The United States of America Seventh National Report for the Convention on Nuclear Safety (2016)

renewal. However, alternate processes, such as the NRC license renewal process, are

considered equally adequate and acceptable.

This section explains how the U.S. regulatory approach provides a continuum of assessment

and review that ensures public health and safety throughout the period of plant operation. Plant

safety is maintained, and aspects are improved, by a combination of the ongoing NRC

regulatory process, oversight of the current licensing basis, backfitting, broad-based

evaluations, license renewal, and licensee initiatives.

14.1.5.1 The NRC’s Robust and Ongoing Regulatory Process and the Current Licensing Basis

Before issuing an operating license, the NRC determines that the design, construction, and

proposed operation of the nuclear power plant satisfy the NRC’s requirements and reasonably

ensure the adequate protection of public health and safety. However, the licensing basis of a

plant does not remain fixed for the 40-year term of the operating license. The licensing basis

evolves throughout the term of the operating license because of the NRC’s continuing

regulatory activities and the licensee’s activities.

The NRC carries out many regulatory activities that, when considered together, constitute a

process providing ongoing assurance that the licensing bases of nuclear power plants provide

an acceptable level of safety. This process includes inspections (both periodic regional

inspections as well as daily oversight by the resident inspectors), audits, investigations,

evaluations of operating experience, regulatory research, and regulatory actions to resolve

identified issues. The NRC’s activities may result in changes to the licensing basis for nuclear

power plants through the issuance of new or revised regulations, orders, or confirmatory action

letters. The agency also publishes the results of operating experience analysis, research, or

other appropriate analyses through generic communication documents such as bulletins, INs,

RISs, and GLs. Licensee responses to these documents may also propose changes to the

plant’s licensing basis when appropriate. In this way, the NRC’s consideration of new

information provides ongoing assurance that the licensing basis for the design and operation of

each nuclear power plant provides an acceptable level of safety. This process continues for

plants that receive a renewed license to operate beyond the original operating license.

In addition to the NRC-required changes in the licensing basis, a licensee may also voluntarily

seek changes to the current licensing basis for its facility. These changes are subject to the

NRC regulations such as those described in 10 CFR 50.54, “Conditions of Licenses,”

10 CFR 50.59, 10 CFR 50.90, and 10 CFR 50.92. These regulations ensure that

licensee-initiated changes to the licensing basis are documented and that the licensee obtains

NRC review and approval, if necessary, before implementing them. In accordance with

10 CFR 50.59(d)(2), the licensee must report to the NRC any changes or modifications it makes

to the licensing basis without prior NRC review at least every 2 years. As stated in

10 CFR 50.71(e), the periodic update ensures the final safety analysis report contains the latest

information. Region-based NRC inspectors perform a sampling inspection of those changes in

accordance with the Reactor Oversight Process to ensure that the licensee has properly

characterized the changes or modifications.

The Reactor Oversight Process is the NRC’s program to inspect, measure, and assess the

safety performance of commercial nuclear power plants and to respond to any decline in

performance. Annually, the Commission devotes a significant amount of resources to the

oversight process. For example, each plant receives 6,000 to 10,000 hours of inspection every

year. Additionally, more than 1,200 hours are spent evaluating licensing tasks at each plant.

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This level of effort gives the Commission the confidence that its oversight process produces a

level of safety comparable to that of the periodic safety review process. Section 6.3.2 of this

report provides a full description the NRC Reactor Oversight Process.

14.1.5.2 The Backfitting Process: Timely Imposition of New Requirements

Decades ago, the NRC recognized the need for a process to determine when to address

generic issues for all plants. The NRC decided to consider new requirements systematically

rather than depending on other regulatory processes to decide on plant upgrades. As a result,

the NRC developed the “backfitting” process and established the Committee to Review Generic

Requirements to review NRC-staff-proposed backfits on licensees.

The Backfitting Rule, 10 CFR 50.109, “Backfitting,” first issued in 1970 and substantially revised

in 1985, applies to both generic and plant-specific backfits for power reactors. The rule defines a

“backfit” as any modification of or addition to (1) facility systems, (2) facility structures, (3) facility

components, (4) facility designs, (5) design approvals, (6) manufacturing licenses, or (7)

procedures or organization required to design, construct, or operate a facility - any of which

may result from the imposition of a new or amended rule or regulatory staff position. Later, in

1989, the NRC extended backfitting-style protections to nuclear power plants licensed under the

new regulatory approval processes in 10 CFR Part 52. These new processes are the early site

permit, standard design certification, and combined license. The NRC also provided backfitting

protection to major fuel cycle facilities, independent spent fuel storage installations, and

gaseous diffusion plants.

In 1988, the NRC amended the Backfitting Rule to state that economic costs will not be

considered through a backfit analysis in cases of backfits imposed to ensure, define, or redefine

adequate protection of public health and safety, or common defense and security. Another

exception to the need to prepare a backfit analysis is when backfits are imposed to ensure

compliance with NRC requirements (i.e., an NRC license, regulation, or order), or conformance

with written commitments by a licensee. These backfits are referred to as adequate protection

and compliance backfits, respectively. For example, 10 CFR 50.109(a)(5) states that the

Commission shall always require the backfitting of a facility if it determines that such regulatory

action is necessary to ensure that the facility provides adequate protection to the health and

safety of the public and is in accord with the common defense and security. For backfits other

than adequate protection and compliance backfits, the NRC must determine that the proposed

backfit will substantially increase the overall protection of public health and safety or the

common defense and security and that the direct and indirect costs for the facility are justified in

view of the increased level of protection. For standard design certifications, there are additional

criteria that must be satisfied to permit backfitting.

Backfitting is permitted only after a formal, systematic review to ensure that changes are

properly justified and suitably defined. The requirements of this process are intended to ensure

order, discipline, and predictability and to optimize the use of NRC staff and licensee resources.

The controls on generic backfitting include a Committee to Review Generic Requirements

review, which is a committee of senior managers from different NRC offices. Established in

1981, this committee operates under a charter that specifically identifies the documents to be

reviewed and the analyses, justifications, and findings to be supplied to this committee by the

NRC staff. Its objectives include eliminating unnecessary burdens on licensees, reducing

radiation exposure to workers while implementing requirements, and optimizing use of NRC and

licensee resources to ensure safe operation. Therefore, the Committee to Review Generic

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Requirements’ charter is a key implementing procedure for generic backfitting, although the

primary responsibility for proper backfit considerations belongs to the initiating organization.

14.1.5.3 The NRC’s Extensive Experience with Broad-Based Evaluations

In the mid-1970s, the NRC recognized the importance of assessing the adequacy of the design

and operation of currently licensed nuclear power plants, and understanding the safety

significance of deviations from applicable current safety standards that may have been

approved after those plants were licensed. It also recognized the importance of providing the

capability to make integrated and balanced decisions about the need for backfit modifications at

those plants.

Consequently, in 1977, the NRC initiated the Systematic Evaluation Program (SEP). From a list

of approximately 800 potential issues and topics related to nuclear safety, the SEP found that

the regulatory requirements for 137 issues had changed sufficiently to warrant evaluation. The

staff compared the designs of 10 of the older plants to the licensing criteria delineated in the

then-recently issued Standard Review Plan.¹⁵ After further review, the staff determined that

27 issues required some corrective action at one or more plants and that resolution of those

issues could lead to safety improvements at other operating plants built at about the same time.

These 27 issues became known as the “27 SEP lessons learned.”

In 1984, NRC staff presented the 27 SEP lessons learned to the Commission as part of a

proposal for an integrated safety assessment program (ISAP). The staff developed this program

to review safety issues for a specific plant in an integrated manner instead of continuing the

SEP at other older operating reactors. In “Commission Policy Statement on the Systematic

Safety Evaluation of Operating Nuclear Power Reactors,” dated November 15, 1984, the

Commission said that issues relating to the safety of operating nuclear power plants can be

more effectively and efficiently implemented in an integrated, plant-specific review. For the first

time, the Commission discussed probabilistic safety analysis as a method to obtain consistent

and comparable results that could be used to enhance a safety assessment. The SEP process

was transformed into the ISAP pilot program.

In May 1985, the NRC initiated the ISAP pilot at two plants, Millstone, Unit 1 and Haddam Neck

(Connecticut Yankee). The ISAP pilot identified some benefits; however, the Commission

deferred extending it beyond the pilot phase until the staff gave an integrated package of

options that clarified the relationship between the proposed follow on program to the ISAP pilot

(ISAP II) and the newly proposed individual plant examination process.

The Commission determined that because ISAP II would be voluntary and the individual plant

examination program, through the NRC’s GL process, would require a licensee response, the

staff should give priority to the individual plant examination program. Many of the same benefits

that might have been derived through the proposed ISAP II were derived instead through the

individual plant examination process (e.g., probabilistic safety analysis).

15 Standard review plans help ensure the quality and uniformity of staff reviews and provide a well-defined base

from which to evaluate a licensee or applicant submittal. Standard review plans are also intended to make

information about regulatory matters widely available, to enhance communication with interested members of the

public and the nuclear power industry, and to improve the understanding of the staff review process.

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In the late 1980s and throughout the 1990s, the NRC continued to strengthen its regulatory

infrastructure and ensure the continued safe operation of commercial nuclear power plants

through inspection, broad-based assessment, and, where appropriate, establishment of new

generic requirements. For example, the Commission determined that licensees should assess

the accessibility and adequacy of their design-basis information and determine whether their

plants needed a design-basis reconstitution program. The Commission expressed its

expectations in “Availability and Adequacy of Design Bases Information at Nuclear Power

Plants; Policy Statement” in the Federal Register on August 10, 1992. The Commission also

expanded the individual plant examination program to consider external events and, recognizing

the relationship between maintenance, equipment reliability, plant risk, and safety, in 1991 the

Commission issued the Maintenance Rule as codified in 10 CFR 50.65.

The Maintenance Rule requires licensees to monitor the performance or condition of SSCs

against licensee-established goals continuously, to give reasonable assurance that these SSCs

are capable of fulfilling their intended functions. The NRC verifies the licensee’s implementation

of the Maintenance Rule through the Reactor Oversight Process, periodic regional inspections,

and daily oversight by the resident inspectors.

As late as 1991, some plants had not definitively resolved the 27 SEP lessons learned. As the

staff considered a process to renew the operating licenses for the operating nuclear power

plants, it assessed the best way to address these 27 issues.

Of the 27 issues, 4 had been completely resolved for all plants. One other issue was of such low

safety significance that it required no additional action. The staff determined that none of the

remaining 22 issues required immediate action to protect public health and safety. The staff

placed these 22 issues into the established regulatory process for determining the safety

significance of generic issues. The Generic Issues Program is discussed in Section 6.3.6 of this

report.

14.1.5.4 License Renewal Confirms Safety of Plants

In developing the License Renewal Rule, the Commission concluded that issues material to the

renewal of a nuclear power plant operating license are limited to those issues that the

Commission determines are uniquely relevant to protecting public health and safety and

preserving the common defense and security during the period of extended operation. Other

issues would, by definition, be relevant to the safety and security of the public during current

plant operation. Given the Commission’s ongoing obligation to oversee the safety and security

of operating reactors, the existing regulatory process within the present 40-year license term

addresses issues related to current plant operation rather than deferring the issues until the time

of license renewal. The NRC manages these issues by implementing the Reactor Oversight

Process, generic communications, and the Generic Safety Issues Program.

The NRC issued the License Renewal Rule in 1995 (in 10 CFR Part 54). The license renewal

process focuses on passive and long-lived SSCs because degradation in active components is

more readily detected by complying with the Maintenance Rule. License renewal applicants are

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required to complete an environmental assessment, an integrated plant assessment¹⁶ and to

evaluate time-limited aging analyses. The current licensing basis must be maintained

throughout the period of extended operation. Section 14.1.4 of this report describes the NRC

license renewal process.

14.1.5.5 Risk-Informed Regulation and the Reactor Oversight Process

The NRC is actively increasing the use of risk insights and information in its regulatory

decisionmaking. A risk-informed approach to regulatory decisionmaking considers risk insights

together with other factors to establish requirements that focus licensee and regulatory attention

on design and operational issues commensurate with their importance to health and safety. For

reactors, risk-informed activities occur in the five broad categories of (1) applicable regulations,

(2) licensing process, (3) Reactor Oversight Process, (4) regulatory guidance, and (5) risk

analysis tools, methods, and data. Activities within these categories include revisions to

technical requirements in the regulations; risk-informed technical specifications; a framework for

inspection, assessment, and enforcement actions; guidance on risk-informed inservice

inspections; and improved standardized plant analysis risk models.

In 2000, the NRC implemented a revised Reactor Oversight Process using risk insights and

lessons learned from more than 30 years of regulating nuclear power plants. The previous

oversight process evolved during a period when the nuclear power industry was less mature

and there was much less operational experience on which to base rules and regulations.

Therefore, very conservative judgments governed the rules and regulations. Significant plant

operating events occurred with some frequency, and the oversight process tended to be

reactive and prescriptive, closely observing plant performance for adherence to the regulations

and responding to operational problems as they occurred.

After nearly 4 decades of operational experience and generally steady improvements in plant

performance, the Reactor Oversight Process now focuses more of the agency’s resources on

the most significant issues and the relatively small number of plants with performance problems.

The process is a way to collect information about licensee performance, assess the information

for its safety significance, and provide for appropriate licensee and NRC response, including

corrective and enforcement actions, when appropriate.

The Reactor Oversight Process is a risk-informed tool that uses direct inspections and objective

performance indicators reported by the licensee to measure and assess plant performance.

Together, the performance indicators and inspection findings give the information needed to

support continuous reviews and assessments of plant performance. The Reactor Oversight

Process also features comprehensive quarterly reviews and expanded annual reviews, which

include inspection planning and a performance report (all posted on the NRC’s public Web site).

The Reactor Oversight Process is more effective at correcting performance or equipment

problems today because the agency’s response to problems is more focused and predictable.

Section 6.3.2 of this report provides a full description of the NRC Reactor Oversight Process.

16 An integrated plant assessment identifies and lists structures and components subject to an aging management

review. These include “passive” structures and components that perform their intended function without moving

parts or without a change in configuration or properties. Examples of these are the reactor vessel, the steam

generators, piping, component supports, and seismic Category I structures. To be in scope, the item must also

be long-lived to be considered during the license renewal process. Long-lived means the item is not subject to

replacement based on a qualified life or specified time period.

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14.1.5.6 Licensee Responsibilities for Safety: Regulations and Initiatives beyond Regulations

As in many countries, U.S. nuclear power plant licensees are responsible for the safety of their

facilities. This responsibility is embedded in their license and in the NRC’s regulatory

infrastructure. Under the regulatory umbrella, licensees routinely assess new technologies,

offnormal conditions, operating experience, and industry trends to make informed decisions

about safety enhancements to their facilities.

Under the U.S. regulatory structure, Appendix B to 10 CFR Part 50 requires nuclear power plant

licensees to maintain a quality assurance program. Quality assurance comprises all those

planned and systematic actions necessary to provide adequate confidence that an SSC will

perform satisfactorily in service. Quality assurance includes quality control, which comprises

those quality assurance actions related to the physical characteristics of a material, structure,

component, or system that provide a means to control their quality to predetermined

requirements.

Licensees carry out a comprehensive system of planned and periodic audits to verify

compliance with all aspects of the quality assurance program and to determine the effectiveness

of the program. The audits are performed in accordance with written procedures or checklists by

appropriately trained personnel not having direct responsibilities in the areas being audited.

Management reviews the audit results and appropriate followup action is initiated.

14.1.5.7 The NRC’s Regulatory Process Compared with International Safety Reviews

The IAEA and the Western European Nuclear Regulators’ Association (WENRA) have

developed guidance¹⁷ and objectives for conducting periodic safety reviews that have much in

common. Consistent with the IAEA guidance, periodic safety reviews are comprehensive

assessments to determine:

the adequacy and effectiveness of the arrangements and the SSCs (equipment) that are

in place to ensure plant safety until the next periodic safety review or, where appropriate,

until the end of planned operation (that is, if the nuclear power plant will cease operation

before the next periodic safety review is due)

the extent to which the plant conforms to current national and/or international safety

standards and operating practices

safety improvements and timescales for their implementation

the extent to which the safety documentation, including the licensing basis, remains valid

17 IAEA guidance appears in Specific Safety Guide SSG-25, “Periodic Safety Review of Nuclear Power Plants

Safety,” issued in 2013. WENRA has published several guidance documents on this subject. One of them is,

“Position Paper on Periodic Safety Reviews (PSR) Taking into Account the Lessons Learnt from the TEPCO

Fukushima Dai-ichi NPP Accident,” WENRA Reactor Harmonization Working Group, dated March 2013.

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After the 2010 International Regulatory Review Service (IRRS) mission in the United States, the

NRC undertook a limited-scope pilot effort and a supplemental evaluation to review a sample of

periodic safety review summary reports from other regulators to identify areas that could

potentially inform the NRC’s regulatory processes. The NRC issued a report titled, “Findings

from the Staff’s Evaluation of Periodic Safety Reviews from Other Countries,” dated

April 24, 2015. Based on the pilot effort and the supplemental evaluation, the NRC staff

concluded that it is reasonable to expect that the U.S. regulatory approach would be sufficient

for detecting and correcting the plant-specific issues documented in the periodic safety review

summary reports, if they were to occur in U.S. plants. Hence, changes to the existing regulatory

processes were deemed unnecessary. Discussions of the findings from other countries’ periodic

safety reviews present a valuable opportunity for NRC to stay apprised of international

experiences in assessing reactor safety. The NRC welcomes such discussions during bilateral

and multilateral exchanges as appropriate. Detailed information on the IRRS mission report can

be found in Section 8.1.5.2 of this report.

For the reasons discussed above and summarized below, the shared objectives associated with

the IAEA and WENRA periodic safety review guidance are substantively accomplished in the

United States on an ongoing basis.

First, the NRC’s regulatory process provides a robust foundation for ongoing assessments,

evaluations, and, when appropriate, imposition of new requirements. Currently, the NRC and

the U.S. nuclear industry consider new information in a more risk-informed manner as it

becomes available; adjust the regulatory oversight and plant safety priority, respectively; and

provide ongoing assurance that the licensing basis for the design and operation of all nuclear

power plants provides an acceptable level of safety. Development of the Maintenance Rule and

License Renewal Rule are two examples of new requirements that serve this purpose. In

addition, the NRC has instituted a series of security enhancements for nuclear power plants

since the September 11, 2001, terrorist attacks. These enhancements include upgraded

physical security plans, enhanced security officer training, increased security patrols, additional

physical barriers, greater stand-off distances for vehicle checks, and more restrictive site access

controls, among others. Sections 10.5, 16.6.4 and 18.3.2.3 of this report provide additional

discussions on the NRC’s response actions after September 11, 2001, in the areas of safety

and security interface, and cyber security. Separately, the NRC has undertaken significant

actions to enhance the safety of nuclear power reactors in the U.S. following the Fukushima

accident in Japan on March 11, 2011. The Fukushima-related actions include, but are not

limited to, required implementation of mitigation strategies to respond to beyond-design-basis

events; ensuring severe accident capable hardened containment vents for boiling water reactors

with Mark I and II containments; reevaluation of seismic and flooding hazards using present-day

guidance, methods, and information; and enhancing spent fuel pool instrumentation.

