEEE analyses did
not document the potential safety impacts of proposed improvements, and plants were notrequired to report completion of proposed improvements to the NRC.
The SEP, IPEEE program, USI A-46, and other regulatory initiatives, including licensing
actions to address vulnerabilities, have resulted in some plant-specific safety enhancements
to address the risk of external events resulting from natural phenomena. However, the staff
has not undertaken a comprehensive reestablishment of the design basis for existing plantsthat would reflect the current state of knowledge or current licensing criteria. As a result,significant differences may exist between plants in the way they protect against design-basis
natural phenomena and the safety margin provided.
With regard to seismic hazards, as discussed above, available seismic data and models show
increased seismic hazard estimates for some operating nuclear power plant sites. The
state of knowledge of seismic hazards within the United States has evolved to the point that
it would be appropriate for licensees to reevaluate the designs of existing nuclear powerreactors to ensure that SSCs important to safety will withstand a seismic event without lossof capability to perform their intended safety function. As seismic knowledge continues toincrease, new seismic hazard data and models will be produced. Thus, the need to evaluate
the implications of updated seismic hazards on operating reactors will recur and need to bereevaluated at appropriate intervals.
With regard to flooding hazards, the assumptions and factors that were considered in flood
protection at operating plants vary. In some cases, the design basis does not consider the
probable maximum flood (PMF). In other cases, the PMF is calculated differently at units
colocated at the same site, depending on the time of licensing, resulting in different designbasis flood protection. The Task Force has observed that some plants have an overrelianceon operator actions and temporary flood mitigation measures such as sandbagging,temporary flood walls and barriers, and portable equipment to perform safety functions. In
addition, potential dam failures have been addressed inconsistently in the establishment of
the design-basis flood. In some cases, emphasis was placed on dam failures coincident withseismic events, while other mechanisms for dam failures were not fully considered. Lastly,
while tsunami hazards are not expected to be the limiting flood hazard for operating plantssited on the Atlantic Ocean and the Gulf of Mexico, plants in these coastal regions do notcurrently include an analysis of tsunami hazards in their licensing basis. Tsunami hazards
have been considered in the design basis for operating plants sited on the Pacific Ocean.
The Task Force has concluded that flooding risks are of concern due to a “cliff-edge”
effect, in that the safety consequences of a flooding event may increase sharply with asmall increase in the flooding level. Therefore, it would be very beneficial to safety for all
licensees to confirm that SSCs important to safety are adequately protected from floods.
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Safety Through Defense-in-DepthThis reevaluation should consider all appropriate internal and external flooding sources,
including the effects from local intense precipitation on the site, PMF on streams andrivers, storm surges, seiches, tsunamis, and dam failures. Similar to seismic hazards, newflooding hazard data and models will be produced from time to time. Thus, there would
be a continuing benefit to having operating reactors reevaluate the implications of updated
flooding hazards at appropriate intervals.
Protection from natural phenomena is critical for safe operation of nuclear power plants due
to potential common-cause failures and significant contribution to core damage frequency
from external events. Failure to adequately protect SSCs important to safety from appropriate
design-basis natural phenomena with appropriate safety margins has the potential forcommon-cause failures and significant consequences as demonstrated at Fukushima.

Recommendation 2
The Task Force recommends that the NRC require licensees to reevaluate and upgrade
as necessary the design-basis seismic and flooding protection of SSCs for each
operating reactor.
The Task Force recommends that the Commission direct the following actions to ensure
adequate protection from natural phenomena, consistent with the current state of knowledge
and analytical methods. These should be undertaken to prevent fuel damage and to ensure
containment and spent fuel pool integrity:
2.1 Order licensees to reevaluate the seismic and flooding hazards at their sites against
current NRC requirements and guidance, and if necessary, update the design basis and
SSCs important to safety to protect against the updated hazards.
2.2 Initiate rulemaking to require licensees to confirm seismic hazards and flooding hazards
every 10 years and address any new and significant information. If necessary, update thedesign basis for SSCs important to safety to protect against the updated hazards.
2.3 Order licensees to perform seismic and flood protection walkdowns to identify andaddress plant-specific vulnerabilities and verify the adequacy of monitoring andmaintenance for protection features such as watertight barriers and seals in the interim
period until longer term actions are completed to update the design basis for external
events.

4.1.2 Protection from Concurrent Related Events
BACKGROUNDThe Task Force evaluated various related concurrent events and determined that fires andinternal floods induced by design-basis earthquakes warranted further Task Force consideration.
Seismically induced fires are frequent after earthquakes in urban areas. Seismic events
have also resulted in fires at nuclear power plants. Seismically induced fires have thepotential to cause multiple failures of safety-related systems and could create fires in
multiple locations at the site. Fire protection systems are not required to be functional
after a seismic event; therefore, efforts to fight seismically induced fires may be impaired bydegraded fire protection equipment. A seismic event may also impede offsite fire crews fromreaching the site, further challenging the capability to respond to such an event.
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Safety Through Defense-in-DepthThis scenario occurred following the July 16, 2007, magnitude 6.6 Niigata-Chuetsu Oki
earthquake that occurred 19 kilometers from the Kashiwazaki-Kariwa nuclear power plant,
located in Niigata, Japan. Following the earthquake, a fire occurred in the KashiwazakiKariwa Unit 3 electrical transformer. Sparks from a short circuit caused by large grounddisplacements of the transformer foundation caused the fire. The sparks ignited oil leaked fromthe transformer. Damage to the onsite fire protection equipment resulting from the seismic
event included multiple failures of the firefighting water system in Units 1, 2, 3, and 4; failure of
one of the fire water storage tanks; and failure of other fire suppression systems. Attempts by
the plant fire brigade to extinguish the fire were unsuccessful. The local municipality fire brigade
eventually extinguished the fire approximately 2 hours after it began. The fire was contained byfire protection walls and did not affect any plant safety equipment. However, the event provided
important insights into vulnerabilities from seismically induced fires.
The 2007 Japanese earthquake event also revealed insights regarding seismically induced
flooding. The plants experienced flooding from sloshing of the spent fuel pool, fire
suppression piping failure outside the Unit 1 reactor building that flowed into the plant
through cable penetrations, and a condenser flexible connection failure. While there were nosafety consequences, these flooding failures led to water flow to various portions of the plantthat could have caused SSC functional failures.

TASK FORCE EVALUATIONThe staff initiated Generic Safety Issue (GSI)-172, “Multiple System Responses Program(MSRP),” to address 21 potential safety concerns that were raised by the Advisory Committee
on Reactor Safeguards (ACRS) during the resolution of USI A-17, “Systems Interactions in
Nuclear Power Plants”; USI A-46, “Seismic Qualification of Equipment in Operating Plants”;
and USI A-47, “Safety Implications of Control Systems.” GSI-172 included the ACRS concernthat the resolution of USI A-46, other seismic requirements, or fire protection regulations did
not adequately address seismically induced fires. This concern was identified as Item 7.4.16
in NUREG/CR-5420, “Multiple System Responses Program—Identification of Concerns
Related to a Number of Specific Regulatory Issues,” published October 1989. ACRS was also
concerned that previous internal flooding studies had examined events such as pipe ruptures(and subsequent flooding) as single events and that the nature of a seismic event could
cause such problems in multiple locations simultaneously. This concern was identified as
Item 7.4.18 in NUREG/CR-5420.
The staff developed guidance for the review of the safety concerns of GSI-172 in the IPE
and IPEEE programs. As a result, the IPEEE program subsumed the issues related toseismically induced fires and floods.
With regard to seismically induced fires, NUREG-1742 states the following:
All of the IPEEE submittals reported that the licensees qualitatively examinedseismically induced fire interaction issues as part of the treatment of Sandia fire risk
scoping study issues. A few licensees performed a PRA study for seismically induced
fire-initiating events; albeit the level of detail varied from a simplistic probabilistic
analysis to inclusion in their plant’s seismic or fire PRA.
In most of the submittals, licensees included seismically induced fire considerations
within the scope of their overall seismic walkdown. The level of effort, scope, anddetail directed toward addressing seismically induced fire issues varied significantly
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Safety Through Defense-in-Depthamong the IPEEE submittals. One licensee did not discuss seismically induced fireevaluations in their IPEEE submittal. In most other cases, licensees limited theirseismically induced fire evaluations exclusively to assessing direct impacts on safe
shutdown equipment.
Seismically induced flooding events can potentially cause multiple failures of safety-related
systems. The rupture of small piping could provide flood sources with the potential to affectmultiple safety-related components simultaneously. Similarly, nonseismically qualified tanks
are a potential flood source of concern. While some licensees proposed plant improvements
to address related issues, NUREG-1742 states that the level of effort, scope, and detail
directed toward addressing seismically induced flooding issues varied significantly among
the IPEEE submittals. Some plants did not provide any information in their IPEEE submittals
to verify this issue.
The GSI-172 issue regarding seismically induced fires and floods was closed based on
the IPEEE results, and the NRC established no new requirements to prevent or mitigate
seismically induced fires or floods. The Task Force concludes that the agency should
reevaluate the closure of GSI-172 in light of the plant experience at the Kashiwazaki-Kariwa
nuclear plant and the potential for common-mode failures of plant safety equipment as theresult of seismically induced fires and floods.

Recommendation 3
The Task Force recommends, as part of the longer term review, that the NRC evaluate
potential enhancements to the capability to prevent or mitigate seismically induced
fires and floods.

4.2

MITIGATIONThe second level of defense-in-depth is mitigation. The Great East Japan Earthquake of2011 and the ensuing tsunami resulted in many mitigation systems at the Fukushima Daiichi Nuclear Power Plant being unable to operate. The subsequent challenges faced by
the operators at Fukushima Dai-ichi were beyond any faced previously at a commercial
nuclear reactor. The Task Force examined the U.S. regulations, guidance, and practices for
mitigating the consequences of accidents similar to those that occurred at Fukushima Daiichi. The following sections discuss the Task Force evaluation of insights from Fukushima
and provide recommendations for enhancing the mitigation capability of U.S. reactors withregard to prolonged loss of ac power, containment overpressure protection, combustible gas
control, spent fuel pool safety, and onsite emergency actions.

4.2.1 Prolonged Loss of Alternating Current PowerBACKGROUND
Alternating current (ac) electrical power is critically important to the safety of nuclear power
plants. Many of the SSCs intended to cool the nuclear fuel in the reactor and in the spent
fuel pools, to maintain radioactive containment systems, and to provide ventilation systems
to minimize release of radioactive materials rely on ac power. These systems depend onelectrical power to drive pumps, fans, and compressors, operate instrumentation and control
systems, and run motors to open and close valves and dampers. For these reasons, the loss
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Safety Through Defense-in-Depthof all ac power both onsite and offsite, as occurred at Fukushima, is highly significant. With
this in mind, the Task Force critically evaluated the design-basis protections to prevent lossof ac power, as discussed above, and the plants’ ability to maintain safety functions following
the loss of ac power as discussed below.
The NRC SBO rule (10 CFR 50.63) requires that each nuclear power plant must be able tocool the reactor core and maintain containment integrity for a specified duration of an SBO(defined in 10 CFR 50.2, “Definitions,” as a complete loss of required onsite and offsite acelectrical power). The specified duration is based on the following factors:• the redundancy of the onsite emergency ac power sources• the reliability of the onsite emergency ac power sources• the expected frequency of loss of offsite power• the probable time needed to restore offsite power
RG 1.155, “Station Blackout,” describes an acceptable means to comply with 10 CFR 50.63.
It primarily addresses three areas: (1) maintaining highly reliable onsite ac electric powersystems, (2) developing procedures and training to restore offsite and onsite emergency ac
power should either one or both become unavailable, and (3) ensuring that plants can copewith an SBO for some period of time based on the probability of occurrence of an SBO, as wellas the capability for restoring ac power to the site in a timely fashion. The RG provides anacceptable method for determining the specified duration for withstanding an SBO considering
the four factors identified in the rule language (10 CFR 50.63(a)(1)(i)–(iv)). The method
described in RG 1.155 results in a minimum acceptable SBO duration capability ranging from
2 to 16 hours. The result for all operating plants was a coping duration of 4 to 8 hours.
In evaluating the expected frequency of loss of offsite power, the guide addresses the
expected frequency of high winds, including those from tornadoes and hurricanes, and the
expected annual snowfall. The impact of other external hazards (e.g., seismic and flooding)on the frequency of loss of offsite power is not addressed. Nor does the guide address theconcurrent consequences on the facility of the external hazards impacting offsite power.
Consequently, common-cause failures of all onsite and offsite power resulting from anaturally occurring external event are not considered.
The analysis supporting the requirements of 10 CFR 50.54(hh) demonstrates, among otherthings, the potential for prolonged SBO conditions arising from terrorist actions including
large aircraft impacts and high explosives.
In a letter from J.E. Dyer (NRC) to A. Pietrangelo (Nuclear Energy Institute (NEI))
dated December 22, 2006, the NRC endorsed industry-developed guidance (NEI 06-02,
Revision 0, “License Amendment Request (LAR) Guidelines,” issued December 2006)
to demonstrate compliance with the requirements reflected in the NRC’s 2002 Interim
Compensatory Measures Order (and, subsequently, in 10 CFR 50.54(hh)(2)). To comply
with the requirements, each nuclear power plant licensee expanded command and control
capabilities and developed strategies and capability to provide cooling to fuel in the reactor
and spent fuel pool and to mitigate releases without reliance on the site’s ac electrical power
distribution system. The following mitigation capabilities were addressed:• for pressurized water reactors (PWRs):»» additional sources of coolant water for the reactor and steam generators
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Safety Through Defense-in-Depth»» methods to reduce pressure in and feed cooling water to the steam generators»» methods to cool the reactor core and minimize releases of radioactive materials
from containment• for BWRs:»» additional sources of coolant water for the reactor»» methods to reduce reactor pressure and feed cooling water to the reactor»» methods to cool the reactor core and reduce releases of radioactive materials
from containment• for spent fuel pools:»» additional sources of coolant water for and methods to inject or spray to thespent fuel pools»» methods to control leakage from damage to the spent fuel pools»» methods to reduce releases of radioactive materials from the spent fuel pools
The strategies, called EDMGs, implemented to meet the requirements of the 2002 Interim
Compensatory Measures Order, subsequent facility-specific license conditions, and ultimately
10 CFR 50.54(hh)(2), did not address external natural hazards (e.g., seismic, flooding,tornadoes, hurricanes) or initiating events other than extensive damage to the facilities causedby large fires or explosions. The EDMGs were designed to assist in the mitigation of one
particular beyond-design-basis scenario (i.e., loss of large areas of the facility due to fires or
explosions) that typically could involve SBO. The implementing guidance for the EDMGs sets
expectations that the amount and capacity of equipment to implement the strategies should be
sufficient to mitigate the consequences of an event that affects only one unit. In addition, theimplementing guidance for the EDMGs specifies that the required mitigation equipment used
under the EDMGs should be stored in an area physically separated by more than 91 meters(300 feet) from the equipment that could have been damaged by a large fire or explosion. The
equipment is not expected to be protected from design-basis or beyond-design-basis external
events, such as floods, earthquakes, or high winds.
Finally, critical instrumentation and control systems typically depend on the availability of
direct current (dc) electrical power at the facility. During a prolonged SBO, when ac power
would not be available and the battery banks become depleted, functional failure would
occur for nearly all instrumentation and control systems for monitoring critical parameters
and operating critical systems that ensure the integrity of the fuel in the reactor and spent
fuel pools and maintaining containment structures.

