Index Manuals The United States of America Seventh National Report for the Convention on Nuclear Safety (2016)
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Capture relevant critical knowledge from employees departing the agency, recapture
knowledge from former employees where possible, communicate leadership
expectations for knowledge sharing, formalize knowledge management values and
principles, and incorporate knowledge management practices within agency work
processes.
A key element to the Knowledge Management Program success is the system of governance
under the agency knowledge management Steering Committee and knowledge management
staff leads with program management provided by the Office of the Chief Human Capital Officer.
These entities oversee and implement activities across the agency ensuring current and future
knowledge management needs of the agency are met. To accomplish this, the NRC uses a
broad and continuously evolving range of knowledge management tools and methods.
A few examples of current knowledge management and knowledge transfer activities include
the following:
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KNOWvember. The month of November is marketed agencywide as KNOWvember to
raise awareness and provide an opportunity to remind employees of the importance of
knowledge management. Every year since its inception in 2010, the program focuses on
a different theme to support knowledge management around the agency. In 2015, the
agency held a number of sessions with NRC experts highlighting critical skills, topics,
and historical events, such as a session on the issuance of NUREG-75/014, “Reactor
Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power
Plants (WASH-1400),“ dated October 1975, and the origins of PRA in the nuclear
industry. In addition, the agency pursued discussion around the influential topic of
multigenerations in the workforce, where a panel session of nine NRC employees from
each of the five generations presented their perspectives on “Leading across
Generations.” All sessions were recorded and preserved for knowledge management.
In 2014, the NRC held a 2-day knowledge management best practice showcase that
was dedicated to sharing best practices around the agency. The showcase exhibited 20
practices from eight offices.
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Ask SME and Learn. In November 2013, the program launched an internal series called,
“Ask SME [subject matter experts] and Learn,” that was initiated to capture and share
critical knowledge and experiences of subject matter experts. The sessions provide an
opportunity for staff across the agency to learn directly from the agency experts on a
particular topic in an open forum. Previous sessions that have been held featured senior
executives sharing their knowledge and experiences prior to retirement, other topics
included cyber security, Price Anderson Act, and low level waste regulation. The
sessions are recorded and preserved for knowledge management.
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Knowledge Management Guidance. In October 2015, the program issued
recommendations and guidance for sharing knowledge and experience gained by NRC
personnel returning from foreign assignments.
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NUREG/KM Series. In 2012, the agency launched a NUREG/Knowledge Management
(NUREG/KM) series to preserve knowledge of historical events that shaped the
regulatory process. The series focuses on collecting and interpreting historical
information on identified topics for the benefit of future generations of NRC professionals
as well as the public. Currently, the series features eight publications. The first of the
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NUREG/KM series is an account of the Three Mile Island accident. The eighth and most
recent series is NUREG/KM-0008, “Reflections on Fukushima: NRC Senior Leadership
Visit to Japan, 2014,” dated December 2014.
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Brown Bag Sessions (Chat n’ Learn). Informal meetings are held by individual offices to
convey information or obtain feedback on common areas of interest, such as
agencywide software system updates and upgrades, or changes to regulations and
policies. More recently, departing senior executives, such as former Executives Directors
for Operations and Office Directors, have used this as a forum to reach different
audiences agencywide to share their knowledge and insight on a variety of technical
topics, issues, and events they have experienced throughout their careers.
8.1.7 Openness and Transparency
Openness has long been one of the NRC’s Principles of Good Regulation as the agency seeks
to carry out its mission of ensuring radioactive materials are used safely and securely. The
agency views nuclear regulation as the public’s business. As such, the NRC makes every effort
to ensure its regulatory activities are open to the public.
Openness also requires the public to be able to participate meaningfully in the NRC’s regulatory
processes. At the same time, the agency must control sensitive information so that it is not
made public.
Openness is one of the five “Principles of Good Regulation” the NRC first established in 1977.
These principles guide all of the agency’s activities. Openness is also one of seven
organizational values, adopted in 1995, to which the agency adheres in all its work. The NRC’s
Strategic Plan emphasizes Open Government principles and includes specific strategies for
ensuring that the regulatory process, decisionmaking, and licensee oversight are all carried out
as transparently as possible. That plan established three specific strategies to achieve
openness, including the need to provide meaningful opportunities for the public and other
stakeholders to participate in NRC activities and to use clear and understandable language in
communicating with the public.
Access to NRC Documents. From its inception, the NRC has made it a priority to maintain a
Public Document Room, to assist the public in finding publicly available NRC information. The
Public Document Room staff comprises of skilled technical and reference librarians who provide
information and research assistance directly to stakeholders, environmental groups, licensees,
the legal community, and concerned citizens. The staff provides assistance in navigating the
agency’s extensive collection of documents on licensing and rulemaking activities, as well as
historical files from the NRC’s predecessor agency, the Atomic Energy Commission.
To ensure the public has access to the information it needs, the NRC makes all nonsensitive
documents available to the public, unless there is a specific reason for them to be withheld. The
agency implemented policies, performance measures, and management controls to ensure this
access. The NRC’s documents database, known as ADAMS, places all final records of publicly
available documents into a searchable library that can be accessed through the NRC’s public
Web site. The database includes documents and correspondence related to license
applications, license renewals, and inspection findings. It does not include security-related,
proprietary, or other sensitive information. In 2015, approximately 69,000 public users accessed
ADAMS more than 247,000 times and requested documents more than 45.2 million times.
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The NRC measures and reports to Congress each year how quickly it releases internal and
external documents, issues notices in advance of public meetings, and responds to requests
filed under the Freedom of Information Act — a Federal law giving the public the right to request
and receive Government documents, with some exceptions.
The NRC also uses traditional tools to keep stakeholders informed. The agency sends copies of
key documents and notifications to Federal, State, local, and Tribal authorities. The NRC also
publishes notices in the Federal Register of Commission meetings, opportunities for hearings,
and opportunities to comment on a variety of the agency’s activities.
Open Government Plan. The Open Government Plan designated one senior NRC manager as
the accountability official, who, together with a senior advisory council, provides guidance on
openness initiatives. The agency established a separate Open Government Advisory Group to
oversee its Open Government program. These leaders work together to ensure that the agency
has a continuing focus on adopting new technologies and making full use of their potential to
reach out and engage the public. The plan lays out specific improvements the NRC is making to
enhance stakeholder engagement.
The NRC is an active participant in several governmentwide programs that promote
source for finding and applying for Federal grants.
The NRC Web Site. The NRC uses its public Web site to share information with stakeholders
and the public. In 2015, the NRC’s Web site had more than 2.5 million individual visitors. The
Web site was visited more than 6.4 million times, and visitors viewed more than 71.4 million
pages. The site provides information on Commission decisions, hearing transcripts, inspection
reports, enforcement actions, petitions, event reports, and daily plant status. It includes a tool
that allows users to locate information easily on facilities the NRC regulates and details on the
performance of reactor licensees. It also provides a great deal of general information and links
to broaden the public’s understanding of the NRC’s mission, goals, and performance, as well as
access to tools and information to help licensees and others conduct business with the agency.
The site makes available all the NRC’s press releases, issued when the agency receives
license applications, makes major licensing decisions, takes enforcement actions, and
announces major public meetings, opportunities for hearings, and other avenues for public
involvement. This information can also be provided automatically to anyone who requests a
subscription. In fact, users may sign up through the Web site to receive automatically a number
of different types of documents, including generic communications, new rulemaking dockets,
speeches, and reports issued by the NRC’s Inspector General. The public also can subscribe to
receive correspondence related to specific facilities. The site includes an Open Government
page with links to high-value datasets, information on the NRC’s openness philosophy, and a
tool allowing the public to suggest ways the agency can improve transparency, public
participation, and collaboration.
The NRC video-streams high-interest Commission meetings over the Internet. More recently,
the agency expanded Webcasting to other high-interest meetings, conferences, and
adjudicatory hearings. These Webcasts are available for viewing live, as they occur, and are
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archived for viewing later. The agency has also made use of webinars as a way of leveraging
technology to communicate with the public.
Social Media. Over the past several years, the NRC has embraced social media as an important
tool for reaching a public audience beyond those with access to traditional communications
media. These social media platforms allow the agency to give information to the public, raise
awareness, explain technical activities, and spotlight accomplishments. They also provide new
vehicles for dialogue, giving the NRC new platforms to participate in two-way communication
with the public. The NRC’s Office of Public Affairs manages these tools, but NRC staff members
at all levels help ensure the agency is meeting the communication needs of all our offices, both
at headquarters and in the regions.
The NRC’s blog, or Web log, allows the agency to communicate with the public in plain
language about high interest or complex topics from a technical or regulatory standpoint. The
NRC posts several blogs each week on a variety of subjects and responds to comments posted
by the public, as appropriate. The blog is a valuable tool for generating discussion about
important matters. As of November 2015, the agency published more than 600 blog posts since
its launch in January 2011. Those posts have been viewed more than 751,000 times. The NRC
received and posted more than 5,735 comments on the blog.
Social media platforms proved to be invaluable crisis communications tools during the events at
Fukushima in 2011. Their value was again proven in October 2012, when several reactors shut
down during Hurricane Sandy, a huge storm that affected the entire east coast of the United
States. The most views the blog has ever had was on October 29, 2012, during the height of the
storm. The blog was viewed more than 6,300 times that day. The blog has been instrumental in
keeping the public informed on matters of great concern. Public comments submitted on the
blog helped the NRC to develop content, as needed, to keep up with public demand for
information.
The NRC’s Twitter account, launched in August 2011, offers an opportunity for the agency to
push out relevant information quickly in a simplified format. For example, the NRC can alert the
public to new press releases, Federal Register notices, licensing decisions, guidance
documents, important personnel changes, and any topic that might emerge. As of November
2015, the NRC had more than 6,800 Twitter followers and continues to add new followers daily.
The agency sent a total of 2,092 tweets over 54 months, for an average of 38 per month. The
NRC’s tweets sent during Hurricane Sandy in October 2012 were retweeted 130 times,
potentially reaching more than 210,000 Twitter users.
The NRC’s YouTube channel and Flickr photo gallery provide a platform for video and image
content and offer a gateway to additional information on the agency’s Web site. The NRC posts
photos and video of special events, important meetings, visits to nuclear facilities, and a variety
of activities carried out by NRC staff. These forums enable the agency to document its work
visually and introduce the people who carry out the agency’s mission. Since launching the
YouTube channel in August 2011, the agency posted about 140 videos, which have received
nearly 104,000 views. More than 700 users subscribe to the NRC YouTube channel and are
notified each time new content is posted. Since February 2012, the NRC has published about
2,200 photos and graphics to its Flickr account which have been collectively viewed more than
1.2 million times.
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The agency launched Facebook in August 2014, rounding out its social media program. Since
that time, its page has gained more than 2,200 likes. With nearly 270 posts, more than
22,000 people have engaged with NRC content on Facebook.
Public Meetings. The public continues to have many different opportunities to be involved in the
NRC’s regulatory decisionmaking process. Stakeholders may participate in a variety of ways
before the agency issues certain licensing actions. To ensure this involvement is meaningful,
the NRC actively communicates with stakeholders so they will understand how the NRC makes
decisions - including the agency’s role, processes, and activities. The NRC holds meetings with
the public and other stakeholders near nuclear facilities, at agency headquarters, and at NRC
regional offices.
The NRC is using a variety of tools to improve public participation. The agency is expanding its
use of Web conferencing to allow participation by anyone with access to a computer, minimizing
travel costs and increasing opportunities for public involvement. The agency actively seeks
feedback from meeting participants to help identify ways the NRC can improve public meetings.
The NRC staff hosts and participates in a number of conferences, workshops, and symposia
each year. The most prominent is the annual Regulatory Information Conference, which brings
together over 3,000 people from more than 30 countries, including members of Congress,
nuclear industry representatives, international counterparts, and other stakeholders. The
conference features presentations by the NRC’s commissioners, NRC staff, licensees, and
other stakeholders. It serves as a communications vehicle to allow open dialogue on research
findings, rulemakings, regulatory and safety issues, regulatory process and procedure
improvements, international activities, and other items of interest. All presentations are available
through the NRC Web site and the NRC Web streams key events.
Details on the NRC’s special programs for public involvement in oversight of operating nuclear
facilities can be found in Section 6.3.11 of this report.
The NRC’s 2012 Open Government Plan describes goals for improving plain writing, high-value
datasets, and services that the Public Document Room offers. The agency will continue efforts
to strengthen social media services, expand the use of virtual meetings, and increase the
visibility of rulemakings and NRC documents open for public comment. Improving the agency’s
use of plain language is an important goal for the immediate future.
The NRC has identified certain types of documents that should be written in plain language.
They include informational brochures, performance assessments, generic communications,
inspection reports, and significant enforcement actions. The agency is encouraging staff
involved in preparing such documents to take plain language training, which the NRC offers
both online and in a 2-day instructor led course.
The agency plans to strengthen the ability of stakeholders who use smart phones or other
mobile devices to engage with the NRC. Under this initiative, the NRC will develop
mobile-friendly Web pages and use quick response codes, enabling interested members of the
public to scan barcodes for quick access to information.