Sections 1.3.1, 1.3.3, 6.4, 14.3, and 16.9 of this report provide additional details on the

Fukushima-related actions.

Second, the NRC and the U.S. nuclear industry have more than 30 years of experience

implementing broad-based plant assessments. The regulatory history of implementing

broad-based assessments is a direct result of an adaptive, probing, and independent regulatory

process. These assessments have included the SEP, the ISAP, and the individual plant

examinations. They provide additional confidence that plant safety continues to be the highest

priority and that the NRC and industry continue to pursue enhancements that improve safety. As

shown in the figure included below, over a period of almost 25 years, broad-based NRC

assessments and regulatory initiatives have provided a continuum of assessment, improvement,

and oversight, which ensures that licensed plants continue to operate safely.

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The NRC’s transition to a more risk-informed regulatory framework and the Reactor Oversight

Process offers an ongoing approach and basis for implementing appropriate safety

improvements, corrective actions, or process improvements, and provides confidence that the

plant can continue to be operated safely. The NRC’s more risk-informed approach helps ensure

that resources are optimally focused on those issues most important to safety.

Finally, U.S. licensees establish performance expectations above the thresholds required by the

NRC. These self-imposed expectations and initiatives - over and above the regulations -

result from the licensee’s self-described motivation to pursue excellence and by the recognition

that safety and economics are directly linked in the competitive, free-market U.S. energy

industry.

14.2 Verification by Analysis, Surveillance, Testing, and Inspection

Licensees are required to verify that they are operating their nuclear installations in accordance

with the plant-specific design and requirements. The technical specifications and national

consensus codes (for testing and periodic inspections) contain some of the requirements for

verification.

In 10 CFR 50.55a, “Codes and Standards,” the NRC enumerates requirements for applying

industry codes and standards to nuclear power reactors during design, construction, and

operation. For example, this section incorporates by reference Section III and Section XI of the

ASME Operation and Maintenance Code for nuclear power plants.

Through analysis, surveillance, testing, and inspection, the licensees verify that the physical

state and operation of nuclear installations continue to be in accordance with the designs,

applicable national safety requirements, and operational limits and conditions. As discussed in

Article 6 of this report, the NRC’s Reactor Oversight Process includes inspections to verify that

licensees are fulfilling their obligations to carry out such surveillances, testing, and inspections

and take corrective action.

Under special circumstances, the Commission may also require under 10 CFR 50.54(f) that

licensees submit written statements to the Commission. If necessary to ensure safe operation,

the Commission can determine whether the license should be modified, suspended, or revoked.

The NRC updates, revises, and improves existing regulatory programs in light of operating

experience and significant new safety information. Article 19 of this report discusses these

activities.

14.3 Fukushima Lessons Learned

The events at Fukushima Dai-ichi highlighted the need for safety improvements for nuclear

power plants related to beyond-design-basis natural hazards. As described in Section 1.3.1 and

1.3.3 of this report, the NRC issued orders and a request for information to its licensees in

response to the accident, and initiated a significant rulemaking to further enhance safety. The

NRC also became involved in a host of international activities related to safety assessment.

Immediately after the event, using the existing Reactor Oversight Process, the NRC conducted

inspections and issued orders, INs, and bulletins to aid in determining the preparedness of U.S.

nuclear power plants to withstand a similar event. Furthermore, the Reactor Oversight Process

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will be used to assess and verify that changes currently being implemented in response to

lessons learned from the accident were completed properly.

14.4 Vienna Declaration on Nuclear Safety

The NRC carries out many regulatory activities that, when considered together, provide for a

comprehensive and systematic assessment and review to ensure public health and safety. One

of the agency’s main programs is the Reactor Oversight Process, which includes the use of

baseline, periodic, special inspections, and daily oversight. Throughout the program, the NRC

inspects, monitors, and assesses safety performance, and solicits feedback. Sections 14.1.1

and 6.3.2 of this report provide more information on the use of the Reactor Oversight Process.

Section 14.1.5 of this report includes more information on how the NRC activities, including

inspections, audits, research, backfitting, license renewal, and probabilistic risk assessment, are

used to ensure that nuclear power plant safety is being continually assessed and assured over

the entire period of operation.

The NRC also recognizes that the effective use of lessons learned from domestic and

international operating experience is important for protecting the health and safety of people and

the environment. The NRC screens operating experience for safety significance and generic

implications, including the need for further action, as delineated in Section 6.3.5 of this report.

The NRC communicates information internally to ensure that technical staff is able to factor

operating experience into their reviews of plant safety. The NRC staff maintains

communications with INPO to ensure that relevant operating experience reviewed by the

industry is also considered by NRC reviews. The NRC communicates in public forums through

the issuance of generic communications to provide NRC operating experience insights to the

industry, the public, and the international community. In addition, the staff can make revisions to

IPs when operating experience reviews indicate potential areas of concern for safety that may

be reviewed through the inspection program. Section 19.7 of this report provides more

information about the operating experience program.

Of significance, in light of the Fukushima accident, the NRC has taken a large number of actions

to strengthen the protection of U.S. nuclear plants against events that could exceed a plant’s

design basis. For example, the NRC issued regulatory requirements in the form of three orders.

These orders are discussed in Sections 1.3.1 and 1.3.3 of this report. The U.S. nuclear industry

has taken a number of actions in response to the accident as well. The U.S. industry actions are

summarized in Part 3.

To a large extent, the international community conducts periodic safety reviews to assess

operating experience, technical developments, and other aspects such as the cumulative effects

of plant aging. In contrast, the NRC uses routine and ongoing comprehensive safety inspection,

audit, and assessment programs that deal with specific safety issues, significant events, and

changes in safety standards and practices as they arise. These programs, as applied by the

NRC with the appropriate scope, frequency, depth, and rigor, achieve the same review

standards and objectives as a periodic safety review. This was demonstrated by the NRC’s

response to Fukushima, which reflects the NRC’s regulatory approach of promptly addressing

new information when discovered and taking appropriate regulatory action in a timely fashion,

rather than awaiting a periodic review.

The effectiveness of the NRC’s regulatory approach was evaluated by the IAEA IRRS mission

and followup mission conducted in the NRC in 2010 and 2014, respectively. During the 2010

IRRS mission, the NRC correlated its regulatory programs to the 14 periodic safety review

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ARTICLE 15. RADIATION PROTECTION

Each Contracting Party shall take the appropriate steps to ensure that, in all operational

states, the radiation exposure to the workers and to the public caused by a nuclear

installation shall be kept as low as reasonably achievable, and that no individual shall be

exposed to radiation doses which exceed the prescribed national dose limits.

This section summarizes the authorities and principles of radiation protection, which include the

regulatory framework, regulations, and radiation protection programs for controlling radiation

exposure for occupational workers and members of the public. This section also discusses

lessons learned from Fukushima. Article 17 of this report discusses radiological assessments

that apply to licensing and facility changes.

15.1 Authorities and Principles

Generally, United States radiation control measures are founded on radiological risk

assessments by the United Nations Scientific Committee on the Effects of Atomic Radiation and

the United States National Academy of Sciences Committee on the Biological Effects of Ionizing

Radiation. The risk management recommendations that the International Commission on

Radiological Protection (ICRP) and the National Council on Radiation Protection and

Measurements issued reflect these assessments. On the basis of these assessments and

recommendations, the EPA develops Federal guidance signed by the President of the United

States and “generally applicable radiation standards” for use by the other Federal agencies,

including the U.S. NRC. The responsible agencies, such as the NRC, then establish regulations

that consider these recommendations and standards. U.S. radiation protection programs are

based on principles generally consistent with the principles espoused by ICRP:

(1) it is known

that large doses of ionizing radiation can be deleterious to human health, and (2) there is an

assumption regarding a direct and proportional relationship between radiation exposure and

cancer risk with all radiation doses (known as the Linear-Non-Threshold model). The U.S.

programs acknowledge, include, and use the ICRP-recommended protection principles of

“limitation,” “justification,” and “optimization,” as appropriate.

Of these principles, “limitation” is the most practicable and most directly included in the

regulatory structure. The regulations establish dose limits. Exceeding these dose limits may

result in the issuance of violations and other sanctions to ensure compliance and prevent

recurrence. There is a lengthy history of the doses being kept within the limits for workers

(NUREG-0713, “Occupational Radiation Exposure at Commercial Nuclear Power Reactors and

Other Facilities 2013: Forty-Sixth Annual Report,” Volume 35, issued in December 2015), and

members of the public living near nuclear power plants (NUREG/CR-2907, “Radioactive

Effluents from Nuclear Power Plants Annual Report 2009,” Volume 15, issued in August 2013.)

More recent effluent release data are available on the NRC Web site at:

http://www.nrc.gov/reactors/operating/ops-experience/tritium/plant-info.html.

“Justification” is the recommendation that any activity involving radiation exposure be shown to

be beneficial before the activity is undertaken. However, the risks or benefits of a new

application of radioactive material can seldom be determined in advance with complete

accuracy. Furthermore, radiation protection considerations are only one contributor to overall

decisions on whether a particular exposure situation is justified. The “justification” activities in

the United States are carried out during the licensing process. In general, the NRC will reject an

application to use or produce radioactive materials if it determines that the application is not

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justified (i.e., that the overall benefit to society is outweighed by the risk of the radiation

exposure associated with the activity). The licensing process under 10 CFR Part 50, “Domestic

Licensing of Production and Utilization Facilities,” does not directly address the justification for

licensing a nuclear power plant. But, when a nuclear power plant is licensed, the environmental

costs and benefits are evaluated in an environmental impact statement. This analysis is

considered in the NRC’s licensing decision.

Rather than using the term “optimization,” the United States has used the term “as low as

reasonably achievable” (ALARA). In most circumstances, these two terms are consistent and

represent the same underlying principle. As a guiding principle, ALARA (with varying

terminology) dates back to 1939, in the United States and is defined in the regulations for

occupational workers and members of the public.

For decades before 1994, 10 CFR Part 20, “Standards for Protection against Radiation,”

addressed the ALARA criterion for occupational radiation exposure, but more as a

recommendation than as a requirement. In 1994, the NRC changed the regulation to require

that all licensees develop, document, and carry out an ALARA program. The NRC judges

compliance with this requirement on the basis of a licensee’s capability to track and, if

necessary, reduce exposures, rather than on whether exposures and doses represented an

absolute minimum or whether the licensee had used all possible methods to reduce exposures.

For control of radiation exposure from nuclear power plants to members of the public, the NRC

modified 10 CFR Part 50 by adding Appendix I, “Numerical Guides for Design Objectives and

Limiting Conditions for Operation to Meet the Criterion ‘As Low as Is Reasonably Achievable’ for

Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents.” Issued in 1975,

this appendix established design objectives to keep radioactive releases from nuclear power

plants ALARA. The ALARA requirement led to the establishment of numerical objectives (for

example, 0.00005 sieverts (Sv) (0.005 rem) in a year for the most highly exposed individual).

Similar EPA requirements for other facilities soon followed. These NRC and EPA requirements

are consistent with ICRP principles and result in public doses that are well below the local

variation in doses from natural sources.

Although U.S. regulations generally are consistent with ICRP recommendations, certain

considerations have limited the extent to which U.S. regulations match those of ICRP. One

important consideration has been the U.S. desire for regulatory stability. Revising the

regulations to incorporate every new ICRP position would impose a serious burden on the

licensees without a commensurate benefit. Furthermore, for nuclear power reactors, new

requirements are constrained by the Backfitting Rule’s requirements that any increase in

regulatory requirements other than those required for compliance with existing regulations or the

statutory standard of “adequate protection” be justified by a commensurate improvement in

safety. Consequently, U.S. regulations were founded on older (rather than the most recent)

ICRP recommendations. Nevertheless, the Commission directed the staff to work closely with

ICRP and other national and international organizations to help develop revised

recommendations. After publication of the new ICRP recommendations (ICRP Publication 103,

“The 2007 Recommendations of the International Commission on Radiological Protection,”

approved March 2007), the NRC staff initiated stakeholder dialogue on key issues, and provided

options for Commission consideration in SECY-12-0064, “Recommendations for Policy and

Technical Direction to Revise Radiation Protection Regulations and Guidance,” dated

April 25, 2012. In its SRM issued in December 2012, the Commission approved the staff

continuing stakeholder dialogue and technical basis development to explore the benefits and

effects of increasing alignment with ICRP. The Commission disapproved any change to the

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occupational limit for effective dose. The Commission initially approved the staff’s development

of the regulatory basis for a revision to 10 CFR Part 20 and parallel alignment of 10 CFR Part

50, Appendix I, to reflect the most recent methodology and use consistent terminology for dose

assessment. However, in light of comments and feedback received on contemplated changes to

10 CFR Part 20, the NRC staff is no longer developing a regulatory basis for the revision to this

rule. The staff has determined that the current NRC regulatory framework continues to provide

adequate protection of the health and safety of workers, the public, and the environment.

Additional information on the options to revise radiation protection regulations and guidance can

be found in the NRC’s public Web site at:

http://www.nrc.gov/about-nrc/regulatory/rulemaking/potential-rulemaking/opt-revise.html.

15.2 Regulatory Framework

The NRC developed requirements for radiation protection to implement three laws that the U.S.

Congress passed: the Atomic Energy Act of 1954, as amended; the Energy Reorganization

Act of 1974; and the Uranium Mill Tailings Radiation Control Act of 1978.

NRC regulations establish the primary direct controls over licensees. Various documents

provide additional guidance and clarification, including RGs, staff reports (NUREG series), GLs,

technical specifications, and license conditions. These documents are supported by

international standards, consensus national standards, and authoritative recommendations

(such as those of ICRP and the National Council on Radiation Protection and Measurements).

However, international standards, consensus national standards, and authoritative

recommendations have no official status unless they are referenced in or adopted by a

regulation or documents providing regulatory guidance, such as RGs or Standard Review Plans.

Of particular importance are NUREG-0800, “Standard Review Plan for the Review of Safety

Analysis Reports for Nuclear Power Plants: LWR Edition,” which guides the staff in reviewing

safety analysis reports, and RG 1.70, “Standard Format and Content of Safety Analysis Reports

for Nuclear Power Plants,” Revision 3, issued in November 1978, which guides the applicant in

writing safety analyses. Chapter 11, “Radioactive Waste Management,” of NUREG-0800

addresses the control of radioactive effluents. Chapter 12, “Radiation Protection,” addresses

radiation protection. Chapter 15, “Transient and Accident Analysis,” details how to calculate

offsite and control room operator doses for design-basis accidents. Under 10 CFR 50.34(h), the

facility must be evaluated against the standard review plan.

As Article 6 of this report discussed, the Reactor Oversight Process has cornerstones for

radiation safety. The cornerstone for public radiation safety focuses on the effectiveness of the

plant’s programs in meeting applicable Federal limits on the exposure, or potential exposure, of

members of the public to radiation and in ensuring that the effluent releases from the plant are

ALARA. The cornerstone for occupational radiation safety focuses on the effectiveness of the

plant’s program(s) in maintaining the worker dose within the regulatory limits and providing

occupational exposures that are ALARA.

15.3 Regulations

The regulations that apply to public and occupational radiation protection from nuclear power

plant operations are 10 CFR Part 20 and 10 CFR Part 50.

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10 CFR Part 20. The NRC regulations in 10 CFR Part 20 establish requirements for radiation

protection for all NRC licensees. The NRC gives additional requirements for specific operations

and specific kinds of licenses in other parts of Title 10: Regulations in 10 CFR Part 30, “Rules

of General Applicability to Domestic Licensing of Byproduct Material”; 10 CFR Part 34,

“Licenses for Industrial Radiography and Radiation Safety Requirements for Industrial

Radiographic Operations”; 10 CFR Part 35, “Medical Use of Byproduct Material”;

10 CFR Part 39, “Licenses and Radiation Safety Requirements for Well Logging”;

10 CFR Part 40, “Domestic Licensing of Source Material”; 10 CFR Part 50; 10 CFR Part 70,

“Domestic Licensing of Special Nuclear Material”; 10 CFR Part 71, “Packaging and

Transportation of Radioactive Material”; and 10 CFR Part 72, “Licensing Requirements for the

Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and

Reactor-Related Greater than Class C Waste.”

The major revision of 10 CFR Part 20, issued in 1991, and fully implemented in 1994, adopted

the recommendations, quantities, and models recommended in ICRP Publication 26,

“Recommendations of the International Commission on Radiological Protection,” issued in

January 1977, and in ICRP Publication 30, “Limits of Intakes of Radionuclides by Workers,”

dated 1978-1982, as well as some recommendations from National Council on Radiation

Protection and Measurements Report No. 91, “Recommendations on Limits for Exposure to

Ionizing Radiation,” issued in June 1987. The 1991 revision to 10 CFR Part 20 also adopted the

same dose limit for a member of the public recommended in ICRP Publication 60,

“1990 Recommendations of the International Commission on Radiological Protection,” issued in

November 1990. The general requirements for radiation protection are provided in

10 CFR Part 20. This part is divided into subparts, with each subpart addressing a specific area

of radiation protection, such as occupational and public dose limits, posting, surveys,

monitoring, waste disposal, and reporting requirements.

The details of the requirements in 10 CFR Part 20 are not entirely consistent with international

standards such as IAEA’s General Safety Requirements Part 3 (GSR-3), “Radiation Protection

and Safety of Radiation Sources: International Basic Safety Standards - General Safety

Requirements,” issued in November 2014. The main areas of difference between

10 CFR Part 20 and the IAEA Basic Safety Standards include:

(1) the use of the effective dose

equivalent in 10 CFR Part 20 versus the use of the effective dose in the IAEA standards, (2) an

annual occupational dose limit on the effective dose equivalent of 0.05 Sv (5 rem) in

10 CFR Part 20 versus 0.02 Sv (2 rem) averaged over 5 years, with a maximum of 0.05 Sv

(5 rem) in any year, in the IAEA standards, and (3) use of the biokinetic models from ICRP

Publication 30 in 10 CFR Part 20 versus the more recent models used in the IAEA standards.

NRC licensees are permitted to use the effective dose in place of the effective dose equivalent

and to use the more recent internal dosimetry models in place of those recommended in ICRP

Publication 30, with prior NRC approval.

In addition, many licensees and agencies have administrative dose limits similar to or lower than

those in the IAEA Basic Safety Standards. Most other licensees operate at occupational doses

far below those limits and standards and therefore are considered ALARA. In some cases, the

occupational doses do exceed 0.02 Sv per year (2 rem per year), but these are a very small

fraction of the total, and efforts are continuing to reduce these doses to lower levels. The current

10 CFR Part 20 provides a level of radiation protection that in almost all situations is

comparable to that provided by international standards.