TASK FORCE EVALUATION
Information available at the time of this report indicates that the earthquake at Fukushima
Dai-ichi on March 11, 2011, caused a loss of all offsite sources of ac power to the six units,
and the ensuing tsunami caused failure of the emergency diesel generators for Units 1
through 4. It appears that Unit 1 also suffered damage to its onsite dc power source becauseof the tsunami. While most of the emergency diesel generators in Units 5 and 6 also failed
due to the tsunami, Unit 6 had one air-cooled emergency diesel generator that was not lost;
operators cross-tied that generator to essential Unit 5 and 6 systems, thus retaining theability to cool the Unit 5 and 6 reactors and spent fuel pools. The scope of the damage to the
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Safety Through Defense-in-Depth
offsite power infrastructure from the earthquake combined with the damage to the site fromthe tsunami resulted in the inability to restore any ac electrical power to Units 1 through4 for many days. The onsite emergency ac power sources have not yet been restored and,while ac power from offsite has been rerouted to the facility, damage to the onsite electrical
distribution and other critical equipment from the tsunami has not allowed full utilizationof this offsite ac power for installed plant equipment in Units 1 through 4. These four units
were in a prolonged SBO for many days and still are significantly challenged regarding thereliable distribution and use of ac electrical power.
Under these circumstances, factors that prevented restoration of power included (1)common-cause failure of onsite emergency ac sources due to flooding, (2) common-cause
failure of ac electrical distribution due to flooding, (3) common-cause failure of offsite(nonemergency) ac power distribution due to earthquake ground motion, and (4) offsite
infrastructure degradation from the earthquake and tsunami, which may have impeded
efforts to restore offsite power.
The Commission’s SBO requirements provide assurance that each nuclear power plant canmaintain adequate core cooling and maintain containment integrity for its approved coping
period (typically 4 or 8 hours) following an SBO. Also, if available, the equipment used forcompliance with 10 CFR 50.54(hh)(2) would provide additional ability to cool either the coreor the spent fuel pool and mitigate releases from primary and secondary containment
during a prolonged SBO. The implementing guidance for SBO focuses on high winds andheavy snowfalls in assessing potential external causes of loss of offsite power, but does notconsider the likelihood of loss of offsite power from other causes such as earthquakes andflooding. Also, the SBO rule does not require the ability to maintain reactor coolant system
integrity (i.e., PWR reactor coolant pump seal integrity) or to cool spent fuel. Further, theSBO rule focuses on preventing fuel damage and therefore does not consider the potential
for the buildup of hydrogen gas inside containment during a prolonged SBO condition andthe potential need to power hydrogen igniters in certain containment designs to mitigate thebuildup of hydrogen. Nor does it consider containment overpressure considerations andthe need to vent the containment in certain designs. Finally, the SBO rule does not requireconsideration of the impact on the station, and particularly on the onsite ac generation anddistribution, of the natural event that caused the loss of offsite ac electrical power.
During the prolonged SBO condition at the Fukushima Dai-ichi units, after the batteries weredamaged or depleted and no ac power was available to operate equipment or recharge thebatteries, the operators faced significant challenges in understanding the condition of thereactors, containments, and spent fuel pools because instrumentation was either lacking or
not functioning properly.
The Task Force concludes that revising 10 CFR 50.63 to expand the coping capability toinclude cooling the spent fuel, preventing a loss-of-coolant accident, and preventing
containment failure would be a significant benefit. The Task Force also concludes that a
strategy is needed to provide these functions for a prolonged period without ac power fromthe normal offsite or emergency onsite sources without the vital ac distribution systems
within the plant. The Task Force developed a three-part strategy to achieve this objective.
First, licensees would need a coping capability to maintain these functions for at least 8hours at each unit. This capability should minimize reliance on operator action during thisperiod in recognition of the potential for adverse work conditions related to the cause of
the prolonged SBO. Also, this 8-hour coping period would provide sufficient time to enable
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the operators to focus their efforts on restoring ac power and deploying onsite equipment
for the extended coping period to follow. Second, licensees would need an extended coping
capability to maintain these functions for at least 72 hours. The Task Force envisions that
establishing this extended coping capability would involve extensive operator actions during
the 8-hour coping period to deploy portable equipment maintained at the site in a mannerthat protects it from the severe natural phenomena that may cause it to be needed. The
purpose of the extended coping capability is to provide a bridge to the third component,
which would be a sustainable cooling capability that would utilize preplanned and prestaged
equipment at an offsite location that could be delivered and installed within 72 hours. The
planning would need to consider the potential for degraded transportation infrastructure
resulting from severe natural phenomena. The Task Force envisions that several sites mightshare the same prestaged equipment, provided that those sites are not susceptible to severe
damage from the same initiating event (e.g., earthquake or hurricane) requiring concurrent
demand for the same equipment.
To achieve the goal of providing an effective level of defense-in-depth for SBOs causedby external events beyond the design basis, the SBO mitigation equipment would need to
be protected from such events. Such protection was not available at Fukushima, where
beyond-design-basis flooding caused a prolonged SBO.
In the Task Force’s proposed risk-informed, defense-in-depth framework(Recommendation 1 in Section 3 of this report), the extent of the beyond-design-basis
external event that would require effective SBO mitigation would be established based on
the likelihood (i.e., estimated frequency) of such events. Extensive work on seismic events
has demonstrated that significant margin exists beyond a well-formulated seismic designbasis SSE. Typically, a margin of 2 exists above an SSE. That is to say, plant equipment
failures are unlikely unless seismic loads are about twice the design-basis SSE loads.
Such loads generally correspond to earthquake frequencies 5 to 10 times less likely thanthe design-basis SSE. The Task Force therefore concludes that SBO equipment designed
to the SSE would likely be sufficiently robust to function following a reasonably foreseeable
beyond-design-basis seismic event.
The Task Force has concluded that the situation is somewhat different in terms of
beyond-design-basis flooding. First, flooding can be caused by a number of different
phenomena: river flooding; dam failure; precipitation; storm surge; tsunami; or internal
failures of pipes, pumps, or tanks within the plant. Second, flooding can have a cliff-edge
effect; that is, a small increase in flooding level can produce a large effect in terms ofequipment failure and potential plant damage. With respect to this issue, the experienceat Fukushima Dai-ichi Units 5 and 6 appears more informative than that at Fukushima
Dai-ichi Units 1, 2, 3, and 4. Units 5 and 6 are sited at an elevation of 13 meters (43 feet)above sea level and, based on the information available, the tsunami reached a level of
14 meters (46 feet), producing about 1 meter (3 feet) of flooding on the site. In contrast,Units 1 through 4 appear to have been inundated by about 5 meters (16 feet) of sea water.
The extensive damage at Units 1 through 4 is therefore not surprising. However, Units 5
and 6 also experienced extensive damage of critical safety equipment as a result of about
1 meter (3 feet) of flooding, leaving the units at significant risk of core damage. Only one
air-cooled diesel generator remained available at Unit 6 and functioned with significant
operator action to maintain cooling at the two units. This circumstance illustrates theconcept of a cliff-edge effect.
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The Task Force also considered external fires, as well as tornado and hurricane winds, as
they relate to the qualification of SBO equipment. The Task Force has concluded that theexisting regulations for both fires and high-wind requirements include sufficient safetymargins to obviate the need for additional special protection for SBO equipment. The Task
Force concludes that neither external fires nor high winds should be characterized as “cliffedge” phenomena.
Based on the preceding considerations, the Task Force concludes that, to have SBO equipment
function effectively as a layer of defense-in-depth, it would need to be protected from flooding
beyond the design basis. The Task Force has also concluded that the safety margin built into
the design-basis flood would not be sufficient to provide the desired level of protection.
The Task Force considers the issue of flood protection for SBO equipment to besignificant and recognizes that flooding protection of such equipment would require
it to be located at a suitable elevation (above the design-basis flood plus a significant
margin) or require it to have an effective watertight enclosure. Establishing suchprotections may be difficult at some sites. Nevertheless, the Task Force concludes thatsuch protection of SBO equipment is essential to implementation of the recommended
framework for reactor safety.
A beyond-design-basis flood could be established through extensive, probabilistic hazards
analysis. As a practical matter, and to prevent undue delays in implementing additional
SBO protections, the Task Force concludes that locating SBO mitigation equipment in
the plant one level above flood level (about 5 to 6 meters (15 to 20 feet)) or in watertightenclosures would provide sufficient enhanced protection for this level of defense-in-depth.
These recommendations for revision to 10 CFR 50.63 would provide additional safetymargins for a prolonged SBO as a part of the overall risk-informed, defense-in-depth
regulatory framework providing adequate protection of public health and safety.
In addition, the EDMGs and associated equipment could be helpful and available promptly
to the operators to mitigate accidents such as those that occurred at Fukushima. However,
the two issues discussed above result in limited effectiveness of the EDMG strategies fornaturally occurring events that typically affect more than one unit.
As an interim measure, the equipment to implement the EDMGs would need to bereasonably protected from external events (i.e., stored in existing locations that arereasonably protected from significant floods and involve robust structures with enhanced
protection from seismic and wind-related events). In addition, this equipment would need
to be expanded to provide sufficient capacity to allow for a multiunit event response.

Recommendation 4
The Task Force recommends that the NRC strengthen SBO mitigation capability at all
operating and new reactors for design-basis and beyond-design-basis external events.
The Task Force recommends that the Commission direct the staff to begin the actions
given below to further enhance the ability of nuclear power plants to deal with the effects ofprolonged SBO conditions at single and multiunit sites without damage to the nuclear fuelin the reactor or spent fuel pool and without the loss of reactor coolant system or primarycontainment integrity.
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4.1 Initiate rulemaking to revise 10 CFR 50.63 to require each operating and new reactor
licensee to (1) establish a minimum coping time of 8 hours for a loss of all ac power, (2)establish the equipment, procedures, and training necessary to implement an “extended
loss of all ac” coping time of 72 hours for core and spent fuel pool cooling and for reactor
coolant system and primary containment integrity as needed, and (3) preplan andprestage offsite resources to support uninterrupted core and spent fuel pool cooling,
and reactor coolant system and containment integrity as needed, including the ability todeliver the equipment to the site in the time period allowed for extended coping, underconditions involving significant degradation of offsite transportation infrastructureassociated with significant natural disasters.
• The purpose of the 8-hour minimum coping capability is to enable operators to restore ac
power or to establish the proposed 72-hour extended coping capability. Core and spent
fuel pool cooling would need to be provided and primary containment isolation capability
and reactor coolant system integrity maintained by equipment (including essential
instrumentation and controls) independent of ac power (e.g., turbine-driven, diesel-driven,
dc-operated, air-operated, or passive systems).
• The 8-hour coping systems and equipment would be protected from damage from all
design-basis events and extended beyond-design-basis events by either locating theequipment one level (i.e., 5 to 6 meters (15 to 20 feet)) above the plant design-basis
flooding level or in water-tight enclosures.
• The 8-hour coping strategy should ensure that core and spent fuel pool cooling ismaintained and unmanageable leakage of coolant does not occur (e.g., from PWR reactor
coolant pump seal failure).
• The 8-hour coping strategy should also ensure that containment integrity can beestablished if needed, including the capability to operate wetwell vents for BWR facilities
with Mark I and Mark II containments, and one train of hydrogen igniters at BWR facilities
with Mark III containments and at PWR facilities with ice condenser containments.
• The design of the systems supporting the 8-hour minimum coping time should bebased on a conservative analysis of the capacity of ac-independent equipment andinstrumentation. The 8-hour coping capability should only rely on permanently installed
equipment. Operator actions relied upon to maintain this coping capability during this8-hour period should be limited to those types of actions consistent with routine types ofoperational activities governed by established procedures and training. This would enable
the operators to focus on actions needed to restore ac power and to deploy and operate
equipment to be used for the extended coping period. The systems and procedures
necessary for this function need not be single failure-proof, but they should be included in
the plant’s licensing basis (i.e., described in the final safety analysis report) and should be
subject to the quality assurance requirements of Appendix B to 10 CFR Part 50.
• The design of the systems supporting the 72-hour extended coping time should cover
the same scope of functions as the 8-hour minimum coping time, but it can be basedon realistic analysis with reasonable operator action using portable or permanently
installed equipment governed by established procedures and training. This extended
coping time will be sufficient to allow time for the effective acquisition, transportation,
installation, and use of preplanned and prestaged offsite resources for continued
achievement of the goals of core and spent fuel pool cooling, and reactor coolant
system and primary containment integrity.
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Safety Through Defense-in-Depth• The prestaged equipment could be used to satisfy these requirements at multiple sites aslong as the sites would not reasonably be expected to experience the same natural eventat the same time.
• As part of the revision to 10 CFR 50.63, the NRC should require that the equipment andpersonnel necessary to implement the minimum and extended coping strategies shall
include sufficient capacity to provide core and spent fuel pool cooling, and reactor coolingsystem and primary containment integrity for all units at a multiunit facility. The staff
should also make the appropriate revisions to the definitions of “station blackout” and“alternate ac source” in 10 CFR 50.2.
4.2 Order licensees to provide reasonable protection for equipment currently provided
pursuant to 10 CFR 50.54(hh)(2) from the effects of design-basis external events and to
add equipment as needed to address multiunit events while other requirements are being
revised and implemented.
• This existing equipment currently provides some of the coping capability that isrecommended for the long term, but current storage requirements do not ensure that it
will be available after a design-basis external event. This requirement would increase thelikelihood that the equipment will be available if called upon.
• The staff should also consider conforming changes to the requirements in
10 CFR 50.54(hh)(2) to address multiunit response capacity.

4.2.2 Containment Overpressure Protection
BACKGROUND
BWRs have robust capability to provide cooling water to the reactor core through many
diverse safety-related and nonsafety-related systems when ac power is available. In
addition, BWRs typically have two ac-independent ways of providing cooling water. In
October 1975, the NRC published NUREG-75/014, “Reactor Safety Study: An Assessment of
Accident Risks in U.S. Commercial Nuclear Power Plants (WASH-1400).” One of the many
insights developed through the study was that the risk of containment failure during severe
accidents was higher at BWRs with Mark I containments because the containment volumeof Mark I containment designs was significantly less than that of the other containment
designs, approximately one-sixth the volume of large dry PWR containments.
In December 1990, the NRC published NUREG-1150, which documented a relatively
high containment failure probability in the event of core damage at a BWR with a Mark I
containment design because of the smaller containment volume. The failure probability
was dominated by accident sequences involving a loss of feedwater with loss of decay heatremoval (due to multiple failures of safety systems) resulting in core damage and subsequentcontainment failure. These sequences are often referred to as “TW” sequences. A prolonged
SBO can also produce these failure sequences.
Concurrent with the development of NUREG-1150, the NRC completed the Containment
Performance Improvement Program for Mark I Containments, identifying a number ofrecommendations to enhance the performance of these containment designs. The NRC
staff described the improvement program recommendations in SECY-89-017, “Mark I
Containment Performance Improvement Program,” issued January 1989. The Commission
evaluated the staff recommendations and determined that the containment improvements
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Safety Through Defense-in-Depthshould be evaluated on a plant-specific basis through the IPE program, with the exceptionof the recommendation for hardened wetwell vents. Specifically, the Commission directed
the staff to approve the installation of hardened vents under the provisions of 10 CFR 50.59,“Changes, Tests and Experiments,” for licensees that, on their own initiative, elected toincorporate that improvement into their plants.
The staff issued GL 89-16, “Installation of a Hardened Wetwell Vent,” on September 1, 1989,
describing the safety benefits of the hardened vent modification and requesting licensees toprovide information on their plans to install that modification, or to provide cost information
should they choose not to install the modification. Eventually, all BWR facilities with Mark I
containment designs voluntarily installed a hardened vent. No regulatory requirement was
imposed. The designs of those vents varied from plant to plant. At some facilities, thehardened vent relied on ac-operated valves, some relied on dc-operated or air-operated
valves, and some incorporated passive rupture disks along with isolation valves. Eachdifferent design has different operational complexities during a prolonged SBO scenario. The
updated final safety analysis report for each facility includes a description of the hardenedvent, but the vent is not a required design feature for that facility.
In BWRs with a Mark II containment design, the containment volume could be approximately
25 percent larger than the volume of Mark I containments. In the resolution of GSI157,“Containment Performance,” the staff concluded that that the need for hardened vents at
BWRs with Mark II containments should be evaluated on a plant-specific basis through theIPE program. Eight BWR units in the United States have Mark II containment designs. Threeof these units have installed hardened vents, and the remaining five units at three sites havenot installed hardened vents.