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8.2 Separation of Functions of the Regulatory Body from Those of Bodies Promoting
Nuclear Energy
The U.S. law, through legislation enacted by the U.S. Congress, ensures the effective
separation between the functions of the regulatory body and those of any other body concerned
with the promotion or utilization of nuclear energy, as well as the independence of the regulatory
body in making its safety-related decisions. Originally, the regulatory and promotional
responsibilities for nuclear energy were combined into a single U.S. agency - the Atomic
Energy Commission. In 1974, the U.S. Congress, through the Energy Reorganization Act of
1974, abolished the Atomic Energy Commission and divided its functions between two new
agencies, the NRC and ERDA, which was succeeded by DOE in 1977. Section 201 of the
Energy Reorganization Act of 1974, established the NRC as an “independent regulatory
commission,” while ERDA, now DOE, was established as a cabinet level agency of the U.S.
President.
Congress conferred upon the NRC the licensing, inspection, and enforcement regulatory
responsibility for all civilian uses of nuclear energy and materials as provided for in the Atomic
Energy Act of 1954, as amended. The promotional and technology development functions were
transferred under the Act to ERDA. This division resulted in the complete separation of
regulatory from promotional responsibilities.
Given the NRC’s status as an independent regulatory agency, the NRC’s Commissioners, in
contrast to the heads of cabinet level agencies like DOE, may be removed by the U.S. President
only for “inefficiency, neglect of duty, or malfeasance in office.” The NRC’s independence
allows it to promulgate regulations governing commercial nuclear power uses without submitting
them to the cabinet-level OMB for review and approval.
Accordingly, the NRC has independent authority to regulate the possession and use of nuclear
materials as well as the siting, construction, and operation of nuclear facilities. The NRC
performs its regulatory mission by issuing regulations, licensing commercial nuclear reactor
construction and operation, licensing the possession of and use of nuclear materials and
wastes, safeguarding nuclear materials and facilities from theft and radiological sabotage,
inspecting nuclear facilities, and enforcing regulations. The NRC regulates commercial nuclear
fuel cycle materials and facilities. The NRC is also responsible for licensing commercial nuclear
waste management facilities, independent spent fuel management facilities, and DOE facilities
for the disposal of high-level radioactive waste and spent fuel.
The enactment of the DOE Organization Act in 1977 subsequently brought a number of Federal
agencies and programs, including ERDA, into a single agency with responsibilities for nuclear
energy technology and nuclear weapons programs (i.e., DOE). Over the ensuing decades, DOE
has expanded its nuclear-related activities to include nonproliferation and the environmental
cleanup of contaminated DOE and certain other legacy sites and facilities. With limited
exceptions, DOE retains authority under the Atomic Energy Act for regulating its nuclear
activities, including the responsibility for activities such as regulating the disposal of its own
low-level radioactive waste.
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8.3 Ethics Rules Applying to NRC Employees and Former Employees
NRC employees must comply with governmentwide ethics rules contained in federal statutes
and regulations. These rules are intended to ensure that every citizen can have confidence in
the integrity of the Federal Government. The rules set standards that federal employees must
follow in situations that may raise ethics concerns. For example, the rules restrict an
employee’s ability to accept gifts from regulated entities; prohibit the employee from working on
matters in which he may have a conflicting financial interest, such as a matter involving a
recent former employer; and preclude the employee from using his public position for private
gain.
In addition to these government-wide rules, the NRC has two of its own ethics rules. The first
rule establishes a Prohibited Securities List consisting of power reactor licensees and certain
other entities that may be affected by the NRC’s regulatory actions. NRC employees in
designated positions cannot own stock in any company appearing on the Prohibited Securities
List. The second rule requires that NRC employees obtain prior approval before accepting
outside employment with certain types of employers, including any organization that operates
in the commercial nuclear field.
When an NRC employee leaves the agency, he must comply with federal post-employment
rules. These rules prevent the former employee from representing a non-federal party before
the federal government on certain matters. The length of the restriction depends on the former
employee’s position while at the NRC and the extent of his participation in the matter on which
he seeks to represent the non-federal party.
In addition to these rules, political appointees at the NRC must currently follow the
commitments listed in the Ethics Pledge that President Obama issued in 2009. The Ethics
Pledge further limits an appointee’s ability to accept gifts, extends certain postemployment
restrictions, and imposes additional requirements.
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ARTICLE 9. RESPONSIBILITY OF THE LICENSE HOLDER
Each Contracting Party shall ensure that prime responsibility for the safety of a nuclear
installation rests with the holder of the relevant license and shall take the appropriate
steps to ensure that each such license holder meets its responsibility.
The U.S. NRC, through the Atomic Energy Act, ensures that the primary responsibility for the
safety of a nuclear installation rests with the licensee. The overall responsibility of licensees for
ensuring safety of their facilities did not change as a result of the Fukushima accident. U.S.
licensees continue to respond to new NRC regulatory requirements and initiatives that confirm
and ensure that adequate measures to protect public health and safety are taken considering
the lessons learned following the accident, as described in Sections 1.3.1 and 1.3.3 of this
report.
Steps that the NRC takes to ensure that each licensee meets its primary responsibility include
the licensing process, discussed in Articles 18 and 19, the Reactor Oversight Process,
discussed in Article 6, and the enforcement program, discussed below. This section provides an
update on the licensee’s responsibility for maintaining openness and transparency and for
maintaining resources for managing accidents.
9.1 Introduction
The NRC’s regulatory programs continue to be based on the premise that the safety of
commercial nuclear power reactor operations is the primary responsibility of NRC licensees.
The agency is responsible for regulatory oversight of licensee activities to ensure that safety is
maintained. The NRC reviews the safety of a reactor design and the capability of an applicant to
design, construct, and operate a facility. If an applicant satisfies the Federal requirements, then
the NRC will issue a license to operate the facility. Such licenses specify the terms and
conditions of operation to which a licensee must conform. If a licensee does not conform to
these license conditions, the NRC may take enforcement action, which can include modifying,
suspending, or revoking the license. The NRC can also order particular corrective actions or
issue civil penalties. The following sections discuss these enforcement mechanisms in greater
detail.
9.2 The Licensee’s Primary Responsibility for Safety
As discussed in Article 7 of this report, the Atomic Energy Act, Section 103, Chapter 10, grants
the NRC authority to issue licenses for production and utilization facilities for commercial or
industrial purposes, which includes nuclear power reactors. Moreover, Section 103 states that
these licenses are subject to such conditions as the NRC may establish by rule or regulation to
implement the purposes and provisions of the Atomic Energy Act. Consistent with the Act,
before issuing a license, the Commission determines that the applicant is (1) equipped and
agrees to observe such safety standards to protect health and minimize danger to life or
property as the Commission may establish by rule and (2) agrees to make available to the
Commission such technical information and data about activities under such license as the
Commission may determine necessary to promote the common defense and security and to
protect public health and safety.
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Embedded in each license is the explicit responsibility for the license holder to comply with the
terms and conditions of the license and the applicable Commission rules and regulations. The
licensee is ultimately responsible for the safety of its activities and the safeguarding of nuclear
facilities and materials used in operation.
This responsibility for safety is implemented, in part, by having trained and qualified operators.
Section 50.54 of 10 CFR, “Conditions of Licenses,” identifies requirements that are conditions in
every nuclear power reactor operating license. This regulation, in part, specifies the minimum
requirements per shift for onsite staffing of the control room by operators and senior operators,
including multiunit sites and shared control rooms (10 CFR 50.54(i) through (m)). Additionally,
10 CFR 50.120, “Additional Standards for Licenses, Certifications, and Regulatory Approvals,”
requires that each licensee establish, implement, and maintain a training program. The training
program must incorporate the instructional requirements necessary to provide qualified
personnel to operate and maintain the facility in a safe manner in all modes of operation. The
training program must be developed to be in compliance with the facility license, including all
technical specifications and applicable regulations. For additional information, see Sections
11.2, 12.3.2, and 12.3.3 of this report. The training program must be periodically evaluated and
revised as appropriate to reflect industry experience as well as changes to the facility,
procedures, regulations, and quality assurance requirements. Additional information on licensee
training and accreditation programs is provided in Part 3 of this report.
Furthermore, licensees have financial responsibilities in the event of an accident. The Atomic
Energy Act, Section 182.a gives the basis for the NRC’s onsite property damage insurance
requirements for operating nuclear power reactors contained in 10 CFR 50.54(w). The license
condition in 10 CFR 50.54(w) requires that licensees obtain insurance in an equivalent amount
of protection covering the licensee’s obligation, in the event of an accident at the licensee’s
reactor, to stabilize and decontaminate the reactor and the reactor site. Licensees are required
to report the current levels of insurance or financial security and the sources of the insurance or
security to the NRC on April 1 of each year. Additionally, licensees are required to have and
maintain financial protection liability insurance for claims arising from accidents. Additional
information on liability insurance can be found in Section 11.1.3 of this report.
When the Commission determines that the licensee is not complying with the Commission’s
rules or regulations, the NRC takes appropriate action to ensure that the facility is returned to a
condition compliant with its license. Details about the NRC’s Enforcement Program are provided
in the next section and in Section 7.2.4 of this report.
9.3 NRC Enforcement Program
As discussed in Article 7, the NRC has enforcement powers. As discussed in Sections 7.2.3 and
7.2.4, the enforcement process complements the Reactor Oversight Process. The NRC uses
enforcement as a deterrent to emphasize the importance of compliance with regulatory
requirements and to encourage prompt identification and prompt, comprehensive correction of
violations.
The NRC identifies violations through inspections and investigations. All violations are subject to
civil enforcement action and may be subject to criminal prosecution. Unlike the burden of proof
standard for criminal actions (beyond a reasonable doubt), the NRC uses the Administrative
Procedure Act standard (preponderance of evidence) in enforcement proceedings. After an
apparent violation is identified, it is assessed in accordance with the Commission’s enforcement
policy, described in the “NRC Enforcement Policy,” last updated on February 4, 2015, which is
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available to NRC licensees and members of the public. The NRC Office of Enforcement
maintains the current policy statement on the NRC’s public Web site. Because it is a policy
statement and not a regulation, the Commission may deviate from it, as appropriate, as the
circumstances of a particular case may dictate.
The NRC has three primary enforcement sanctions available: notices of violation, civil
penalties, and orders.11 A notice of violation identifies a requirement and how it was violated;
formalizes a violation pursuant to 10 CFR 2.201, “Notice of Violation”; requires corrective action;
and normally requires a written response. A civil penalty is a monetary fine issued under
authority of the Atomic Energy Act, Section 234, or the Energy Reorganization Act, Section 206.
Section 234 of the Atomic Energy Act provides for penalties of up to $100,000 per violation per
day; however, that amount is adjusted every 4 years by the Federal Civil Penalties Inflation
Adjustment Act of 1990, as amended by the Debt Collection Improvement Act of 1996, and is
currently $140,000. Section 161 of the Atomic Energy Act gives the Commission broad authority
to issue orders; this authority extends to any area of licensed activity that affects public health
and safety or the common defense and security. Orders modify, suspend, or revoke licenses, or
they may require specific actions by licensees or persons. The NRC issues notices of violations
and civil penalties on the basis of violations. The agency may issue orders for violations or, in
the absence of a violation, because of a concern involving public health and safety or the
common defense and security.
After identifying a violation, the NRC assesses its significance by considering the following
factors:
x
actual safety consequences
x
potential safety consequences
x
potential for impacting the NRC’s ability to perform its regulatory function
x
any willful aspects of the violation
Given those factors, the NRC takes one of the following actions based on the significance of the
violation:
x
assigns a severity level, ranging from Severity Level IV (more than minor concern) to
Severity Level I (the most significant)
x
associates the violation with findings assessed through the Reactor Oversight Process
significance determination process (described in Article 6) and assigns a color code of
green, white, yellow, or red based on increasing risk significance
The Commission recognizes that there are violations of minor safety or environmental concern
that are below Severity Level IV violations, as well as below violations associated with green
findings. These minor violations are not assigned a severity level category or a color
assessment.
11 The NRC also uses administrative actions, such as notices of deviation, notices of nonconformance,
confirmatory action letters, and demands for information, to supplement its enforcement program.
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The NRC may hold a predecisional enforcement conference or a regulatory conference with a
licensee before making an enforcement decision if (1) escalated enforcement action (i.e., a
Severity Level III or higher notice of violation or a greater-than-green Reactor Oversight Process
finding) appears warranted, (2) the NRC decides a conference is necessary, or (3) the licensee
requests it. The purpose of the conference is to obtain information to assist the NRC in
determining whether an enforcement action is necessary and, if so, what the appropriate
enforcement action is. The conference focuses on areas such as (1) a common understanding
of facts, root causes, and missed opportunities associated with the apparent violation and (2) a
common understanding of the corrective actions taken or planned.
At several junctions during the enforcement process involving cases of discrimination or willful
violation of NRC regulations, the agency offers its licensees (including their contractors) or
individuals the opportunity to participate in the Alternative Dispute Resolution Program.
Alternative Dispute Resolution is also offered as an option for nonwillful (traditional)
enforcement cases with the potential for civil penalties. Alternative dispute resolution is a
general term encompassing various techniques for resolving conflict outside of court using a
neutral third party. The NRC uses mediation, a technique in which a neutral mediator with no
decisionmaking authority helps parties clarify issues, explore settlement options, and evaluate
how best to advance their respective interests. Neutral mediators are selected from a roster of
experienced mediators provided by a neutral program administrator who is under contract with
the NRC. The mediator assists the parties in reaching an agreement. However, the mediator
has no authority to impose a resolution upon the parties. Mediation is a confidential and
voluntary process. If the parties to the process (the NRC and the licensee or individual) agree to
use alternative dispute resolution, they select a mutually agreeable neutral mediator and share
the cost of the mediator’s services equally. In cases in which the NRC and the other party reach
an agreement, the agency issues a confirmatory order reflecting the terms of the agreement.