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10 CFR Part 50. Although 10 CFR Part 50 is the principal regulation addressing the safety of

nuclear power plants, only a small section of it directly addresses radiation protection. Even so,

the sections of 10 CFR Part 50 that affect radiation protection are significant. Of particular

importance are 10 CFR 50.34a, “Design Objectives for Equipment to Control Releases of

Radioactive Material in Effluents—Nuclear Power Reactors”; Appendix I, and

10 CFR 50.34(b)(3) and (h), which require NRC review of the plant radiation sources and

protection programs. In 10 CFR 50.36a, “Technical Specifications on Effluents from Nuclear

Power Reactors,” the NRC also requires licensees to limit effluents from nuclear power reactors

to the values in Appendix I to 10 CFR Part 50. The revised dose criteria, in effective dose

equivalent, appear in 10 CFR 50.34(a)(1)(ii)(D) for evaluating design basis accidents associated

with licensing actions that have been submitted to the NRC since 1997. (The pre-1997 dose

criteria for siting and determining the exclusion area low population zone and population center

distance for nuclear power reactors appear in 10 CFR 100.11(a).) In light of comments and

feedback received on the contemplated changes to 10 CFR Part 50, Appendix I, the NRC staff

is no longer developing a regulatory basis for the revision to 10 CFR Part 50, Appendix I. The

current NRC regulatory framework continues to provide adequate protection of the health and

safety of workers, the public, and the environment.

15.4 Radiation Protection Activities

Radiation protection activities apply to occupational workers and to members of the public.

15.4.1 Control of Radiation Exposure of Occupational Workers

In addition to focusing on personnel qualifications for licensing, the NRC’s oversight and

regulation of radiation protection programs ensure that the safety analysis report and radiation

protection plan properly address each item in 10 CFR Part 20, as well as the provisions for

instructions to workers in 10 CFR Part 19, “Notices, Instructions, and Reports to Workers:

Inspection and Investigations.” Guidance is provided in relevant RGs, such as RG 1.8,

“Qualification and Training of Personnel for Nuclear Power Plants,” Revision 3, issued in

May 2000, and RG 8.8, “Information Relevant to Ensuring that Occupational Radiation

Exposures at Nuclear Power Stations Will Be as Low as Is Reasonably Achievable,” Revision 3,

issued in June 1978.

The NRC maintains an active regulatory inspection program that includes routine baseline

inspections and supplemental inspections, as needed. Significant health physics problems can

trigger reactive regional inspections or a generic communication to the industry.

The NRC staff has been collecting the annual occupational exposure data for light-water

reactors since 1969. Because the amount and type of maintenance performed strongly

influence the doses, the individual plant collective doses fluctuate from year to year. Before the

nuclear plant accident in 1979 at Three Mile Island, Unit 2, the average collective dose per

reactor varied substantially. After the accident, the collective worker doses increased because

of the extensive modifications required of all nuclear power plants in response to new NRC

requirements. The average collective dose reached a peak of 7.91 person-Sv (791 person-rem)

per reactor in 1980. Since then, collective doses have declined steadily by more than a factor of

10, to the current level of 0.71 person-Sv (71 person-rem) per reactor. The average collective

dose for each BWR in 2014 was 1.09 person-Sv (109 person-rem). The average collective dose

for each PWR in 2014 was 0.51 person-Sv (51 rem). The collective dose in a BWR is

approximately a factor of 2 higher than PWRs, in part because of the larger work force at

BWRs.

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In 2014, 174,851 workers at nuclear plants were monitored for radiation exposure. Of these,

70,844 workers received a collective measureable dose of 71.24 person-Sv (7,124 person-rem)

for an average of 0.0010 Sv (0.10 rem) per worker.

15.4.2 Control of Radiation Exposure of Members of the Public

The regulations in 10 CFR 20.1301, “Dose Limits for Individual Members of the Public,” and

10 CFR 20.1302, “Compliance with Dose Limits for Individual Members of the Public,” control

radiation exposures to members of the public. In addition to the 1.0 millisievert (100 millirem)

annual dose limit in 10 CFR Part 20, the EPA regulations in 40 CFR Part 190, “Environmental

Radiation Protection Standards for Nuclear Power Operations,” establish a regulatory standard

such that the annual dose to a member of the public from exposures to sources associated with

the entire uranium fuel cycle does not exceed 0.25 millisievert (25 millirem).

The regulations in 10 CFR 20.1406, “Minimization of Contamination,” 10 CFR 50.34a,

10 CFR 50.36a, and Appendix I to 10 CFR Part 50, define the ALARA objectives for effluents

from nuclear power plants. The regulations in 10 CFR 20.1406 require the minimization of

contamination by conducting operations to minimize the introduction of residual radioactivity into

the site, including the subsurface. Appendix I to 10 CFR Part 50 and 10 CFR 50.36a also

establish regulations on effluent monitoring, environmental monitoring, investigations, land-use

censuses, and reporting. Section IV.B of Appendix I requires the licensee to establish an

appropriate surveillance and monitoring program that will accomplish the following:

Provide data on quantities of radioactive material released in liquid and gaseous

effluents.

Provide data on measurable levels of radiation and radioactive materials in the

environment to evaluate the relationship between quantities of radioactive

material released in effluents and resultant radiation doses to individuals from

principal pathways of exposure.

Identify changes in the use of unrestricted areas (e.g., for agricultural purposes)

to permit modifications in monitoring programs for evaluating doses to individuals

from principal pathways of exposure.

Appendix I requirements for ALARA are complemented by 10 CFR Part 20.1501, “General,”

which requires, in part, that a licensee perform surveys, including the subsurface, to evaluate

potential radiological hazards and to demonstrate compliance with the public dose limits in

10 CFR 20.1301 and 10 CFR 20.1302. Therefore, a licensee is responsible for performing

radiation surveys at its facility for radioactive materials that have the potential to affect workers

and members of the public.

The NRC staff continues to provide the public with current information on control of radiation

exposure to members of the public on its Web site at http://www.nrc.gov/about-

nrc/regulatory/rulemaking/potential-rulemaking/opt-revise.html. Information posted on the NRC

Web site includes the annual radiological effluent reports for each nuclear site, the annual

environmental monitoring report for each site, a radioactive effluent summary report by calendar

years, and a list of the plant sites with licensed radioactive material in ground water.

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ARTICLE 16. EMERGENCY PREPAREDNESS

(i)

Each Contracting Party shall take the appropriate steps to ensure that there are

onsite and offsite emergency plans that are routinely tested for nuclear

installations, and cover the activities to be carried out in the event of an

emergency.

(ii)

For any new nuclear installation, such plans shall be prepared and tested

before it [the installation] commences operation above a low power level

agreed [to] by the regulatory body.

(iii)

Each Contracting Party shall take appropriate steps to ensure that, insofar as they

are likely to be affected by a radiological emergency, its own population and the

competent authorities of the States in the vicinity of the nuclear installation are

provided with appropriate information for emergency planning and response.

(iv)

Contracting Parties that do not have a nuclear installation on their territory,

insofar as they are likely to be affected in the event of a radiological emergency at

a nuclear installation in the vicinity, shall take the appropriate steps for the

preparation and testing of emergency plans for their territory that cover the

activities to be carried out in the event of such an emergency.

This section discusses (1) the background of emergency planning in the United States,

(2) offsite emergency planning and preparedness, (3) emergency classification system and

emergency action levels, (4) recommendations for protective action in severe accidents,

(5) inspection practices and regulatory oversight, (6) response to an emergency,

(7) communications with neighboring states and international arrangements,

(8) communications with the public, and (9) lessons learned from the Fukushima event.

16.1 Background

The responsibilities of the U.S. NRC for radiological emergency preparedness stem from the

agency’s licensing functions under the Atomic Energy Act and the Energy Reorganization Act.

Both statutes authorize the Commission to issue regulations that it deems necessary to fulfill its

responsibilities under the acts. After the accident at Three Mile Island, Unit 2 in March 1979, the

NRC amended the regulations to require significant changes in emergency planning and

preparedness for U.S. commercial nuclear power plants.

The NRC’s emergency planning regulations are an important part of the regulatory framework

for protecting public health and safety and have been adopted as an added conservatism in the

NRC’s defense-in-depth safety philosophy of multiple-barrier containment and redundant safety

systems. Before a full-power operating license can be issued, NRC regulations require a finding

that there is reasonable assurance that adequate measures to protect public health and safety

can and will be taken in a radiological emergency (10 CFR Section 50.47(a)).

Emergency planning in the United States recognizes that a spectrum of accidents could exceed

the design-basis accidents that nuclear plants are required to accommodate without significant

public health and safety effects. For design-basis accidents, the small releases that might occur

would not likely require responses such as evacuating or sheltering the general public. These

actions become important only when considering accidents that are much less probable than

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design-basis accidents. 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 in December 1978, and NUREG-0654/FEMA-REP-1 (NUREG-0654), “Criteria

for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness

in Support of Nuclear Power Plants,” Revision 1, issued in November 1980, describe the

emergency planning basis. NUREG-0654/FEMA-REP-1 is being revised to align with the NRC

emergency preparedness rule changes, which became effective in December 2011, and with

the revised Federal Emergency Management Agency (FEMA) Radiological Emergency

Preparedness Program manual issued in 2011. The NRC significantly enhanced its emergency

preparedness regulations to address lessons learned from the September 11, 2001, terrorist

attacks and security events. This is further discussed in Section 16.6.4 of this report.

16.2 Offsite Emergency Planning and Preparedness

The accident at Three Mile Island, Unit 2 revealed that better coordination and more

comprehensive emergency plans and procedures were needed if the NRC and the public were

to have confidence in the readiness of onsite and offsite emergency response organizations to

respond to a nuclear emergency. Before the accident at Three Mile Island, Unit 2, there was no

clear obligation for State and local governments to develop emergency plans for radiological

accidents, and the Federal role was one of assistance and guidance. After the accident, the

NRC amended its emergency planning regulations to require, as a condition of licensing, that

each applicant or licensee submit the radiological emergency response plans of the State, Tribal

and local governments that are within the plume exposure zone, as well as the plans of State

governments within the ingestion pathway zone (10 CFR 50.33(g) and 10 CFR 50.54(s)).

In December 1979, the President directed FEMA to take the lead in ensuring the development

of acceptable State, Tribal, and local offsite emergency plans and activities for nuclear power

plants. The NRC and FEMA regulations, as well as a memorandum of understanding between

the two agencies, contained in Appendix A, “Memorandum of Understanding between Federal

Emergency Management Agency and Nuclear Regulatory Commission,” to 44 CFR Part 350,

“Review and Approval of State and Local Radiological Emergency Plans and Preparedness,”

dated June 17, 1993, subsequently codified FEMA’s role and responsibilities.

FEMA provides its findings on the acceptability of the offsite emergency plans and

preparedness to the NRC, which has the ultimate responsibility for determining the overall

acceptability of radiological emergency plans and preparedness for a nuclear power reactor.

The NRC will not issue a license to operate a nuclear power reactor unless it finds that the

condition of onsite and offsite emergency preparedness provides reasonable assurance that

protective measures can and will be taken in a radiological emergency. The NRC bases its

decision on a review of the FEMA findings and determinations on whether State and local

emergency plans are adequate and can be carried out, and on its own assessment of whether

the onsite emergency plans are adequate and can be implemented (10 CFR 50.47(a)).

The principal guidance for preparing and evaluating radiological emergency plans for licensee,

State, and local government emergency planners is NUREG-0654/FEMA-REP-1, a joint NRC

and FEMA document. NUREG-0654/FEMA-REP-1 gives evaluation criteria for an acceptable

way to meet the emergency planning standards in the NRC and FEMA regulations

(10 CFR 50.47(b) and 44 CFR Part 350, respectively). These criteria provide a basis for

licensees, States, Tribal, and local governments to develop acceptable emergency plans.

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The NRC and FEMA coordinate their evaluation of periodic emergency response exercises and

require all operating nuclear power plant sites to conduct an exercise every 2 years, as outlined

in Section IV.F.2(b) of Appendix E to 10 CFR Part 50, “Emergency Planning and Preparedness

for Production and Utilization Facilities.” These mandatory full-participation exercises are

integrated efforts by the licensee, State, Tribal, and local radiological emergency response

organizations that have a role in support of the licensee’s emergency plan. The NRC evaluates

the licensee’s performance, while FEMA evaluates State, Tribal, and local agencies’ responses.

In some cases, other Federal response agencies also participate in these exercises. Any

weaknesses or deficiencies that the NRC or FEMA identify because of the exercise must be

corrected through appropriate remedial actions. Section IV.F.2(d) of Appendix E to 10 CFR Part

50, requires the offsite response agencies to participate in biennial exercises of their plume

exposure pathway plans every 2 years, and for the State to participate in an ingestion pathway

exercise with a nuclear power plant located within its State every 8-year exercise cycle.

Through the Steering Committee for Emergency Planning, established under the NRC-FEMA

memorandum of understanding, both agencies discuss and coordinate on the interpretation and

implementation of existing regulations and guidance; the consistent evaluation of each

respective agency’s radiological emergency preparedness programs and resolution of identified

deficiencies; and the development and implementation of proposed changes to radiological

emergency preparedness-related regulations and guidance.

16.3 Emergency Classification System and Emergency Action Levels

A licensee or applicant at a U.S. nuclear power plant is required to develop a standard

emergency classification and action level scheme based on facility system and effluent

parameters (10 CFR 50.47(b)(4)). Appendix E (Section IV.C.1) of 10 CFR Part 50 defines four

emergency classification levels in order of increasing severity:

(1) notification of unusual

events, (2) alert, (3) site area emergency, and (4) general emergency. The specific class of

emergency is declared on the basis of plant conditions that trigger the emergency action levels.

Licensees and State, Tribal, and local agencies have established specific procedures for

carrying out emergency plans for each emergency classification level. The event classification,

declared by the licensee, initiates appropriate actions for that class, including notification of

offsite authorities, activation of onsite and offsite emergency response organizations, and,

where appropriate, protective action recommendations for the public.

Emergency action level development guidance was initially established in GL 79-50,

“Emergency Plans Submittal Dates,” and was subsequently established in

NUREG-0654/FEMA-REP-1, which was endorsed as an approach for the development of an

emergency action levels scheme through RG 1.101, “Emergency Planning and Preparedness

for Nuclear Power Reactors,” Revision 2, dated October 1981. NUREG-0654/FEMA-REP-1

defines and gives examples of initiating conditions for the four emergency classification levels.

These conditions form the basis for each licensee to establish specific thresholds and

indicators, known collectively as emergency action levels. These action levels reflect specific

plant conditions (e.g., plant system status, inplant and effluent radiological parameters, and

other inplant hazards) or external events (e.g., flooding, earthquakes, high winds, security

events) for each of the four emergency classification levels.

As industry and regulatory experience was gained with the implementation and use of

emergency action levels schemes, the industry issued revised emergency action levels scheme

development guidance to reflect lessons learned. To date, NUMARC/NESP-007, “Methodology

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for Development of Emergency Action Levels,” dated January 1992, and NEI 99-01,

“Methodology for Development of Emergency Action Levels,” Revisions 4, 5, and 6, were

provided to the NRC for review and endorsement as generic (nonplant-specific) emergency

action levels development guidance. RG 1.101, Revisions 3 and 4, endorsed

NUMARC/NESP-007 and NEI 99-01, Revision 4, dated January 2003, as acceptable

alternatives for licensees to consider in the development of their plant-specific emergency action

levels schemes, and allowed licensees to develop plant-specific emergency action levels based

upon an alternative approach. The NRC endorsed NEI 99-01, Revision 5, dated February 2008,

through a letter dated February 22, 2008. The NRC endorsed NEI 99-01, Revision 6, through a

letter dated March 28, 2013. Emergency action levels development guidance for the Advanced

Passive (AP)1000 and the General Electric-Hitachi’s Economic Simplified Boiling Water Reactor

(ESBWR) reactor designs is provided in NEI 07-01, “Methodology for Development of

Emergency Action Levels Advanced Passive Light Water Reactors,” Revision 0, dated

July 2009.

The emergency action level development guidance contained in GL 79-50,

NUREG-0654/FEMA-REP-1, NUMARC/NESP-007, and NEI 99-01, Revisions 4, 5, and 6, are

all considered generic emergency action level scheme development guidance, as they are not

plant-specific and may not be entirely applicable for some reactor designs (note that NEI 07-01

is only applicable to the AP1000 and ESBWR designs). However, the guidance contained in

these documents bounds the most typical accident or event scenarios for which emergency

response is necessary, in a format that allows for industry standardization and consistent

regulatory oversight. Most licensees choose to develop plant-specific emergency action level

schemes using the latest NRC-endorsed guidance with appropriate plant-specific alterations, as

applicable. Under 10 CFR Part 50, Appendix E, Section IV.B (2), a revision to an emergency

action level must be approved by the NRC before implementation if the licensee is changing

action level schemes.

Although not required under existing U.S. emergency preparedness regulations contained in

10 CFR 50.47(b) and Appendix E to 10 CFR Part 50, several procedures guide onsite licensed

reactor operator actions depending on the nature and extent of events at the plant. These

events, such as a loss of offsite electrical power, are within the plant’s design basis and

addressed by various plant procedures, typically abnormal operating procedures, alarm

response procedures, and emergency operating procedures. These procedures instruct the

plant operators on the steps necessary to take the plant from full-power operation to a safe

shutdown condition, if necessary, based on the severity of the event. Emergency operating

procedures have long been part of the NRC’s safety requirements. Numerous regulatory guides

and technical reports address the development of emergency operating procedures and their

use (e.g., NUREG-0660, “NRC Action Plan Developed as a Result of the TMI-2 Accident” and

NUREG-0737, “Clarification of TMI Action Plan Requirements,” issued in November 1980).

The nuclear industry developed SAMGs in response to the Three Mile Island accident based on

extensive research on severe accident phenomena. Their purpose is to enhance the ability of

plant operators to manage accident sequences that progress beyond emergency operating

procedures and other applicable plant procedures. Although not required under U.S. emergency

preparedness regulations, SAMGs are intended for use by plant technical staff, usually in

emergency support facilities activated under the emergency plan, in support of onshift control

room operators. In GL 1988-20, “Accident Management Strategies for Consideration in the

Individual Plant Examination Process,” Supplement 2, dated April 4, 1990, the NRC

encouraged, but did not require, licensees to develop and implement SAMGs. Because SAMGs

are voluntary, formal training and licensing of plant operators and emergency preparedness

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regulations do not require them to be addressed. Nevertheless, operating power reactor

licensees committed to implementing SAMGs following guidance that was agreed upon with the

NRC and completed that implementation by the end of 1998.