TASK FORCE EVALUATION
Information available at the time of this report indicates that, during the days following
the Fukushima Dai-ichi prolonged SBO event, primary containment (drywell) pressure in
Units 1, 2, and 3 substantially exceeded the design pressure for the containments. The
operators attempted to vent containment, but they were significantly challenged operating
the wetwell (suppression pool) vents because of complications from the prolonged
SBO. Units 1, 2, 3, and 4 use the Mark I containment design; however, because Mark II
containment designs are only slightly larger in volume than Mark I containment designs,
it can reasonably be concluded that a Mark II under similar circumstances would havesuffered similar consequences.
The process at Fukushima Dai-ichi Units 1, 2, 3, and 4 for venting the wetwell involves
opening one ac-powered motor-operated valve to permit air pressure to open air-operated
valves in the vent line, and then opening another ac-powered motor-operated valve in linewith the air-operated valves, permitting containment pressure to impact a rupture disk
designed to open if containment pressure is significantly above design pressure. If all ofthese actions are successful, the containment would vent directly to the plant stack, andcontainment integrity could be reestablished by closing either the in-line ac-powered motoroperated valve or the air-operated valves. In a prolonged SBO situation, these actions would
not be possible from the control room because of the loss of ac power and the depletion of
the batteries providing dc control power for the valves. It is unclear whether the operators
were ever successful in venting the containment in Unit 1, 2, or 3. The operators’ inability
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to vent the containments complicated their ability to cool the reactor core, challenged thecontainment function, and likely resulted in the leakage of hydrogen gas into the reactor
building, precipitating significant explosions in Units 1, 3, and 4.
Ensuring that BWR Mark I and Mark II containments have reliable hardened venting
capability would significantly enhance the capability of those BWRs to mitigate serious
beyond-design-basis accidents. A reliable venting system could be designed to be
independent of ac power and to operate with limited operator actions from the control room.
Alternatively, a reliable venting capability could be provided through a passive containment
venting design, such as rupture disks with ac-independent isolation valves to reestablish
containment following rupture of the disk. The Task Force concludes that the addition orconfirmation of a reliable hardened wetwell vent in BWR facilities with Mark I and Mark II
containment designs would have a significantly safety benefit.
During the longer term review, the staff needs to reevaluate the design of other containment
structures for operating reactors to reaffirm the past conclusion that hardened vents are notnecessary to mitigate certain beyond-design-basis accident scenarios.

Recommendation 5
The Task Force recommends requiring reliable hardened vent designs in BWR facilities withMark I and Mark II containments.
The Task Force recommends that the Commission direct the staff to take the following
actions to ensure the effectiveness of hardened vents:
5.1 Order licensees to include a reliable hardened vent in BWR Mark I and Mark II
containments.
• This order should include performance objectives for the design of hardened vents toensure reliable operation and ease of use (both opening and closing) during a prolonged
SBO.
5.2 Reevaluate the need for hardened vents for other containment designs, considering theinsights from the Fukushima accident. Depending on the outcome of the reevaluation,
appropriate regulatory action should be taken for any containment designs requiring
hardened vents.

4.2.3 Combustible Gas Control
BACKGROUND
The NRC regulations in 10 CFR 50.44, “Combustible Gas Control for Nuclear Power
Reactors,” require reactors either to operate with their containment atmosphere inerted,
resulting in the lack of oxygen to support combustion, or to have the capability for controlling
combustible gas generated from a metal-water reaction involving approximately threequarters of the fuel cladding so that there is no impact on the containment structural
integrity. RG 1.7, “Control of Combustible Gas Concentrations In Containment,” provides
further guidance on regulatory expectations for combustible gas control.
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PWR facilities with ice condenser containments are required to have hydrogen igniters insidecontainment to control the buildup of hydrogen gas. These igniters are operated in two
redundant trains, with each train powered by one of the redundant safety-grade dc electrical
power systems. GSI-189 raised questions about the effectiveness of these igniter systems
during a prolonged SBO scenario. In response to the issues raised in GSI-189, licensees
operating BWRs with Mark III containments and PWRs with ice condenser containments
voluntarily installed nonsafety-grade backup electrical power to one train of the igniters,
independent of the safety-grade ac and dc onsite power systems. PWR facilities with large
dry containments do not control hydrogen buildup inside the containment structure because
the containment volume is sufficient to keep the pressure spike of potential hydrogendeflagrations within the design pressure of the structure.

TASK FORCE EVALUATION
Information available at the time of this report indicates that, during the days following
the Fukushima prolonged SBO event, Units 1, 3, and 4 experienced explosions, causing
significant damage to the reactor buildings for those units. It is believed that the explosions
in Units 1 and 3 resulted from hydrogen gas that was liberated inside the drywell duringhigh-temperature zirconium fuel cladding reactions with water and that hydrogen gas
migrated to the reactor building. The migration route of the hydrogen gas from the primarycontainment to reactor building has not yet been determined definitively; however, thefailure to prevent, through containment venting, the primary containment pressure fromsignificantly exceeding the design pressure likely contributed to the transport of hydrogengas. It is believed that the explosion in the Unit 4 reactor building also resulted fromhydrogen gas, but the source of the gas in Unit 4 is not yet clear. Unit 2 may also haveexperienced a hydrogen explosion in its suppression pool inside containment. However, themechanism for suppression pool failure remains unclear.
The method of combustible gas control in BWR Mark I and Mark II containments (i.e., containment
inerting with nitrogen) will prevent hydrogen fires or explosions as long as containment remains
isolated, but it will not eliminate the hydrogen resulting from an accident damaging the core. In
contrast, other designs eliminate hydrogen through controlled burning (BWR Mark III and PWR ice
condenser containment designs) or by accommodating an associated hydrogen explosion (PWR
large dry containments). This means that in a BWR Mark I or Mark II containment, the hydrogenmust be kept in containment by controlling containment pressure without venting (i.e., through heatremoval from the containment when possible) or by venting to a safe location.
Implementation of Task Force Recommendation 4, associated with prolonged SBO, would
reduce the likelihood of core damage and hydrogen production. In addition, implementationof Recommendation 5 to enhance the containment venting capabilities for Mark I and Mark II
containments, while primarily intended for overpressure protection, would also provide forthe reliable venting of hydrogen to the atmosphere. These two steps would greatly reduce
the likelihood of hydrogen explosions from a severe accident.
Sufficient information is not yet available for the Task Force to reasonably formulate any
further specific recommendations related to combustible gas control based on insights fromthe Fukushima Dai-ichi accident.

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Recommendation 6
The Task Force recommends, as part of the longer term review, that the NRC identify insights
about hydrogen control and mitigation inside containment or in other buildings as additional
information is revealed through further study of the Fukushima Dai-ichi accident.

4.2.4 Spent Fuel Pool Safety
BACKGROUND
SSCs for spent fuel storage and handling have safety classifications that reflect theirimportance to safety. SSCs essential to retaining the inventory of spent fuel pool water
covering the spent fuel and to maintaining a substantial margin to criticality are typicallyclassified as safety related. Such safety-related SSCs include the spent fuel poolstructure and penetrations, the spent fuel storage racks, the neutron-absorbing panelsin the racks, and the spent fuel itself. Some fuel handling equipment is also safety
related. Because the consequences of many fuel handling events and loss of spent fuelforced cooling events have been evaluated and found to be small, these events are notclassified as design-basis events. Consequently, other spent fuel storage and handlingequipment and spent fuel pool water inventory makeup and cooling systems may not
be classified as safety related. At U.S. reactors, some of the spent fuel pool cooling andmakeup systems are powered by safety-grade ac electrical power and some are poweredby nonsafety-grade ac electrical power.
Licensees developed the EDMGs, described in Section 4.2.1 of this report, to maintain or
restore spent fuel pool cooling and mitigate releases under the circumstances associatedwith the loss of large areas of the plant due to fire or explosion. These requirements arosefrom analyses conducted by the NRC staff after September 11, 2001, that demonstrated,among other things, the potential for prolonged SBO conditions.
The NRC regulations in 10 CFR 50.36, “Technical Specifications,” describe therequirements for technical specifications included in each facility operating license. The
technical specifications describe requirements, called limiting conditions for operation(LCOs), for the characteristics of safety systems and components that must be available
during various reactor operating modes. Specifically for electrical power systems, when
the reactor is operating, the LCOs typically require two independent onsite and at least
two offsite ac electrical power systems to be operable to provide electrical power to thesafety equipment. When the reactor is shut down and defueled for maintenance work andall of the fuel is placed in the spent fuel pool, the LCOs do not require any electrical powersystems to be operable. In this condition, the heat load in the spent fuel pool is highest,
and the time margin to boil could be as little as several hours immediately after transfer ofthe fuel from the reactor to the spent fuel pool. The fuel could be at risk of uncovery in asa few as 10 hours without pool cooling; ac electrical power is important to the cooling thenuclear fuel in the spent fuel pools and to providing ventilation systems to minimize therelease of radioactive materials. The regulation in 10 CFR 50.36(b) authorizes the NRC to
include additional technical specifications as it finds appropriate.
The spent fuel pools at Fukushima Dai-ichi Units 1 through 4 contained many fewer
assemblies than typically stored in U.S. reactor unit spent fuel pools. Unit 4 contained more
assemblies than the other units because Unit 4 was in an extended outage and the reactor
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was defueled, transferring all assemblies to the spent fuel pool to facilitate maintenanceactivities associated with the reactor systems.
Fuel Assemblies in the Reactor Cores and Spent Fuel Pools at Fukushima Dai-ichi
Units 1 through 4
Reactor Core

Spent Fuel Pool

Unit 1

400

292

Unit 2

548

587

Unit 3548

514

Unit 40

1,331

The storage capacity of U.S. reactor unit spent fuel pools ranges from less than
2,000 assemblies to nearly 5,000 assemblies, with an average storage capacity of
approximately 3,000 spent fuel assemblies. Typically, the U.S. spent fuel pools are filledwith spent fuel assemblies up to approximately three-quarters of their capacity. In addition
to the unit-specific spent fuel pools at the Fukushima Dai-ichi facility, a separate spent fuelstorage building existed onsite with wet pool storage of spent fuel containing 6,291 spent fuelassemblies. U.S. reactor facilities do not typically have an additional spent fuel wet storage
building like that at Fukushima Dai-ichi.
A spent fuel assembly just removed from a BWR after an operating cycle generates
approximately 10 kilowatts of heat, and that heat level diminishes rapidly over time. Withina short period of time after removal from the reactor, spent fuel assemblies are coolable
with water sprays, and within a few years, the assembly rate of heat generation is reduced
to a few percent of its original heat generation rate. Consequently, the number of spent
fuel assemblies in the spent fuel pool does not significantly affect the ability to cool the
spent fuel in the pool.
Many U.S. reactors have dry spent fuel storage capability in addition to wet storage in the
spent fuel pools. These dry storage facilities are called independent spent fuel storage
installations, and fuel stored in such an installation first must be placed into a dry cask.
The NRC reviews and certifies the designs for these dry casks. Typically, 5 years must passfrom the time an assembly is removed from the reactor before it can be placed into a dry
cask. After 5 years, the heat generation rate is very low.
In addition to responding to prolonged SBO conditions, the EDMGs also addresscontingencies for cases when the spent fuel pools are unable to retain water above the topof the fuel. This situation could also occur with significant beyond-design-basis seismic
events. One effective strategy for mitigation is to blanket the spent fuel with a water spray.
Current spent fuel pool instrumentation provides limited indication and typicallydepends on the availability of dc electrical power at the facility. That power is provided
either through inverters powered by ac electrical power or by the station’s safety-grade
redundant battery banks. Direct spent fuel pool level indication is rarely provided in
the control room for the current nuclear fleet. Typically, level is measured using a levelswitch in the skimmer surge tank. During a prolonged SBO, ac power would not be
available and the battery banks would be depleted, resulting in functional failure of nearly
all instrumentation and control systems for monitoring spent fuel pool parameters andoperating systems ensuring the integrity of the fuel in the spent fuel pools.

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TASK FORCE EVALUATION
Information available at the time of this report indicates that the earthquake at Fukushimacaused a loss of all offsite sources of ac power to the six units, and the ensuing tsunami
caused the failure of the emergency diesel generators for Units 1 through 4. The scopeof the damage to the offsite power infrastructure from the earthquake, combined with thedamage to the site from the tsunami, resulted in the inability to restore any ac electrical
power to Units 1 through 4 for many days.
During this protracted SBO condition, no ac power was available to operate equipment, andthe batteries were depleted. This resulted in having no onsite capability to provide water
inventory or cooling to the spent fuel pools, and the operators were significantly challenged
in understanding the condition of the spent fuel pools because of the lack of instrumentation
or because of instrumentation that was not functioning properly. Eventually, spent fuelcooling was provided by pumper trucks employing high booms to spray water from a distanceinto the spent fuel pools.
The reliability and availability of U.S. spent fuel pool makeup systems would be better
ensured if the NRC had a requirement for those systems to have safety-related ac power thatis controlled under a technical specification LCO.
Substantial additional defense-in-depth would be provided, and cooling the spent fuel ina prolonged SBO would have been substantially simplified, with an installed seismicallyqualified means to spray water into the spent fuel pools, including an easily accessible
connection to supply the water (e.g., using a portable pump or pumper truck) at grade
outside the building.
The lack of information on the conditions of the fuel in the Fukushima spent fuel pools wasa significant problem in monitoring the course of the accident and contributed to a poor
understanding of possible radiation releases and to confusion about the need and priorities
for support equipment. The Task Force therefore concludes that reliable information on theconditions in the spent fuel pool is essential to any effective response to a prolonged SBO orother similarly challenging accident.
The Task Force concludes that clear and coherent requirements to ensure that the plant staff
can understand the condition of the spent fuel pool and its water inventory and coolability
and to provide reliable, diverse, and simple means to cool the spent fuel pool under various
circumstances are essential to maintaining defense-in-depth. The Task Force seessignificant value in ensuring that sufficient cooling capacity exists under various designbasis natural phenomena.
Final understanding of the detailed sequence of events and the condition of the spent fuelpools will not be fully developed for some time. Based on the information to date, it is clearthat the two most cogent insights from the Fukushima accident related to spent fuel poolsafety concern (1) the instrumentation to provide information about the condition of the pooland the spent fuel and (2) the plant’s capability for cooling and water inventory management.
The Task Force’s recommendations address these two critical areas.