The agency considers civil penalties for Severity Level I, II, and III violations, as well as knowing
and conscious violations of the reporting requirements of Section 206 of the Energy
Reorganization Act and the release of Safeguards Information by an individual. Although not
normally used for violations associated with the Reactor Oversight Process, civil penalties
(and the use of severity levels) are considered for issues that are willful, that have the potential
to affect the regulatory process, or that have actual consequences.
Although each severity level may have several associated considerations, the outcome
of the assessment process for each violation or problem (absent the exercise of discretion)
results in one of three outcomes, which may involve no civil penalty, a base civil penalty,
or twice the base civil penalty. A base civil penalty has been established for each escalated
severity level violation and for each type of licensee. Specific Commission approval is required
for proposals to impose a civil penalty for a single violation or problem that is greater than three
times the Severity Level I civil penalty value for that type of licensee.
The NRC may issue orders to modify, suspend, or revoke a license; issue orders to cease and
desist from a given practice or activity; or take other action as may be proper. The agency may
issue orders in place of, or in addition to, civil penalties. Additionally, the NRC may issue an
order to impose a civil penalty when a licensee refuses to pay a civil penalty or an order to an
unlicensed person (including vendors) when the agency has identified deliberate misconduct.
By statute, a licensee or individual may request a hearing upon receiving an order. Orders are
normally effective after a licensee or individual has had an opportunity to request a hearing
(i.e., 30 days). However, orders can be made immediately effective without prior opportunity for
a hearing when the agency determines it is in the best interest of public health and safety to do
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so. Subsequent to the hearing process, a licensee or individual may appeal the administrative
hearing decision to the Commission and, if desired, appeal the Commission’s decision to a U.S.
court of appeals.
Providing interested stakeholders with enforcement information is very important to the NRC.
Conferences that are open to public observation appear in the listing of public meetings on the
NRC’s public Web site. The agency issues a press release for each proposed civil penalty or
order. All orders are published in the Federal Register. Significant enforcement actions
(including actions to individuals) are included in the enforcement document collection in the
NRC’s public Web site.
In the last 3 calendar years, the NRC issued the following significant enforcement actions to
operating power reactors:
Calendar Year
2013
2014
2015
Notices of violation without civil penalties
26
30
28
Civil penalties
1
1
3
Orders without civil penalties
8
7
2
Total enforcement actions
35
38
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9.4 Openness and Transparency
U.S. nuclear power plant licensees are required to demonstrate that the appropriate
governmental authorities have the capability to alert the public of a nuclear power plant event
(e.g., sirens, tone alert radios, and route alerting) and provide prompt, clear instructions on
protective actions. At least annually, licensees provide members of the public located within the
plume exposure pathway emergency planning zone information on how they would be notified
and what their initial actions should be in an emergency as described in Section 16.8 of this
report. Educational information on radiation, contact(s) for additional information, information on
protective measures (e.g., evacuation routes and relocation centers, sheltering, respiratory
protection, and radioprotective drugs), and direction to those needing assistance during an
emergency is provided. A licensee’s public information program includes the use of signs,
notices, or other means, placed in areas such as motels, stores, and recreational venues for
transient populations.
Each licensee has established a Joint Information Center that serves as a focal point for the
coordination and dissemination of information from the licensee and Federal, State and local
authorities to the public and media during an incident. In February 2011, the NRC published
NUREG/CR-7032, “Developing an Emergency Risk Communication (ERC)/Joint Information
Center (JIC) Plan for a Radiological Emergency,” and NUREG/CR-7033, “Guidance on
Developing Effective Radiological Risk Communication Messages: Effective Message
Mapping and Risk Communication with the Public in Nuclear Plant Emergency Planning Zones,”
which address Joint Information Center enhancements to account for changes in media
practices, advances in communications technology, and changes in public access to information
and to address message mapping to support concise and consistent messaging.
The NRC’s openness and transparency objectives are described in Section 8.1.7 of this report.
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ARTICLE 10. PRIORITY TO SAFETY
Each Contracting Party shall take the appropriate steps to ensure that all organizations
engaged in activities directly related to nuclear installations shall establish policies that
give due priority to nuclear safety.
Policies of the U.S. NRC that give due priority to safety covered under this article are PRA
policy statements and policies that apply to licensee safety culture and safety culture at the
NRC. Other articles (e.g., Articles 6, 14, 18, and 19) also discuss activities undertaken to
achieve nuclear safety at nuclear installations.
The NRC has not made specific changes to the priority of its safety programs addressed under
this article as a result of the Fukushima accident. However, the actions that the NRC has taken
are broadly supportive of the due priority to safety and a strong safety culture. For example, as
part of the larger effort discussed in Sections 1.3.1 and 1.3.3, the NRC has ensured that
licensees have adequate staffing to respond to emergencies and clearly defined roles and
responsibilities for those responders. Similarly, the NRC requested that nuclear power plant
licensees reevaluate potential seismic and flooding hazards and perform a mitigation strategies
assessment of the reevaluated hazards.
10.1 Background
The NRC has a longstanding goal to move toward more risk-informed and performance-based
approaches in its regulatory programs. In SRM-SECY-98-144, “White Paper on Risk-Informed
and Performance-Based Regulation,” dated March 1, 1999, the Commission approved defining
the terminology and expectations for evaluating and implementing initiatives related to
risk-informed, performance-based approaches. A risk-informed approach is an approach
whereby risk results and insights from a PRA that addresses a broad range of plant conditions
are used, in a complementary manner with the traditional (deterministic) engineering concepts
of defense-in-depth and safety margin, to establish requirements. In contrast, a solely
deterministic approach would only address a limited number of design basis conditions and
relies on conservatisms in the analyses. The risk-informed approach better focuses licensee
and regulatory attention on design and operational issues commensurate with their importance
to public health and safety. A performance-based approach is an approach that establishes
measurable (or calculable) outcomes to be met, instead of using prescriptive requirements that
specify particular features, actions, or programmatic elements to be included in the design or
process. Therefore, the performance-based approach provides more flexibility to the licensee in
meeting the design or process objective. Implemented together, the risk-informed, performance-
based approach uses risk insights, engineering analyses, judgment, the principles of defense-
in-depth and safety margins, and performance history to:
x
focus attention and resources on the most important activities and issues
x
establish objective criteria for evaluating performance
x
develop measurable or calculable parameters for monitoring system and licensee
performance
x
provide flexibility to determine how to meet the established performance criteria in a way
that encourages and reward improved outcomes
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x
focus on the results as the primary basis for regulatory decisionmaking
The United States has made progress in developing and using risk information to give due
priority to nuclear safety and establishing policies, programs, and practices that apply to
licensee safety culture.
The NRC has developed extensive guidance on the role of PRA in U.S. regulatory programs
and applies risk insights gained from PRAs to complement traditional engineering analyses. The
increased use of risk information has improved issue-specific safety regulation and risk
information has been used to evaluate proposed changes to the current licensing bases for
individual plants. For example, 10 CFR, Section 50.69, “Risk-Informed Categorization and
Treatment of Structures, Systems, and Components for Nuclear Power Reactors,” allows
licensees to use a risk-informed approach to categorize SSCs, and assign special treatment
requirements, according to their safety significance. As another example, 10 CFR 50.48(c)
allows an operating nuclear power plant licensee to adopt a risk-informed, performance-based
fire protection program. The NRC continues to evaluate ways that risk insights can be used to
enhance its regulatory framework in a risk-informed, performance-based manner.
On June 14, 2011, the NRC issued its “Final Safety Culture Policy Statement” and identified
traits of a positive safety culture. All U.S. nuclear power plants have committed to conducting a
safety culture self-assessment every 2 years and have committed to conducting monitoring
panels as described in Nuclear Energy Institute (NEI) 09-07, “Fostering a Healthy Nuclear
Safety Culture,” dated March 2014.
10.2 Probabilistic Risk Assessment Policy
Three policy statements form the basis of the NRC’s current treatment of PRA and the related
regulatory safety goals and objectives: the “Policy Statement on Severe Reactor Accidents
Regarding Future Designs and Existing Plants,” dated August 8, 1985; the “Safety Goals for the
Operation of Nuclear Power Plants; Policy Statement; Republication,” dated August 21, 1986;
and the “Use of Probabilistic Risk Assessment Methods in Nuclear Activities; Final Policy
Statement,” dated August 16, 1995.
10.3 Applications of Probabilistic Risk Assessment
The NRC uses risk insights gained from PRA in conjunction with traditional engineering
analyses to resolve emergent issues, evaluate new and existing requirements and programs,
and to provide a technical basis for risk-informed regulation. In addition, the NRC conducts
research to improve data and methods used in risk analysis. The NRC also engages in
cooperative activities with industry (such as pilot programs for 10 CFR 50.69 and
10 CFR 50.48(c)), and in activities that assess risk in determining plant-specific changes to the
licensing basis. To assess the technical adequacy of the supporting PRA for risk-informed
applications, the NRC staff uses RG 1.200, “An Approach for Determining the Technical
Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,” Revision 2,
issued in March 2009; and NUREG-0800, “Standard Review Plan for the Review of Safety
Analysis Reports for Nuclear Power Plants: LWR Edition,” Section 19.1, “Determining the
Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed License
Amendment Requests after Initial Fuel Load,” Revision 3, issued in December 2015.
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The NRC maintains a risk-informed and performance-based plan, updated annually, which sets
forth the agency’s planned actions to make its regulatory activities risk informed and
performance based. In the past, the Risk-Informed Regulation Implementation Plan, focused
largely on risk-informed initiatives. The current improved plan has expanded the objectives to
more fully achieve a risk-informed and performance-based regulatory structure. The NRC has
created a public Web site for the risk-informed and performance-based plan with links to
documents that specifically describe activities and status:
http://www.nrc.gov/about-nrc/regulatory/risk-informed/rpp.html.
The NRC and industry representatives have cooperated in several areas and piloted programs
to develop and apply risk-informed methodologies for specific regulatory applications. The staff
uses the lessons learned from these activities to develop and publish detailed implementation
guidance. These activities, described in the sections below, include special treatment, inservice
inspection, technical specification changes, and standards development.
For new reactors licensed under 10 CFR Part 52, “Licenses, Certifications, and Approvals for
Nuclear Power Plants,” the NRC requires applicants to describe the design-specific PRA and its
results for a design certification application and a plant-specific PRA and its results for a
combined license application. In addition, the NRC requires the holder of a combined license to
develop a Level 1 and a Level 2 PRA before initial fuel load. This PRA must cover those
initiating events and modes for which NRC-endorsed consensus standards on PRA exist 1 year
before the scheduled date for initial loading of fuel into the reactor. Each holder of a combined
license must maintain and update the PRA every 4 years with upgraded consensus standards in
effect 1 year prior to each required upgrade until operations permanently cease. Finally, before
any application for license renewal, a combined license holder must upgrade the PRA to cover
all modes and all initiating events.
10.3.1 Risk-Informed Special Treatment
The agency has approved applications of risk-informed inservice testing, of generally limited
scope. Special treatment requirements for SSCs go beyond industry-established requirements
for equipment classified as commercial grade. Special treatment requirements provide
increased assurance that the SSCs are capable of meeting their functional requirements under
design basis conditions. These special treatment requirements include additional design
considerations, qualification, change control, documentation, reporting, maintenance, testing,
surveillance, and quality assurance requirements. In August 2001, the staff granted the licensee
of the South Texas Project a risk-informed exemption request, which included an exemption
from the prescriptive inservice testing requirements, regarding special treatment requirements
for low-risk and nonrisk-significant safety-related nuclear components. Having successfully
implemented this exemption at the South Texas Project, the staff developed a new rule,
10 CFR 50.69, to allow the application of risk insights to assign the special treatment
requirements in 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities,”
for SSCs according to their safety significance.
The Commission approved the final rule in October 2004. The final rule was published in the
Federal Register on November 22, 2004. The NRC staff issued RG 1.201, “Guidelines for
Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their
Safety Significance,” Revision 1, in May 2006, for trial use.
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A topical report, WCAP-16308-NP, “Pressurized Water Reactor Owners Group 10 CFR 50.69
Pilot Program - Categorization Process - Wolf Creek Generating Station,” Revision 0, dated
July 2006, proposed a categorization process to support implementation of 10 CFR 50.69. The
staff completed its review of the topical report and issued its final safety evaluation in
March 2009. The staff found the categorization process described in the topical report to be
acceptable, but it did not approve or endorse any specific treatment process. Treatment
programs being implemented under 10 CFR 50.69 do not require prior approval from the NRC
as part of the license amendment review process.
The staff has also developed guidance for sample inspections to be conducted at plants
voluntarily choosing to implement 10 CFR 50.69. The performance of sample inspections is
consistent with the statement of considerations accompanying the final 10 CFR 50.69 rule. The
staff has issued draft guidance to obtain stakeholder input and has addressed those comments
with the issuance of the final guidance. Inspection efforts will be focused on the most risk
significant aspects related to implementation of 10 CFR 50.69 (i.e., proper categorization of
SSCs and treatment of Risk-Informed Safety Class (RISC)-1 and RISC-2 SSCs). Additionally,
the inspections are expected to be performance based, with SSCs with a lower safety significant
function, such as those classified RISC-3, not receiving a major portion of inspection focus
unless adverse performance trends are observed.