Following the Fukushima Dai-ichi accident, the nuclear industry and the NRC revisited the issue

of SAMGs. Immediately following the accident, the NRC conducted inspections of the

voluntarily-implemented SAMGs under TI 2515/184, “Availability and Readiness Inspection of

Severe Accident Management Guidelines (SAMGs).” The inspectors observed inconsistent

implementation of SAMGs and attributed it to the voluntary nature of the initiative. As a result of

this effort, the Commission once again considered whether SAMGs should be required of

licensees, but concluded that the potential risk of a severe accident, which has diminished since

the Three Mile Island accident through the imposition of additional pre-core damage safety

requirements over the years, does not rise to the level for which it would be appropriate to

require SAMGs. In parallel with this NRC reexamination of the need to require SAMGs, the

nuclear industry made a number of changes to the technical basis document that supports

development of the guidelines in order to appropriately treat the lessons learned from that

event. These revisions were used by the owners groups to revise and update their generic

guidelines, which in turn are being used by the industry to update the facility-specific guidelines.

These improvements to the industry’s SAMG program, including revised commitments on the

part of the nuclear industry for a process under which the SAMGs would be maintained and

updated, contributed to the NRC’s conclusion that continued status as a voluntary program is

appropriate. In addition, the NRC initiated changes to the Reactor Oversight Process to include

inspection of the resulting programs as voluntarily-imposed standards.

Additionally, the NRC is in the process of developing regulatory guidance for the Mitigation of

Beyond-Design-Basis Events Rulemaking that would provide for appropriate coordination of the

voluntarily-maintained SAMGs with the guidance and strategies required by the rulemaking for

beyond-design-basis events resulting from natural phenomena as well as the extensive damage

mitigating guidelines required of the licensees after the terrorist events of September 11, 2001.

On August 27, 2015, the Commission issued SRM-SECY-15-0065, “Proposed Rulemaking:

Mitigation of Beyond-Design-Basis Events,” directing the staff to update the Reactor Oversight

Process to explicitly provide periodic oversight of industry’s voluntary implementation of the

SAMGs. Because this is a longer-term activity, the NRC staff envisions that the Reactor

Oversight Process and relevant inspection procedures will be updated by December 31, 2020,

to allow for oversight of the site-specific voluntary incorporation of SAMGs in generic guidance

revisions. The boiling-water reactor owners group and the pressurized-water reactor owners

group have also issued revised versions of their generic guidance for SAMGs based on the

industry’s technical basis document update and continue to improve this guidance to reflect

lessons learned in the implementation.

16.4 Recommendations for Protective Action in Severe Accidents

The technical basis and guidance for developing protective action strategies for use during a

nuclear power plant event resulting in a general emergency classification in the United States

appear in NUREG-0654/FEMA-REP-1, Revision 1, Supplement 3, issued in November 2011,

and EPA 400-R-92-001, “Manual of Protective Action Guides and Protective Actions for Nuclear

Incidents,” issued in May 1992, which is under revision. NUREG-0654/FEMA-REP-1,

Supplement 3, “Guidance for Protective Action Strategies,” reflects the conclusions developed

from analysis of a spectrum of nuclear power plant core melt accident scenarios. These

analyses are documented in NUREG/CR-6953, “Review of NUREG-0654, Supplement 3,

“Criteria for Protective Action Recommendations for Severe Accidents,” Volumes 1, 2, and 3.

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Although a general emergency is a serious event and warrants protective action, it is not

necessarily synonymous with a “severe accident” as that term is used in U.S. nuclear power

plant accident analyses. NUREG-0654/FEMA-REP-1, Supplement 3, recognizes the disparity

between a severe accident with early release and other general emergency conditions, and

provides scenario-specific protective action decision guidance. Additionally, it provides guidance

for the consideration of evacuation time estimates and for the immediate evacuation of those

closest to the nuclear power plant and criteria for the expansion of initial protective actions.

The NRC considers evacuation and sheltering to be the two primary protective actions. A

supplemental protective action for the general population is using the thyroid-blocking agent

potassium iodide. In 2001, the NRC amended its regulations for emergency planning associated

with potassium iodide, 10 CFR 50.47(b)(10). This amendment requires that each State consider

giving potassium iodide to the general public as a protective measure, supplementing the

evacuation and sheltering protective actions. The NRC found that potassium iodide is a

reasonable, prudent, and inexpensive supplement to evacuation and sheltering for specific local

conditions. In January 2002, the NRC, in cooperation with the cognizant agencies, updated the

Federal policy statement on potassium iodide prophylaxis to reflect the changes in NRC

regulations.

The agency provides guidance for response procedures and training manuals for NRC staff in

NUREG/BR-0150, “Response Technical Manual 96,” Volume 1, Revision 4, issued in March

1996. The NRC’s guidance on evacuation and sheltering in the event of a nuclear power plant

accident is consistent with guidance in IAEA TECDOC-953, “Method for the Development of

Emergency Response Preparedness for Nuclear or Radiological Accidents,” and IAEA

TECDOC-955, “Generic Assessment Procedures for Determining Protective Actions during a

Reactor Accident,” both issued in 1997.

16.5 Inspection Practices - Reactor Oversight Process for Emergency Preparedness

The NRC’s Reactor Oversight Process addresses emergency preparedness. The process

allows licensees to manage their own emergency preparedness programs, including corrective

actions, as long as the performance indicators and inspection findings are within an acceptable

performance band. The NRC handles inspection findings through its significance determination

process. Article 6 of this report discusses the NRC’s Reactor Oversight Process and

significance determination process.

Emergency preparedness is one of the Reactor Oversight Process’ seven cornerstones of

safety. The objective of this cornerstone is to “ensure that the licensee is capable of

implementing adequate measures to protect the public health and safety during a radiological

emergency.” Oversight of this cornerstone is achieved through three performance indicators

and a supporting risk-informed inspection program. The performance indicators are drill and

exercise performance, emergency response organization drill participation, and alert and

notification system reliability. The performance indicator for drill and exercise performance

monitors timely and accurate licensee performance in drills, exercises, and actual events when

presented with opportunities to classify emergencies, notify offsite authorities, and recommend

protective actions. The indicator for emergency response organization drill participation

measures the percentage of key members of the licensee’s emergency response organization

who have participated in proficiency-enhancing drills, exercises, training opportunities, or an

actual event over a determinant amount of time. The alert and notification system reliability

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indicator monitors the reliability of the offsite alert and notification system, which is a critical link

for communicating with the public.

The emergency preparedness cornerstone of the Reactor Oversight Process includes the

following areas for inspection:

Maintenance of Emergency Preparedness Program - Inspectors evaluate the licensees’

efforts to identify and resolve program weaknesses, adequacy of internal program

assessment activities, emergency plan change process, maintenance of equipment

important to emergency preparedness, evacuation time estimate population monitoring,

and implementation of emergency response facility maintenance.

Drill Evaluation - Inspectors evaluate drills and simulator-based training evolutions in

which shift operating crews and licensee emergency response organization members

participate.

Exercise Evaluation - Inspectors independently observe the licensee’s performance in

classifying, notifying, and developing recommendations for protective actions, and other

activities during the exercise. Evaluated exercise scenarios are varied over an 8-year

exercise cycle to include a hostile action event, no radiological release, or minimal

release not requiring public protective actions, and a rapidly progressing event. The

inspectors assess whether the licensee’s self-critique is consistent with their

observations. The emergency preparedness performance indicators for drill and exercise

performance rely upon the accurate determination of successful performance and the

correction of identified weaknesses during the conduct of drills and exercises. If a

licensee either fails to properly critique performance or correct identified weaknesses,

then the validity of the drill and exercise performance indicators come into question.

Performance problems with classification, notification, dose assessment and protective

action recommendations are the highest priority inspection areas. Exercise evaluation

results are provided in inspection reports available on the NRC’s public Web site. These

inspection reports identify findings associated with a licensee’s failure to either properly

critique or correct weaknesses observed during the conduct of a licensee’s drill and

exercise program.

Alert and Notification System Evaluation - Inspectors verify how well the testing program

complies with program procedures.

Emergency Action Level and Emergency Plan Changes - Inspectors review all of the

licensee’s changes to emergency action levels and a sample of changes to the

emergency plan to determine if any of the changes have decreased the effectiveness of

the emergency plan.

Emergency Response Organization Staffing and Augmentation System - Inspectors

review the augmentation system to determine whether, as designed, it will support

augmentation of the emergency response organization in accordance with the goals for

activating the emergency response facility.

Reactor Safety/Emergency Preparedness - Inspectors verify that the data reported for

the performance indicator values are valid.

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It is important to note, however, that even though FEMA has no direct regulatory authority over

State or local governments and their full-participation exercise evaluations are not considered

inspections, FEMA’s exercise findings carry substantial weight in the NRC regulatory process.

FEMA notifies the State Government and the NRC of any significant deficiencies in offsite

performance shortly after the exercise. FEMA also issues a formal exercise report within

90 days of the exercise’s completion describing the FEMA exercise findings. Because of the

potential effect of deficiencies on offsite emergency preparedness, findings are expected to be

corrected within 120 days of the exercise. Failure of offsite organizations to correct deficiencies

promptly could lead FEMA to withdraw its finding of “reasonable assurance.” This would cause

the NRC to assess the continued operation of the facility.

16.6 Responding to an Emergency

Fundamental changes in the response to national emergencies have occurred as a result of the

publication of the National Response Framework in May 2013 and the update of its associated

annexes. Additionally, the U.S. Department of Homeland Security (DHS) has revised and

republished the National Incident Management System (NIMS) document in December 2008.

This section explains the roles of the NRC, other Federal agencies, licensees, States, and local

governments during the response to an incident. It also explains the security issues associated

with supporting the response efforts.

16.6.1 Federal Response

The Federal response structure was revamped in the aftermath of the events of

September 11, 2001, with the creation of DHS, the implementation of Homeland Security

Presidential Directive 5 (HSPD-5), “Management of Domestic Incidents,” dated March 4, 2003,

and the implementation of Presidential Policy Directive 8 (PPD-8) “National Preparedness,”

dated March 30, 2011. HSPD-5 establishes the Secretary of Homeland Security as the primary

Federal official for managing domestic incidents. Under the Homeland Security Act of 2002,

DHS is responsible for coordinating Federal operations within the United States to prepare for,

respond to, and recover from terrorist attacks, major disasters, and other emergencies. PPD-8

directed the development of a national preparedness goal that identifies the core capabilities

necessary for preparedness and a national preparedness system to guide activities that will

enable the Nation to achieve the goal.

DHS will assume overall Federal incident management coordination responsibilities when any

one of the following four conditions applies:

(1)

A Federal department or agency acting under its own authority has requested DHS

assistance.

(2)

The resources of State and local authorities are overwhelmed, and the appropriate State

and local authorities have requested Federal assistance.

(3)

More than one Federal department or agency has become substantially involved in

responding to the incident.

(4)

The President of the United States has directed the Secretary to assume incident

management responsibilities.

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In 2008, 2011, and 2013, the governing documents outlining the responsibilities of the Secretary

of Homeland Security, DHS, and other Federal, State, and local entities were updated. These

documents were related to NIMS and the National Response Framework and its associated

annexes.

NIMS is a comprehensive, national approach to incident management that is applicable at all

jurisdictional levels and across functional disciplines. NIMS enables Federal, State, and local

entities to work together to prevent, protect against, respond to, recover from, and mitigate the

effects of incidents, regardless of cause, size, location, or complexity, to reduce the loss of life

and property and harm to the environment. NIMS provides an organized set of scalable and

standardized operational structures that is critical for allowing various organizations and

agencies to work together in a predictable, coordinated manner.

NIMS works hand-in-hand with the National Response Framework. NIMS provides the template

for the management of incidents, while the National Response Framework describes the

structures and mechanisms for national-level policy for incident management. The five National

Planning Frameworks (i.e., prevention, protection, mitigation, response, and disaster recovery)

provide guidance on Federal coordinating structures and processes to prevent, prepare for,

respond to, and recover from domestic incidents such as terrorist attacks, major disasters, and

other emergencies.

The Federal response to a potential nuclear or radiological incident is designed to support the

efforts of the facility operator and offsite officials. For such emergencies, Federal response

activities are carried out in accordance with the National Response Framework’s

Nuclear/Radiological Incident Annex, which describes the roles of DHS, coordinating agencies

(e.g., the NRC during an incident with one of its licensees), and other supporting Federal

agencies. During an incident that meets the criteria of HSPD-5 (invoked during a

terrorist-related incident or at a general emergency level for an NRC licensee), DHS is

responsible for the overall domestic incident management, while the coordinating agency

coordinates the Federal onscene actions and helps State and local governments determine

measures to protect life, property, and the environment. The coordinating agency may respond

as part of the Federal response as requested by DHS under the framework, or in accordance

with its own authorities. During less severe incidents, coordinating agencies will oversee the

onsite response, monitor and support owner or operator activities (when there is an owner or

operator), provide technical support to the owner or operator if asked, serve as the principal

Federal source of information about onsite conditions, and, if asked, advise the State and local

government agencies on implementing protective actions. The coordinating agency also will

provide a hazard assessment of onsite conditions that might have significant offsite effects and

ensure that onsite measures are taken to mitigate offsite consequences.

16.6.2 Licensee, State, and Local Response

The NRC recognizes the nuclear power plant operator (licensee) and the State or local

government as the two primary decisionmakers during a radiological incident at a licensed

power reactor. The licensee is primarily responsible for mitigating the consequences of an

incident on site and recommending timely protective actions to State and local authorities. The

States or local governments are ultimately responsible for implementing appropriate protective

actions for public health and safety.

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16.6.3 The NRC’s Response

In fulfilling its legislative mandate to protect the public health and safety, the NRC has

developed a plan and procedures detailing its response to incidents involving licensed material

and activities (NUREG-0728, “NRC Incident Response Plan,” Revision 4, issued in

April 14, 2005). In accordance with that plan, the NRC will initially assess any reported event

and decide whether or how it will respond as an agency. To meet its statutory and regulatory

obligations, the NRC will usually dispatch a team to the site for all serious incidents. The team

may help the State interpret and analyze technical information, update other responding Federal

agencies on event conditions, and coordinate any multiagency Federal response.

Once the NRC has decided to respond as an agency, it activates the NRC headquarters

Operations Center near Washington, DC, and the associated regional incident response center.

The NRC headquarters Operations Center will then take the following actions:

(1) maintain

continuous communications with the facility, (2) assess the incident, (3) advise the facility

operator and offsite officials, (4) coordinate the Federal radiological response with other Federal

agencies, and (5) respond to inquiries from the national media. The staff at the NRC

headquarters Operations Center includes emergency preparedness and response experts and

personnel experienced with liaison activities. Because regional office personnel usually have

firsthand knowledge of the details of the affected facility, early in an incident the Regional

Administrator provides operational authority from the affected regional office and, if necessary,

from the regional incident response center. When the NRC’s onsite presence is required, the

agency will dispatch a team from the affected regional office.

As soon as the NRC site team arrives at the facility and is ready to assume the agency’s

leadership role, it may be delegated certain responsibilities that may include the authority to

direct the agency’s onsite response.

The NRC site team consists of many technical specialists and representatives who respond to

the designated response centers that the facility and offsite officials use to coordinate the

response. These response centers include the affected State’s emergency operations center,

the first-responder’s incident command post, the joint information center, established by the

facility or local government to interact with the media, and, if necessary, the joint field office (the

primary Federal incident management field structure that is usually established 48 to 72 hours

after an incident). Through participation in these response centers, the NRC site team has

access to wide-ranging State and Federal response assets, as well as to extensive radiological

monitoring capabilities through the U.S. Department of Energy (i.e., field teams and aerial

monitoring).

The NRC regularly participates in nuclear power plant and Federal interagency exercises each

year to ensure its readiness to respond. The NRC also participates in the planning and conduct

of the annual continuity of operations exercise and National Level Exercises each year. The

NRC’s participation in such exercises gives the agency a valuable perspective on multievent

response. This perspective improves interagency cooperation and imparts a better

understanding of response roles during emergencies.

16.6.4 Aspects of Security that Support Response

Before September 11, 2001, the physical security measures implemented at NRC-licensed

nuclear facilities provided for the protection of public health and safety against the design-basis

threat for radiological sabotage as described in 10 CFR 73.1, “Purpose and Scope.” Following

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the events of September 11, 2001, the nuclear industry significantly enhanced its defensive

capability through voluntary actions by the licensees in response to NRC advisories and

additional actions required by security orders issued in 2002 and 2003. These enhancements

included a revised design-basis threat for radiological sabotage and security measures against

threats from an insider, waterborne attack, vehicle bomb attack, and land-based assault. The

NRC subsequently codified its revised design-basis threat regulations on March 19, 2007, and

updated the power reactor security regulations on March 27, 2009. These updated regulations

incorporated provisions of the security orders and lessons learned during the implementation of

the orders.

The NRC receives security-related information from the national intelligence community, law

enforcement, and licensees, and it continually evaluates this information to assess threats to

regulated facilities or activities. The NRC works with other Federal agencies, particularly DHS

and the Federal Bureau of Investigation, to ensure that security around nuclear power plants is

well coordinated and that law enforcement responders are prepared for a significant event. If an

event were to occur, the NRC would have significant resources accessible to it and as many as

18 Federal agencies available to help mitigate the radiological consequences of a serious

accident or successful attack.

16.7 Communications with Neighboring States and International Arrangements

The NRC has agreements with the United States’ geographical neighbors, Canada and Mexico.

The NRC’s bilateral arrangements with nonneighboring countries also address and promote

sharing of information on emergency preparedness and resources.

Since 2001, the United States has participated in the International Nuclear Event Scale by

evaluating operating reactor events and reporting to IAEA any events resulting in a

categorization of International Nuclear Event Scale Level 2 or higher. The United States has

also played a significant role on the IAEA’s International Nuclear and Radiological Event Scale

Advisory Committee, including supporting the negotiations that resulted in the expanded use of

the International Nuclear and Radiological Event Scale for rating radiation and transport events.

The NRC participates in the IAEA’s Unified System for Information Exchange for Incidents and

Events as the method for rapidly sharing nuclear or radiological event information with IAEA and

its member countries. To meet the U.S. commitment under the IAEA Convention on Early

Notification of a Nuclear Accident, the NRC will promptly notify IAEA if a serious accident occurs

at a commercial nuclear power plant. Afterward, the NRC will work with the U.S. Department of

State to update IAEA frequently regarding the emergency event.

Under its bilateral agreements with Canada and Mexico, the NRC will promptly notify and

exchange information in the event of an emergency that has the potential for transboundary

effects. The arrangement with Canada, the “Arrangement between the United States of America

Nuclear Regulatory Commission and the Canadian Nuclear Safety Commission for the

Exchange of Technical Information and Cooperation in Nuclear Safety Matters,” was most

recently renewed in 2012 for a period of 5 years. The arrangement with Mexico, the

“Arrangement between the Nuclear Regulatory Commission of the United States of America and

the Comision Nacional de Seguridad Nuclear y Salvaguardias of the United Mexican States for

the Exchange of Technical Information and Cooperation in Nuclear Safety and Research

Matters,” was most recently renewed in 2012 for a period of 5 years.