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Recommendation 7
The Task Force recommends enhancing spent fuel pool makeup capability andinstrumentation for the spent fuel pool.
The Task Force recommends that the Commission direct the staff to do the following:
7.1 Order licensees to provide sufficient safety-related instrumentation, able to withstanddesign-basis natural phenomena, to monitor key spent fuel pool parameters (i.e., waterlevel, temperature, and area radiation levels) from the control room.
7.2 Order licensees to provide safety-related ac electrical power for the spent fuel poolmakeup system.
7.3 Order licensees to revise their technical specifications to address requirements to have
one train of onsite emergency electrical power operable for spent fuel pool makeup
and spent fuel pool instrumentation when there is irradiated fuel in the spent fuel pool,
regardless of the operational mode of the reactor.
7.4 Order licensees to have an installed seismically qualified means to spray water into thespent fuel pools, including an easily accessible connection to supply the water (e.g., using
a portable pump or pumper truck) at grade outside the building.
7.5 Initiate rulemaking or licensing activities or both to require the actions related to thespent fuel pool described in detailed recommendations 7.1–7.4.

4.2.5 Onsite Emergency Actions
BACKGROUND
A number of guidelines and procedures guide onsite emergency actions by reactor
operators, depending on the nature and extent of events at the plant. As discussed inprevious sections, nuclear reactors are designed to handle the loss of offsite electrical powerwith multiple onsite emergency diesel generators. Events such as a loss of offsite power are
within the plants’ design basis and are addressed by plant procedures (typically abnormal
operating procedures, alarm response procedures, and EOPs). These procedures instruct
the plant operators on the steps necessary to take the plant from full-power operation to asafe shutdown condition.
EOPs have long been part of the NRC’s safety requirements. The NRC regulations address
them through 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 in
the administrative controls section of the technical specifications for each plant. Numerous
RGs and technical reports (e.g., NUREG-0660, “NRC Action Plan Developed as a Resultof 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 February 2004) also address EOPs. In addition, the EOPs
are the subject of a national consensus standard (American National Standards Institute/
American Nuclear Society 3.21994, “Administrative Controls and Quality Assurance for
the Operational Phase of Nuclear Power Plants”). The training and both the written andsimulator exams for licensing reactor operators and senior reactor operators also include
the EOPs. While implementing EOPs, the event command and control functions remain in
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the control room under the direction of the shift supervisor and plant manager, both of whom
have senior reactor operator licenses.
An SBO is considered to be beyond the plant’s design basis. This means that the regulatory
requirements stated above do not apply. Instead, specialized requirements included within
10 CFR 50.63 cover SBO. In addition to addressing the loss of offsite electrical power,
10 CFR 50.63 requires reactor licensees to address the simultaneous loss of onsite andoffsite electrical power (i.e., SBO) by providing an additional source of electrical power(referred to as an “alternate ac” source) or by showing that the plant could cope with acomplete loss of ac power by removing decay heat for a specified period of time. In thecase of an SBO, the operators would follow a set of procedures (usually abnormal operatingprocedures) required by 10 CFR 50.63(c)(ii) and (iii). These procedures would instruct theoperators in maintaining safety functions using the alternate ac power source or throughcoping strategies. In addition, procedures would direct operators to take steps to restore theonsite and offsite sources of ac power.
In addition, the nuclear industry developed SAMGs during the 1980s and 1990s in responseto the TMI accident and followup activities. These followup activities included extensive
research and study (including several PRAs) on severe accidents and severe accident
phenomena. The SAMGs are intended for use by plant technical support staff, usually
located in the plant’s Technical Support Center (TSC), and are meant to enhance the ability ofthe operators to manage accident sequences that progress beyond the point where EOPs andother plant procedures are applicable and useful. EOPs typically cover accidents to the pointof loss of core cooling and initiation of inadequate core cooling (e.g., core exit temperatures
in PWRs greater than 649 degrees Celsius (1,200 degrees Fahrenheit)). As stated in theWestinghouse SAMG documentation, “the SAMG is designed to fill the void between the EOPs
and the E-Plan [the procedure and guidance for emergency response].” While implementing
SAMGs, the accident command and control functions shift to the TSC and typically to theemergency coordinator or shift technical advisor or both.
In GL 88-20, Supplement 2, “Accident Management Strategies for Consideration in the
Individual Plant Examination Process,” dated April 4, 1990, the NRC encouraged but did not
require licensees to develop and implement SAMGs. Since the SAMGs are voluntary andtargeted to technical support staff, the formal training and licensing of plant operators does
not address them.
Following the terrorist events of September 11, 2001, the NRC issued security advisories,
orders, license conditions, and ultimately a new regulation (10 CFR 50.54(hh)) to require
licensees to develop and implement guidance and strategies intended to maintain or
restore core cooling and containment and spent fuel pool cooling capabilities under thecircumstances associated with the loss of large areas of the plant due to fire or explosion.
These requirements have led to the development of EDMGs at all U.S. nuclear power plants.
The guidelines and strategies included in the EDMGs are NRC requirements, and the NRC
inspected EDMG implementation following the events at Fukushima under TI 2515/183,“Followup to the Fukushima Daiichi Nuclear Station Fuel Damage Event,” dated March 23,
2011. The results of this TI are available on the NRC Web site at http://www.nrc.gov/NRR/OVERSIGHT/ASSESS/follow-up-rpts.html. In addition, the NRC added the requirementsof 10 CFR 50.54(hh) to the agency’s routine inspection program as part of the triennial fire
protection inspections.

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As stated in the industry’s guidance document NEI 06-12, “B.5.b Phase 2&3 Submittal
Guideline,” Revision 2, issued December 2006 (Agencywide Documents Access andManagement System (ADAMS) Accession No. ML070090060), “the initial EDMGs are not a
type of emergency operating procedures (EOPs), nor are they intended to be a replacement
for EOPs. They are, in fact, intended to be used when the normal command and control
structure is disabled and the use of EOPs is not feasible.” In terms of command and control,
either control room, plant, TSC, or emergency operations facility (EOF) staff could make
EDMG decisions. The EDMGs do not play a large role in the formal training and licensing ofplant operators.

TASK FORCE EVALUATION
The accidents at Fukushima highlight the importance of having plant operators who are well
prepared and well supported by technically sound and practical procedures, guidelines, andstrategies. In addition, it is clear that a preplanned approach to command and control anddecisionmaking during an emergency is vital.
Each of the onsite emergency action programs (the abnormal operating procedures, alarm
response procedures, EOPs, SAMGs, and EDMGs) contributes to overall emergency response
capability of plant and operators to mitigate accidents. It is clear that the SAMGs and
EDMGs complement the EOPs in an important way. The NRC and industry have establishedthe command and control responsibilities for each of these programs, although notnecessarily in a consistent manner. Each of these programs was developed at a different
time to serve a different purpose, and each of these programs is treated differently in
the NRC’s regulations, inspection program, and licensing process, as well as in licensee
programs and organizations.
Soon after its establishment, the Task Force considered whether to include the SAMGs
as a regulatory requirement and whether to require the integration of the SAMGs into
a coherent and holistic program for onsite emergency response activities. This was
based on the important role the SAMGs would play in onsite emergency response—justas important as that of EOPs and EDMGs, both required programs. To gain insights into
the current implementation of the SAMGs, the Task Force requested that NRC inspectors
collect information on how each licensee had implemented that industry voluntary initiative.
The inspectors collected information on the initial implementation, ongoing training, andmaintenance of the SAMGs under TI 2515/184. The results of the inspection under the
SAMG TI reinforced the value of making SAMGs a requirement. The inspectors observed
inconsistent implementation of SAMGs and attributed it to the voluntary nature of thisinitiative. The results of this TI are available on the NRC Web site at http://www.nrc.gov/
NRR/OVERSIGHT/ASSESS/SAMGs.html.
In order to ensure the effectiveness of these onsite emergency action programs and to support
the effectiveness of mitigation, sound training as well as other readiness measures areimportant. The inspections under TI 2515/184 identified that some licensee programs include
extensive classroom and simulator training and testing on SAMGs, while others do not.
The Task Force concludes that all U.S. plants have addressed all of the elements of onsite
emergency actions that need to be accomplished by reactor operators. However, the overall
effectiveness of those programs could be substantially enhanced through further integration,
including clarification of transition points, command and control, decisionmaking, and
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Safety Through Defense-in-Depththrough rigorous training that includes conditions that are as close to real accident
conditions as feasible.
The Task Force also concludes that action is warranted to confirm, augment, consolidate,
simplify, and strengthen current regulatory and industry programs in a manner that
produces a single, comprehensive framework for accident mitigation, built around NRCapproved licensee technical specifications. These modified technical specifications
would consolidate EOPs, SAMGs, EDMGs, and other important elements of emergency
procedures, guidance, and tools in a manner that would clarify command and control anddecisionmaking during accidents.
Integration of these accident support functions in a logical and coherent manner and withappropriate regulatory treatment to ensure the effectiveness of operator actions during
events would substantially increase the effectiveness of the overall event mitigation. Since
the current requirements in this area apply only to normal operation and emergencies
within the plant’s design basis, they appear outdated and inconsistent with Commission
decisions in policy statements and rulemakings to regulate accident mitigation in otherareas beyond the plant’s design basis. The Task Force concludes that an expansion of theregulatory requirements to include procedures for beyond-design-basis events is warranted,
and that such an expansion would redefine the scope of such activities to include them in
the regulatory framework to provide defense-in-depth and to ensure adequate protection of
public health and safety.
The new treatment of accident procedures would also include the authority to implement
SAMGs and EDMGs as necessary and as described or referenced in the plant technicalspecifications without the need to seek NRC permission or to invoke 10 CFR 50.54(x) and (y).
This change would further clarify authority, streamline decisionmaking, and prevent potential
delays in taking important emergency actions.
The effectiveness of onsite emergency actions is a very important part of the overall safetyof nuclear power plants. The NRC could strengthen the current system substantially byrequiring more formal, rigorous, and frequent training of reactor operators and other onsite
emergency response staff on realistic accident scenarios with realistic conditions.

Recommendation 8
The Task Force recommends strengthening and integrating onsite emergency response
capabilities such as EOPs, SAMGs, and EDMGs.
The Task Force recommends that the Commission direct the staff to further enhance thecurrent capabilities for onsite emergency actions in the following ways:
8.1 Order licensees to modify the EOP technical guidelines (required by Supplement 1,“Requirements for Emergency Response Capability,” to NUREG-0737, issued
January 1983 (GL 82-33), to (1) include EOPs, SAMGs, and EDMGs in an integrated
manner, (2) specify clear command and control strategies for their implementation, and(3) stipulate appropriate qualification and training for those who make decisions during
emergencies.
• The Task Force strongly advises that the NRC encourage plant owners groups toundertake this activity rather than have each licensee develop its own approach. In
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addition, the Task Force encourages the use of the established NRC practice of publishing
RG (rather than NUREGs, supplements to NUREGs, or GLs) for endorsing any acceptableapproaches submitted by the industry.
8.2 Modify Section 5.0, “Administrative Controls,” of the Standard Technical Specifications foreach operating reactor design to reference the approved EOP technical guidelines for thatplant design.
8.3 Order licensees to modify each plant’s technical specifications to conform to the above changes.
8.4 Initiate rulemaking to require more realistic, hands-on training and exercises on SAMGs
and EDMGs for all staff expected to implement the strategies and those licensee staff
expected to make decisions during emergencies, including emergency coordinators andemergency directors.

4.3

EMERGENCY PREPAREDNESS

If mitigation is not successful in preventing a release of radioactive materials from theplant, EP ensures that adequate protective actions are in place to protect public health andsafety. Protective actions are taken to avoid or reduce radiation dose. As a condition of theirlicense, operators of nuclear power plants must develop and maintain EP plans that meetcomprehensive NRC EP requirements.
Following the TMI accident in 1979, the NRC revised its regulations to substantiallyenhance EP requirements. The regulations in 10 CFR 50.47, “Emergency Plans,” include
the 16 planning standards of 10 CFR 50.47(b), and Appendix E, “Emergency Planningand Preparedness for Production and Utilization Facilities,” to 10 CFR Part 50 describes
information needed to demonstrate compliance with EP requirements. The NRC considers
the evaluation criteria in NUREG-0654, “Criteria for Preparation and Evaluation ofRadiological Emergency Response Plans and Preparedness in Support of Nuclear Power
Plants,” Revision 1, issued November 1980, to be guidance and an acceptable means fordemonstrating compliance with the Commission’s regulations.
After the events of September 11, 2001, the NRC staff reviewed the EP basis for nuclear
power plants, considering the impact of hostile actions unanticipated at the time the basis
was established. The staff concluded that the EP basis remains valid. The staff based its
conclusion on studies that confirmed that the timing and magnitude of releases related
to hostile actions would be no more severe than those associated with the other accident
sequences considered in the EP basis. However, hostile actions could present uniquechallenges to EP programs since they differ from operational events for which licensees andoffsite response organizations typically plan, train, and exercise. The accident at Fukushima
provides its own unique challenges to EP; however, they are rooted in more traditional
accident progression sequences.
The agency is in the final stages of a comprehensive rulemaking effort to revise EP
regulations, as outlined in SECY-11-0053, “Final Rule: Enhancements to Emergency
Preparedness Regulations (10 CFR Part 50 and 10 CFR Part 52),” dated April 8, 2011. These
revisions codify the hostile-action-based enhancements, among others. The Task Forcehas reviewed the draft final rule, particularly the provisions that provide enhancementsthat would also address challenges to EP programs caused by an accident like that at
Fukushima. The subsections below discuss these provisions.
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In analyzing the accident and its impact on EP in the United States, the Task Force focused
primarily on EP considerations for multiunit and prolonged SBO events, as discussed in thefollowing section. Section 4.3.2 captures other EP insights beyond that focus.

4.3.1	Emergency Preparedness Considerations for Multiunit Events andProlonged Station Blackout
The accident at Fukushima has shown that prolonged SBO and multiunit events are realities
that must be addressed as part of EP. While of low probability, these events have the potential
for severe consequences that require an effective EP response. The Task Force’s evaluation inthis section focuses on a licensee’s capability to respond during these types of events.
Currently, the United States has 29 single-unit sites, 33 dual-unit sites, and 3 triple-unit
sites. The agency is currently reviewing new reactor applications that may add units toexisting sites; however, no applicant has requested to bring the total number of units at asingle site to more than four. In most cases, proposed quadruple-unit sites have physicalseparation between the two existing and the two proposed units.