The staff recognizes the need for an effective, stable, and predictable regulatory climate for the
implementation of 10 CFR 50.69. Inspection guidance developed with industry stakeholder input
is viewed as an efficient vehicle for reaching a common understanding of what constitutes an
acceptable treatment program for SSCs, because the NRC does not review specific treatment
plans as part of a licensee’s application to implement 10 CFR 50.69.
On December 17, 2014, the NRC issued a license amendment approving the Vogtle Electric
Generating Plant, Units 1 and 2, pilot application of 10 CFR 50.69. As necessary, the lessons
learned from this pilot will be incorporated into future revisions of the industry guidance and the
NRC’s regulatory and inspection guidance.
10.3.2 Risk-Informed Inservice Inspection
The NRC uses the guidance in RG 1.178, “An Approach for Plant Specific Risk-Informed
Decisionmaking for Inservice Inspection of Piping,” Revision 1, and NUREG-0800, Chapter 3
Section 3.9.8, “Risk-Informed Inservice Inspection of Piping,” both issued in September 2003, to
evaluate applications of risk-informed inservice inspections. The agency has approved industry
methodologies, one developed by the Westinghouse Owners Group and the other by the
Electric Power Research Institute (EPRI), for alternatives to the ASME Boiler and BPV Code,
Section XI, Inservice Inspection Program.
ASME has also developed Code Case N-716-1, “Alternative Piping Classification and
Examination Requirements, Section XI, Division 1.” Code Case N-716-1 is founded, in large
part, on the risk-informed inservice inspection process as described in NRC-approved EPRI
Topical Report 112657, “Revised Risk-Informed Inservice Inspection Evaluation Procedure,”
Revision B-A, issued in December 1999. NRC-approved EPRI Topical Report TR-1021467-A,
“Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for
Risk-Informed In-Service Inspection Programs,” dated December 2008, provides additional
guidance to ensure that the PRA quality is sufficient to support a risk-informed inservice
inspection program. Code Cases provide alternatives to existing ASME BPV Code requirements
that ASME has developed and approved. RG 1.147, “Inservice Inspection Code Case
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Acceptability, ASME Section XI, Division 1,” Revision 17, issued in August 2014, identifies the
Code Cases that the NRC has determined to be acceptable alternatives to applicable parts of
the ASME BPV Code, Section XI. RG 1.147 endorses Code Case N-716-1.
The NRC regularly participates in the ASME BPV Code development process to resolve issues
on risk-informed inservice inspection methodology.
10.3.3 Risk-Informed Technical Specification Changes
Since the mid-1980s, the NRC has reviewed and granted improvements to technical
specifications that are based, at least in part, on PRA insights. In its “Final Policy Statement on
Technical Specification Improvements for Nuclear Power Reactors,” published in the
Federal Register on July 22, 1993, the Commission stated that it expects licensees to use a
plant-specific probabilistic safety assessment (currently referred to as PRA by the NRC) or risk
survey in preparing submittals related to technical specifications. The Commission reiterated
this point when it revised 10 CFR 50.36, “Technical Specifications,” in July 1995.
The NRC uses RG 1.177, “An Approach for Plant-Specific, Risk-Informed Decisionmaking:
Technical Specifications,” Revision 1, and RG 1.174, “An Approach for Using Probabilistic Risk
Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,”
Revision 2, both issued in May 2011, as guidance to improve plant technical specifications.
Guidance for evaluating the technical basis for proposed risk-informed changes is provided in
NUREG-0800, Chapter 19, Section 19.2, “Review of Risk-Information Used to Support
Permanent Plant-Specific Changes to the Licensing Basis: General Guidance.” Guidance on
evaluating PRA technical adequacy is provided in NUREG-0800, Chapter 19, Section 19.1,
Revision 3. More specific guidance related to risk-informed technical specification changes is
provided in NUREG-0800, Chapter 16, Section 16.1, “Risk-informed Decision-making:
Technical Specifications,” Revision 1, dated March 2007, which includes changes to
surveillance frequencies and completion times as part of risk-informed decisionmaking.
The industry and the NRC continue to increase the use of PRA in developing improvements to
technical specifications. As discussed in a letter from NEI to the NRC dated June 8, 2001
(ADAMS Accession No. ML011690233), the industry proposed eight initiatives to improve
existing technical specification configuration control requirements through the use of risk
insights. The NRC worked with the industry Technical Specifications Task Force (TSTF) to
develop and approve technical specification “change travelers” associated with seven of the
proposed initiatives. The approved initiatives allow licensees to:
(1) modify end states for
some technical specifications required actions to allow certain equipment to be repaired during
hot shutdown instead of cold shutdown, (2) eliminate shutdown requirements for unintentionally
missed surveillances, (3) increase mode change flexibility, and (4) permit a risk-informed delay
time before entering limiting condition for operation actions for inoperability attributable to a loss
of support function provided by equipment not addressed in technical specifications (i.e.,
snubbers and other hazard barriers). The NRC also approved up to a 24-hour completion time
for a very limited scope of technical specification systems (e.g., pressurizer heaters) when both
safety trains are inoperable for Combustion Engineering plants. Additionally, the two initiatives
considered to be the most significant expansion of the use of risk information for improving
technical specifications are summarized below.
Initiative 4b, “Risk-Informed Completion Times.” Initiative 4b modifies technical specification
completion times, in accordance with NRC-approved topical report NEI 06-09, “Risk-Managed
Technical Specifications (RMTS) Guidelines,” Revision 0, issued in November 2006, to reflect a
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configuration risk-management approach that is more consistent with the 10 CFR 50.65,
“Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.”
Industry guidance and the South Texas Project pilot were approved in 2007. The associated
technical specification change traveler can be found in TSTF-505, “Provide Risk-Informed
Extended Completion Times—RITSTF [Risk-Informed Technical Specifications Task Force]
Initiative 4B,” Revision 1. The NRC received a second pilot application in September 2012. The
NRC staff is nearing completion of its review of the application, and is actively working to
resolve any remaining technical issues and provide clarifying guidance. Four additional
applications to implement this program for currently operating reactors have been received and
are currently being reviewed by the technical staff. This program is expected to be widely
adopted by licensees.
Initiative 5b, “Risk-informed Method for Control of Surveillance Frequencies.” Initiative 5b
relocates most periodic frequencies of technical specification surveillances to a licensee
controlled program in accordance with NRC-approved topical report NEI 04-10, “Risk-Informed
Method for Control of Surveillance Frequencies,” Revision 1, dated April 2007. The associated
technical specification change traveler can be found in TSTF-425, “Relocate Surveillance
Frequencies to Licensee Control—RITSTF Initiative 5b,” Revision 3, dated March 18, 2009. This
program has already been implemented by more than half of the U.S. licensees and the NRC
continues to receive and review applications for this initiative.
10.3.4 Development of Standards
The NRC worked with ASME and the American Nuclear Society (ANS) to update the national
consensus standard for PRA quality. In February 2009, ASME and ANS issued their joint PRA
quality standard, ASME/ANS-RA-Sa-2009, “Standard for Level 1/Large Early Release
Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications,” and the NRC
endorsed it in RG 1.200, Revision 2, in March 2009. This PRA quality standard addresses all
hazards at full power operations for core damage frequency (Level 1 PRA) and large early
release frequency (aspect of Level 2 PRA) for current light water reactor designs.
Continuing to work with ASME and ANS, the agency is in the process of supporting issuance of
the next revision to the PRA standard, which will be accompanied by Revision 3 of RG 1.200.
The next revision of the PRA standard will include low power and shutdown modes, Level 2 and
Level 3 PRA, and advanced light water and nonlight water reactor designs, in addition to
updates to internal and external events portions, as needed.12
10.3.5 Level 3 Probabilistic Risk Assessment Project
As directed in Staff Requirements Memorandum (SRM)-SECY-11-0089, “Options for
Proceeding with Future Level 3 Probabilistic Risk Assessment (PRA) Activities,” dated
September 21, 2011, the staff is conducting a full-scope site Level 3 PRA that addresses all
internal and external hazards, plant operating modes, reactor units, spent fuel pools, and dry
cask storage.
12 While an intermediate version, ASME/ANS RA-Sb-2013, was issued in September 2013, this version was not
endorsed nor has it been used in place of ASME/ANS-RA-Sa-2009, which remains in effect today until the cited
next revision is endorsed via RG 1.200, Revision 3.
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The full-scope site Level 3 PRA project includes the following objectives:
x
Develop a Level 3 PRA, generally based on current state-of-practice methods, tools, and
data, that (1) reflects technical advances since completion of the NUREG-1150 study,
titled, “Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,”
dated December 1990, and (2) addresses scope considerations that were not previously
considered (e.g., shutdown and low-power operations, multiunit risk, and spent fuel
storage).
x
Extract new risk insights to enhance regulatory decisionmaking and help focus limited
agency resources on issues most directly related to the agency’s mission to protect
public health and safety.
x
Enhance and improve the PRA staff’s capability and expertise, and documentation to
make PRA information more accessible, retrievable, and understandable.
x
Obtain insight into the technical feasibility and cost of developing new Level 3 PRAs.
Consistent with the objectives of this project, the Level 3 PRA study is largely being carried out
using current PRA state-of-practice methods, tools, and data. However, there are several gaps
in PRA technology, along with other challenges, that require advancement in the PRA
state-of-practice. To address these gaps and challenges for the Level 3 PRA study, the general
approach is to rely primarily on existing research and the collective expertise of the NRC’s
senior technical advisors and contractors, with limited new research for a few specific technical
areas (e.g., multiunit risk).
Based on a set of site selection criteria and support from NEI, Vogtle Electric Generating Plant,
Units 1 and 2, were selected as the volunteer site for the Level 3 PRA study. To enhance the
efficiency in performing the study, the Level 3 PRA project team is leveraging the existing and
available information on the Vogtle facility, the licensee’s PRAs, and related research efforts.
The Level 3 PRA project team is using the following NRC tools for performing the Level 3 PRA
study:
x
Systems Analysis Programs for Hands-on Integrated Reliability Evaluation (SAPHIRE)
x
MELCOR Severe Accident Analysis Code
x
MELCOR Accident Consequence Code System
In addition, the Level 3 PRA study is being made consistent with many of the modeling
conventions used for the standardized plant analysis risk models, which are plant-specific PRA
models used by the staff to support risk-informed regulatory activities.
10.4 Safety Culture
This section covers the policies, programs, and practices that apply to licensee safety culture.
10.4.1 Safety Culture Policy Statement
Industry experience has shown the value of establishing and maintaining a positive safety
culture. The NRC’s Safety Culture Policy Statement outlines the Commission’s expectation that
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all licensees maintain a positive safety culture at their facilities. The NRC defines nuclear safety
culture as the core values and behaviors resulting from a collective commitment by leaders and
individuals to emphasize safety over competing goals to ensure protection of people and the
environment. This policy statement applies to all licensees, certificate holders, permit holders,
authorization holders, holders of quality assurance program approvals, vendors and suppliers of
safety-related components, and applicants for a license, certificate, permit, authorization, or
quality assurance program approval, subject to NRC authority. Safety and security are the
primary pillars of the NRC’s regulatory mission and consideration of both is an underlying
principle of the Safety Culture Policy Statement.
The NRC has identified the following traits of a positive safety culture:
x
Leadership safety values and actions—leaders demonstrate a commitment to safety in
their decisions and behaviors
x
Problem identification and resolution—issues potentially affecting safety are promptly
identified, fully evaluated, and promptly addressed and corrected commensurate with
their significance
x
Personal accountability—all individuals take personal responsibility for safety
x
Work processes—the process of planning and controlling work activities is implemented
so that safety is maintained
x
Continuous learning—opportunities to learn about ways to ensure safety are sought out
and implemented
x
Environment for raising concerns—a safety conscious work environment is maintained in
which personnel feel free to raise safety concerns without fear of retaliation, intimidation,
harassment, or discrimination
x
Effective safety communication—communications maintain a focus on safety
x
Respectful work environment—trust and respect permeate the organization
x
Questioning attitude—individuals avoid complacency and continuously challenge
existing conditions and activities in order to identify discrepancies that might result in
error or inappropriate action
After publication of the policy statement, the NRC engaged the Institute of Nuclear Power
Operations, NEI, and external stakeholders in the reactor community to develop a common
safety culture language using the NRC’s Safety Culture Policy Statement’s traits as a basis.
This language, which was finalized in early 2013, better aligns the industry’s previous safety
culture language with the NRC’s previous safety culture language to allow for more clarification
and enhance understanding of licensee performance. A 10th safety culture trait,
“Decisionmaking—decisions that support or affect nuclear safety are systematic, rigorous, and
thorough,” was added in this common language effort for the reactor community. The NRC
updated all guidance and inspection documents appropriately with the new common safety
culture language and published NUREG-2165, “Safety Culture Common Language,” in
March 2014.