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Because both bilateral agreements’ most recent renewals occurred after the Fukushima

accident, the NRC and its Canadian and Mexican counterparts have placed increased focus on

their commitment to share information not only in the event of an accident, but on a regular

basis as part of an effort to enhance their respective emergency preparedness programs. The

NRC and the Canadian Nuclear Safety Commission have conducted several technical bilateral

meetings in 2013, 2014, and 2015. Most recently, senior staff and managers from the Canadian

regulator and the Mexican regulator observed a full Federal emergency exercise at the H.B.

Robinson Nuclear Power Plant in South Carolina as a part of the U.S. Government’s Southern

Exposure 2015 multiple-day exercise. During the exercise, the NRC hosted 52 foreign

regulators, government officials, and nuclear utility staff from 13 countries, as well as observers

from the IAEA and the Nuclear Energy Agency.

The NRC also routinely communicates with the IAEA and its Canadian and Mexican

counterparts during its emergency drills. In addition, the NRC regularly participates in IAEA

emergency preparedness and response conferences, technical meetings and consultancies in

Vienna, Austria. The NRC also hosts several bilateral exchanges every year regarding

emergency preparedness and response activities and emergency exercise observation with

foreign regulatory bodies at the NRC Headquarters in Rockville, MD, and at U.S. nuclear power

plants around the country.

16.8 Communications with the Public

One of the emergency planning standards for U.S. nuclear power reactors requires that

information be made periodically available to the public on how they would be notified and what

their initial actions should be in an emergency (e.g., listening to a local broadcast station and

remaining indoors), that the principal points of contact with the news media for dissemination of

information during an emergency (including the physical location or locations) be established in

advance, and that procedures have been established for coordinated dissemination of

information to the public. If an emergency were declared, another emergency planning standard

requires that the content of initial and followup public messages has been established; and that

a means has been established to provide early notification and clear instruction to the

population within the plume exposure pathway emergency planning zone. NUREG-0654/FEMA-

REP-1 outlines the evaluation criteria for complying with the requirements of these emergency

planning standards.

Appendix E (Section IV.D) to 10 CFR Part 50 describes licensee requirements for the prompt

notification of the public in the event of a declared emergency and for the yearly dissemination

of basic emergency planning information to the public located within the plume exposure

pathway emergency planning zone, such as:

the methods and times required for public notification and the planned protective actions

if an accident were to occur

general information on the nature and effects of radiation

a listing of local broadcast stations that would be used to disseminate information during

an emergency

the use of signs or other measures to disseminate appropriate information to transient

populations in the event of an accident

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The NRC performs continuous outreach with licensees and respective State, Tribal, and local

emergency response organizations to facilitate stakeholder interface and involvement on

existing and proposed radiological emergency preparedness activities. The NRC outreach effort

consists of:

(1) attending nuclear industry and radiological emergency preparedness-related

conferences and forums, (2) conducting public meetings on proposed changes to radiological

emergency preparedness-related regulations and guidance, and (3) using the NRC Web site,

blog posts, and periodic newsletters for outreach.

16.9 Fukushima Lessons Learned

After the Fukushima event, the NRC undertook actions to enhance emergency preparedness for

licensees with respect to communications and staffing given a multiunit event and a prolonged

SBO. The accident highlighted the need for licensees to identify the staff needed to respond to a

multiunit event given a prolonged SBO. In addition, the accident highlighted that communication

equipment relied upon during an emergency must be operable to coordinate the event response

during a prolonged SBO.

On March 12, 2012, the NRC issued an RFI to all power reactor licensees and holders of

construction permits to obtain information that would help the staff to evaluate the NTTF

Recommendation 9.3 on assessing staff needs and communications to effectively respond to a

multiunit event. This recommendation was identified as an activity that should begin without

unnecessary delay (i.e., Tier 1).

The addressees were requested to assess both staffing and communications. Licensees were

asked to evaluate their current communications systems and the equipment that would be used

during an emergency event assuming that a large-scale natural event resulted in a loss of all

alternating current power on site, to consider enhancements regarding the communications

requirement in NRC regulations (10 CFR 50.47, “Emergency Plans,” and Appendix E to

10 CFR Part 50) and in NUREG-0696, “Functional Criteria for Emergency Response Facilities,”

issued in February 1981, and to assume the event resulted in extensive damage to normal and

emergency communications systems both on site and in the area surrounding the site and that

cellular and other communications infrastructures were unavailable. Addressees also were

asked to evaluate:

how communications equipment used during an emergency event would be powered

assuming a prolonged SBO

their emergency response organization staffing following the occurrence of a large

scale natural event that altered the normal access routes to the site, thereby affecting

the response time for the emergency response organization

their current staffing levels and the appropriate staff and positions to respond to a

multiunit event given a beyond-design-basis natural event and to determine if

enhancements were needed

the minimum staffing that would be on site at the time the event occurred and to assess

the need for additional onsite staff as the event unfolded, since this could affect a

licensee’s assessment capabilities

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All licensees submitted their communications assessments by October 31, 2012. NRC staff

issued safety assessments documenting the staff’s review to each licensee by July 2013.

Phase 1 of the staffing assessments had licensees evaluate their ability to respond to a multiunit

SBO event utilizing existing processes and procedures. At single unit sites, addressees were

required to provide information on staffing necessary to cope with an extended loss of all

alternating current power if access to the site was impeded. The licensee responses to the RFI

for the Phase 1 staffing assessments were received and evaluated by the NRC staff. The NRC

staff issued acknowledgement letters to all licensees with multiunit sites.

Phase 2 of the staffing assessments has licensees assess staffing needs associated with

implementation of mitigating strategies for beyond-design-basis external events. As such, the

Phase 2 assessments have a dependency on NRC Order EA-12-049, “Order Modifying

Licenses with Regard to Requirements for Mitigating Strategies for Beyond-Design-Basis

External Events,” issued on March 12, 2012. The majority of these assessments have been

received. The remaining four assessments, with the exception of the FitzPatrick site that it is

requesting relaxation until 2017, have been received by the agency. The NRC’s review of these

assessments will be completed by the end of 2016.

Additionally, the NRC staff identified lessons learned applicable to the NRC Incident Response

Program not covered under the NTTF recommendations. One of these items was associated

with the challenges faced in communicating with States and regional stakeholders. The staff has

put considerable effort into improving its communication strategy with all response stakeholders

and will continue to make enhancements when improvement opportunities are identified.

The NRC is focusing its efforts on the implementation of the Tier 2 NTTF Recommendation 9.3,

which includes the following:

adding guidance to licensees’ emergency plans that describes the ability to perform a

multiunit dose assessment (including releases from spent fuel pools) using the

licensees’ site-specific dose assessment software and approach

conducting periodic training and exercises for multiunit and prolonged SBO scenarios

ensuring that emergency preparedness equipment and facilities are sufficient for dealing

with multiunit and prolonged SBO scenarios

The NRC staff determined that the mitigating strategies recommendation (NTTF

Recommendation 4.2) addresses periodic training and exercises for multiunit and prolonged

SBO scenarios and ensures that emergency preparedness equipment and facilities are

sufficient.

The NRC staff considered options regarding how licensees should perform a multiunit dose

assessment using the licensee’s site-specific dose assessment software and approach, and the

NRC determined that all licensees had full capability to perform these dose assessments by

May 2015.

In addition to the Tier 3 emergency preparedness items discussed in Sections 1.3.1 and 1.3.3 of

this report, the NRC staff also identified recommendations for lessons learned from the

Fukushima event that may warrant regulatory action, but were not specifically included with the

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NTTF recommendations. The recommendations that require further staff study to support

regulatory action (i.e., Tier 3) include:

Evaluate the basis of the emergency planning zone size.

Evaluate whether potassium iodide should be prestaged beyond the current 10 mile

zone.

In October 2015, the staff issued SECY-15-0137, “Proposed Plans for Resolving Open

Fukushima Tier 2 and 3 Recommendations,” recommending that these two items be closed for

the reasons stated below. In Staff Requirement Memorandum (SRM)-SECY-15-0137, dated

February 8, 2016, the Commission approved closing these items.

The staff had conducted an extensive analysis of the emergency planning zone size in response

to a petition for rulemaking numbered 50-104, “Petition for Rulemaking Requesting

Amendments Regarding Emergency Planning Zone Size,” which can be found in

www.regulations.gov (Docket ID: NRC-2012-0046). The NRC staff concluded that the current

size of the emergency planning zones is appropriate for existing reactors (including multiunit

sites) and proposed new reactors and that emergency plans will provide an adequate level of

protection of the public health and safety in the event of an accident at a nuclear power plant.

Furthermore, the staff noted that the current emergency planning zones provide for a

comprehensive emergency planning framework that would allow expansion of the response

efforts beyond the designated distances should events warrant such an expansion. In

SRM-SECY-13-0135, “Denial of Petition for Rulemaking Requesting Amendments Regarding

Emergency Planning Zone Size (PRM-50-104),” dated February 27, 2014, the Commission

agreed with the staff’s proposal to deny the petition and directed the staff to publish the denial in

the Federal Register.

Regarding the prestaging of potassium iodide beyond the current 10-mile zone, the staff plans

to continue to monitor the studies being conducted by the World Health Organization, United

Nations Scientific Committee on the Effects of Atomic Radiation, and the Fukushima Health

Management Survey, and engage stakeholders as appropriate.

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ARTICLE 17. SITING

Each Contracting Party shall take the appropriate steps to ensure that appropriate

procedures are established and implemented for

(i)

evaluating all relevant site-related factors that are likely to affect the safety of a

nuclear installation for its projected lifetime

(ii)

evaluating the likely safety impact of a proposed nuclear installation on

individuals, society, and the environment

(iii)

re-evaluating, as necessary, all relevant factors referred to in subparagraphs (i)

and (ii) so as to ensure the continued safety acceptability of the nuclear

installation

(iv)

consulting Contracting Parties in the vicinity of a proposed nuclear installation,

insofar as they are likely to be affected by that installation and, upon request,

providing the necessary information to such Contracting Parties, in order to

enable them to evaluate and make their own assessment of the likely safety

impact on their own territory of the nuclear installation

This section explains the responsibilities of the U.S. NRC for siting, which include site safety,

environmental protection, and emergency preparedness. This article discusses the regulations

applying to site safety and their implementation, emphasizing regulations applying to seismic,

geological, hydrological, meteorological, and radiological assessments. It explains

environmental protection and reevaluation of site-related factors. It also addresses the Vienna

Declaration on Nuclear Safety, which was issued in February 2015. Article 16 of this report

discusses emergency preparedness and international arrangements, which would apply to

Contracting Parties in obligation (iv) above. Finally, no changes to the current NRC practices

associated with siting were identified as part of the NRC’s Fukushima lessons learned

initiatives.

The United States reviewed the results of the CNS 2015-2016 consultancy meetings that

developed a template to support drafting Articles 17 and 18 of the contracting parties’ National

reports. The group of CNS experts helped correlate each subsection of Articles 17 and 18 with

relevant IAEA safety requirements. The United States has taken into consideration the template

and its supporting information. No changes to the U.S. National report were made as a result of

this effort.

17.1 Background

The NRC’s siting responsibilities stem from the Atomic Energy Act, the Energy Reorganization

Act, and the National Environmental Policy Act. These statutes confer broad regulatory powers

on the Commission and authorize the NRC to issue regulations that it deems necessary to fulfill

its responsibilities under the acts.

As discussed in Article 7 of this report, in 1989 the NRC revised the regulatory approach

governing the licensing of new nuclear power plants. This approach provides for certified

standard designs and combined licenses that resolve design issues before construction, and

early site permits that resolve most siting and environmental issues years before construction.

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The NRC’s siting regulations are integral to protecting public health and safety and the

environment. Siting away from densely populated centers has been, and will continue to be, an

essential component of the NRC’s defense-in-depth safety philosophy (see Article 18 of this

report), which also includes multiple-barrier containment and redundant and diverse safety

systems. The primary factors that determine public health and safety are reactor design and

construction and operation of the facility. However, siting factors and criteria are important to

ensure that radiological doses from normal operation and postulated accidents will be

acceptably low, natural phenomena and manmade hazards will be properly accounted for in the

design and operation of the plant, and the human environment will be protected during the

construction and operation of the plant.

17.2 Safety Elements of Siting

This section explains the safety elements of siting. After providing a short background,

it explains the basic framework for assessing nonseismic, seismic, and other geological factors

important to siting. Finally, it discusses radiological assessments performed for initial licensing,

as a result of facility changes, and according to regulatory developments since the licensing of

all U.S. operating plants.

17.2.1 Background

The NRC’s site safety regulations consider societal and demographic factors, manmade

hazards (such as airports and dams), and physical characteristics of the site (such as

hydrological, seismological, and meteorological factors) that could affect the design or operation

of the plant. Siting requirements for applications submitted after January 10, 1997, are specified

in Subpart B, “Evaluation Factors for Stationary Power Reactor Site Applications on or after

January 10, 1997,” to 10 CFR Part 100, “Reactor Site Criteria.” Siting factors that must be

considered are specified in 10 CFR 100.20, “Factors To Be Considered When Evaluating Sites,”

and include population distributions, proximity to man-related hazards, and the physical

characteristics of the proposed site. Nonseismic siting criteria in 10 CFR 100.21, “Nonseismic

Site Criteria,” restrict occupancy around the site and establish limits on radiological releases and

dose consequences from normal operations and postulated accidents. Geologic and seismic

siting criteria in 10 CFR 100.23, “Geologic and Seismic Siting Criteria,” require evaluation of all

factors that might affect the design and operation of the proposed facility, and establish design

bases for seismic and other naturally occurring phenomena.

To meet applicable regulatory requirements, the license applicant’s safety analysis report must

describe the physical characteristics in and around the site and contain accident analyses that

are relevant to evaluating the suitability of a site. The NRC has developed numerous RGs to

provide guidance on approaches that applicants can use to address issues of site safety and

meet applicable requirements. The specifics of applicable RGs are discussed in subsequent

sections of Article 17 of this report. RG 4.7, “General Site Suitability Criteria for Nuclear Power

Stations,” Revision 2, issued in April 1998, provides a general set of safety and environmental

criteria that the NRC staff has found useful in assessing candidate site identification in specific

licensing cases. NUREG-0800, “Standard Review Plan for the Review of Safety Analysis

Reports for Nuclear Power Plants: LWR Edition,” guides the staff in reviewing the site safety

content of the applicant’s safety analysis report. Review Standard (RS)-002, “Processing

Applications for Early Site Permits,” dated May 3, 2004, identifies parts of NUREG-0800 that

apply to the review of early site permits.

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17.2.2 Assessments of Nonseismic Aspects of Siting

Siting facilities away from densely populated areas is a principal component of NRC’s

defense-in-depth safety philosophy. The evaluation of population distributions and the creation

of restricted-use zones around a proposed facility are essential elements of compliance with

regulatory requirements in 10 CFR Part 100. The dimensions of an inner “exclusion zone” and

an outer “low population zone” will depend on plant design aspects such as the reactor power

level and allowable containment leak rate, as well as the atmospheric dispersion characteristics

of the site. In addition, the distance to a population center of more than about 25,000 residents

must be at least 1.3 times the distance from the reactor to the outer boundary of the “low

population zone.” Radiological doses for postulated accidents are calculated using methods

presented in Section 17.2.4 of this report. These doses are used to evaluate the effectiveness of

the proposed restricted-use zones.

Accidents at nearby civilian or military facilities, or from nearby transportation routes, might

produce missiles, shock waves, flammable vapor clouds, toxic chemicals, or incendiary

fragments. These phenomena might affect the nuclear power plant itself or the plant operators

in a way that jeopardizes the safety of the facility. As established in 10 CFR 100.21(e), potential

hazards associated with these manmade features must be evaluated and site parameters

established such that potential hazards from such routes and facilities will pose no undue risk to

the proposed nuclear power plant. Additional information on the evaluation of these hazards is

given in RG 1.78, “Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control

Room During a Postulated Hazardous Chemical Release,” Revision 1, issued in December

2001; RG 1.91, “Evaluations of Explosions Postulated to Occur at Nearby Facilities and on

Transportation Routes Near Nuclear Power Plants,” Revision 2, issued in April 2013; and

RG 1.217, “Guidance for the Assessment of Beyond-Design-Basis Aircraft Impacts,” Revision 0,

issued in August 2011.

Radiological dose calculations must use meteorological data from the site. The site’s

atmospheric characteristics, combined with engineered safety features, must keep potential

radiological doses from postulated accidents below the regulatory limits established in

10 CFR 50.34, “Contents of Applications; Technical Information.” Acceptable approaches for

obtaining meteorological data are given in RG 1.23, “Meteorological Monitoring Programs for

Nuclear Power Plants,” Revision 3, issued in March 2007. These meteorological data also are

used in safety analyses or to establish plant design bases for phenomena such as wind loads or

impacts from tornado-generated missiles. RG 1.76, “Design-Basis Tornado and Tornado

Missiles for Nuclear Power Plants,” Revision 1, issued in March 2007, and RG 1.221,

“Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants,” Revision 0, issued

in October 2011, provide additional information on assessing these phenomena.

In siting a nuclear power plant, a highly dependable system of water supply sources should be

available under postulated occurrences of natural phenomena and site-related accident

phenomena. Considerations for water supply are addressed in RG 1.27, “Ultimate Heat Sink for

Nuclear Power Plants,” Revision 2, issued in January 1976. Because of the likely proximity to

water, many sites need to be evaluated for flood hazards from precipitation, wind, tsunami, or

human-related hazards such as dam failure events. Acceptable approaches for conducting

flood-hazard evaluations are given in RG 1.59, “Design Basis Floods for Nuclear Power Plants,”

Revision 2, issued in August 1977. RG 1.59 is being revised and is expected to be issued for

public comment in 2016.

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Site characteristics also are an important component of emergency and security planning. For

emergency planning, 10 CFR 100.21 requires the site evaluation to determine whether there are

any characteristics that would pose a significant impediment to taking protective actions to

protect the public in the event of emergency. In addition, 10 CFR 100.21 also requires that site

characteristics must allow for the development of adequate security plans and measures.

17.2.3 Assessments of Seismic and Geological Aspects of Siting

The NRC’s siting regulations listed in Section 17.2.1 of this report detail the assessments

applying to seismic and geologic aspects of siting. In simple terms, all geologic factors that

might affect the design or operation of the nuclear power plant must be assessed. Recent

developments in these geologic assessments include a performance-based approach for

determining the site-specific ground motion response spectrum and the safe-shutdown

earthquake. The performance-based approach described in RG 1.208, “A Performance-Based

Approach to Define the Site-Specific Earthquake Ground Motion,” issued in March 2007,

combines the site seismic hazard curves and seismic fragility curves for nuclear structures to

meet a specified performance target. RG 1.208 also incorporates recent developments in

seismic hazard assessment, including the use of cumulative absolute velocity filtering in place of

a lower-bound magnitude cutoff and guidance on the development of earthquake time histories,

site response analysis, and the location of the ground motion response spectrum within the soil

profile.