BACKGROUND
Requirements for emergency response personnel, staffing, and emergency worker protection
are codified in 10 CFR 50.47(b)(1), 10 CFR 50.47(b)(2), and 10 CFR 50.47(b)(11).
The proposed EP rulemaking amends Appendix E to 10 CFR Part 50 to address concerns
about the assignment of tasks or responsibilities to on-shift emergency responseorganization (ERO) personnel that would potentially overburden them and prevent the timely
performance of their functions under the emergency plan. Licensees must have enoughon-shift staff to perform specified tasks in various functional areas of emergency response
24 hours a day, 7 days a week.
The regulation in 10 CFR 50.47(b)(2) requires adequate on-shift staffing levels but gives
no clear definition of “adequate.” This provides some flexibility in how licensees assign
emergency plan implementation duties to on-shift personnel. The proposed EP rule better
ensures sufficient on-shift staffing in the threat environment after September 11, 2001, bylimiting the assignment of responsibilities that on-shift ERO members would likely performconcurrently with their emergency plan functions.
The regulations in 10 CFR 50.47(b)(1) and 10 CFR 50.47(b)(2) require the assignment anddefinition of emergency response responsibilities to address decisionmaking and commandand control.
Dose assessment, or dose projection, as required by 10 CFR 50.47(b)(9), is the primarymeans for assessing the potential consequences of a radiological emergency. Appendix E to
10 CFR Part 50 describes the required content of emergency plans, including assessment of
the impact of the release of radioactive materials.
After declaration of an emergency, licensees must notify the NRC and State and local
response organizations. In addition, licensees are responsible for providing notificationand instruction to the public within the plume exposure pathway emergency planning zone
(EPZ), as required in 10 CFR 50.27(b)(5) and Section IV.D.3 of Appendix E to 10 CFR Part 50.
The predominant method used in the United States for alerting the public is an alert andnotification system based on sirens to provide an acoustic warning signal. Some sites
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employ other means, such as tone alert radios and route alerting, as either primary or
supplemental alerting methods. The public then typically receives information aboutan event and offsite protective actions via emergency alert system broadcasts or othermeans, such as mobile loudspeakers. The State and local governments are responsible foractivating systems to alert and notify the public, and FEMA evaluates this capability.
The regulations in 10 CFR 50.47(b)(6) and 10 CFR 50.47(b)(8) cover the communications
equipment needed during an emergency.
As a result of the TMI accident, the NRC and others recognized a need to substantially
improve the NRC’s ability to acquire data on plant conditions during emergencies. The
Commission has defined the NRC’s role in the event of an emergency primarily as oneof monitoring the licensee to ensure that appropriate recommendations are made withrespect to offsite protective actions. Other aspects of the NRC role include supporting thelicensee with technical analysis and logistic support, supporting offsite authorities (including
confirming the licensee’s recommendations to offsite authorities), keeping other Federal
agencies and entities informed of the status of the incident, and keeping the media informed
based on the NRC’s knowledge of the status of the incident, including coordination with otherpublic affairs groups.
To fulfill its role, the NRC requires accurate and timely data on four types of parameters:(1) core and coolant system conditions must be known well enough to assess the extent or
likelihood of core damage, (2) conditions inside the containment building must be known
well enough to assess the likelihood and consequence of its failure, (3) radioactivity release
rates must be available promptly to assess the immediacy and degree of public danger, and(4) data from the plant’s meteorological tower are necessary to assess the likely patterns ofpotential or actual impact on the public.
Experience with the voice-only emergency communications link, previously used for datatransmission, has demonstrated that excessive amounts of time are needed for the routine
transmission of data and for verification or correction of data that appear questionable.
Therefore, the NRC selected the Emergency Response Data System (ERDS) to fulfill theagency’s data collection needs. ERDS allows for the direct electronic transmission of
selected parameters from the electronic data systems that are currently installed at licensee
facilities. ERDS was designed for use only during emergencies and would be activated by
the licensees during declared emergencies classified at the Alert lever or higher to begin
transmission to the NRC Operations Center. ERDS would be supplemented with voice
transmission for essential data not available on the licensee’s systems, rather than require a
modification to the existing system to transmit that data.
Section IV of Appendix E to 10 CFR Part 50 codifies the requirements for ERDS. The NRC
requires power reactor licensees to transmit ERDS data to a server at NRC Headquarters.
Many licensees currently use analog modulator/demodulators (modems) to establish
point-to-point data connections. Although this technology was state of the art when ERDS
was first implemented, it is now obsolete, and replacement equipment is no longer readily
available. In addition, the use of modems inherently introduces a cyber security vulnerability
to the systems to which the modems are attached.
As part of the current effort to modernize the ERDS infrastructure, the NRC has beenworking with individual licensees to develop an acceptable solution to replace the existing
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to-point data pathway between the licensee site and NRC Headquarters. This technology
permits all ERDS-enabled facilities to connect to the NRC simultaneously, thereby enhancing
the NRC’s ability to respond to incidents that may affect multiple licensees simultaneously,
such as grid instability events. Some licensees are currently in the process of replacing
the existing analog modems with VPN devices on a voluntary basis, while others have yet tocommit to the initiative.
The regulation in 10 CFR 50.47(b)(14) requires periodic drills and exercises to develop key
skills and evaluate major portions of emergency response capabilities, and 10 CFR 50.47(b)(15) requires radiological emergency response training. NUREG-0654 includes further detailon the implementation of these regulations.
The regulation in 10 CFR 50.47(b)(8) requires the provision and maintenance of adequate
emergency facilities and equipment to support the emergency response. Studies of the TMI
accident identified the need for extensive improvements in the overall response to accidents
at nuclear plants, including enhanced facilities and systems to support the control room in
mitigating the consequences of accidents and to support the licensee’s capability to respondto abnormal plant conditions. NUREG-0696, “Functional Criteria for Emergency Response
Facilities,” issued February 1981, describes the facilities and systems that licensees can use
to improve emergency response to accidents, such as the TSC, operational support center(OSC), and EOF. The document also provides guidance on the functional criteria for thesefacilities and on the integrated support these facilities offer to the control room.
Licensees must have the capability to augment the on-shift staff within a short time after
the declaration of an emergency. To accomplish this, licensees typically staff an onsite TSC,
which relieves the control room of emergency response duties and allows operators to focus
on reactor plant safety. Responders also staff an onsite OSC to provide an assembly areafor damage repair teams. Finally, licensees establish an EOF to function as the center forevaluation and coordination activities related to the emergency and as the focal point forproviding information to Federal, State, Tribal, and local authorities involved in the response.
The proposed EP rulemaking amends Section IV.E.8 of Appendix E to 10 CFR Part 50 toaddress concerns about ERO augmentation during hostile action by requiring licensees to
identify alternative facilities as staging areas for responding staff. This alternate site musthave the capability for communication with the EOF, control room, and plant security; thecapability to perform offsite notifications; and the capability for engineering assessment
activities, including damage control team planning and preparation.
Licensees have submitted several requests to the NRC to consolidate EOFs for plants they
operate within a State or in multiple States. Previous regulatory standards did not address
the capabilities and functional requirements for a consolidated EOF, such as capabilities forhandling simultaneous events at two or more sites. The NRC is revising, via the proposed EP
rule, the regulations and associated guidance to reflect a performance-based approach for
EOFs and to provide functional requirements for these facilities.

TASK FORCE EVALUATIONThe events in Japan have highlighted the importance of the human element during responseto an emergency. The current regulatory approach provides for staffing and protectionof staff during postulated reactor accidents. External events on the scale of Fukushima
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Safety Through Defense-in-Depthcause concerns that the existing framework would present challenges to personnel andtheir safety. In particular, the presence of staff onsite during the initiation of the emergency
condition, staff needed to augment the current onsite staff, and staff needed for crew relief
poses significant challenges during large external events.
The proposed EP rulemaking will require an analysis of the duties of on-shift personnel
to ensure the fulfillment of all required functions. The accident at Fukushima supports
an expansion of this analysis to include whether enough responders are available formultiunit events. During traditional accident scenarios, most required response personnel
would be summoned to the site within a certain required time. During a catastrophic
natural disaster, the local infrastructure may challenge the timely augmentation of on-shift
personnel. Should licensees choose not to add more on-shift personnel, they would needto have a viable notification and transportation strategy for ensuring that the staff needed
to augment the site response would be available to respond effectively. In addition, sincemore responders are needed during a multiunit event, additional protective equipment
would be needed.
Through discussions with inspectors and technical experts, the Task Force has found thatthe current framework for command and control during an accident has been developed andpracticed over the years and that authorities for decisionmaking during an event are welldefined. However, in this area as well, multiunit events create a nuance to the command andcontrol structure that is not yet fully developed.
During an accident like that at Fukushima, it may be necessary to make difficult decisions toprioritize limited response resources. The EOF remains the primary facility for interaction
with offsite authorities, and the TSC remains the primary facility for technical responseto the accident. Currently, during a General Emergency, the licensee’s emergency
director assigned with the authority to lead the licensee response is located in the EOF.
An emergency director in the TSC remains in command of the technical assessment
and damage control aspects of the response. During a multiunit event, the lead TSC
official would be in the best position to address the triage and prioritization of resource
requirements for each unit.
Ensuring that the response framework contains the correct level of authority, knowledge,
and experience is paramount to successful response. In light of the Fukushima accident,
the staff should explore concepts such as whether decisionmaking authority is in the correct
location (i.e., at the facility), whether currently licensed operators need to be integral to the
ERO outside of the control room (i.e., in the TSC), and whether licensee emergency directors
should have a formal “license” qualification for severe accident management in addition totheir existing qualification requirements, and different than a reactor operator license.
The accident at Fukushima also presented challenges with respect to dose assessment
capability because of the multiunit nature of the release. While monitoring the accident,
dose assessors at the NRC had to use makeshift, ad hoc methods to consider the source
term from these multiple concurrent releases and overlay the release points to arrive at a
final sitewide dose projection. The difficulty of conducting this dose assessment highlights
the gap in capability for U.S. plants to perform multiunit dose assessment. Currently,
dose assessment software (such as the NRC’s RASCAL) is not designed to model multiunit
accidents. The presence of releases from multiple units and spent fuel pools at Fukushima
has highlighted the need for the ability to project doses from releases at multiple units.
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In the absence of a software solution, licensees would need to develop emergency plan
guidance that outlines how to conduct a multiunit dose assessment, including spent fuelpools as release points, to ensure that this capability exists.
Following a large natural disaster, such as the accident at Fukushima, the ability to notify
both government authorities and the public could be challenged. Licensees use traditional
telephones, cellular telephones, satellite telephones, short wave radio, and the Internet tocommunicate with the NRC and State and local governments. The Task Force believes thatlicensees have enough redundant and diverse methods to communicate with the NRC andState and local governments that it is reasonable to expect the successful communication ofa declared emergency onsite. Because of the diverse nature of these methods, it is unlikelythat a common-cause failure would disable all means of communication.
With respect to alerting the public within the plume exposure pathway EPZ, the Task Forcefinds that the provisions being incorporated by the EP rulemaking will enhance the existing
public alerting framework. The proposed final rule addresses provisions for a backup to
the primary alert and notification system. The backup measures would be implemented if
the primary means of alert and notification were unavailable during an emergency. Theseenhancements will provide an additional layer of preparedness that will be useful during alarge-scale natural disaster.
Communications equipment plays a critical role in the effective conduct of any incident
response effort. Currently, licensees use an array of different telecommunications
devices to communicate onsite, including hardwired telephones, cellular telephones,
satellite telephones, radios, and pagers. Since many of these devices depend onelectrical power, most by battery, their use during a prolonged SBO would be limited bybattery life. Additionally, since hardwired telephones and cellular telephones rely upon
offsite infrastructure (i.e., phone switches and cell towers) that could be damaged or
destroyed by the event, their use may be limited. Onsite radios may be hampered by alack of power to radio repeaters.
During the accident at Fukushima, numerous organizations assessed the evolution of
the situation onsite and sought data to aid decisionmakers during the response. Having
accurate, real-time data from the site allows for the performance of a multitude of analyses.
In addition, the more data that can be provided using automated sources, the less burden
placed on the licensee to provide information. Having data provided directly from automatedsources at the site also gives confidence to government authorities and the public thatthe plant operator is not filtering the details of an evolving accident. However, the current
regulatory approach and requirements do not ensure that ERDS data would be available
during a prolonged SBO or during other natural disasters when power supplies could be lost
and transmission capability may be affected.
Another challenge evident from Fukushima is the need for archived data to aid in thereconstruction of events after an accident. When ERDS data are transmitted to the
NRC servers, the NRC stores the incoming data. Ensuring that licensees have reliable
power sources and transmission methods will therefore ensure that data are archived for
accident reconstruction.
Effective onsite and offsite response necessitates abundant communication. During a
prolonged SBO, an awareness of plant conditions becomes even more critical. Ensuring that
the NRC and State stakeholders continue to receive plant data, especially during the most
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severe accidents, adds another layer of both technical assessment capability and confidence
that onsite conditions are being communicated in a timely and accurate way.
Effective and timely ERDS modernization is necessary to ensure that during multiunit events,
the NRC is able to receive data from all affected nuclear units. Given the current limited
modem connections available for simultaneous use, the NRC should ensure that licensees
upgrade to the VPN solution in an expedient manner. Once the current modernization effort
is complete, the staff should then evaluate additional modernization to address redundanttransmission capability, completeness of data, and continuous monitoring.
The conduct of training and exercises provides confidence that emergency plans areworkable and that personnel would be successful in mitigating the consequences of an
accident. However, EP drills and exercises currently do not consider prolonged SBO or
multiunit accident scenarios. Therefore, training and exercises should explore and practicethe concepts of command and control, decisionmaking, prioritization, and contingency
planning under these conditions. Since licensees will likely rely heavily on offsite support
during this type of event, exercising the steps necessary to identify and acquire offsite
equipment and support is an important element to practice.
Both prolonged SBO and multiunit events present new challenges to EP facilities that were
not considered when the NRC issued NUREG-0696. The accident at Fukushima has clearlyshown that these events are a reality. While several utilities have implemented combined
EOFs that are capable of handling multiunit events, licensee onsite TSCs and OSCs havenot been designed or drilled for multiunit events. TSCs and OSCs are not as conducive toconsolidation since most job functions in these facilities would be specific to a particularunit, and attempts to have one person manage the technical response for multiple units
could introduce opportunities for errors. The proposed EP rulemaking requires theidentification of alternate facilities. During a multiunit event, licensees could use both thenormal and alternate facilities to provide enough capacity for response. The proposed EP
rulemaking codifies performance-based requirements for combined EOFs. EOFs are moresuited to handling multiunit events since they represent the interface with offsite responders,regardless of the number of onsite units in emergency status. The NRC should use aperformance-based approach to TSCs, OSCs, and EOFs to ensure flexibility for licensees.

Recommendation 9
The Task Force recommends that the NRC require that facility emergency plans address
prolonged SBO and multiunit events.
The Task Force recommends that the Commission direct the staff to do the following:
9.1 Initiate rulemaking to require EP enhancements for multiunit events in the following areas:• personnel and staffing• dose assessment capability• training and exercises• equipment and facilities

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9.2 Initiate rulemaking to require EP enhancements for prolonged SBO in the following areas:• communications capability• ERDS capability• training and exercises• equipment and facilities
9.3 Order licensees to do the following until rulemaking is complete:• Determine and implement the required staff to fill all necessary positions for responding
to a multiunit event.
• Add guidance to the emergency plan that documents how to perform a multiunit dose
assessment (including releases from spent fuel pools) using the licensee’s site-specificdose assessment software and approach.
• Conduct periodic training and exercises for multiunit and prolonged SBO scenarios. Practice(simulate) the identification and acquisition of offsite resources, to the extent possible.
• Ensure that EP equipment and facilities are sufficient for dealing with multiunit andprolonged SBO scenarios.
• Provide a means to power communications equipment needed to communicate
onsite (e.g., radios for response teams and between facilities) and offsite (e.g., cellular
telephones, satellite telephones) during a prolonged SBO.
• Maintain ERDS capability throughout the accident.
9.4 Order licensees to complete the ERDS modernization initiative by June 2012 to ensure
multiunit site monitoring capability.

Recommendation 10
The Task Force recommends, as part of the longer term review, that the NRC should pursue
additional EP topics related to multiunit events and prolonged SBO.
The Task Force recommends that the Commission direct the staff to do the following:
10.1 Analyze current protective equipment requirements for emergency responders andguidance based upon insights from the accident at Fukushima.
10.2 Evaluate the command and control structure and the qualifications of decisionmakers toensure that the proper level of authority and oversight exists in the correct facility for along-term SBO or multiunit accident or both.
• Concepts such as whether decisionmaking authority is in the correct location (i.e., at thefacility), whether currently licensed operators need to be integral to the ERO outside of thecontrol room (i.e., in the TSC), and whether licensee emergency directors should have aformal “license” qualification for severe accident management.
10.3 Evaluate ERDS to do the following:• Determine an alternate method (e.g., via satellite) to transmit ERDS data that does not relyon hardwired infrastructure that could be unavailable during a severe natural disaster.
• Determine whether the data set currently being received from each site is sufficient formodern assessment needs.
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Safety Through Defense-in-Depth• Determine whether ERDS should be required to transmit continuously so that no operator
action is needed during an emergency.