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10.4.2 NRC Monitoring of Licensee Safety Culture
10.4.2.1 Background
Section 6.3.2 of this report describes the Reactor Oversight Process. Based on lessons learned
from the Davis-Besse reactor pressure vessel head degradation event and other considerations,
the NRC enhanced the Reactor Oversight Process to more fully address safety culture and
identify safety culture problems earlier so that corrective steps can be taken to address the
problems and prevent further plant performance degradation.
10.4.2.2 Enhanced Reactor Oversight Process
Licensees perform periodic, voluntary self-assessments of safety culture in accordance with
industry guidelines. There are no regulatory requirements for licensees to perform safety culture
assessments routinely. However, depending on the extent of deterioration of licensee
performance, the NRC has a range of expectations about regulatory actions and licensee safety
culture assessments, as described below.
The Reactor Oversight Process uses a graded approach, such that plants that are performing in
a specified manner warrant only a routine level of inspection and oversight. However, as
licensee performance deteriorates, inspection and oversight become increasingly more intrusive
to ensure safe plant operation. The Reactor Oversight Process safety culture enhancements
continue to allow licensees to self-diagnose and implement corrective actions for their
performance problems before the NRC performs followup inspections.
The Reactor Oversight Process applies the safety culture traits and attributes of NUREG-2165
to the inspection and assessment of licensee performance as described in Inspection Manual
Chapter 0310, “Aspects within the Cross-Cutting Areas,” dated February 23, 2010. For most
licensees (i.e., those listed in the Licensee Response column, Column 1, of the Reactor
Oversight Process Action Matrix), the NRC performs the baseline inspection program. In the
routine or baseline inspection program, the inspector will develop an inspection finding and then
identify whether an aspect of safety culture (e.g., a cross-cutting aspect) is a significant causal
factor of the finding. The NRC communicates the inspection findings to the licensee along with
the associated cross-cutting aspect.
When performing inspections using IP 71152 “Problem Identification and Resolution,” dated
February 26, 2015, inspectors have the option to review licensee self-assessments of safety
culture. This IP also instructs inspectors to be aware of safety culture attributes when selecting
samples. In addition, the procedure contains enhanced questions related to a safety-conscious
work environment.
IP 71153, “Followup of Events and Notices of Enforcement Discretion,” dated
December 17, 2015, directs inspection teams to consider contributing causes related to the
safety culture attributes as part of their efforts to fully understand the circumstances surrounding
an event and its probable cause(s).
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As part of the assessment process, the NRC considers the aspects of safety culture
components associated with inspection findings to determine whether common themes exist at
a plant. If, over three consecutive assessment periods (i.e., 18 months), a licensee has the
same safety culture issue with the same common theme, the NRC may ask the licensee to
conduct a safety culture self-assessment.
As licensee performance declines (Regulatory Response column, Column 2, of the Reactor
Oversight Process Action Matrix), the inspectors, through a specific supplemental IP, verify that
the licensee’s root cause, extent of condition, and extent of cause evaluations for the
risk-significant finding(s) appropriately considered the safety culture attributes.
If the licensee performance degrades further (Degraded Cornerstone column, Column 3, of the
Reactor Oversight Process Action Matrix), the NRC expects that the licensee’s root cause
evaluation for the risk-significant finding(s) will determine whether any safety culture attribute
contributed to the risk-significant performance issues. If, through the conduct of supplemental
inspection using IP 95002, “Inspection for One Degraded Cornerstone or any Three White
Inputs in a Strategic Performance Area,” dated February 9, 2011, the NRC determines that the
licensee did not recognize that existing or suspected safety culture attributes caused or
significantly contributed to the risk-significant performance issues, the NRC may request the
licensee to complete an independent assessment of its safety culture.
Finally, for licensees with more significant performance degradation (Multiple/Degraded
Cornerstone column, Column 4, of the Reactor Oversight Process Action Matrix), the NRC will
expect the licensee to conduct a third-party independent assessment of its safety culture. The
NRC will review the licensee’s assessment and will conduct an independent assessment of the
licensee’s safety culture through a specific supplemental IP 95003, “Supplemental Inspection for
Repetitive Degraded Cornerstones, Multiple Degraded Cornerstones, Multiple Yellow Inputs or
One Red Input,” that was substantially revised in December 2015, to provide guidance for these
assessments.
Considerations of safety culture within the Reactor Oversight Process provide the NRC staff
with (1) better opportunities to consider safety culture weaknesses and to encourage licensees
to take appropriate actions before significant performance degradation occurs, (2) a process to
determine the need to specifically evaluate a licensee’s safety culture after performance
problems have resulted in the placement of a licensee in the Degraded Cornerstone column of
the Reactor Oversight Process Action Matrix, and (3) a structured process to evaluate the
licensee’s safety culture assessment and to independently conduct a safety culture assessment
for a licensee in the Multiple/Repetitive Degraded Cornerstone column of the Action Matrix.
By using the existing Reactor Oversight Process framework, the NRC’s safety culture oversight
activities are based on a graded approach and remain transparent, understandable, objective,
risk-informed, performance-based, and predictable. These activities range from requesting the
licensee perform a safety culture self-assessment to a meeting between senior NRC managers
and a licensee’s Board of Directors to discuss licensee performance issues and actions to
address persistent and continuing safety culture cross-cutting issues.
10.4.3 The NRC Safety Culture
The NRC fosters a culture in which all employees may live the NRC’s values, demonstrate a
positive safety culture, and adhere to the Principles of Good Regulation to support the NRC’s
mission to protect public health, safety, and the environment. The NRC culture includes a
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system of shared values, beliefs, and behaviors that demonstrate our collective commitment to
emphasize safety as the overriding priority in our regulatory decisionmaking, and that
recognizes the important role each employee plays in the NRC’s success. The NRC is
committed to creating and sustaining a positive work environment to ensure we remain a model
regulator.
The NRC acknowledges the nature and purpose of a regulatory body is distinct from that of its
licensees; therefore, the practical applications of ensuring a positive safety culture are slightly
different and unique in some sense. Although many similarities regarding safety culture exist in
any organization, the NRC emphasizes and relays the importance of safety culture as an
inherent component of the broader NRC organizational culture that is complementary to, but
distinct from, the NRC’s regulatory oversight of licensees’ safety culture.
The NRC emphasizes the notion that safety is every employee’s responsibility. When each NRC
employee demonstrates a level of responsibility for their behaviors and attitudes, which support
a positive safety culture, it produces immeasurable gains that lead to higher operating margins
across the board. Previous studies conducted at the NRC have revealed that high levels of key
safety culture indices result in an engaged, enabled, and energized workforce- all of which
comprise sustainable engagement. Thus, when safety culture indices increase, employee
engagement increases. For this reason, the NRC has focused on achieving a positive safety
culture and considers it to be a key driver of sustainable engagement.
Three key components of the NRC’s safety culture include:
(1)
Creating an environment that encourages all NRC employees and contractors to raise
concerns and differing views promptly, without fear of reprisal. The free and open
exchange of views or ideas conducted in a nonthreatening environment provides the
ideal forum where concerns and alternative views can be considered and addressed in
an efficient and timely manner that improves decisionmaking and supports the agency’s
safety and security mission.
(2)
The NRC’s commitment to the free and open discussion of professional views is
illustrated by providing multiple ways for employees and contractors to raise mission-
related concerns and differing views. Although all NRC employees and contractors are
expected to discuss their views and concerns with their immediate supervisors on a
regular, ongoing basis, there are times when informal discussions are not sufficient to
resolve issues. The NRC uses a three-tiered approach for addressing concerns and
differing views, including the processes described in Management Directive (MD)
10.160, “Open Door Policy,” dated October 26, 2015, MD 10.158, “NRC Non-
Concurrence Process,” dated March 14, 2014, and MD 10.159, “The NRC Differing
Professional Opinions Program,” dated May 16, 2004. These directives provide
increasing levels of formality to air differences: the broad Open Door Policy is least
formal and does not require documentation, the Non-Concurrence Process requires
documentation, and the Differing Professional Opinions Program is most formal and
provides for a high level of agency review. The NRC believes that the existence of
multiple channels for expressing disagreement helps create a positive environment for
raising concerns by reducing barriers to expressing differing opinions. The
Non-Concurrence Process and Differing Professional Opinions Program also support
our openness value, in that when the process is complete, an employee can request to
make the records public.
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(3)
The NRC conducts assessments of our safety culture and continually reviews results
and develops action plans to improve. In addition, the NRC recognizes the need for
continuous improvement to maintain a positive safety culture. Complacency lends itself
to a degradation in safety culture when new information and historical lessons are not
processed and used to enhance the NRC and its regulatory products.
The agency uses the Office of the Inspector General’s triennial Safety Culture and Climate
Survey, as well as postsurvey assessment activities (e.g., focus groups, and employee
interviews), to assess the effectiveness of new and existing safety culture efforts. In 1998, the
Office of the Inspector General conducted the first in a continuing series of Safety Culture and
Climate Surveys to identify areas for additional organizational improvements. The surveys are
voluntary, provide for anonymity, and are offered to all NRC employees, supervisors, and
managers. The Office of the Inspector General has conducted the Safety Culture and Climate
Surveys six times —in 1998, 2002, 2005, 2009, 2012, and 2015.
The Government-administered Federal Employee Viewpoint Survey provides an annual check
on topics such as leadership, employee engagement and job satisfaction. The U.S. Office of
Personnel Management has conducted the Federal Employee Viewpoint Survey since 2002 and
annually since 2010. A survey such as this makes it possible to compare results over time to
assess increasing or declining trends. Action plans are developed at the agency, office, and
Region levels to address areas needing improvement, and those plans are evaluated each year
and updated, as necessary.
10.5 Managing the Safety and Security Interface
Safety and security have always been the primary pillars of the NRC’s regulatory programs.
Safety and security activities are closely intertwined, and it is critical that consideration be given
to the integration of safety and security activities so as not to diminish or adversely affect either.
Although many safety and security activities complement each other, there is the potential for
security measures to inadvertently affect plant safety, or, for safety activities to inadvertently
affect security. Recognizing the potential for adverse impact, the NRC maintains its attention to
the interfaces between safety and security during both normal (day-to-day operations) and
emergency conditions.
The NRC’s mission statement and strategic goals are achieved, in part, through a regulatory
framework that stresses the importance of maintaining both safety and security under all site
conditions. The NRC continues its efforts in the areas of rulemaking, licensing, emergency
planning, and inspection to recognize, establish, and improve this interface. For example, the
NRC has been working multilaterally with the International Atomic Energy Agency and bilaterally
with its international counterparts to promote this concept. In March 2009, the NRC also issued
10 CFR 73.58, “Safety/Security Interface Requirements for Nuclear Power Reactors,” which
requires licensees to assess and manage changes to safety and security activities. In addition,
the NRC issued RG 5.74, “Managing the Safety/Security Interface,” in June 2009, describing
acceptable methods that could be used to meet the safety and security interface requirements
of 10 CFR 73.58. Revision 1 of RG 5.74 was issued in April 2015, to include cyber security as
part of the safety and security assessment.
From 2004 to 2012, as part of the NRC’s increased focus on security events after the events of
September 11, 2001, security issues were considered in the Reactor Oversight Process through
a different assessment process than safety issues. To enhance consistency in the assessment
process, on July 20, 2011, the Commission issued SRM-SECY-11-0073, “Staff Proposal to
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Reintegrate Security into the Action Matrix of the Reactor Oversight Process Assessment
Program,” approving the reintegration of the security cornerstone in the reactor assessment
process. As described in RIS 2012-03, “Reintegration of Security into the Reactor Oversight
Process Assessment Program,” dated March 14, 2012, this reintegration became effective on
July 1, 2012.
Satisfactory licensee performance in the Reactor Oversight Process cornerstones provides
reasonable assurance of safe and secure facility operation during both normal and emergency
conditions and assurance that the NRC’s safety and security missions are being effectively
accomplished. Like the other cornerstones, the security cornerstone contains IPs and
performance indicators to ensure that its objectives are being met. The NRC evaluates safety
and security interface issues relative to their implications among the cornerstones and in the
cross-cutting areas of human performance, safety conscious work environment, emergency
planning, and problem identification and resolution. Safety and security activities are integrated
into the NRC’s regulatory framework and evaluated by the NRC staff using an integrated
assessment process. To ensure licensees are complying with the regulations, the NRC has
incorporated the evaluation of the licensee’s safety and security interface processes into its IPs.
The section of this report on nuclear programs and Section 6.3.2 of this report discuss the
Reactor Oversight Process in more detail.
The NRC also recognizes the impact that organizational safety culture has on both safety and
security, as well as on the interface between the two areas. The ongoing effort to implement the
NRC’s Safety Culture Policy Statement is described in detail in Section 10.4 of this report.
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ARTICLE 11. FINANCIAL AND HUMAN RESOURCES
1.
Each Contracting Party shall take the appropriate steps to ensure that adequate
financial resources are available to support the safety of each nuclear installation
throughout its life.
2.
Each Contracting Party shall take the appropriate steps to ensure that sufficient
numbers of qualified staff with appropriate education, training, and retraining are
available for all safety-related activities in or for each nuclear installation,
throughout its life.
This section explains the requirements about financial resources that licensees must have to
support the nuclear installation throughout its life, and the regulatory requirements for qualifying,
training, and retraining personnel.
There have been no changes in licensee financial resource considerations as a result of the
Fukushima nuclear accident. Training related to the orders and the proposed rulemaking related
to Fukushima lessons learned are described in Sections 1.3.1 and 1.3.3 of this report.