In 2012, a new seismic source model was completed for the central and eastern United States

(NUREG-2115, “Central and Eastern United States Seismic Source Characterization for Nuclear

Facilities,” issued in January 2012), which built upon previous seismic source models. The new

seismic source model used a Senior Seismic Hazard Analysis Committee Level 3 assessment

process to represent the center, body, and range of technically defensible interpretations of the

available data, models, and methods (NUREG/CR-6372, “Recommendations for Probabilistic

Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts,” issued in April 1997).

The updated model provides a consistent and stable basis for evaluating seismic source zones

in probabilistic seismic hazards assessments for the central and eastern United States.

The NRC reviews and certifies new and advanced reactor designs under 10 CFR Part 52,

“Licenses, Certifications, and Approvals for Nuclear Power Plants.” The seismic capacity of

the certified designs is determined independent of any specific site but capable of being sited in

most currently existing sites. Because a seismic probabilistic risk assessment requires

site-specific hazards information, the NRC requires all new and advanced reactor designs to

conduct a seismic margin analysis. This analysis evaluates the sequence-level ability of plant

structures, systems, and components to withstand an earthquake with high confidence

(i.e., 95 percent) of low probability (i.e., five percent) of failure capacities and fragilities for all

sequences leading to core damage or containment failures. A design has an acceptably low

level of seismic risk if the design-specific seismic capacity of the plant can withstand at least

1.67 times the ground motion acceleration of the design-basis safe shutdown earthquake.

17.2.4 Assessments of Radiological Consequences from Postulated Accidents

The Reactor Site Criteria Rule, 10 CFR Part 100, contains provisions for assessing whether

radiological doses from postulated accidents will be acceptably low. The NRC has issued the

following regulatory guidance for licensees to implement the current requirements for dose

assessments from postulated accidents:

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RG 1.145, “Atmospheric Dispersion Models for Potential Accident Consequence

Assessments at Nuclear Power Plants,” Revision 1, reissued in February 1983

RG 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis

Accidents at Nuclear Power Reactors,” Revision 0, issued in July 2000

RG 1.195, “Methods and Assumptions for Evaluating Radiological Consequences of

Design Basis Accidents at Light-Water Nuclear Power Reactors,” issued in May 2003

In addition to regulatory guides, the NRC staff review guidance in NUREG-0800, Chapter 15,

“Transient and Accident Analysis,” provides additional information on analysis methods

acceptable to the staff.

NUREG-1465, “Accident Source Terms for Light-Water Nuclear Power Plants,” issued in

February 1995, provides updated information on light-water reactor accident source terms. In

supplying guidance on the implementation of NUREG-1465, RG 1.183 presents one method

that may be used to show compliance with 10 CFR 50.67, “Accident Source Term,” or the

accident dose assessment requirements in 10 CFR 50.34 and 10 CFR Part 52 for new

light-water reactor licensing.

Regulations also require that, in addition to the analysis of internally initiated accident

sequences, the potential hazards associated with nearby transportation routes and industrial

and military facilities must be evaluated. Site parameters must be established so that potential

hazards from such routes and facilities will pose no undue risk to the proposed nuclear power

plant.

Although applicants analyze dose primarily to support reactor siting, licensees are required to

evaluate the potential increase in the consequences of accidents that might result from

modifying facility structures, systems, and components. Commitments (including the radiological

acceptance criteria) the applicant made during siting and documented in its final safety analysis

report remain binding until modified. A licensee must evaluate the potential consequences of

design changes against these radiological criteria to demonstrate that the changes will result in

a design that still conforms to the regulations and commitments. If the consequences increase

more than minimally, as outlined in 10 CFR 50.59, “Changes, Tests and Experiments,” or

require a change to the technical specifications, as discussed in Article 14 of this report, the

licensee must obtain NRC approval before implementing the proposed modification.

Requirements in 10 CFR 50.67 allow licensees to use an alternative source term in place of the

accident source term used in the original licensing and siting of the operating facility.

If a licensee has not implemented the alternative source term approach in 10 CFR 50.67,

RG 1.195 provides an acceptable approach for assessing the potential significance of changes

to plant design and licensing bases. Thus, RG 1.195 provides an alternative approach to the

dose assessment methods in RG 1.3, “Assumptions Used for Evaluating the Potential

Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors,”

Revision 2, dated June 1974, and RG 1.4, “Assumptions Used for Evaluating the Potential

Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors,”

Revision 2, dated June 1974.

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The NRC has applied the 1996 revision to 10 CFR Part 100, along with the alternative source

term as described in RG 1.183, in its design certification review for a passive light-water reactor,

the AP600 design. More recently, the agency has applied the practice to the AP1000 and

Economic Simplified Boiling-Water Reactor designs with similar results and is applying it for all

contemplated light-water reactor design certification application reviews, including the U.S.

Evolutionary Power Reactor, the Mitsubishi Heavy Industries, Ltd.’s U.S. Advanced

Pressurized-Water Reactor, and the APR1400. For other than light-water reactor designs and

advanced reactors, applicants will have to describe their rationale for an appropriate accident

source term characterization that will be subject to NRC independent review.

The industry continues to explore the use of the alternative source term in implementing

cost-beneficial licensing actions at operating reactors. Some of these applications resulted in

improved safety equipment reliability calculations and reduced occupational exposures,

providing the licensee regulatory margin. Since the issuance of 10 CFR 50.67 in 1999, the

majority of operating reactor licensees requested either full implementation of the alternative

source term or selective implementation for certain regulatory applications. Operating plant

licensees also have used the alternative source term to analyze the adequacy of certain

engineered safety features in meeting the operability requirements in their operating reactor

technical specifications.

17.3 Environmental Protection Elements of Siting

This section explains the environmental protection elements of siting. It covers the governing

documents and site approval process. Since the first operating plants in the United States

received licenses, issues have arisen that must be considered in siting reviews for new facilities.

This section explains the effect of these issues.

17.3.1 Governing Documents and Process

The environmental protection elements of siting consist of the plant’s demands on the

environment (e.g., water use and effects of construction and operation). These elements are

addressed in 10 CFR Part 51, “Environmental Protection Regulations for Domestic Licensing

and Related Regulatory Functions,” which implements the National Environmental Policy Act

consistent with the NRC’s statutory authority and reflects the agency’s policy to voluntarily apply

the regulations of the President’s Council on Environmental Quality, subject to certain

conditions. Integrating environmental reviews into its routine decisionmaking, the NRC

considers environmental protection issues and alternatives before taking any action that may

significantly affect the human environment.

The site approval process leading to the construction or operation of a nuclear power plant

requires the NRC to prepare an environmental impact statement. The updated and revised

environmental standard review plans (NUREG-1555, “Standard Review Plans for Environmental

Reviews for Nuclear Power Plants,” issued in March 2000) guide the staff’s environmental

reviews for a range of applications, including site reviews for construction permits and operating

licenses under 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities,” for

early site permits under 10 CFR Part 52, Subpart A, “Early Site Permits,” and for combined

licenses under 10 CFR Part 52, Subpart C, “Combined Licenses,” when the application does not

reference an early site permit. The NRC issued updates to review practices in 2007 and 2010 to

reflect experience gained from early site permit reviews, account for the changes resulting from

the amendment to the limited work authorization rule (discussed later in this section), and

include consideration of the environmental effects of greenhouse gas emissions and climate

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change. On September 3, 2014, the NRC issued COL/ESP-ISG-026, “Interim Staff Guidance on

Environmental Issues Associated with New Reactors,” to encompass the 2007 and 2010

updates. COL/ESP-ISG-026 will be incorporated into the next revision of the NUREG-1555.

Article 19 of this report discusses these governing documents and processes for combined

license reviews.

Environmental standard review plans are also appropriate for environmental reviews of

applications for combined licenses under 10 CFR Part 52, Subpart C, when the applications

reference an early site permit. Reviews of early site permit applications are limited because the

reviews focus on the environmental effects of nuclear power plant construction and operation

that have characteristics that fall within the postulated site parameters and because the reviews

need not assess benefits (e.g., the need for power) or alternative energy sources. The

environmental information in applications for combined licenses that reference an early site

permit is limited to (1) information to demonstrate that the design of the facility falls within the

parameters specified in the early site permit, (2) new and significant information on issues

previously considered in the early site permit proceeding, and (3) any significant environmental

issue not considered in any previous proceeding on the site or design.

The environmental standard review plans in Supplement 1 to NUREG-1555 guide the staff’s

environmental review for license renewal applications under 10 CFR Part 54, “Requirements for

Renewal of Operating Licenses for Nuclear Power Plants.” Article 14 of this report discusses

the license renewal process in more detail.

Several other NRC actions on siting and site suitability require environmental reviews, including

issuance of limited work authorizations (10 CFR 50.10(e); 10 CFR 52.25, “Extent of Activities

Permitted”; and 10 CFR 52.91, “Authorization to Conduct Site Activities”), early partial decisions

(10 CFR 2.600, “Scope of Subpart,” in Subpart F, “Additional Procedures Applicable to Early

Partial Decisions on Site Suitability Issues in Connection with an Application for a Permit to

Construct Certain Utilization Facilities,” of 10 CFR Part 2, “Agency Rules of Practice and

Procedure”), and preapplication reviews of site suitability issues (Appendix Q, “Pre-Application

Early Review of Site Suitability Issues,” to 10 CFR Part 50).

With its 2007 amendment to the limited work authorization licensing framework (10 CFR 50.10,

“License Required, Limited Work Authorization”), the Commission limited its authority to

construction activities that have a “reasonable nexus to radiological health and safety or

common defense and security” and defined “construction” within the context of its authority. The

effect of this change is not restricted to limited work authorizations. Other activities related to

building the plant that do not require NRC approval (but may require a permit from other

regulatory agencies) may occur before, during, or after NRC-authorized construction activities.

These activities, called “preconstruction” in 10 CFR 51.45(c), may be regulated by other local,

State, Tribal or Federal agencies. On September 12, 2008, the NRC and the U.S. Army Corps

of Engineers signed an updated memorandum of understanding to enhance the effectiveness of

reviews of nuclear power plant license applications that would require multiple Federal permits

under separate statutes. The NRC and the U.S. Army Corps of Engineers are participating as

cooperating agencies in the preparation of many environmental impact statements.

17.3.2 Other Considerations for Environmental Reviews

The NRC’s environmental standard review plan was first published in the 1970s. Since the

1970s, many changes to the regulatory environment have affected both the NRC and applicants

seeking site approvals. These include new environmental laws and regulations, changes in

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policies and procedures resulting from decisions of courts and administrative hearing boards,

and changes in the types of authorizations, permits, and licenses issued by the NRC. This

section highlights some of these changes and subsequent revisions to environmental standard

review plans.

In the late 1980s, the NRC issued regulations for an alternative licensing framework to

10 CFR Part 50, which required a construction permit followed by an operating license. The

framework in 10 CFR Part 52 introduced the concepts of approving nuclear power plant designs

independent of sites and approving sites independent of these designs, and then efficiently

linking these approvals to approve construction and operation of the facility. As discussed in the

introduction of this report, the NRC has approved five early site permits and four combined

license applications (for a total of seven licenses) under 10 CFR Part 52 and is actively

conducting additional siting and new plant licensing reviews.

As part of the revisions to the licensing framework, the NRC issued RS-002, which incorporates

the environmental guidance in NUREG-1555, the environmental standard review plan, and the

outcome of interactions with stakeholders. In addition, in 2007, the NRC revised 10 CFR Part 52

to reflect experience gained in its use and to provide guidance on the preparation of combined

license applications. As part of that rulemaking the NRC issued RG 1.206, “Combined License

Applications for Nuclear Power Plants,” in June 2007, which includes guidance on the

assessment of environmental issues.

Since 1984, the NRC has considered the environmental impacts of spent nuclear fuel storage

after the licensed lifetime of reactor operations to be a generic issue that is best addressed

through rulemaking. Several technical concerns were identified in the analyses supporting the

regulation that addressed this issue (10 CFR 51.23, “Temporary Storage of Spent Fuel after

Cessation of Reactor Operation—Generic Determination of No Significant Environmental

Impact”), which resulted in the U.S. Court of Appeals vacating this regulation in June 2012. The

NRC developed an environmental impact statement that addresses the technical concerns

raised by the Court and provides the National Environmental Policy Act analyses needed to

support a revision to 10 CFR 51.23. In September 2014, the NRC issued a revised rule at

10 CFR 51.23 and its associated NUREG-2157, “Generic Environmental Impact Statement for

Continued Storage of Spent Nuclear Fuel.” The revised rule adopts the generic impact

determinations made in NUREG-2157 and codifies the NRC’s generic determinations regarding

the environmental impacts of continued storage of spent nuclear fuel beyond a reactor’s

operating license.

As described in previous U.S. National Reports, other relevant regulatory developments include

the following:

Presidential Executive Order 12898, “Federal Actions To Address Environmental Justice

in Minority and Low-Income Populations,” issued in February 1994, which instructed

Federal agencies to make “environmental justice” part of each agency’s mission by

addressing disproportionately high and adverse human health or environmental effects

of Federal programs, policies, and activities on minority and low-income populations

the 1978 decision on the Tennessee Valley Authority Yellow Creek Nuclear Plant, which

determined that the authority of the NRC is limited in matters that are expressly assigned

to U.S. Environmental Protection Agency (EPA)

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changes in the economic regulation of utilities that have expanded the options to be

addressed in considering the need for power in environmental impact statements

design alternatives to mitigate the consequences of severe accidents

EPA rules about cooling water intake structures (Federal Water Pollution Control Act,

Section 316(b))

increased emphasis on greenhouse gases and climate change impacts

EPA’s Clean Power Plan rule, dated October 23, 2015 (80 FR 64966)

17.4 Re-evaluation of Site-Related Factors

Although operating nuclear power plants are not reevaluated periodically for site-related factors,

the continued safety of nuclear plants and the adequate protection of a licensed plant are

imperative. If there is a significant change in any hazard to an already licensed nuclear plant,

then the NRC will determine whether a backfit action under 10 CFR 50.109, “Backfitting,” is

necessary. The NRC will always require the backfitting of a nuclear power plant if it determines

that such regulatory action is necessary to ensure that the plant provides adequate protection to

the health and safety of the public and is in accordance with the common defense and security.

In response to the Fukushima accident, the NRC used its existing regulatory processes to

request that licensees reevaluate the seismic and flooding hazards at their sites using

present-day regulatory guidance and methodologies and, if necessary, perform a risk

evaluation. The results of these reevaluations will be used to determine whether additional

regulatory actions are necessary to ensure plants are adequately protected from seismic and

flooding events.

Periodic seismic requalification of equipment is not necessary, because databases are available

for equipment already qualified or tested to fragility levels. IEEE standard 344, “IEEE

Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power

Generating Stations,” provides criteria to determine the appropriate level of equipment

ruggedness. Using this standard, a licensee is able to determine whether equipment needs to

be requalified or replaced.

17.5 Consultation with Other Contracting Parties To Be Affected by the Installation

At this time, the NRC does not have any specific international arrangements with neighboring

countries for siting new builds. The agency’s current arrangements with its Canadian and

Mexican regulatory counterparts for the exchange of information and experience serves as the

mechanism for cooperative dialogue.

17.6 Vienna Declaration on Nuclear Safety

Consistent with the goals of the Vienna Declaration on Nuclear Safety, NRC regulatory

requirements for siting have long reflected a defense-in-depth approach that requires all natural

phenomena and manmade hazards at a potential site to be identified and properly accounted

for in the siting, design and operation of the plant. This approach avoids siting of plants at

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problematic sites, and ensures that radiological doses from normal operation and postulated

accidents will be acceptably low. As discussed in Section 17.2.1 of this report, siting regulations

are implemented primarily through requirements in 10 CFR Part 100, Supbart B. In addition, the

General Design Criteria in Appendix A to 10 CFR Part 50 further embody the defense-in-depth

philosophy. General Design Criterion 2, “Design Bases for Protection against Natural

Phenomena,” requires that the plant’s design basis reflects appropriate consideration of severe

natural phenomena, including effects of normal and accident conditions. Integration of

Criterion 2 with the other General Design Criteria provides assurance that the NRC’s approach

to siting is consistent with the safety goals of the Vienna Declaration. This defense-in-depth

approach is discussed further in Section 18.1 of this report.

The safety goals of the Vienna Declaration on Nuclear Safety are consistent with the NRC’s

siting regulations. These regulations favor siting of nuclear power plants in areas of relatively

low population density, with restricted-use zones around the plant that reflect the design

characteristics of the plant (e.g., power level) and the atmospheric dispersion characteristics of

the site. The plant’s design and operations must be protected from the effects of accidents at

nearby civilian or military facilities, or from nearby transportation routes. Siting regulations also

contain provisions to ensure that radiological doses from postulated accidents will be acceptably

low. In addition, all natural phenomena that might affect the design or operation of the plant

must be appropriately characterized, so that the plant’s design basis appropriately considers the

most severe natural phenomena at the site, with sufficient margin for the limited accuracy,

quantity, and period of time in which historical data have been accumulated. By taking this

approach to protection against external hazards, the NRC’s regulations effectively discourage

the siting of new plants at locations where there is an unacceptable risk of long-term offsite

contamination or large releases requiring long-term protective actions.

If there is a significant change in any hazard to an already licensed nuclear plant, then the NRC

will determine whether a backfit action under 10 CFR 50.109 is necessary. The NRC will always

require the backfitting of a nuclear power plant if it determines that such regulatory action is

necessary to ensure that the plant provides adequate protection to the health and safety of the

public and is in accordance with the common defense and security. In response to the

Fukushima accident, the NRC used its existing regulatory processes to request that licensees

reevaluate the seismic and flooding hazards at their sites using present-day regulatory guidance

and methodologies and, if necessary, to perform a risk evaluation. The results of these

reevaluations will be used to determine whether additional regulatory actions are necessary.

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ARTICLE 18. DESIGN AND CONSTRUCTION

Each Contracting Party shall take the appropriate steps to ensure that:

(i)

the design and construction of a nuclear installation provides for several reliable

levels and methods of protection (defense in depth) against the release of

radioactive materials, with a view to preventing the occurrence of accidents and to

mitigating their radiological consequences should they occur

(ii)

the technologies incorporated in the design and construction of a nuclear

installation are proven by experience or qualified by testing or analysis

(iii)

the design of a nuclear installation allows for reliable, stable, and easily

manageable operation, with specific consideration of human factors and the

man-machine interface

This section explains the defense-in-depth philosophy and how it is embodied in the general

design criteria of U.S. regulations. It explains how applicants meet the defense-in-depth goals

and how the U.S. NRC reviews applications and conducts inspections before issuing licenses to

ensure that this philosophy is implemented in practice. Next, this section discusses measures

for ensuring that the applications of technologies are proven by experience or qualified by

testing or analysis. This section discusses requirements for reliable, stable, and easily

manageable operation, specifically considering human factors and the man-machine interface,

lessons learned from Fukushima, and addresses the Vienna Declaration on Nuclear Safety,

which was issued in February 2015. Article 12 of this report also provides information on the

human factors obligations.