4.3.2 Other Emergency Preparedness Insights
In addition to prolonged SBO and multiunit events, the Task Force has identified other EP
insights from the accident at Fukushima. An overarching lesson is that major damage toinfrastructure in the area surrounding the plant might challenge an effective emergency
response. In addition, the NRC should learn from the real-world implementation ofprotective actions and further develop concepts such as recovery and reentry. A third lesson
relates to the need for the public to be educated on radioactivity and radiological hazards
before an incident occurs.

BACKGROUND
Natural disasters, such as hurricanes, floods, large fires, and tornados, occur often in the
United States. State and local governments constantly manage these types of events andensure that emergency plans are workable. Likewise, the NRC works closely with FEMA
to ensure that the state of offsite preparedness continues to ensure reasonable assurance
that the public can be protected should an accident occur. While these disasters do notsignificantly impact nuclear plants because of the plants’ robust design features, they oftencause significant damage in the surrounding community and could challenge the effective
implementation of an offsite emergency plan should it be required. These natural disasters
may alter the preplanned emergency framework by changing evacuation routes (e.g., bridge
washed out or tree down in the roadway), disabling emergency sirens, or making evacuation
dangerous (e.g., icy roads). Local government decisionmakers take factors such as these
into consideration when formulating protective action decisions to determine how best
to protect the public. A catastrophic natural disaster would create additional challenges
beyond those routinely experienced by state and local emergency planners.
Currently, if a nuclear power plant shuts down as a result of a natural disaster and thedisaster is of such severity that damage or changes to the offsite emergency response
infrastructure may be substantial or are in question, the NRC and FEMA determine thestatus of offsite EP and coordinate approval of plant restart activities.
Evacuation time estimates (ETEs) are used as a tool to (1) develop and improve evacuationplans in advance of an accident and (2) to decide whether sheltering or evacuation is the more
protective response during an accident. Evacuation is preferred if a dose in excess of protective
action guides is probable, but it is not always more effective in reducing public exposure. ETEs
are a planning tool and do nothing to affect conditions during an actual evacuation. Draft
NUREG/CR-7002, “Criteria for Development of Evacuation Time Estimate Studies,” issued
May 2010, provides the latest guidance for licensees to develop a comprehensive set of ETEs.
Infrastructure damage could also affect the ability of licensees to bring equipment from
offsite sources to the site. The regulation in 10 CFR 50.47(b)(3) requires licensees tomaintain agreements with offsite organizations needed to support emergency response.
Typically, this includes local fire departments, law enforcement, and medical support. As
part of the requirements of 10 CFR 50.54(hh)(2)(i), licensees must identify and seek readily
available agreements for additional offsite resources (both local and regional) that could
support fire fighting, electrical power, core cooling, and other needs.
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The regulation in 10 CFR 50.47(b)(10) requires licensees to make protective action
recommendations. To facilitate a preplanned strategy for protective actions during an
emergency, two EPZs surround each nuclear power plant. The exact size and shape ofeach EPZ is determined through detailed planning that includes consideration of thespecific conditions at each site, unique geographical features of the area, and demographic
information. This preplanned strategy for an EPZ provides a substantial basis to support
activity beyond the planning zone in the extremely unlikely event it would be needed. The
plume exposure pathway EPZ has a radius of about 16 kilometers (10 miles) from the reactor
site. Predetermined protective action plans in place for this EPZ are designed to avoid or
reduce dose from potential exposure to radioactive materials. These action plans include
sheltering, evacuation, and the use of potassium iodide (KI) where appropriate. The ingestion
exposure pathway EPZ has a radius of about 80 kilometers (50 miles) from the reactor site.
Predetermined protective action plans in place for this EPZ are designed to avoid or reducedose from the potential ingestion of radioactive materials. These action plans include a banof contaminated food and water.
NUREG-0396, “Planning Basis for the Development of State and Local Government
Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants,”
issued December 1978, establishes the concept and basis for the 10-mile and 50-mile
EPZs, describes the basis for the types of reactor accidents that should be considered in an
emergency plan, states that emergency planning is not based on probability but on public
perceptions of the problem and what could be done to protect public health and safety (as amatter of prudence rather than necessity), and concludes that the objective of an emergency
plan should be to provide dose savings for a spectrum of accidents that could produce offsite
doses in excess of the U.S. Environmental Protection Agency’s (EPA’s) protective action guides.
The regulations in 10 CFR 50.47(b)(13) also address general plans for recovery and reentry
and the means by which decisions to relax protective measures (e.g., allow reentry into
an evacuated area) are reached. While requirements for the licensees, States, andlocal governments appear in 10 CFR 50.47(b)(13) and NUREG-0654, the approach of the
U.S. Government to recovery and reentry involves numerous other Federal partners.
Appendix E to 10 CFR Part 50 and 10 CFR 50.47(b)(9) require radiation monitoring, whichrefers to the measurement of actual radiation dose. NUREG-0654 provides guidance tolicensees and offsite response organizations on topics including field monitoring, monitoring
equipment, and airborne plume tracking.
In January 2001, the Commission published a rule change to the NRC EP regulations toinclude consideration of the use of KI. If taken properly, KI may help reduce the dose ofradiation to the thyroid gland from radioactive iodine and thus reduce the risk of thyroid
cancer. If a person takes in radioactive iodine after consuming KI, the radioactive iodine will
be rapidly excreted from the body. The NRC has supplied KI tablets to States requesting
it for the population within the 10-mile EPZ. If warranted, KI is to be used to supplement
evacuation or sheltering, not to take the place of these actions.
The population closest to the nuclear power plant that is within the 10-mile EPZ is atgreatest risk of exposure to radiation and radioactive materials. When the population is
evacuated from the area and potentially contaminated foodstuffs are removed from themarket, the risk from further radioactive iodine exposure to the thyroid gland is essentiallyeliminated. Beyond the 10-mile EPZ, the major risk of radioiodine exposure is from the
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ingestion of contaminated foodstuffs, particularly milk products. Both EPA and the Foodand Drug Administration have published guidance to protect consumers from contaminatedfoods. These protective actions are preplanned in the 50-mile ingestion pathway EPZ.
In the unlikely event of a nuclear power plant accident, it is important to follow the directionof State or local governments in order to make sure that protective actions, such as taking
KI pills, are implemented safely and effectively for the affected population. While use of KI at
the correct time and dosage can help prevent the uptake of radioactive iodine in the thyroid
gland, it does not offer any protection from other radioisotopes or for other organs. KI is a
medical tool for a limited, targeted situation.

TASK FORCE EVALUATIONThe current regulatory approach for the evaluation of offsite EP following a natural disasteris robust and has proven its effectiveness following recent hurricanes, including Hurricane
Katrina. An NRC task force examined the lessons learned from the active 2005 hurricane
season in a report dated March 30, 2006 (ADAMS Accession No. ML060900005).
ETEs are currently recalculated when the population around a nuclear plant either increases
or decreases significantly. As supported by the proposed EP rule, the scenarios described
in NUREG/CR-7002 provide a basis for licensees to develop a comprehensive set of ETEs.
Performing additional time estimates for natural disasters with unpredictable damage would
offer no corresponding benefit to licensee personnel in providing appropriate protective
action recommendations to offsite officials or to offsite emergency planners in developing
evacuation and other protective action strategies. With regard to seismic events impacting
a plant site and causing a severe accident, consideration of the estimated time to evacuate
the area would require assessment of the overall condition of dwellings, buildings, roadways,
and other critical infrastructure at the time of the event. Additionally, when dwellings andother buildings are compromised and can no longer provide effective sheltering, evacuation
would be conducted regardless of the time needed. If sheltering is not an option, ETE values
are not relevant to decisionmaking. With regard to extreme weather conditions such as
significant hurricanes, evacuation might involve an area much larger than a nuclear power
plant site and begins days in advance of landfall. Therefore, the evacuation of the populationin the plant vicinity would likely be completed well before the hurricane affects that area.
The accident at Fukushima has illustrated the potential increased need for offsite
assistance to the licensee. In the case of large natural disasters such as earthquakes,
hurricanes, and floods, the phenomena challenging the plant will also have affected the
local community. In these cases, prearranged resources may not be available becauseof their inability to reach the plant site, other (potentially lifesaving) priorities within thecommunity, or the destruction of those resources.
During the emergency at Fukushima, conditions deteriorated such that Japanese officials
required additional protective actions up to and beyond a 20-kilometer (16-mile) area around
Fukushima (i.e., beyond the equivalent of the U.S. plume exposure pathway EPZ). The
possibility of making protective action recommendations for areas beyond the plume exposure
pathway EPZ has been a program consideration since the inception of the EPZ concept in the
United States. The emergency planning basis for U.S. plants, as discussed in NUREG-0654,
states, “…detailed planning within 10 miles would provide a substantial base for expansion of
response efforts in the event that this proved necessary.” NUREG-0654 goes on to state that it
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would be unlikely that any protective actions for the plume exposure pathway would be required
beyond the plume exposure pathway EPZ. It further stated that the plume exposure EPZ is ofsufficient size for actions within this zone to provide for substantial reduction in early severe
health effects (injuries or deaths) in the event of a worst case core melt accident.
While the U.S. EP framework has always noted that the plume exposure pathway EPZ
provides a basis for expansion, insights from real-world implementation at Fukushima,including the realities of multiunit events, might further enhance U.S. preparedness
for such an event. The Task Force acknowledges that every situation will differ, so
detailed preplanning in this area is not plausible. As information and insights emerge
about the challenges faced by Japanese officials while implementing protective actions
around Fukushima, the NRC and its partners should evaluate those insights to identify
enhancements to the decisionmaking framework in the United States.
Similarly, licensees and States are required to have plans for recovery and reentry; however,these plans remain largely conceptual and are rarely practiced. Since recovery and reentry
have proven to present challenges at Fukushima, the NRC should continue work in this area
to forward the U.S. Government approach.
In the area of radiation monitoring, licensees have numerous fixed radiation monitors locatedinside the plant on the property of the site. During an emergency, licensee field teams,
in addition to Federal, State, and local teams, are dispatched to take real-time radiation
readings offsite. While multiunit events do not specifically change radiation monitoring
concepts, a long-term SBO could challenge radiation monitors that rely upon an external
power supply. As long as field teams are adequately staffed, equipped, and capable of transit
given the nature of the natural disaster, field monitoring remains an effective method toacquire radiation data. In additional, real-time radiation monitoring helps to validate dose
projections, quantify actual dose, and provide confidence to the public and stakeholders thatthe conditions around the site are being accurately characterized.
During a radiological release, accurate and timely dose data are critical to validating dose
projections and ensuring dose reduction for the public. Having publicly available dose
data provides a level of public confidence. The staff should explore the concepts of ac
independence, survivability of equipment during a natural hazard, and ability to transmit
real-time radiation readings publicly via the Internet.
The concepts of radiation and radiation safety continue to be challenging topics to teach tothose not involved in nuclear fields. This is underscored by the invisible nature of radiation andthe potential chronic, in addition to acute, health affects that could be attributed to radiation
exposure. While the NRC and various other Federal, State, and local government agencies andother medical organizations have conducted education on the uses, benefits, and drawbacks of
KI, recent events have shown a continued gap in the public knowledge with respect to KI.
Based on the observed gaps in public awareness following the accident at Fukushima,an effort to increase education and outreach in the vicinity of each nuclear power plantis warranted. Misinformation and hysteria during a nuclear emergency challenge theagency’s goal of public confidence. Training should be targeted to the areas surrounding
each nuclear power plant in order to reach those who could be affected should an
emergency at a nuclear power plant occur. In addition to public participation, the
NRC should make extra effort to involve local response personnel, health officials,
decisionmakers, media, and local politicians.
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Recommendation 11
The Task Force recommends, as part of the longer term review, that the NRC should pursue
EP topics related to decisionmaking, radiation monitoring, and public education.
The Task Force recommends that the Commission direct the staff to do the following:
11.1 Study whether enhanced onsite emergency response resources are necessary tosupport the effective implementation of the licensees’ emergency plans, including theability to deliver the equipment to the site under conditions involving significant natural
events where degradation of offsite infrastructure or competing priorities for response
resources could delay or prevent the arrival of offsite aid.
11.2 Work with FEMA, States, and other external stakeholders to evaluate insights fromthe implementation of EP at Fukushima to identify potential enhancements to the U.S.
decisionmaking framework, including the concepts of recovery and reentry.
11.3 Study the efficacy of real-time radiation monitoring onsite and within the EPZs (including
consideration of ac independence and real-time availability on the Internet).
11.4 Conduct training, in coordination with the appropriate Federal partners, on radiation,
radiation safety, and the appropriate use of KI in the local community around eachnuclear power plant.

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5.

IMPLICATIONS FOR NRC PROGRAMS

Section 3 of this report presented the most significant Task Force finding andrecommendation relative to NRC programs (i.e., the NRC regulatory framework and how it
could be strengthened). This section will discuss impacts on the NRC inspection program,the management of NRC records and information, and NRC participation in international
activities.

5.1

NRC INSPECTION PROGRAMBACKGROUND
For many decades, the NRC has had a vigorous reactor inspection and oversight programfor operating reactors and vendor inspection program for new plant construction. The
ROP is a formal process integrating the NRC’s inspection, assessment, and enforcement
programs, and it has been in place for the last decade. The ROP evaluates the overall safety
performance of operating nuclear power reactors and communicates this information tolicensee management, members of the public, and other stakeholders. In addition, since2004, the NRC has included safety culture in the ROP for operating reactor licensees. Overthe past three decades, the safety performance of the nuclear fleet in the United States has
improved, as a result in some part of the effectiveness of the NRC’s inspection, assessment,
and enforcement programs.
Regarding reactor protection and mitigation systems, a fundamental characteristicof the ROP is that inspection activities or “samples” are selected for the relative risk
significance of the activity or equipment being examined based on its effect on coredamage frequency. Further, the NRC evaluates inspection findings in these areas anduses the significance determination process to determine significance based on risk.
The ROP’s reliance on risk undervalues the safety benefit of defense-in-depth andconsequently reduces the level of NRC resources focused on inspecting defense-indepth characteristics that contribute to safety.
In addition, the ROP does not consider the industry’s voluntary safety enhancements.
Consequently, the staff devotes limited inspection effort to voluntary initiatives such as
the implementation and adequacy of SAMGs or the implementation of the Groundwater
Protection Initiative. If it inspects them at all, the agency would likely evaluate voluntary
initiatives through nonrecurring inspections guided by TIs. Finally, the structure of the riskbased inspection program under the ROP focuses on licensee compliance with regulations
and requirements and leaves very limited opportunity for inspection staff to evaluate theadequacy of the licensing basis at a given facility.