11.1 Financial Resources
Currently, the financial qualification regulations of the U.S. NRC are codified in 10 CFR Part 50,
“Domestic Licensing of Production and Utilization Facilities,” and 10 CFR Part 52, “Licenses,
Certifications, and Approvals for Nuclear Power Plants.” They require applicants for a
construction permit, operating license, or combined license to provide reasonable assurance of
adequate funds to safely construct and operate nuclear production and utilization facilities. This
means that applicants must provide information specifying their legal and financial relationships
with stakeholders, corporate affiliates, or financial institutions upon which the applicant is relying
for financial assistance, and information to support the financial capability of each such entity to
meet its financial commitment to the applicant. After closely examining the current financial
qualification regulations, the NRC has determined that the details of these arrangements go well
beyond the NRC’s mandate of ensuring public health and safety. Therefore, the NRC is
considering the conformance of the existing 10 CFR Part 50 standard to be consistent with a 10
CFR Part 70, “Domestic Licensing of Special Nuclear Material,” standard requiring a licensee to
only demonstrate that it “appears to be financially qualified” to construct and operate a facility
safely.
Additionally, the NRC’s regulations at 10 CFR 50.54(w) and 10 CFR Part 140, “Financial
Protection Requirements and Indemnity Agreements,” require licensees to maintain financial
protection in the form of onsite and offsite liability insurance. This insurance provides the
licensee with financial assistance for any claims of bodily injury and property damage resulting
from a nuclear incident, and helps pay onsite recovery costs. Additional information can be
found in Sections 11.1.3 and 11.1.4 of this report.
The NRC also maintains decommissioning funding and related reporting requirements under
10 CFR 50.75, “Reporting and Recordkeeping for Decommissioning Planning,” and
10 CFR 50.82, “Termination of License,” throughout the life of a reactor facility, and regularly
reviews the status of licensees’ decommissioning trust funds. These detailed reviews provide
NRC reasonable assurance that licensees maintain adequate funds to safely decommission
their facilities.
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11.1.1 Financial Qualifications for Construction and Operations
This section explains the financial qualifications program for construction and operations and
describes NRC reviews for construction permits, operating licenses, combined licenses,
postoperating nontransferred licenses, and license transfers.
Section 182.a of the Atomic Energy Act provides that “each application for a license … shall
specifically state such information as the Commission, by rule or regulation, may determine to
be necessary to decide such of the technical and financial qualifications of the applicant … as
the Commission may deem appropriate for the license.” To implement this provision, the NRC
has developed the regulations and guidance discussed below.
On April 24, 2014, the Commission issued SRM-SECY-13-0124, “Policy Options for Merchant
(Non-Electric Utility) Plant Financial Qualifications,” approving the staff’s recommendation to
conduct a rulemaking to amend the financial qualifications requirements in 10 CFR Part 50 to
conform to standards contained in 10 CFR Part 70. The rulemaking would permit the inclusion
of a license condition to assure applicant financial qualifications reflecting the revised standards
for review, and require the applicant to submit a plan for how it will proceed to finance the
construction and operation of the facility to ensure that the applicant has a well-articulated
understanding of the size of the project it is undertaking and the financial capacity to obtain the
necessary financing when the applicant is ready to start construction. The NRC staff is currently
revising the draft regulatory basis for the rulemaking, based on public comments, and
anticipates that the rulemaking activities will continue through calendar year 2017.
11.1.1.1 Construction Permit Reviews
As required by 10 CFR 50.33(f)(1), applicants for construction permits must submit information
that “demonstrates that the applicant possesses or has reasonable assurance of obtaining the
funds necessary to cover estimated construction costs and related fuel cycle costs.”
Appendix C, to 10 CFR Part 50, “A Guide for the Financial Data and Related Information
Required to Establish Financial Qualifications for Facility Construction Permits,” provides more
specific directions for evaluating the financial qualifications of applicants.
NUREG-1577, “Standard Review Plan on Power Reactor Licensee Financial Qualifications and
Decommissioning Funding Assurance,” Revision 1, provides staff guidance for its review and
approval of an applicant’s and licensee’s financial qualification during initial plant construction
and operations.
11.1.1.2 Operating License Reviews
An “electric utility” as defined in 10 CFR 50.2, “Definitions,” is “any entity that generates or
distributes electricity and which recovers the cost of this electricity, either directly or indirectly,
through rates established by the entity itself or by a separate regulatory authority.” Electric
utilities are exempt under 10 CFR 50.33(f) from reviews of financial qualifications of applications
for operating licenses. The reason for this exemption is that cost-of-service rate regulation, as it
has existed in the United States, has ensured that ratepayers provide a source of funds for the
safe operation of nuclear power plants. Applicants for operating licenses that are not electric
utilities are required under 10 CFR 50.33(f)(2) to submit information that demonstrates that they
possess or have reasonable assurance of obtaining the necessary funds to cover estimated
operating costs. Nonelectric-utility applicants for operating licenses are also required to submit
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estimates of the total annual operating costs for each of the first 5 years of operation of their
facilities including the sources of funds to cover these costs.
11.1.1.3 Combined License Application Reviews
As authorized in 10 CFR Part 52, applicants may apply for a combined construction permit and
operating license. Under 10 CFR 52.77, “Contents of Applications; General Information,” such
applications must contain all of the information required under 10 CFR 50.33, including
information about financial qualifications. Under the requirements in 10 CFR 50.33(f)(4), each
application for a combined license submitted by a newly-formed entity organized for the primary
purpose of constructing or operating a facility must include information showing:
(1) the legal
and financial relationships it has or proposes to have with its stockholders or owners, (2) the
stockholders’ or owners’ financial ability to meet any contractual obligation to the entity that they
have incurred or proposed to incur, and (3) any other information considered necessary by the
Commission to enable it to determine the applicant’s financial qualification.
11.1.1.4 Postoperating License Nontransfer Reviews
The NRC does not systematically review the financial qualifications of power reactor licensees
once it has issued an operating license, other than for license transfers as described below.
However, as provided in 10 CFR 50.33(f)(5), the NRC can seek additional information on
licensees’ financial resources if the agency considers such information appropriate. For
example, the staff may review financial and industry trade press as well as other publicly
available information, such as Securities and Exchange Commission and Federal Energy
Regulatory Commission submissions, to identify potential changes in licensees’ financial health.
If the review of any of these sources indicates that a licensee’s financial health may be
deteriorating, the NRC can request additional financial information from the licensee as
authorized by 10 CFR 50.33(f)(5) to confirm that a licensee has the financial resources to
operate the facility safely. On March 4, 2015, the NRC published OL/FR-ISG-2014-01, “Interim
Staff Guidance - Reviewing and Assessing the Financial Condition of Operating Power Reactor
Licensees, Including Requests for Additional Information,” on its process for reviewing and
assessing licensee financial conditions.
11.1.1.5 Reviews of License Transfers
The NRC regulations in 10 CFR 50.80, “Transfer of Licenses,” require agency review and
approval of transfers of operating licenses, including licenses for nuclear power plants owned or
operated by electric utilities. The NRC performs these reviews to determine whether a proposed
transferee or new owner is technically and financially qualified to hold the license.
NUREG-1577 provides staff guidance for its review and approval of applicants’ and licensees’
financial qualifications during initial plant construction and operations, including license
transfers. Specifically, NUREG-1577 requests staff to determine whether, in the case of a direct
transfer, a proposed transferee is qualified to hold the license, or whether, in the case of an
indirect transfer, the holder of the license is qualified to hold the license. The regulations at 10
CFR 50.80(b) require license transfer applicants to include as much of the information with
respect to, among other things, the financial qualifications of the proposed holder of the license
as required in section 10 CFR 50.33(f). The reviewer should evaluate the financial qualifications
associated with these transfers by:
(1) determining whether the proposed holder of the license
will remain an electric utility following the direct or indirect transfer; (2) for nonelectric-utility
applicants, reviewing the recent financial performance of the proposed transferee, or, if the
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proposed transferee is a new entity such as an operating, generating, or service company
subsidiary, evaluating the ownership or participation agreement with its owners or other
responsible party; and (3) identifying all parent companies that are not licensed by the NRC or
did not undergo a 10 CFR 50.80 review.
11.1.2 Financial Qualifications Program for Decommissioning
The Atomic Energy Act establishes the basis for the NRC’s regulations and guidance on
decommissioning funding assurance. The NRC’s regulations at 10 CFR 50.75 and
10 CFR 50.82 require an applicant or licensee to provide the NRC with reasonable assurance of
its plan to safely decommission a facility, including a cost estimate, the mechanism
(e.g., establishment of a dedicated trust fund) and schedule to pay for decommissioning, and a
certification that financial assurance for decommissioning will be, or has been provided.
Additionally, the NRC has a comprehensive decommissioning funding oversight program in
place to provide reasonable assurance that sufficient funds will be available for radiological
decommissioning of all U.S. commercial nuclear reactors. Under 10 CFR 50.75, this program
requires operating reactor licensees to submit biennial Decommissioning Funding Status
Reports, which includes, at a minimum:
x
the amount of decommissioning funds estimated to be required pursuant to
10 CFR 50.75(b) and (c)
x
the amount of decommissioning funds accumulated to the end of the calendar year
preceding the date of the report
x
a schedule of the annual amounts remaining to be collected
x
the assumptions used regarding rates of escalation in decommissioning costs, rates of
earnings on decommissioning funds, and rates of other factors used in funding
projections
x
any contracts upon which the licensee is relying
x
any modifications occurring to a licensee’s current method of providing financial
assurance since the last submitted report
x
any material changes to trust agreements
For power reactors that have ceased operations and are in decommissioning, similar reports are
submitted on annual basis under 10 CFR 50.82, and includes information regarding the amount
of decommissioning funds spent over the calendar year and the amount of remaining funds
needed to complete decommissioning.
NRC-required decommissioning trust funds are designed in such a way as to protect the funds
from withdrawals for expenditures other than those specifically authorized by NRC regulations.
The intent of the trust funds is to cover the costs associated with the radiological
decommissioning of the reactor facility, resulting from the termination of the NRC-issued
license.
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11.1.3 Financial Protection Program for Liability Claims Arising from Incidents
The Price-Anderson Act of 1957, which became Section 170 of the Atomic Energy Act, governs
the U.S. financial protection program. Along with related definitions in Section 11, Section 170
supplies the financial and legal frameworks to compensate those who suffer bodily injury or
property damage as a result of incidents at nuclear facilities covered by the law. The NRC
regulations implementing the provisions of Section 170 for NRC licensees are codified in
10 CFR Part 140.
The Price-Anderson Act was enacted to (1) remove the deterrent to private-sector participation
in atomic energy presented by the threat of potentially enormous liability claims in the event of a
catastrophic nuclear incident and (2) ensure that adequate funds are available to the public to
satisfy liability claims if such an incident were to occur.
The Price-Anderson Act was most recently revised in 2005, when Congress renewed the
Commission’s authority to regulate insurance requirements for nuclear facilities until 2020.
Under the current law, power reactors of 100,000 kilowatts electric or more must contribute to a
funding pool that is enacted if the primary layer of financial protection (in the form of private
liability insurance — now at $375 million) is exhausted. The NRC is required by Section 170(t)
to adjust these amounts for inflation every 5 years based on the aggregate change in consumer
price index.
Reactor operators must pay into a funding pool for the secondary layer of protection, called the
“retrospective premium pool” in maximum annual installments not to exceed $18.963 million, up
to a total of $121.255 million each. However, payment is required only if an incident exhausts 15
percent of the first layer of financial protection, that is currently set at $375 million, and only if
and to the extent that, additional funds are needed to pay the damages. With 10213 reactors
currently participating in the system, the total financial protection available under the
Price-Anderson Act for any one incident is approximately $13.3 billion (i.e., $375 million of
primary coverage plus $121.255 million per reactor times 102 reactors), which is also the limit
on liability. The limit of insurance coverage fluctuates as reactor licensees join or withdraw from
the retrospective premium pool. A change in the limit also may occur when the amount of
insurance coverage in either the primary or secondary tier is adjusted for inflation, as must be
done every 5 years. In any potential incident, Congress will address any damages exceeding
the total sum that reactor operators must contribute to the pool and will decide upon the next
steps needed for compensation.
The public benefits significantly from another feature of the Price-Anderson Act. Neither proof of
fault, nor proof of what caused the incident, is necessary to issue a claim. This feature helps to
ensure that potential claims are settled without delay from deliberation in the court system.
As of 2015, claims for more than 240 alleged incidents involving nuclear material have been
filed under various liability policies since the inception of the Price-Anderson Act in 1957. To
date, the insured losses and expenses paid are approximately $507 million. Insurance pools
13 The number or reactors participating in the Price-Anderson Act system depends on granted insurance
exemptions for the respective reactors, and it is not dependent on the number of reactors currently in operation
or reactors in decommissioning. In the U.S., there are currently 100 nuclear reactors in operation.
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paid out a total of approximately $71 million in claims and litigation costs in association with the
Three Mile Island incident.
Separate from the Price-Anderson Act, the U.S. is a party to the Convention on Supplementary
Compensation for Nuclear Damage, which was developed under the auspices of the IAEA to be
the basis for a global nuclear liability regime. The Convention on Supplementary Compensation
for Nuclear Damage was adopted on September 12, 1997, and was opened for signature on
September 29, 1997. The United States signed at that time. The United States deposited its
instrument of ratification in May 2008. The Convention on Supplementary Compensation for
Nuclear Damage entered into force on April 15, 2015. The current parties include Argentina,
India, Japan, Montenegro, Morocco, Romania, the United Arab Emirates, and the United States.