Finally, the United States reviewed the results of the CNS 2015-2016 consultancy meetings that

developed a template to support drafting Articles 17 and 18 of the contracting parties’ National

reports. The group of CNS experts helped correlate each subsection of Articles 17 and 18 with

relevant IAEA safety requirements. The United States has taken into consideration the template

and its supporting information. No changes to the U.S. National report were made as a result of

this effort.

18.1 Defense-In-Depth Philosophy

This section explains the defense-in-depth philosophy followed in regulatory practice, governing

documents and regulatory process for designing and constructing a nuclear power plant. It also

discusses relevant experience and examples.

18.1.1 Governing Documents and Process

The defense-in-depth philosophy, as applied in regulatory practice, requires that nuclear plants

contain a series of independent, redundant, and diverse safety systems. The physical barriers

for defense-in-depth in a light-water reactor are the fuel matrix, the fuel rod cladding, the primary

coolant pressure boundary, and the containment. The levels of protection in defense-in-depth

are (1) a conservative design, quality assurance, and safety culture, (2) control of abnormal

operation and detection of failures, (3) safety and protection systems, (4) accident management,

including containment protection, and (5) emergency preparedness.

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Appendix A to 10 CFR Part 50 embodies the defense-in-depth philosophy. General design

criteria cover protection by multiple fission product barriers, protection and reactivity control

systems, fluid systems, containment design, and fuel and radioactivity control. The NRC staff

amplified its defense-in-depth philosophy in RG 1.174, which provides guidance on using a PRA

in risk-informed decisions on plant-specific changes. The general design criteria establish the

minimum requirements for the principal design criteria, which in turn establish the necessary

design, fabrication, construction, testing, and performance requirements for SSCs that are

important to safety.

To ensure that a plant is properly designed and built as designed, that proper materials are

used in construction, that future design modifications are controlled, and that appropriate

maintenance and operational practices are followed, a good quality assurance program is

needed. To meet this need, General Design Criterion 1 of Appendix A to 10 CFR Part 50, and

its implementing regulatory requirements specified in Appendix B to 10 CFR Part 50, establish

quality assurance requirements for all activities affecting the safety-related functions of the

SSCs.

In accordance with the two-step licensing process under 10 CFR Part 50, an applicant for a

construction permit must present the principal design criteria for a proposed facility in its

preliminary safety analysis report. For guidance in writing a safety analysis report, the applicant

may use RG 1.70. The safety analysis report also must contain design information for the

proposed reactor and comprehensive data on the proposed site. The report must also discuss

various hypothetical accident situations and the safety features to prevent accidents or, if

accidents occur, to mitigate their effects on both the public and the facility’s employees.

After obtaining a construction permit under 10 CFR Part 50, the applicant must submit a final

safety analysis report to support an application for an operating license, unless it submitted the

report with the original application. This report should give the details of the final design of the

facility, plans for operation, and procedures for coping with emergencies. The preliminary and

final safety analysis reports are the principal documents the applicant provides for the staff to

determine whether the proposed plant can be built and operated without undue risk to the health

and safety of the public. Current applications to build new nuclear power plants have been

submitted using the combined license process under 10 CFR Part 52, although applicants are

not precluded from using the two-step licensing process under 10 CFR Part 50. Applications

submitted under 10 CFR Part 52 must meet all of the 10 CFR Part 50 requirements as well as

the applicable requirements referenced in other regulations (e.g., 10 CFR Part 20, Part 30,

“Rules of General Applicability to Domestic Licensing of Byproduct Material,” Part 40, “Domestic

Licensing of Source Material,” Part 70, Part 73, and Part 100). The NRC issued guidance for the

content and format of a combined license application in RG 1.206. A significant difference in the

10 CFR Part 52 process is that the final safety analysis report must be submitted before

authorization is granted to begin construction. Article 19 of this report describes the combined

license review process.

The NRC staff reviews safety analysis reports according to NUREG-0800 to ensure that the

applicant has satisfied the general design criteria and other applicable regulations. The staff

reviews each application to determine whether the plant design meets the Commission’s

regulations (10 CFR Parts 20, 50, 73, and 100). These reviews include, in part, the

characteristics of the site. In addition, each application for a nuclear installation must include a

comprehensive environmental report that provides a basis for evaluating the environmental

impact of the proposed facility. RG 4.2, Revision 2, gives applicants information on writing

environmental reports. The NRC staff reviews the environmental reports according to

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NUREG-1555. In reviewing an application, the staff, supported by outside experts, conducts

independent technical studies to review certain safety and environmental matters. The staff

states its conclusions in an environmental impact statement and a safety evaluation report,

which it may update before granting the license. Under the two-step licensing process in

10 CFR Part 50, the NRC does not issue an operating license until construction is complete and

the Commission makes the findings required under 10 CFR 50.57, “Issuance of Operating

License.” For combined license applications submitted under 10 CFR Part 52, the Commission

must make a finding in accordance with 10 CFR 52.97, “Issuance of Combined License,” to

issue the combined license. With issuance of the combined license, construction of the facility

may begin; however, the Commission must make a finding in accordance with

10 CFR 52.103(g) that all acceptance criteria in the combined license are met to authorize

operation of the facility.

The NRC monitors nuclear power plant construction to ensure compliance with the agency’s

regulations to protect public health and safety and the environment. The NRC has developed an

inspection program for nuclear plants licensed under 10 CFR Part 52. The new inspection

program revises the 10 CFR Part 50 Construction Inspection Program. It incorporates

inspections, tests, analyses, and acceptance criteria (ITAAC) from 10 CFR Part 52, as well as

lessons learned from the inspection program used in the previous construction era (1970-1980).

It also considers modular construction at remote locations.

Before the combined license is issued, the NRC inspection program verifies that the applicant’s

quality assurance program is adequately implemented and that any pre-construction activities

meet specified requirements in Appendix B to 10 CFR Part 50. Inspection Manual Chapter

2502, “Construction Inspection Program: Pre-Combined License (Pre-COL) Phase,” dated

December 13, 2010, lists the inspections for this phase.

The NRC also interacts with manufacturers and suppliers of safety-related components through

the NRC vendor inspection programs that inspect compliance with quality assurance and defect

reporting requirements. Vendor inspections are conducted at vendor shops principally to

examine whether the vendor has been complying with Appendix B to 10 CFR Part 50, as

required by procurement contracts with applicants and licensees. Inspection Manual

Chapter 2507, “Vendor Inspections,” dated October 3, 2013, lists inspections for vendors.

During construction, NRC inspectors sample the spectrum of the applicant’s activities related to

the ITAAC in the combined license to confirm that the applicant is adhering to quality and

program requirements. Inspection Manual Chapter 2503, “Construction Inspection Program:

Inspections of Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Related Work,”

dated July 5, 2012, describes these inspections. The NRC staff will verify successful ITAAC

completion based on these inspections and will review all ITAAC closure notifications from the

licensee. The NRC will publish notices in the Federal Register of completed ITAAC.

In addition to inspections of ITAAC related work, the NRC inspection program addresses

inspections of programs that support construction activities (e.g., quality assurance and

preoperational testing) as well as programs that support eventual operation of the facility

(e.g., fire protection, security, training, radiation protection, and startup testing), and programs

that enable the transition of the organization from construction to power operations. Inspection

Manual Chapter 2504, “Construction Inspection Program—Inspection of Construction and

Operational Programs,” dated October 24, 2012, lists inspections for this phase.

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18.1.2 Experience

The agency’s recent review of the Watts Bar Nuclear Plant, Unit 2, operating license application

is an example of how the NRC design and construction process for a 10 CFR Part 50

application (described in Section 18.1.1) is currently implemented.

The Watts Bar Nuclear Plant, owned by Tennessee Valley Authority, is in southeastern

Tennessee. The site has two Westinghouse designed PWRs. Watts Bar, Unit 1, received a full-

power operating license in early 1996, and it was the last new power reactor licensed in the

United States under 10 CFR Part 50. Tennessee Valley Authority stopped construction at Watts

Bar, Unit 2, in the mid-1980s; however, in 2007, Tennessee Valley Authority notified the NRC of

its plans to resume construction.

In its regulatory framework for the completion of Unit 2, the Commission approved

(SRM-SECY-07-0096, “Possible Reactivation of Construction and Licensing Activities for the

Watts Bar Nuclear Plant, Unit 2,” dated July 25, 2007) a licensing review approach that uses the

current licensing basis for Watts Bar, Unit 1, as the reference basis for review and licensing of

Unit 2. This approach ensures safety while preserving design and operational consistency

between the units. However, considering the construction status of the unit, the NRC

encouraged Tennessee Valley Authority to adopt updated standards wherever feasible and look

for opportunities to resolve any generic safety issues in which the unirradiated state of Unit 2

makes the issue easier to resolve before plant operation. The NRC’s operating license review

included safety design, environmental review, and inspection of construction activities.

Tennessee Valley Authority updated its initial 1970s operating license application. The NRC

published a notice of the updated application for an operating license in the Federal Register to

provide public notice and an additional opportunity for a hearing. Tennessee Valley Authority

submitted its final supplemental environmental impact statement for the completion and

operation of Watts Bar, Unit 2. The staff published its draft supplemental environmental

statement for completion and operation of Watts Bar, Unit 2, in late 2011 for public comment.

The final environmental statement was published in May 2013. The NRC also held public

outreach meetings in the vicinity of the site to inform the public about its licensing and inspection

activities, including how the public can monitor and participate in the licensing process.

To complete its licensing review in a timely and comprehensive manner, the NRC established

dedicated teams at both its headquarters and regional offices for review and inspection of the

Unit 2 activities. The staff independently reviewed Tennessee Valley Authority’s regulatory

framework and documented its results in a safety evaluation report, NUREG-0847,

Supplement 21, “Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant,

Unit 2,” issued in February 2009. The review identified the items that needed to be completed

before issuance of an operating license. The staff published five additional supplements to

NUREG-0847 documenting its review of the open items laid out in Supplement 21.

The NRC Region II office performed the necessary inspections and oversight activities for Watts

Bar, Unit 2. It developed Inspection Manual Chapter 2517, “Watts Bar Unit 2 Construction

Inspection Program,” issued in February 2008, to provide guidance for these inspection

activities. The NRC Region II office examined historical inspection records, employee concerns,

operating experience, scope of new or rework, and construction deficiency reports. On January

30, 2015, the NRC issued IP 94302, “Status of Watts Bar Unit 2 Readiness for an Operating

License.” The objective of IP 94302 is to inform the Director of the Office of Nuclear Reactor

Regulation of the following items:

(1) completion of inspections necessary to support the

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findings required by 10 CFR 50.57(a)(1), 50.57(a)(2), and 50.57(a)(3)(ii), (2) any incomplete

inspections or open items at the time that item (1) is communicated, and (3) any significant

issues in the construction or testing that could affect the conclusions associated with item (1).

The Region II Administrator provided notification of the above items to the Director of the Office

of Nuclear Reactor Regulation on October 15, 2015.

Based on the findings of the NRC staff’s review of the operating license application as

documented in NUREG-0847 and its supplements, the NRC staff’s inspection activities, the

IP 94302 results, and the May 2013 final environmental impact statement, the operating license

for Watts Bar Nuclear Plant, Unit 2, was issued on October 22, 2015, and is valid for 40 years.

Initial criticallity at Watts Bar, Unit 2, was achieved on May 23, 2016.

18.2 Technologies Proven by Experience or Qualified by Testing or Analysis

In 10 CFR 50.43(e), the NRC requires that new technologies are demonstrated to be proven.

This rule requires demonstration of new technologies through analysis, appropriate test

programs, experience, or a combination thereof. In its safety analysis reports for the AP600 and

AP1000 standard plant designs, Westinghouse used separate effects tests, integral systems

tests, and analyses to demonstrate that its passive safety systems will perform as predicted.

Section 14.2 of this report discusses the qualification of currently used technologies.

18.3 Design for Reliable, Stable, and Easily Manageable Operation

The NRC specifically considers human factors and the human-system interface in the design of

nuclear installations. For safety analysis reports, the NRC reviews the human factors

engineering design of the main control room and the control centers outside of the main control

room. Article 12 of this report also discusses human factors.

18.3.1 Governing Documents and Process

To support its reviews of the human factors engineering issues associated with the certification

and licensing of new plant designs, the NRC uses NUREG-0800, Chapter 18, Revision 2, and

NUREG-0700, “Human-System Interface Design Review Guidelines,” Revision 2, issued in

May 2002. The NRC used NUREG-0711, “Human Factors Engineering Program Review

Model,” Revision 2, issued in February 2004, for evaluating the design of next-generation main

control rooms listed in Section 18.1.2.2. In November 2012, the NRC issued NUREG-0711,

Revision 3, “Human Factors Engineering Program Review Model,” to address lessons learned

from these reviews. NUREG-0800, Section 14.3.9, “Human Factors Engineering - Inspections,

Tests, Analyses, and Acceptance Criteria,” issued in March 2007, provides additional guidance.

The NRC has recently initiated work to update these review guidelines. Additionally, the NRC

developed guidance for reviewing combined license applications, RG 1.206, which includes

sections that address the human factors engineering review of combined license applications.

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18.3.2 Experience

The NRC is actively reviewing new plant designs and combined license applications.

18.3.2.1 Human Factors Engineering

The NRC has completed the evaluation of the human factors engineering sections of the design

certification reviews of the ESBWR and AP1000 applications as well as the Vogtle, V.C.

Summer, Fermi, and South Texas Project combined license applications. Reviews continue on

the U.S. Advanced Pressurized-Water Reactor (US APWR) and the APR1400 certification

submittals and on the remaining combined license applications. The NRC’s human factors

engineering reviews for design certification applications principally focus on evaluating

implementation plans for the design of the control facilities to ensure that the design process will

be carried out consistent with state-of-the-art human factors principles. The NRC will verify

acceptable implementation of these plans through specified ITAAC (i.e., design acceptance

criteria).

The completed staff reviews identified the following weaknesses in the previous revision of

NUREG-0711:

The “human reliability analysis” element did not address manual actions credited in the

Standard Review Plan, Chapters 7 and 15.

The technical support facility, emergency operating facility, and local control stations are

included in the human factors engineering program scope, but it was unclear which

elements applied to them.

The “verification and validation” element was complex and created confusion on how

performance measurement criteria were meant to be applied.

The content of Implementation Plans and Results Summary Reports were not

adequately defined, resulting in insufficient detail in applications, confusion on which

design products could be deferred, and difficulty in establishing ITAACs with sufficient

scope.

NUREG-0711, Revision 3, was issued to address these issues.

18.3.2.2 Digital Instrumentation and Controls

Chapter 7 of NUREG-0800 provides guidance to the NRC staff in reviewing the instrumentation

and control design of the nuclear power reactors. This guidance assists the staff in determining

whether the design complies with the applicable regulatory requirements and whether the

applicant has demonstrated with reasonable assurance that the design provides adequate

protection of public health and safety. All of the new reactor designs contain highly integrated

digital instrumentation and control systems, which present issues that are not relevant to analog

systems. Examples of these issues include:

A common-cause failure attributable to software errors was not possible with analog

systems. This possible failure mode may require consideration of diversity and

defense-in-depth in the application of digital instrumentation and control systems.

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Digital system architectures raise issues such as interchannel communication,

communication between nonsafety and safety systems, and cyber security that must be

addressed to ensure that public safety is preserved.

Highly integrated control room designs with safety and nonsafety displays and controls

are the norm for new reactor designs. Human factors design and quality assurance

during all phases of software development, control, and validation and verification are

critical.

The NRC developed several interim staff guidance documents for review of new and innovative

digital instrumentation and control systems found in new reactor designs. The guidance also

provided the industry with the expectations and criteria the staff uses to evaluate their designs

and determine compliance with NRC regulations. The staff has been using this guidance, along

with other existing guidance such as NUREG-0800, in its review of applications for design

certifications and combined licenses. The staff has incorporated some of the interim staff

guidance into formal NRC staff guidance in NUREG-0800 and associated RGs. All interim staff

guidance documents on digital instrumentation and control can be found at

http://www.nrc.gov/reading-rm/doc-collections/isg/digital-instrumentation-ctrl.html.

The staff has completed its safety reviews of the instrumentation and control systems for the

AP1000, ESBWR, and Advanced Boiling-Water Reactor (ABWR) reactor designs as well as

those for the Fermi, Unit 3, and South Texas Project, Units 3 and 4, combined licenses. The

staff is in the process of reviewing the instrumentation and control design for the US APWR and

APR1400 reactor designs and multiple combined license applications. The staff also has

initiated the instrumentation and control ITAAC inspection activities for the AP1000 combined

licenses, and preapplication activities on a small modular reactor design.

To prepare for the review of applications for small modular reactor design certifications and

combined licenses, the NRC staff is developing a design-specific review standard. This design-

specific review standard chapter reflects a number of important lessons the staff learned when

using NUREG-0800 to review new large light water reactor designs. The staff has incorporated

the following lessons learned into this guidance to:

Emphasize fundamental instrumentation and control design principles such as

independence, redundancy, determinism, and diversity and defense-in-depth, as derived

through design and analysis, such as hazard analysis, to prevent loss or impairment of a

safety function. This guidance aims to address all of the significant aspects of the

instrumentation and control design in a unified manner through this framework.

Reflect an integrated instrumentation and control design using digital technology, which

is common in new and advanced reactor designs. In addition, the topical areas most

significant to safety are discussed first. The NUREG-0800 guidance is system-based;

therefore, many regulatory requirements and their supporting guidance are repeated in

multiple subsections. The approach of this design-specific review standard minimizes

such repetition.

Introduce the use of an integrated hazards analysis approach, which is a

well-established safety engineering practice. This approach consolidates the various

methods discussed in NUREG-0800 and provides a consistent, comprehensive, and

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systematic way to address the potential hazards associated with instrumentation and

control systems in a unified framework.

Address various new sources, such as the Multinational Design Evaluation Program

common positions and lessons learned from other countries.

Encompass all relevant branch technical positions contained in the current

NUREG-0800. This guidance also clarifies the interface between the instrumentation

and control area and other disciplines, such as human factors engineering, quality, and

reactor systems.

The NRC participates in the Multinational Design Evaluation Program, an international assembly

of nuclear regulators addressing common issues with the licensing of new reactors. The NRC

chairs the Digital Instrumentation and Control Issue-Specific Working Group, which is looking at

ways to harmonize requirements, standards, and guidance for instrumentation and control. The

NRC is also working with the US EPR and ABWR instrumentation and control technical expert

subgroups, which are an international collaboration of regulatory agencies engaged in review of

the US EPR and ABWR instrumentation and control designs. The Multinational Design

Evaluation Program allows the NRC to share digital instrumentation and control information to

support regulatory infrastructure improvements and licensing decisions.