TASK FORCE EVALUATION
Following the Fukushima accident, the NRC recognized the more general relevance andimportance of the SAMGs and EDMGs that had been developed, and later required, following
the accident at TMI and the terrorist events of September 11, 2001. Therefore, the NRC
issued TI 2515/183 to its resident inspectors to ensure licensee compliance with existing
requirements and to collect information on the readiness of these measures for use undervarious external challenges.
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Implications for NRC Programs
The Task Force has had the benefit of the results of the inspections under TI 2515/183 andconcludes that the effort was well planned and worthwhile. In addition, in an April 22, 2011,
memorandum from Charles L. Miller to Eric J. Leeds, “Task Force Request Regarding
Inspection of Severe Accident Management Guidelines,” the Task Force asked that inspectors
collect information on the implementation of the voluntary SAMG initiative. In response, the
NRC issued TI 2515/184. The Task Force has factored the information from both of thesestaff efforts into its evaluation and recommendations.
Through these two inspection activities, the Task Force also had the opportunity to compareindustry activities under a required program and a similar voluntary initiative (i.e., EDMGs
and SAMGs). Both programs had been effectively implemented, including initial program
formulation and licensee staff training. Those programs are now 10 to 20 years old, andsome licensees have maintained both programs in a manner expected for an importantsafety activity, including in terms of maintenance, configuration control, training, andretraining. However, some licensees have treated the industry voluntary initiative (the SAMG
program) in a significantly less rigorous and formal manner, so much so that the SAMG
inspection would have resulted in multiple violations had it been associated with a required
program. The results of the SAMG inspection do not indicate, nor does the Task Forceconclude that, the SAMGs would not have been effective if needed. However, indications ofprogrammatic weaknesses in the maintenance of the SAMGs are sufficient to recommend
strengthening this important activity.
On the basis of its evaluation, the Task Force concludes that enhancements to the inspection
program would improve its focus on safety. Since these modifications are all staff activities
not involving any backfit considerations, the NRC can pursue them based solely on theirvalue in improving the staff’s safety focus.

Recommendation 12
The Task Force recommends that the NRC strengthen regulatory oversight of licensee safety
performance (i.e., the ROP) by focusing more attention on defense-in-depth requirements
consistent with the recommended defense-in-depth framework.
The Task Force recommends the Commission direct the staff to strengthen the ROP by doing
the following:
12.1 Expand the scope of the annual ROP self assessment and biennial ROP realignment tomore fully include defense-in-depth considerations.
12.2 Enhance NRC staff training on severe accidents, including training resident inspectors on
SAMGs.

5.2

MANAGEMENT OF NRC RECORDS AND INFORMATIONBACKGROUND
In the 1990s, the NRC decided to prepare for a transition from paper records management
to electronic records management and began a process that resulted in the deployment of
ADAMS in the late 1990s and early 2000s. At present, essentially every document generated
by the agency is effectively integrated into a powerful electronic records management system
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that is searchable in a variety of ways. However, at the time of the original transition to
ADAMS, individual offices decided what records generated before the deployment of ADAMS
would be incorporated into the electronic system and what records would remain in paper
and microfiche format.
In SECY-06-0164, “The NRC Knowledge Management Program,” dated July 25, 2006,
the staff described to the Commission the plans for developing and implementing anagency knowledge management program. The program recognized that the agency is aknowledge-centric organization that depends on its staff to make sound regulatory decisions
to accomplish the agency’s mission of protecting people and the environment. The staff
informed the Commission about the NRC’s ongoing programs for maintaining explicit,
implicit, and tacit knowledge and new initiatives to leverage information technology solutions
to transfer knowledge. The knowledge management program focused on a variety of areas
to enhance identifying, retaining, transferring, and using critical information and knowledge
in the agency’s ongoing work. The staff updated the Commission on the implementation
status of knowledge management initiatives in SECY-07-0138, “NRC Knowledge
Management Program Status Update,” dated August 14, 2007.

TASK FORCE EVALUATION
During the course of data collection and analysis, the Task Force searched for and reviewed
hundreds of documents, including some issued 40 or more years ago. The Task Force alsointeracted with scores of NRC staff and other individuals to understand issues and evaluate
insights from the Fukushima event.
During the course of these activities, the Task Force struggled to access electronically avariety of critical historical documents, including Commission papers and staff requirements
memoranda, GLs, bulletins, NUREGs, RGs, and other key documents that significantly
contribute to the knowledge base for technical and regulatory evaluation and decisions
made by the agency. Many documents dated before 2000 are not available electronically.
Consequently, from the perspective of integrated knowledge management, it would be equallydifficult for staff to access these foundational documents without personal paper files.
In addition, during a variety of interviews with staff from multiple offices on similar subjects,
it became clear that organizationally influenced knowledge gaps may exist, in that staff
members in one organization may not be aware of information in another organization thatis relevant to their work. In addition, staff members working on operating reactor licensing
noted that, in order to support a licensing action or decision, they often need to read,evaluate, and refer to critical documents that must be viewed on microfiche.
The NRC has made great strides in electronic document management. Because of financial
constraints, legacy documents are not all accessible electronically through ADAMS, andno strategy exists to further enhance the availability of historical documents underpinning
agency positions and decisions. The Task Force believes that the staff should strongly
consider continued retrofitting of its electronic document library in a methodical andcoherent way to incorporate all key legacy documents.
Further, the Task Force believes that the NRC would enhance its organizational and technicalcapacity by establishing an electronic framework, similar to Wikipedia and managed by staff
technical experts, that links data on various subjects so that the staff members could access
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the information easily, allowing them to develop the well-founded knowledge required toaccomplish their tasks and support their regulatory decisions.
As time passes, maintaining historical institutional knowledge becomes more dependent on
a comprehensive and accessible information management system. The Task Force therefore
encourages the staff to do the following:• Add key legacy documents to the electronic records system (ADAMS).
• Develop an information management platform to enhance organizational and technicalcapacity by providing subject matter in an electronic format managed by staff technicalexperts.

5.3

INTERNATIONAL COOPERATION AND COORDINATIONBACKGROUND
The NRC is actively involved in numerous international activities related to the Fukushima
accident, both bilaterally and multilaterally, through such organizations as IAEA, NEA, andthe Multinational Design Evaluation Program (MDEP). The regulatory authority of Japan(the Nuclear and Industrial Safety Agency with its support organization, the Japan Nuclear
Energy Safety Organization) is also a member of each of these international organizations.
Through each of the venues, the NRC has received information on the accident in Japan andon the followup activities of each nation.
In addition to its frequent contacts through the NRC Operation Center, the NRC staff has
participated in the following activities, all of which included discussions of the Fukushima
accident and followup actions planned or underway:• April 2011 NEA Steering Committee meeting• April 2011 MDEP Steering Technical Committee meeting• May 2011 MDEP EPR reactor design working group meeting• May 2011 NEA Committee on Nuclear Regulatory Activities (CNRA) planning meeting• June 2011 CNRA discussions on Fukushima• June 2011 NEA, IAEA, France G20-G8 Forum on the Fukushima Accident• June 2011 IAEA Ministerial Meeting on Fukushima
These international interactions have helped the world’s nuclear safety authorities share
information and develop plans to enhance safety following the Fukushima accident. In part
because of these interactions, the nuclear regulatory authorities around the world havedeveloped a consistent understanding of the important elements of the accident and similarroadmaps forward, each appropriate to the individual country and its regulatory approach.
The staff has also been well informed of the activities of the European regulatory authorities
in the Western European Nuclear Regulators’ Association and the European Nuclear Safety
Regulators Group through discussions within MDEP, NEA/CNRA, and IAEA.

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TASK FORCE EVALUATIONThe NRC’s longstanding commitment to international cooperation has benefitted the agency
during the monitoring and followup activities related to the Fukushima accident. It is clearthat the NRC’s active participation in international activities has served and continues toserve nuclear safety in the United States and worldwide. It is also clear that the NRC’s longterm activities, such as further information collection, event analysis, and comparisons of
the effectiveness of actions taken, would benefit from international cooperation both in termsof effectiveness and cost efficiency.
International cooperation to date has been very effective in collecting and sharing
information and in developing and confirming an understanding of the important elementsof the Fukushima accident relevant to the activities of the Task Force. NRC participation inthese international efforts also contributes to the safety of nuclear reactors throughout theworld by providing forums for sharing the agency’s best understanding of events and issues
and effectively addressing them. These international forums also enable the NRC to provide
leadership in the resolution of important safety concerns.
In addition, participation in international efforts generally requires a minimal commitment of
resources and in some cases, such as the collection and analysis of accident data, actually
results in a resource savings.
The Task Force endorses continued international cooperation and coordination, including thefollowing:• participation in collaborative, international efforts to determine and analyze the
Fukushima accident sequence of events• participation in international efforts to update IAEA Fundamental Safety Standard (NSR-1)
and other related standards to reflect insights from the Fukushima event• continued cooperation and coordination with other national regulatory authorities on
insights from the Fukushima event as well as their plans, actions, and findings

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6.

SUMMARY OF OVERARCHING RECOMMENDATIONSThis section presents the Task Force’s recommendations for improving the safety of both
operating and new nuclear reactors. It also addresses recommended improvements in
the NRC programs for the oversight of reactor safety. The recommendations are based on
the Task Force’s evaluations of the relevant issues identified from the Fukushima accident.
Appendix A of this report proposes an implementation strategy and offers further details onthese recommendations.
The Task Force makes the following overarching recommendations, as stated in thepreceding sections of this report:

Clarifying the Regulatory Framework
1.

The Task Force recommends establishing a logical, systematic, and coherent regulatory
framework for adequate protection that appropriately balances defense-in-depth andrisk considerations. (Section 3)

Ensuring Protection
2.

The Task Force recommends that the NRC require licensees to reevaluate and upgrade
as necessary the design-basis seismic and flooding protection of SSCs for each
operating reactor. (Section 4.1.1)

3.

The Task Force recommends, as part of the longer term review, that the NRC evaluate
potential enhancements to the capability to prevent or mitigate seismically induced fires
and floods. (Section 4.1.2)

Enhancing Mitigation
4.

The Task Force recommends that the NRC strengthen SBO mitigation capability at all
operating and new reactors for design-basis and beyond-design-basis external events.
(Section 4.2.1)

5.

The Task Force recommends requiring reliable hardened vent designs in BWR facilities
with Mark I and Mark II containments. (Section 4.2.2)

6.

The Task Force recommends, as part of the longer term review, that the NRC identify
insights about hydrogen control and mitigation inside containment or in other buildings
as additional information is revealed through further study of the Fukushima Dai-ichi
accident. (Section 4.2.3)

7.

The Task Force recommends enhancing spent fuel pool makeup capability andinstrumentation for the spent fuel pool. (Section 4.2.4)

8.

The Task Force recommends strengthening and integrating onsite emergency response
capabilities such as EOPs, SAMGs, and EDMGs. (Section 4.2.5)

Strengthening Emergency Preparedness
9.

The Task Force recommends that the NRC require that facility emergency plans address
prolonged SBO and multiunit events. (Section 4.3.1)

10. The Task Force recommends, as part of the longer term review, that the NRC pursue
additional EP topics related to multiunit events and prolonged SBO. (Section 4.3.1)

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Summary of Overarching Recommendations
11. The Task Force recommends, as part of the longer term review, that the NRC should
pursue EP topics related to decisionmaking, radiation monitoring, and public education.
(Section 4.3.2)

Improving the Efficiency of NRC Programs
12. The Task Force recommends that the NRC strengthen regulatory oversight of licensee
safety performance (i.e., the ROP) by focusing more attention on defense-in-depth
requirements consistent with the recommended defense-in-depth framework.
(Section 5.1)

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Applicability and Implementation Strategy for New Reactors

7 .	APPLICABILITY AND IMPLEMENTATION STRATEGYFOR NEW REACTORSThe Task Force has considered the applicability and implementation of its recommendations
for new reactors, including certified designs, designs in the certification process, certified
designs applying for renewal, early site permits, and applications for operating licenses
and COLs. Since no new reactors are currently licensed, the recommendations involving
orders are not applicable. However, the Task Force has assessed the rulemaking
recommendations, and they would be equally applicable to new reactors.
Recommendation 8 for the integration of EOPs, SAMGs, and EDMGs and for controlling
accident decisionmaking under technical specifications would be applicable to COLs. For
near-term COLs (i.e., those expected to be licensed before the NRC completes the proposed
rulemakings), the Task Force recommends that the agency impose those requirements
through inspections, tests, analyses, and acceptance criteria (ITAAC). The Task Forcerecognizes that, with the sole exception of fire protection programs, the Commission has
expressed (in the staff requirements memorandum dated September 11, 2002, for SECY02-0067, “Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) for Operational
Programs (Programmatic ITAAC),” dated April 5, 2002) a desire to resolve programmatic
issues (e.g., training, quality assurance, fitness for duty) before COL issuance and handle
the issues through existing oversight programs rather than resolve them through the ITAAC
process. Specifically, the Commission stated (in the staff requirements memorandum
for SECY-02-0067) that, “They [ITAAC] should encompass only those matters that, by theirnature, cannot be resolved prior to construction.” Further, the Commission stated, “…theCommission is not prepared to dismiss the possibility that programmatic ITAAC may benecessary in some very limited areas.” The Task Force suggests that this would be one ofthose areas in which it is not practical to resolve the issue before COL issuance, in that the
integration of EOPs, SAMGs, and EDMGs could require a few years of effort by licensees, theindustry, and the NRC staff. However, this strategy would ensure implementation and NRC
oversight before plant operation.
The Task Force concludes that all of the current early site permits already meet therequirements of detailed recommendation 2.1, relating to the design-basis seismicand flooding analysis, and all of the current COL and design certification applicants are
addressing them adequately in the context of the updated state-of-the-art and regulatory
guidance used by the staff in its reviews.
The Task Force concludes that Recommendation 4, with new requirements for prolonged
SBO mitigation, and Recommendation 7, about spent fuel pool makeup capability andinstrumentation, should apply to all design certifications or to COL applicants if therecommended requirements are not addressed in the referenced certified design. The Task
Force recommends that design certifications and COLs under active staff review address thisrecommendation before licensing.
The Task Force notes that the two design certifications currently in the rulemaking process(i.e., the AP1000 and the economic simplified boiling-water reactor (ESBWR)) have passive
safety systems. By nature of their passive designs and inherent 72-hour coping capability for
core, containment, and spent fuel pool cooling with no operator action required, the ESBWR
and AP1000 designs have many of the design features and attributes necessary to address
the Task Force recommendations. The Task Force supports completing those design
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Applicability and Implementation Strategy for New Reactors
certification rulemaking activities without delay. However, COL applicants referencing these
designs would have to address prestaging of any needed equipment for beyond 72 hours, and
ITAAC should be established to confirm effective implementation of minimum and extended
coping, as described in detailed recommendation 4.1.
Since the Task Force recommends the SBO additions on the basis of adequate protection,
the NRC should impose them as new requirements in accordance with 10 CFR 52.59(b)(1)
as part of the staff’s review of the recently docketed Advanced Boiling-Water Reactor design
certification renewal applications.
The recommendations related to expanding 10 CFR 50.54(hh) and the EP requirements tofully address multiunit accidents and SBO conditions should apply to COL applicants. Nearterm COLs could implement these recommendations through ITAAC.
For the two plants with reactivated construction permits (Watts Bar Unit 2 and Bellefonte
Unit 1), the Task Force recommends that those operating license reviews and the licensing
itself include all of the near-term actions and any of the recommended rule changes thathave been completed at the time of licensing. Any additional rule changes would be imposedon the plants in the same manner as for other operating reactors.

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Appendix A

APPENDIX A
Summary of Detailed Recommendations by Implementation Strategy
In this appendix, the Task Force proposes an integrated strategy for implementing its
recommendations in a coherent manner. The structure of this process includes (1) apolicy statement, (2) rulemaking activities, (3) orders, (4) staff actions, and (5) actions forlong-term evaluation.
As described in Section 3 of the report, in light of the low likelihood of an event beyond
the design basis of a U.S. nuclear power plant and the current mitigation capabilities
at those facilities, the Task Force concludes that continued operation of these plants
and continued licensing activities do not pose an imminent risk to the public healthand safety. Further, the Task Force concludes that the current regulatory approach andregulatory requirements continue to serve as a basis for the reasonable assurance ofadequate protection of public health and safety until the actions set forth below havebeen implemented.