11.1.4 Insurance Program for Onsite Property Damages Arising from Incidents
Among other sections of the Atomic Energy Act, Section 182.a gives the basis for the NRC’s
onsite property damage insurance requirements for operating nuclear power reactors contained
in 10 CFR 50.54(w). Onsite insurance provides financial protection to stabilize and
decontaminate the reactor and reactor station site at which the reactor experiencing an incident
is located.
The U.S. nuclear power industry has not experienced an incident involving offsite radioactive
release within the scope of this program since the Three Mile Island, Unit 2, event in 1979.
11.2 Regulatory Requirements for Qualifying, Training, and Retraining Personnel
This section explains the regulatory requirements for qualifying, training, and retraining
personnel. It discusses the governing documents, the process for implementing requirements,
and experience. It also discusses INPO accreditation activities.
11.2.1 Governing Documents and Process
The NRC regulates the training requirements for licensed operators and licensed senior
operators under 10 CFR Part 55, “Operators’ Licenses,” which allows facility licensees to have
operator requalification program content that is derived using a systems approach to training
(SAT), as defined in 10 CFR 55.4, “Definitions,” or that meets the requirements outlined in
10 CFR 55.59(c). Subpart D, “Applications,” of 10 CFR Part 55 requires that operator license
applications must contain information about an individual’s training and experience, unless the
facility licensee certifies that the applicant has successfully completed a Commission-approved
training program that is SAT-based and uses an acceptable simulation facility.
Both initial licensing and requalification training include training conducted on a control room
simulator. Although the NRC does not mandate specific simulator training requirements
(i.e., simulator training is determined by each facility licensee through the SAT process), typical
initial licensing classes include 200 or more hours of simulator training, whereas requalification
training includes 40 or more hours per year of simulator training. Simulator training includes
normal integrated plant operations (e.g., startups, shutdowns, heat ups, cool downs, refueling,
testing, technical specifications); abnormal, alarm, and transient response; and emergency
response, including safety function challenges.
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Associated with emergency response, operators and other plant staff are trained and examined
on aspects of the facility’s emergency plan, including requirements for maintaining sufficient
staff during all modes of plant operation. For operating crews, routine emergency response
training is conducted in the simulator using short (approximately 1-2 hour) scenarios. A facility’s
complete emergency response organization is exercised once every 2 years using scenarios
lasting several hours during drills that the NRC observes.
The operator licensing process at power reactors includes a generic fundamentals examination
covering the theoretical knowledge required to operate a nuclear power plant. License
applicants must pass the generic fundamentals examination before they can take a site-specific
examination. The site-specific examination consists of a written examination and an operating
test that includes a plant walkthrough and a dynamic performance demonstration on a
simulation facility.
The NRC staff has transferred most of the responsibility for developing site-specific licensing
examinations to facility licensees. In 1999, the NRC amended 10 CFR Part 55 to allow nuclear
power reactor licensees to prepare the written examinations and operating tests that the agency
uses to evaluate the competence of applicants for operators’ licenses at those facilities.
Licensees that elect to prepare their own examinations are required to establish procedures to
control examination security and integrity. They prepare and submit proposed examinations and
operating tests to the NRC according to the guidance in NUREG-1021, “Operator Licensing
Examination Standards for Power Reactors,” Revision 10, issued in December 2014. The NRC
reviews the facility-prepared examinations, prepares examinations for facility licensees upon
request, administers all operating tests, makes the final licensing decisions, and issues the
licenses.
As required by 10 CFR 50.120, “Training and Qualification of Nuclear Power Plant Personnel,”
licensees must establish, implement, and maintain training programs using a SAT approach for
eight categories of nonlicensed workers at nuclear power plants and for the shift supervisor,
who is licensed in accordance with 10 CFR Part 55. These provisions complement the
requirements for training based on a systems approach for the requalification of licensed
operators and licensed senior operators. RG 1.8, “Qualification and Training of Personnel for
Nuclear Power Plants,” Revision 3, issued in May 2000, contains guidance to implement the
regulations.
The NRC continues to endorse the training accreditation process that INPO manages. The staff
recognizes that training programs developed in accordance with INPO guidelines and
accredited by the National Nuclear Accrediting Board are SAT based; therefore, accredited
programs are considered to be consistent with the regulations in 10 CFR Part 55 and
10 CFR 50.120. The NRC also recognizes that INPO-managed accreditation and associated
training evaluation activities are an acceptable way of self-improvement in training. Such
recognition encourages industry initiative and reduces NRC evaluation and inspection activities.
In accordance with its memorandum of agreement with INPO, the NRC monitors INPO
accreditation activities as part of its continuing assessment of the effectiveness of the industry’s
training programs. Specifically, the NRC staff observes selected accreditation team visits and
NRC managers periodically observe National Nuclear Accrediting Board meetings. These
observations are intended to monitor the implementation of programmatic aspects of the
accreditation process, and they also give an opportunity to assess the selected performance
areas of facility licensees.
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If the National Nuclear Accrediting Board has concerns about the performance of an accredited
training program, it will place the program on probation. This does not necessarily place a
training program in noncompliance with either 10 CFR Part 55 or 10 CFR 50.120 because
training programs are accredited to a standard of excellence rather than to a minimum level of
regulatory compliance. However, the NRC does review the circumstances leading to the
probation to ensure safe operations and continued compliance with the regulations.
The National Nuclear Accrediting Board may also withdraw accreditation in response to major
deficiencies in a licensee’s accredited training program. If accreditation is withdrawn, the NRC
would ask that the licensee report the circumstances of the withdrawal for the staff to determine
the significance of the issues related to the withdrawal. If the NRC determines that compliance
with the regulations is not affected, it may not be necessary to take any further action. If the
withdrawal is linked to a breakdown in the training process or a safety-significant issue, the NRC
will conduct an immediate inspection focused on the process problem or safety issues. If
appropriate, the agency would take further action, such as issuing confirmatory action letters or
orders.
The NRC monitors industry performance in implementing the training requirements of
10 CFR Part 50 and 10 CFR Part 55 by (1) reviewing licensee event reports and inspection
reports for training issues, (2) observing the accreditation process, and (3) reviewing the results
of operator licensing activities. Guidance for periodically inspecting the licensed operator
requalification training program at every facility is given in IP 71111.11, “Licensed Operator
Requalification Program and Licensed Operator Performance,” dated September 24, 2014.
When appropriate for cause, the NRC will also use IP 41500, “Training and Qualification
Effectiveness,” dated June 13, 1995, which references the guidance in NUREG-1220, “Training
Review Criteria and Procedures,” Revision 1, issued in January 1993, to verify compliance with
SAT requirements.
11.2.2 Experience
The NRC continually reviews operating experience information (e.g., event reports, inspection
reports, reactor scrams, safety system actuations and failures, and forced plant outages) and
monitors for trends concerning human performance, decisionmaking, and training, among other
areas. Since the last CNS report was issued in 2013, there has been no notable increase in the
trends associated with training deficiencies and operator errors. However, the NRC has noticed
increased examples of nonconservative decisions that facility licensee personnel have made
over the past few years, and the NRC has provided additional inspector guidance when
reviewing certain decisions (i.e., equipment operability determinations) that facility licensees
have made.
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ARTICLE 12. HUMAN FACTORS
Each Contracting Party shall take the appropriate steps to ensure that the capabilities
and limitations of human performance are taken into account throughout the life of a
nuclear installation.
This section explains the program on human performance run by the U.S. NRC. This program
has seven major areas:
(1) human factors engineering, (2) emergency operating procedures
and plant procedures, (3) staffing, (4) fitness for duty, (5) the Human Factors Information
System, (6) support to event investigations and for-cause inspections, and (7) training. This
section also discusses lessons learned from Fukushima.
12.1 Goals and Mission of the Program
The NRC has a comprehensive program for ensuring that human performance is properly
addressed in a risk-informed regulatory framework for maintaining reactor safety. The NRC
developed the program based on reviewing risk information and activities in the domestic and
international nuclear industry.
12.2 Program Elements
The Reactor Oversight Process (discussed in Article 6) focuses on safety cornerstones that are
assessed through a combination of performance indicators and risk-informed inspections that
focus on risk-significant activities and systems related to the cornerstones. The three elements
that cut across the cornerstones are human performance, a safety-conscious work environment,
and problem identification and resolution. The Human Performance Program has contributed
directly to the development of a supplemental IP related to the human performance
cross-cutting element. The Human Performance Program is also engaged in the other two
elements, as a safety-conscious work environment and many of the actions involved in
corrective action programs result from human performance problems.
The Human Performance Program also supports the risk-informed and performance-based plan
by generating, collecting, and evaluating data on human performance for use in human
reliability analysis models. The staff evaluates information to gain insights supporting
risk-informed regulation and to find human performance data for human reliability analysis. The
NRC is working with industry to develop and implement the Scenario Authoring,
Characterization, and Debriefing Application database to collect licensed operator simulator
training and experimental data to support regulatory applications in human reliability analysis
and human factors.
The Human Performance Program monitors technological developments and emerging issues
to help prepare the NRC for the future. Because a number of licensees are replacing analog
controls and displays with digital components, the NRC must be prepared to review safety
issues for human-system interfaces resulting from such new designs and technologies. The
NRC has been processing a few industry requests to transfer operating licenses due to changes
of ownership of nuclear power plants. Changes in ownership often involve changes in
organizational structure. Some organizational changes may have the potential to affect human
performance, especially if operations are prioritized in the structure over safety (e.g., safety
organizations are marginalized or devalued when they report to senior management through an
operational portion of the organization).
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12.3 Significant Regulatory Activities
The NRC performs significant regulatory activities in the following areas to address human
performance:
x
human factors engineering
x
emergency operating procedures and plant procedures
x
staffing
x
fitness for duty
x
Human Factors Information System
x
support to event investigations and for-cause inspections
x
training
The following sections cover the first six activities. Article 11 of this report describes training.
12.3.1 Human Factors Engineering
This section discusses human factors activities related to plant design.
Governing Documents and Process. The NRC evaluates the human factors engineering design
of the main control room and control centers outside of the main control room using
NUREG-0800, Chapter 18, “Human Factors Engineering,” Revision 2, issued in March 2007;
NUREG-0700, “Human System Interface Design Review Guideline,” Revision 2, issued in
May 2002; and NUREG-0711, “Human Factors Engineering Program Review Model,”
Revision 3, issued in November 2012. These documents provide guidance for the review of
human-system interface issues in connection with the design certification of nuclear installations
and the NRC’s inspection program. The NRC also uses NUREG-1764, “Guidance for the
Review of Changes to Human Actions,” Revision 1, issued in September 2007, to review license
amendment requests that credit the use of manual actions. Moreover, Information Notice 97-78,
“Crediting of Operator Actions in Place of Automatic Actions and Modifications of Operator
Actions, Including Response Times,” dated October 23, 1997, identifies references that the
NRC uses to review the completion times of operator manual actions and how the actions will
be reflected in the licensee’s emergency procedures and operator training. In October 2007, the
staff published NUREG-1852, “Demonstrating the Feasibility and Reliability of Operator Manual
Actions in Response to Fire,” for use in evaluating exemptions from fire protection requirements
that assume credit for timely manual actions.
To make some of the current human factors guidance simpler, clearer, and more relevant to the
digital environment, the staff issued an interim staff guidance (ISG) DI&C-ISG-05,
“Highly-Integrated Control Rooms—Human Factors Issues (HICR-HF),” Revision 1, dated
November 3, 2008. This ISG addresses computer-based procedures, minimum inventory of
controls and displays to support plant shutdown, and crediting manual operator actions in
diversity and defense-in-depth analyses. The crediting of manual operator actions in diversity
and defense-in-depth analyses interim guidance has been incorporated into permanent
regulatory guidance through Appendix A, “Crediting Manual Operator Actions in Diversity and
Defense-in-Depth (D3) Analyses,” of Chapter 18 to NUREG-0800, “Standard Review Plan for
the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition.” The NRC
plans to issue a regulatory guide that will endorse, in part, the IEEE Standard 1786-2011, “IEEE
Guide for Human Factors Applications of Computerized Operating Procedure Systems (COPS)
at Nuclear Power Generating Stations and Other Nuclear Facilities,” dated September 22, 2011.
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Experience. The NRC reviews licensees’ requests that involve aspects of human factors
engineering. Examples include crediting operator manual actions in amendments to plant
technical specifications, transferring facility operating licenses, and increasing the reactor’s
authorized power level (i.e., power uprates).
The NRC reviews and approves requests for power uprates from currently licensed plants. For
such requests, the NRC examines the effect of the power uprate on plant procedures, controls,
displays, and alarms, and required operator actions using Section 2.11.1 of Review Standard
(RS-001), “Review Standard for Extended Power Uprates,” issued in December 2003. The
agency recently reviewed and approved extended power uprates for Monticello Nuclear
Generating Plant, Unit 1, in December 2013, and Peach Bottom Atomic Power Station, Units 2
and 3, in August 2014.