18.3.2.3 Cyber Security

After September 11, 2001, the NRC issued two security-related orders, NRC Order EA-02-026,

“Issuance of Order for Interim Safeguards and Security Compensatory Measures,” issued in

February 2002, and NRC Order EA-03-086, “Issuance of Order Requiring Compliance with

Revised Design Basis Threat for Operating Power Reactors,” issued in April 2003, which require

power reactor licensees to implement measures to enhance cyber security. These security

measures required immediate identification and assessment of computer-based systems

deemed to be critical to the operation and security of the facility. From 2006 through

February 2009, cyber security design reviews were performed solely based on the guidance in

RG 1.152, Revision 2, “Criteria for Use of Computers in Safety Systems of Nuclear Power

Plants.”

Subsequently, in March 2009, the NRC issued a new rule on cyber security, 10 CFR 73.54,

“Protection of Digital Computer and Communication Systems and Networks,” and RG 1.152 was

revised to remove cyber security guidance. The cyber security rule requires operating power

reactor licensees to provide high assurance that nuclear power plants’ safety-related,

important-to-safety, security, and emergency preparedness functions are protected from cyber

attacks up to and including the design-basis threat. To meet the cyber security rule

requirements, operating power reactor licensees had to submit a cyber security plan, including a

proposed implementation schedule with interim milestones, to the NRC for review and approval

by November 23, 2009, and combined license applicants are required to submit a plan in

accordance with their overall license application. All operating nuclear power plant licensees

met that submission deadline, and the NRC reviewed and approved all the plans. Essential

elements of a plan include describing the process for finding critical digital assets, describing the

defensive model (i.e., protective strategy), referencing a comprehensive set of security controls,

and describing the process for addressing each control. The cyber security plan also must

acknowledge a commitment to maintain the cyber security program and provide adequate

documentation of how that will be accomplished.

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In January 2010, the NRC published RG 5.71, “Cyber-Security Programs for Nuclear Facilities,”

which provides implementation guidance to licensees and applicants on an acceptable method

for satisfying the requirements of 10 CFR 73.54. This guidance describes an acceptable method

licensees can follow to address potential security vulnerabilities in each life-cycle phase of

critical digital assets that perform safety-related, important-to-safety, security, and emergency

preparedness functions. It is equally applicable to both combined license applicants and the

current fleet of operational reactors. The guidance embodies recommended practices from

standards organizations such as the International Society of Automation, the Institute of

Electrical and Electronics Engineers, the National Institute of Standards and Technology, and

the U.S. Department of Homeland Security.

In 2010, the NRC and the North American Electric Reliability Corporation entered into a 5-year

memorandum of understanding to address nuclear plant cyber security roles, responsibilities,

and areas of coordination between the two organizations. The 5-year memorandum of

understanding with the North American Electric Reliability Corporation was renewed in 2015.

Subsequent to the memorandum of understanding with the North American Electric Reliability

Corporation, the NRC determined that 10 CFR 73.54 should be interpreted to include SSCs that

have a nexus to radiological health and safety at NRC-licensed nuclear power plants. The

Federal Energy Regulatory Commission and the North American Electric Reliability Corporation

found this policy decision acceptable and they, likewise, found the NRC’s regulatory framework

sufficient to meet the North American Electric Reliability Corporation cyber security

requirements for power generation plants. In accordance with the memorandum of

understanding, the staff will continue to coordinate with the North American Electric Reliability

Corporation to share relevant operating experience and other related technical information. In

2010, the NRC entered into a 5-year memorandum of agreement with the Federal Energy

Regulatory Commission to facilitate a continuing and cooperative relationship and the exchange

of experience, information, and data related to the reliability of the U.S. bulk electricity supply.

The 5-year memorandum of agreement was renewed in 2015.

The NRC has developed an oversight program for cybersecurity that includes an inspection

program, inspector training, and a process for evaluating the significance of inspection findings.

This was accomplished collaboratively with stakeholders, including members of industry and

representatives from the U.S. Department of Homeland Security, the Federal Energy Regulatory

Commission, and the National Institute of Standards and Technology. The NRC completed

inspection activities related to the interim milestones in calendar year 2015. In 2016, the NRC,

along with industry, is preparing for full implementation inspection activities that will begin in

calendar year 2017.

18.4 New Reactor Construction Experience Program

The nuclear industry in the United States faced many construction quality and design issues in

the 1970s and 1980s. In 1984, the NRC issued NUREG-1055, “Improving Quality and the

Assurance of Quality in the Design and Construction of Nuclear Power Plants,” to document the

lessons learned from plant construction. Since then, the NRC has revised some of its licensing

review processes and construction oversight programs to implement recommendations made in

NUREG-1055. In 2007, the NRC began developing a construction experience (ConE) program

to focus on collecting, analyzing, and applying lessons learned from the design and construction

of new reactors. To achieve this goal, the NRC staff developed a risk-informed process to

obtain, screen, evaluate, communicate, and incorporate construction experience insights into its

new reactor licensing and construction oversight activities. In recent years, the ConE program

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was expanded to include reviews of events at operating reactors that were related to latent

design and construction issues.

The staff’s review of an operating experience or ConE event occasionally results in the issuance

of INs or RIS. Recent examples include IN 2015-09, “Mechanical Dynamic Restraint (Snubber)

Lubricant Degradation Not Identified Due to Insufficient Service Life Monitoring,” dated

September 24, 2015; IN 2015-04, “Fatigue in Branch Connection Welds,” dated April 24, 2015;

RIS 2015-08, and RIS 2014-11, “Information on Licensing Applications for Fracture Toughness

Requirements for Ferritic Reactor Coolant Pressure Boundary Components,” dated

October 14, 2014.

The NRC staff values close cooperation with the international community for the exchange of

information on design and construction of new reactors and continues to work closely with

several countries that are currently building new nuclear power plants. As an example, the

NRC ConE program staff is participating in a NEA Working Group on Regulation of New

Reactors. The NRC staff also visits international sites under construction and participates in

joint vendor inspections with regulators from other nations. The NRC values these

partnerships and is committed to continuing its collaborative relationship with the international

community.

The NRC also exchanges relevant construction related information with our international

counterparts via the NEA’s Construction Experience Program. Recent international events taken

from the Construction Experience Program include issues associated with reactor vessel

manufacturing anomalies, misinstallation of containment vertical tendon sheaths, and design

mismatch of control room display windows of the plant monitoring and alarm system. In 2015,

the NRC uploaded construction experience events to the Construction Experience Program

database for issues related to the potential safety hazards caused by pipe support coating

deviations and inadvertent damage to the V. C. Summer, Unit 2, AP1000 containment vessel.

18.5 Fukushima Lessons Learned

The NRC has long recognized that protection from natural phenomena is an important means to

prevent core damage and ensure the integrity of containment and the SFP. As described in

Sections 1.3.1 and 1.3.3 of this report, the NRC has issued requests for information for

licensees to reevaluate their seismic and flooding hazards. Before the Fukushima accident, the

NRC was evaluating two generic issues that involved natural hazards: Generic Issue (GI)-204,

“Flooding of Nuclear Power Plant Sites Following Upstream Dam Failure,” and GI-199,

“Implications of Updated Probabilistic Seismic Hazard Estimations in Central and Eastern

United States on Existing Plants.” Subsequently, the evaluation of these generic issues was

incorporated into the Fukushima lessons learned reevalautions. As an additional safety

enhancement against beyond-design-basis natural hazards, the NRC also issued an order

requiring licensees to have mitigating strategies that preserve core cooling, SFP cooling, and

containment. The staff’s review of the Near Term Task Force recommendations identified areas

for further evaluation to enhance the regulations and cope with events beyond the current

design basis. Essentially, all of the actions the NRC is pursuing relate to events beyond the

current design basis, as discussed in Sections 1.3.1 and 1.3.3 of this report.

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18.6 Vienna Declaration on Nuclear Safety

Consistent with the first principle of the Vienna Declaration on Nuclear Safety, new nuclear

power plants licensed in the United States must meet safety, security, technical and financial

qualification requirements in the NRC’s regulations in 10 CFR Chapter I, including

10 CFR Parts 20; 50; 52; 30; 40; 70; 73; and 100; as well as 10 CFR Part 21, “Reporting of

Defects and Noncompliance”; and 10 CFR Part 55, “Operators’ Licenses.” These NRC

requirements govern the design, siting, construction, and operation of nuclear power plants.

These requirements address the prevention and mitigation of accidents through the

establishment of criteria for control and safety systems, such as the containment, reactor

coolant systems, and emergency core cooling systems. Regulatory requirements exist to ensure

adequate emergency planning to protect populations living within a 50-mile radius of nuclear

power plants, and to evacuate populations living within a 10-mile radius of nuclear power plants

in the unlikely event of a radioactive release. Each of these requirements were established with

consideration of uncertainties to ensure that adverse consequences to the public are acceptably

low. These regulations serve to prevent accidents and mitigate adverse consequences in a

manner that effectively minimizes the potential for (and therefore addresses the risk of adverse

consequences associated with) long-term offsite contamination.

In evaluating the design of a new reactor, the NRC assesses whether the prevention of

accidents and mitigation of consequences provided for in the design meets the Commission’s

safety goals. In addition to addressing design basis events, the NRC has imposed a

requirement on applicants seeking design certification under 10 CFR Part 52 to perform a PRA

for their proposed design and requires them to provide a description and analysis of design

features for the prevention and mitigation of severe accidents (e.g., challenges to containment

integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection,

hydrogen combustion, and containment bypass.) Combined license applicants are also

required to address the design features for prevention and mitigation of severe accidents by

considering site-specific conditions and factors, as well as a complete plant-specific PRA.

Because NRC requirements protect public health and safety through prevention of accidents

and by mitigating releases in the event of an accident, including severe, beyond design-basis

accidents, the risk of offsite contamination is rendered acceptably low.

The NRC uses deterministic and risk-informed requirements, as well as defense-in-depth

principles, to achieve this goal. Defense-in-depth embraces a broad set of principles and

requirements, including:

(1) the need to prevent accidents from occurring and mitigating

accidents if they occur (including robust emergency preparedness requirements), (2) the

concept of multiple barriers against radioactive releases, (3) the application of the principles of

independence, redundancy and diversity, which is implemented through requirements such as

the “single failure” assumption, and (4) siting new nuclear power plants in lower population

areas and areas with less adverse natural phenomenon characteristics. Section 18.1 of this

report provides additional details regarding the NRC’s defense-in-depth philosophy. Thus, the

NRC’s current regulatory approach provides reasonable assurance that there is a low likelihood

of offsite contamination requiring long term protective measures and actions.

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ARTICLE 19. OPERATION

Each Contracting Party shall take appropriate steps to ensure that:

(i)

the initial authorization to operate a nuclear installation is based upon an

appropriate safety analysis and a commissioning program demonstrating that the

installation, as constructed, is consistent with design and safety requirements

(ii)

operational limits and conditions derived from the safety analysis, test, and

operational experience are defined and revised as necessary for identifying safe

boundaries for operation

(iii)

operation, maintenance, inspection, and testing of a nuclear installation are

conducted in accordance with approved procedures

(iv)

procedures are established for responding to anticipated operational occurrences

and to accidents

(v)

necessary engineering and technical support in all safety related fields is

available throughout the lifetime of a nuclear installation

(vi)

incidents significant to safety are reported in a timely manner by the holder of the

relevant license to the regulatory body

(vii)

programs to collect and analyze operating experience are established, the results

obtained and the conclusions drawn are acted upon and that existing

mechanisms are used to share important experience with international bodies and

with other operating organizations and regulatory bodies

(viii)

the generation of radioactive waste resulting from the operation of a nuclear

installation is kept to the minimum practicable for the process concerned, both in

activity and in volume, and any necessary treatment and storage of spent fuel and

waste directly related to the operation and on the same site as that of the nuclear

installation take into consideration conditioning and disposal

The U.S. NRC relies on regulations in 10 CFR and internally developed associated programs in

granting the initial authorization to operate a nuclear installation and in monitoring its safe

operation throughout its life. This section describes the most significant regulations and

programs corresponding to each obligation of Article 19. It also includes a discussion on the

Vienna Declaration on Nuclear Safety, which was issued in February 2015.

Immediately after the accident at Fukushima in Japan, the NRC took actions that verified

nuclear power plant operators’ preparedness to respond to and mitigate the consequences of

beyond-design-basis events. These actions are discussed in Sections 1.3.1 and 1.3.3 of this

report.

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19.1 Initial Authorization to Operate

All currently operating reactors in the United States received licenses under the two-step

process in 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities.” This

licensing process requires both a construction permit and an operating license. The additional

licensing processes in 10 CFR Part 52, “Licenses, Certifications, and Approvals for Nuclear

Power Plants,” provide for site approvals and design approvals in advance of construction

authorization. In addition, 10 CFR Part 52 includes a process that combines a construction

permit and an operating license with conditions into one license (a combined license). Both the

two-step and the combined license processes require NRC approval to construct and operate a

nuclear power plant.

The Advisory Committee on Reactor Safeguards, an independent statutory committee

established to advise the NRC on reactor safety, reviews each application to construct or

operate a nuclear power plant. The committee begins its review early in the licensing process by

selecting the proper stages at which to meet with the applicant and NRC staff. Upon completing

its review, the committee reports to the Commission.

The public also has an opportunity to have its concerns addressed. The Atomic Energy Act and

the NRC’s regulations implementing this Act require the NRC to hold a public hearing before it

may issue a construction permit, early site permit, or combined license for a nuclear power

plant. Three-member Atomic Safety and Licensing Boards, which consist of one legal judge who

acts as the chairperson and two technically qualified judges from the Atomic Safety and

Licensing Board Panel, conduct public hearings for applications for construction permits and

early site permits. For combined licenses, the Commission conducts the uncontested mandatory

public hearing, while Atomic Safety and Licensing Boards conduct any contested hearings on

these license applications if a request for such a hearing is filed and granted. Members of the

public may submit written statements as part of these hearings, or they may petition for leave to

intervene as full parties in the hearing.

To obtain NRC approval to construct or operate a nuclear power plant, an applicant must submit

safety analysis and environmental reports. Article 18 describes the final safety analysis report

and the NRC’s review of the application for an operating license. Unlike the process for an

application for a construction permit, early site permit, or combined license, a public hearing is

neither mandatory nor automatic for an application for an operating license under

10 CFR Part 50. However, soon after the NRC accepts the application for review, it publishes a

notice in the Federal Register stating that it is considering issuing the license. This notice states

that any person whose interest might be affected by the proceeding may petition the NRC for a

hearing. Similar to the public hearings on applications for construction and early site permits,

three-member Atomic Safety and Licensing Boards conduct any public hearings on applications

for operating licenses. A licensing board will also determine whether to grant or deny the

request for a hearing.

An early site permit issued under Subpart A, “Early Site Permits,” to 10 CFR Part 52, provides

for resolution of site safety, environmental protection, and emergency preparedness issues,

independent of a specific nuclear plant design review. The application for an early site permit

must address the safety and environmental characteristics of the site and evaluate potential

physical impediments to the development of an acceptable emergency plan or security plan.

The applicant may submit additional information on emergency preparedness issues up to a

complete emergency plan. The staff documents its findings on site safety characteristics and

emergency planning in a safety evaluation report and its findings on environmental protection

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issues in an environmental impact statement. The early site permit may also allow limited

construction activities in accordance with 10 CFR 50.10, “License Required; Limited Work

Authorization,” subject to redress, before the issuance of a combined license. The NRC will

issue a Federal Register notice for a mandatory public hearing, and the Advisory Committee on

Reactor Safeguards will perform an independent safety review. The duration of an early site

permit is 10 - 20 years, and the permit may be renewed. A construction permit or combined

license application may reference the early site permit. To date, the NRC has issued five early

site permits. According to this process, environmental and siting issues that have been resolved

in the early site permit proceedings cannot be reopened during a combined license proceeding.

The NRC also may certify a standard plant design through a rulemaking under Subpart B,

“Standard Design Certifications,” to 10 CFR Part 52. The design certification process resolves

final design information for an essentially complete plant, independent of a specific site, and the

Advisory Committee on Reactor Safeguards performs an independent safety review. The

duration of a design certification is 15 years, and the certification may be renewed. The NRC

has certified five standard plant designs under the design certification process:

(1) General

Electric’s ABWR, (2) Westinghouse Electric Company, LLC’s System 80+ (originally designed

by Combustion Engineering), (3) Westinghouse’s AP600 design, (4) Westinghouse’s AP1000,

and most recently, in October 2014, (5) General Electric-Hitachi’s ESBWR. In December 2011,

the NRC staff issued amendments to the AP1000 and ABWR design certification rules. The

NRC staff is currently performing the following two design certification reviews:

(1) Korea

Hydro and Nuclear Power’s APR1400 and (2) Mitsubishi’s U.S. APWR. In addition, the NRC

staff has received two applications to renew the ABWR design certification. The NRC received

one renewal application from General Electric Hitachi Nuclear Energy and a separate

application from Toshiba Corporation. The NRC is actively reviewing the General Electric

Hitachi renewal application. Toshiba requested withdrawal of its renewal application in

June 2016.

A combined license, issued under Subpart C, “Combined Licenses,” to 10 CFR Part 52

authorizes construction of a facility in a manner similar to a construction permit under

10 CFR Part 50. An application for a combined license may incorporate by reference an early

site permit, design certification, both, or neither. The advantage of referencing an early site

permit or design certification is that issues resolved during those processes are not considered

again at the combined license stage. Just as for a construction permit, the NRC must hold a

hearing before deciding whether to issue a combined license. However, the combined license

will specify the inspections, tests, and analyses that the licensee must perform and the

acceptance criteria that must be met (collectively referred to as ITAAC) to provide reasonable

assurance that the facility has been constructed and will be operated in conformity with the

license and the applicable regulations. In 2012, the NRC issued its first combined licenses

authorizing construction and operation of new nuclear power plants at two sites in the United

States. The NRC issued combined licenses referencing the AP1000 certified design to Southern

Nuclear Operating Company for two units at the Vogtle Nuclear Plant in Georgia and to South

Carolina Electric and Gas Company and South Carolina Public Service Authority for two units at

the V.C. Summer Nuclear Plant in South Carolina. In May 2015, the NRC issued a combined

license referencing the ESBWR certified design to DTE Energy Company for one ESBWR unit

at the Fermi 3 site in Michigan. In February 2016, NRC issued combined licenses to Nuclear

Innovation North America for two ABWR units at the South Texas site in Texas.

After issuing a combined license, the NRC staff will verify that the licensee has performed the

required ITAAC, and before operation of the facility the Commission must find whether the

licensee has met the acceptance criteria. The licensee must submit notifications to the NRC

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The United States of America Seventh National Report for the Convention on Nuclear Safety (2016) - page 4