Policy Statement
Draft a Commission policy statement that articulates a risk-informed defense-indepth framework that includes extended design-basis requirements in the NRC’s
regulations as essential elements for ensuring adequate protection. (Section 3—detailed
recommendation 1.1)

Recommended Rulemaking Activities
The Task Force recommends that the Commission direct the staff to initiate these important
rulemaking activities, including concurrent development of associated guidance, andcomplete them as a soon as possible.
• Initiate rulemaking to implement a risk-informed, defense-in-depth framework consistent
with the above recommended Commission policy statement. (Section 3—detailed
recommendation 1.2)
• Initiate rulemaking to require licensees to confirm seismic hazards and flooding hazards
every 10 years and address any new and significant information. If necessary, update
the design basis for SSCs important to safety to protect against the updated hazards.
(Section 4.1.1—detailed recommendation 2.2)
• Initiate rulemaking to revise 10 CFR 50.63 to require each operating and new reactor
licensee to (1) establish a minimum coping time of 8 hours for a loss of all ac power, (2)establish the equipment, procedures, and training necessary to implement an “extended
loss of all ac” coping time of 72 hours for core and spent fuel pool cooling and for reactor
coolant system and primary containment integrity as needed, and (3) preplan andprestage offsite resources to support uninterrupted core and spent fuel pool cooling, andreactor coolant system and containment integrity as needed, including the ability to deliver
the equipment to the site in the time period allowed for extended coping, under conditions
involving significant degradation of offsite transportation infrastructure associated withsignificant natural disasters. (Section 4.2.1—detailed recommendation 4.1)
• Initiate rulemaking or licensing activities or both to require the actions related to thespent fuel pool described in detailed recommendations 7.1–7.4. (Section 4.2.5—detailed
recommendation 7.5)
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Appendix A• Initiate rulemaking to require more realistic, hands-on training and exercises on SAMGs
and EDMGs for all staff expected to implement the strategies and those licensee staff
expected to make decisions during emergencies, including emergency coordinators andemergency directors. (Section 4.2.6—detailed recommendation 8.4)
• Initiate rulemaking to require EP enhancements for multiunit events in the following
areas: personnel and staffing, dose assessment capability, training and exercises, andequipment and facilities. (Section 4.3.1—detailed recommendation 9.1)
• Initiate rulemaking to require EP enhancements for prolonged SBO in the following areas:
communications capability, ERDS capability, training and exercises, and equipment andfacilities. (Section 4.3.1—detailed recommendation 9.2)

Recommended Orders
The Task Force recommends that the Commission use orders to ensure that licensees takethe near-term actions described below. In some cases, these are interim actions to be taken
until requirements associated with future rulemakings can be implemented.
• Order licensees to reevaluate the seismic and flooding hazards at their sites against
current NRC requirements and guidance, and, if necessary, update the design basis and
SSCs important to safety to protect against the updated hazards. (Section 4.1.1—detailed
recommendation 2.1)
• Order licensees to perform seismic and flood protection walkdowns to identify andaddress plant-specific vulnerabilities and verify the adequacy of monitoring andmaintenance for protection features such as watertight barriers and seals in the interim
period until longer term actions are completed to update the design basis for external
events. (Section 4.1.1—detailed recommendation 2.3)
• Order licensees to provide reasonable protection for equipment currently provided
pursuant to 10 CFR 50.54(hh)(2) from the effects of design-basis external events and to
add equipment as needed to address multiunit events while other requirements are being
revised and implemented. (Section 4.2.1—detailed recommendation 4.2)
• Order licensees to include a reliable hardened vent in boiling-water reactor (BWR) Mark I
and Mark II containments. (Section 4.2.3—detailed recommendation 5.1)
• Order licensees to provide sufficient safety-related instrumentation, able to withstanddesign-basis natural phenomena, to monitor key spent fuel pool parameters (i.e., waterlevel, temperature, and area radiation levels) from the control room. (Section 4.2.5—
detailed recommendation 7.1)
• Order licensees to provide safety-related ac electrical power for the spent fuel poolmakeup system. (Section 4.2.5—detailed recommendation 7.2)
• Order licensees to revise their technical specifications to address requirements tohave one train of onsite emergency electrical power operable for spent fuel poolmakeup and spent fuel pool instrumentation when there is irradiated fuel in the spent
fuel pool, regardless of the operational mode of the reactor. (Section 4.2.5—detailed
recommendation 7.3)
• Order licensees to have an installed, seismically qualified means to spray water into thespent fuel pools, including an easily accessible connection to supply the water (e.g., using
a portable pump or pumper truck) at grade outside the building. (Section 4.2.5—detailed
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Appendix A
recommendation 7.4)
• Order licensees to modify the EOP technical guidelines (required by Supplement 1,“Requirements for Emergency Response Capability,” to NUREG-0737, issued
January 1983 (GL 82-33), to (1) include EOPs, SAMGs, and EDMGs in an integrated
manner, (2) specify clear command and control strategies for their implementation, and(3) stipulate appropriate qualification and training for those who make decisions during
emergencies. (Section 4.2.6—detailed recommendation 8.1)
• Order licensees to modify each plant’s technical specifications to conform with detailed
recommendation 8.2. (Section 4.2.6—detailed recommendation 8.3)
• Order licensees to do the following until rulemaking is complete: determine andimplement the required staff to fill all necessary positions for responding to a multiunit
event, conduct periodic training and exercises for multiunit and prolonged SBO scenarios,
ensure that EP equipment and facilities are sufficient for dealing with multiunit andprolonged SBO scenarios, provide a means to power communications equipment needed
to communicate onsite and offsite during a prolonged SBO, and maintain ERDS capability
throughout the accident. (Section 4.3.1—detailed recommendation 9.3)
• Order licensees to complete the ERDS modernization initiative by June 2012 to ensure
multiunit site monitoring capability. (Section 4.3.1—detailed recommendation 9.4)

Recommended Staff Actions
The Task Force recommends that the staff begin the actions given below.
• Modify the Regulatory Analysis Guidelines to more effectively implement the defensein-depth philosophy in balance with the current emphasis on risk-based guidelines.
(Section 3—detailed recommendation 1.3)
• Evaluate the insights from the IPE and IPEEE efforts as summarized in NUREG1560,“Individual Plant Examination Program: Perspectives on Reactor Safety and Plant
Performance,” issued December 1997, and NUREG-1742, “Perspectives Gained fromthe Individual Plant Examination of External Events (IPEEE) Program,” issued April
2002, to identify potential generic regulations or plant-specific regulatory requirements.
(Section 3—detailed recommendation 1.4)
• Modify Section 5.0, “Administrative Controls,” of the Standard Technical Specifications foreach operating reactor design to reference the approved EOP technical guidelines for thatplant design. (Section 4.2.6—detailed recommendation 8.2)
• Expand the scope of the annual ROP self assessment and biennial ROP realignment
to more fully include defense-in-depth considerations. (Section 5.1—detailed
recommendation 12.1)• Enhance NRC staff training on severe accidents, including training resident inspectors on
SAMGs. (Section 5.1—detailed recommendation 12.2)

Recommended Actions for Long-Term Evaluation
The Task Force recommends that the staff pursue the longer term review activities
described below to further evaluate insights from the Fukushima event and to enhance thesafety of U.S. plants.
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Appendix A
induced fires and floods. (Section 4.1.2—detailed recommendation 3)• Reevaluate the need for hardened vents for other containment designs, considering theinsights from the Fukushima accident. Depending on the outcome of the reevaluation,
appropriate regulatory action should be taken for any containment designs requiring
hardened vents. (Section 4.1.3—detailed recommendation 5.2)
• Identify insights about hydrogen control and mitigation inside containment or in otherbuildings as additional information is revealed through further study of the Fukushima
Dai-ichi event. (Section 4.1.4—detailed recommendation 6)• Analyze current protective equipment requirements for emergency responders andguidance based upon insights from the accident at Fukushima. (Section 4.3.1—detailed
recommendation 10.1)• Evaluate the command and control structure and the qualifications of decisionmakers
to ensure that the proper level of authority and oversight exists in the correct
facility for a long-term SBO or multiunit accident or both. (Section 4.3.1—detailed
recommendation 10.2)• Evaluate ERDS to do the following: determine an alternate method (e.g., via satellite)
to transmit ERDS data that does not rely on hardwired infrastructure that could be
unavailable during a severe natural disaster, determine whether the data set currently
being received from each site is sufficient for modern assessment needs, and determine
whether ERDS should be required to transmit continuously so that no operator action isneeded during an emergency. (Section 4.3.1—detailed recommendation 10.3)• Study whether enhanced onsite emergency response resources are necessary tosupport the effective implementation of the licensees’ emergency plans, including theability to deliver the equipment to the site under conditions involving significant natural
events where degradation of offsite infrastructure or competing priorities for response
resources could delay or prevent the arrival of offsite aid. (Section 4.3.2—detailed
recommendation 11.1)• Work with FEMA, States, and other external stakeholders to evaluate insights fromthe implementation of EP at Fukushima to identify potential enhancements to the
U.S. decisionmaking framework, including the concepts of recovery and reentry.
(Section 4.3.2—detailed recommendation 11.2)• Study the efficacy of real-time radiation monitoring onsite and within the EPZs
(including consideration of ac independence and real-time availability on the Internet).
(Section 4.3.2—detailed recommendation 11.3)• Conduct training, in coordination with the appropriate Federal partners, on radiation,
radiation safety, and the appropriate use of KI in the local community around each nuclear
power plant. (Section 4.3.2—detailed recommendation 11.4)

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Appendix B

APPENDIX B
Tasking Memorandum

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Appendix C

APPENDIX C
Memo Transmitting Charter for the Nuclear Regulatory Commission Task
Force to Conduct a Near-Term Evaluation of the Need for Agency Actions
Following the Events Iin Japan
March 30, 2011

MEMORANDUM TO:

Martin J. Virgilio
Deputy Executive Director
for Reactor and Preparedness Programs
Executive Director for Operations
Charles L. Miller, Director
Office of Federal and State Materials
and Environmental Management Programs

FROM:

R. W. Borchardt /RA/
Executive Director for Operations

SUBJECT:

AGENCY TASK FORCE TO CONDUCT NEAR-TERM
EVALUATION OF THE NEED FOR AGENCY ACTIONSFOLLOWING THE EVENTS IN JAPANOn March 11th, 2011, Japan experienced a severe earthquake resulting in the shutdown ofmultiple reactors. It appears that the reactors’ response to the earthquake went according todesign. At the Fukushima Daiichi site, the earthquake caused the loss of normal AC power. In
addition, it appears that the ensuing tsunami caused the loss of emergency AC power at the
Fukushima Daiichi site. Subsequent events caused damage to fuel and radiological releases
offsite.
The purpose of this memorandum is to task the Deputy Executive Director for Reactor andPreparedness Programs (DEDR) to convene an agency task force of U.S. Nuclear Regulatory(NRC) senior leaders and experts. The task force should conduct a methodical and systematic
review of relevant NRC regulatory requirements, programs, and processes, and theirimplementation, to recommend whether the agency should make near-term improvements toour regulatory system. The task force should also identify a framework and topics for review
and assessment for the longer-term effort.
Attached is a charter for the task force. The charter defines the objective, scope, coordinationand communication, expected products, schedule, staffing, and Executive Director forOperations interface. The task force should update the Commission on the near-term review atapproximately 30 and 60 days, and provide its observations, findings, and recommendations in
the form of a written report and briefing at the completion of the near-term effort occurring at
approximately 90 days.
The review should be conducted in accordance with Tasking Memorandum –COMGBJ-11-0002, “NRC Actions Following the Events in Japan.”
Enclosure: As stated
CONTACT: Nathan T. Sanfilippo, OEDO
301-415-3951

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Appendix C

CHARTER FOR THE NUCLEAR REGULATORY COMMISSION TASK FORCETO CONDUCT A NEAR-TERM EVALUATION OF THE NEED FOR AGENCY ACTIONSFOLLOWING THE EVENTS IN JAPANObjective
The objective of this task force is to conduct a methodical and systematic review of relevant
NRC regulatory requirements, programs, and processes, and their implementation, torecommend whether the agency should make near-term improvements to our regulatory
system. This task force will also identify a framework and topics for review and assessment for
the longer-term effort.
Scope
The task force review will include the following:a. A near-term review to:••

Evaluate currently available technical and operational information from the events
that have occurred at the Fukushima Daiichi nuclear complex in Japan to identify
potential or preliminary near-term/immediate operational or regulatory actions
affecting domestic reactors of all designs, including their spent fuel pools. The task
force will evaluate, at a minimum, the following technical issues and determine
priority for further examination and potential agency action:• External event issues (e.g. seismic, flooding, fires, severe weather)• Station blackout• Severe accident measures (e.g., combustible gas control, emergency
operating procedures, severe accident management guidelines)• 10 CFR 50.54 (hh)(2) which states, “Each licensee shall develop andimplement guidance and strategies intended to maintain or restore corecooling, containment, and spent fuel pool cooling capabilities under thecircumstances associated with loss of large areas of the plant due toexplosions or fire, to include strategies in the following areas: (i) Fire fighting;(ii) Operations to mitigate fuel damage; and (iii) Actions to minimize
radiological release.” Also known as B.5.b.
• Emergency preparedness (e.g., emergency communications, radiologicalprotection, emergency planning zones, dose projections and modeling,protective actions)
Develop recommendations, as appropriate, for potential changes to NRC’s regulatory
requirements, programs, and processes, and recommend whether genericcommunications, orders, or other regulatory actions are needed.

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Appendix C

b. Recommendations for the content, structure, and estimated resource impact for
the longer-term review.
Coordination and Communications
The near-term task force will:
•
•
•
•
•
•
•
•

Solicit stakeholder input as appropriate, but remain independent of industry
efforts.
Coordinate and cooperate where applicable with other domestic andinternational efforts reviewing the events in Japan for additional insights.
Provide recommendations to the Commission for any immediate policy issues
identified prior to completion of the near-term review.
Provide recommendations to program offices for any immediate actions not
involving policy issues, prior to completion of the near-term review.
Identify resource implications of near-term actions.
Consider information gained from Temporary Instruction 2515/183, “Followup
to the Fukushima Daiichi Nuclear Station Fuel Damage Events.”
Develop a communications plan.
Update and brief internal stakeholders, as appropriate.

Expected Product and Schedule
The task force will provide its observations, conclusions, and recommendations in theform of a written report to the Deputy Executive Director for Reactor and Preparedness
Programs at the completion of the 90-day near-term review.
During the development of its report, the task force will brief the Commission on thestatus of the review at approximately the 30- and 60-day points.
The report will be transmitted to the Commission via a SECY paper, and the task force
will brief the Commission on the results of the near-term effort at approximately the 90day point. The report will be released to the public via normal Commission processes.
The task force will recommend a framework for a longer-term review as a part of thenear-term report. The longer-term review will begin as soon as the NRC has sufficienttechnical information from the events in Japan (with a goal of beginning by the end of thenear-term review).

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Appendix C

Staffing
The task force will consist of the following members:
LeaderSenior Managers
Senior Staff
Administrative AssistantCharles Miller
Daniel Dorman
Jack Grobe
Gary Holahan
Amy Cubbage
Nathan Sanfilippo
Cynthia Davidson

FSME
NMSS
NRR
NRO
NRO
OEDO
OGC

Additional task force members will be added as needed. For the near-term review, otherstaff members may be consulted on a part-time basis.
EDO Interface
The task force will keep agency leadership informed on the status of the effort andprovide early identification of significant findings. The task force will report to Martin J.
Virgilio, Deputy Executive Director for Reactor and Preparedness Programs.

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