The NRC has also evaluated requests to transfer facility operating licenses, which affected
management and organization, staffing, and technical qualifications. The NRC used
NUREG-0800, Chapter 13, “Conduct of Operations,” as the principal guidance for these
reviews.
12.3.2 Emergency Operating Procedures and Plant Procedures
Licensees must have programs to develop, implement, and maintain emergency operating and
plant procedures. Article 16 discusses emergency preparedness; the discussion here is limited
to the human factors aspect of emergency operating procedures.
Governing Documents and Process. On December 17, 1982, the NRC issued GL 1982-33,
“Requirements for Emergency Response Capability,” which transmitted NUREG-0737,
Supplement 1, “Requirements for Emergency Response Capability,” which requires each
licensee to submit a set of documents for developing emergency operating procedures. In
addition, Criterion V, “Instructions, Procedures, and Drawings,” of Appendix B, “Quality
Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” to 10 CFR Part 50,
“Domestic Licensing of Production and Utilization Facilities,” requires licensees to have
operating procedures.
Experience. In 2010, a fire and subsequent complicated reactor trip at H.B. Robinson had
complications in part because emergency operating procedures were inadequate. This resulted
in the loss of reactor coolant pump seal cooling, which operators did not recognize. The NRC
staff describes the violation and the overall event in “H.B. Robinson Steam Electric Plant - NRC
Integrated Inspection Report 05000261/2010004 and 05000261/2010501: Assessment Follow-
up Letter,” dated November 12, 2010. A followup inspection found that the plant’s emergency
operating procedures were structured in a nonstandard manner, and that, as part of the
corrective actions, they would update the procedures to standard 2-column Westinghouse
format. The findings are documented in Inspection Report 05000261/2011010, dated July 6,
2011.
On September 9, 2011, the NRC issued SECY-11-0124, “Recommended Actions To Be Taken
without Delay from the Near Term Task Force Report,” regarding lessons learned from
Fukushima. Recommendation 8 was for the “strengthening and integration of emergency
operating procedures, severe accident management guidelines [SAMGs], and extensive
damage mitigation guidelines.” Emergency operating procedures, SAMGs, and extensive
damage mitigation guidelines had been developed in the United States at different times and for
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different purposes, without explicit requirements for their integration. In addition, SAMGs were
implemented at U.S. nuclear power plants as part of an industry voluntary initiative and,
subsequent to the Fukushima accident, the NRC found that their maintenance was inconsistent
from site-to-site.
On November 13, 2015, the NRC published for public comment proposed requirements related
to the mitigation of beyond-design-basis events. New strategies for the mitigation of
beyond-design-basis external events (also known as diverse and flexible coping strategies or
FLEX guidelines) are integrated with emergency operating procedures, such that they support
an integrated accident response capability. The integrated response capability would include the
capability to respond to beyond-design-basis accidents, such as the earthquake and tsunami
that affected Fukushima. In addition, the proposed Mitigation of Beyond-Design-Basis Events
rule would require that each licensee’s integrated accident management response capability
include sufficient staffing to support implementation of the strategies and guidelines in
conjunction with the emergency operating procedures and a supporting organizational structure
for directing and performing the strategies and guidelines. Additional information on the NRC
Fukushima lessons learned activities can be found in Sections 1.3.1 and 1.3.3 of this report.
12.3.3 Shift Staffing
Governing Documents and Process. In 10 CFR 50.54(m), the NRC establishes minimum onsite
staffing requirements for licensed operators and senior operators at nuclear power reactor
facilities. Appendix R, “Fire Protection Program for Nuclear Power Facilities Operating prior to
January 1, 1979,” and Appendix E, “Emergency Planning and Preparedness for Production and
Utilization Facilities,” to 10 CFR Part 50 contain the NRC staffing requirements for fire brigades
and emergency response personnel.
In September 2002, the NRC began work on a process to evaluate exemption requests from the
requirements in 10 CFR 50.54(m) resulting from the changing demands and new technologies
presented by advanced reactor control room designs and significant light-water reactor control
room upgrades. In July 2005, the NRC issued NUREG-1791, “Guidance for Assessing
Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing Requirements
Specified in 10 CFR 50.54(m).” The purpose of reviewing the exemption requests is to ensure
public health and safety by verifying that the applicant’s staffing plan and supporting analyses
sufficiently justify the requested exemption. NUREG/CR-6838, “Technical Basis for Assessing
Exemptions from Nuclear Power Plant Licensed Operator Staffing Requirements in
10 CFR 50.54(m),” issued in February 2004, explains the justification for the recommended
process.
SECY-10-0034, “Potential Policy, Licensing, and Key Technical Issues for Small Modular
Nuclear Reactor Designs,” dated March 28, 2010, discusses appropriate requirements for
operator staffing for small or multimodule (advanced reactor) facilities. The NRC’s regulations
do not currently address the possibility of more than two reactors being controlled from one
control room; some applications are expected to include control of more than two modules in a
single control room. In addition, small modular reactor designers have stated that they are
considering whether their designs can operate with a staffing complement that is less than what
the Commission regulations currently require. Other small modular reactor policy issues include
the possible need for requirements on control room staffing during refueling operations, reactor
staff that interact with an interconnected manufacturing plant, supervisory staff, shift work, and
training. The NRC staff has stated in previous reports that it believes that operator crew staffing
may be design dependent and intended to review the justification for a smaller crew size for the
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advanced reactors by evaluating the function and task analyses for normal operation and
accident management. The staff made revisions to its RG, as needed, to address control room
staffing for small modular reactors in preparation for receipt of the NuScale small modular
reactor design certification application.
Experience. As noted previously, NUREG-1791 provides guidance for the review of staffing
exemption requests for new generations of advanced reactors, as well as the increased use of
advanced, automated, and digital systems in existing plants. A key element is the review of the
staffing plan validation, an evaluation using performance-based tests to determine whether the
staffing plan meets performance requirements and acceptably supports safe operation. Recent
license application review experience indicates that applicants may be challenged to establish
simulation capabilities to support such validation activities while they are finalizing other aspects
of the plant design.
12.3.4 Fitness for Duty
This section discusses the NRC’s requirements pertaining to the fitness for duty of nuclear
power plant workers, including requirements regarding drug and alcohol testing, behavioral
observation, and management of worker fatigue.
Governing Documents and Process. As required by 10 CFR Part 26, “Fitness for Duty
Programs,” each licensee authorized to operate or construct a nuclear power reactor must
implement a fitness for duty program for all personnel who have unescorted access to the
protected area of its plant or who perform the duties specified in 10 CFR 26.4, “FFD Program
Applicability to Categories of Individuals.” This rule also applies to licensees and permit
holders authorized to construct a nuclear power plant to cover personnel performing certain
construction, management, security, and quality control activities. All fitness for duty programs
must meet the following performance objectives:
(1) provide reasonable assurance that
nuclear power plant personnel perform their tasks in a reliable and trustworthy manner as
demonstrated by the avoidance of substance abuse; (2) provide reasonable assurance that
persons are not under the influence of any substance, legal or illegal, or mentally or physically
impaired from any cause; (3) provide reasonable measures for the early detection of persons
who are not fit to perform activities; (4) provide reasonable assurance that workplaces subject to
10 CFR Part 26 are free from the presence and effects of illegal drugs and alcohol; and
(5) provide reasonable assurance that nuclear power plant management is managing the effects
of fatigue on an individual’s ability to safely and competently perform his or her duties.
On March 31, 2008, the NRC amended 10 CFR Part 26, Subpart I, “Managing Fatigue,” to
include requirements for the management of worker fatigue. Subpart I supersedes the
Commission’s “Policy on Factors Causing Fatigue of Operating Personnel at Nuclear Reactors,”
documented in GL 82-12, “Nuclear Power Plant Staff Working Hours,” dated June 15, 1982. It
also strengthens the effectiveness of fitness for duty programs for protecting public health and
safety by establishing enforceable requirements for the management of worker fatigue. In
addition to the rulemaking and its associated analyses, the NRC issued RG 5.73, “Fatigue
Management for Nuclear Power Plant Personnel,” in March 2009, to provide guidance on how
to implement the rule.
Experience. After implementation of the rule, the NRC received several petitions for rulemaking
from members of the public, requesting the NRC to alleviate alleged impacts adverse to safety
that were introduced when the rule was implemented. The petitioners asserted that
implementation of the rule had impeded some beneficial safety practices. The NRC worked with
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the industry and other external stakeholders to develop an alternative method for managing
cumulative fatigue. The alternative method limits work hours to a weekly average of 54 hours
worked, with work hours being averaged over a rolling period of up to 6 weeks. As a result, the
alternative method limits work hours to levels comparable to the original requirements while
adding the simplicity and flexibility desired by the industry. The rule codifying the alternative
method was published on July 21, 2011, and the rule was effective on August 22, 2011. To
date, several licensees have adopted the alternative method and feedback indicates that it has
allowed the beneficial safety practices to be reinstituted at those facilities that adopted that
alternative.
The NRC has issued reports on statistical data and lessons learned from licensee’s fitness for
duty program performance reports. The latest report is titled “Summary of Fitness-for-Duty
Program Performance Reports for Calendar Year 2013,” and can be located at
reports.html.
12.3.5 Human Factors Information System
Governing Documents and Process. The Human Factors Information System is designed to
store, retrieve, sort, and analyze human performance information extracted from NRC
inspection and licensee event reports. Initiated in 1990, this automated information
management system can generate a variety of specialized reports that are not readily available
from other NRC sources. In 2006, the NRC improved this system to better align the coding
scheme with the Reactor Oversight Process and to enhance the system’s search capabilities.
The Human Factors Information System now captures information related to training,
procedures and reference documents, fitness for duty, oversight, problem identification and
resolution, communications, human-system interface and environment, and work planning and
practices. Currently, the database is being updated to include data with a safety culture
perspective.
Experience. The NRC responds to stakeholder and public inquiries and data requests on this
system on a regular basis. For example, inspectors use the data this system generates in
preparing inspection activities related to human performance. In addition, the NRC’s Office of
Nuclear Regulatory Research uses the data to support activities in human performance and
human reliability analysis. Other NRC program offices use the data to gain insights about
human performance, to monitor the frequency of human performance issues, and to inform
several types of reports, such as internal operating experience reports. The NRC also uses a
Web site to disseminate information on human performance issues at individual nuclear power
plant sites.
12.3.6 Support to Event Investigations and For-Cause Inspections and Training
Governing Documents and Process. NRC staff members with human factors expertise often
participate in special inspections, incident investigation team inspections, augmented team
inspections, event investigations, and supplemental inspections. Human factors experts have
assessed management effectiveness, procedures, training issues, staffing issues,
human-machine interfaces, personnel performance issues, safety-conscious work environment,
and safety culture.
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For training issues, inspectors use IP 41500, “Training and Qualification Effectiveness,” dated
June 13, 1995. For procedure issues, inspectors use IP 42001, “Emergency Operating
Procedures,” dated June 28, 1991, and IP 42700, “Plant Procedures,” dated
November 15, 1995. For baseline inspections under the Reactor Oversight Process, inspectors
use IP 71152, “Problem Identification and Resolution,” which is intended to establish confidence
that each licensee is detecting and correcting problems in a way that limits the risk to the public
and includes a review of the licensee’s safety-conscious work environment. A key premise of
the Reactor Oversight Process is that weaknesses in problem identification and resolution
programs will manifest themselves as performance issues that can be identified during the
baseline inspection program or by crossing predetermined indicator thresholds.
IP 95003, “Supplemental Inspection for Repetitive Degraded Cornerstones, Multiple Degraded
Cornerstones, Multiple Yellow Inputs or One Red Input,” provides the supplemental response
for repetitive degraded cornerstones, multiple degraded cornerstones, multiple yellow inputs, or
one red input to the NRC Assessment Action Matrix. IP 95003 was revised in February 2011, to
include requirements for the NRC staff to review the licensee’s third-party safety culture
assessment and independently assess the licensee’s safety culture. Staff members with
technical expertise in human factors and safety culture perform the safety culture assessment
activities. The NRC first implemented the revised IP 95003 at the Palo Verde Nuclear
Generating Station in October 2007. Based on the lessons learned from the 2007 NRC
inspection and on input from the industry and the public, the staff updated Inspection Manual
Chapter 0305, “Operating Reactor Assessment Program,” in 2009 and again in 2012.
Subsequent inspections conducted in 2013, which are briefly discussed in the following
paragraphs, continued to evaluate performance initiatives and safety performance at the sites.
Experience. In 2007, NRC staff with human factors expertise participated in an IP 95003
inspection at Palo Verde to assess human performance at the site. The inspectors determined
that some findings related to procedure adherence had strong human performance
contributions. The NRC discussed its safety concerns, and how and when these issues were
identified with Palo Verde. Palo Verde made a commitment to take action to improve their
performance.
The NRC increased its plant oversight and conducted numerous inspections. The results of
these inspections demonstrated that performance at Palo Verde had improved substantially.
The NRC determined that the commitments that Palo Verde previously made had been
completed and decided to reduce its oversight at this site.
In 2013, human factors experts participated in IP 95003 inspection activities at Browns Ferry
utilizing the guidance in the 2011 procedure. The overall result and conclusion of the inspection
was that the plant was being operated safely and that the licensee had to aggressively continue
the implementation of the licensee’s integrated improvement plan to achieve substantial
performance improvement.
The NRC has decided to reduce its oversight at Browns Ferry and Palo Verde based on
improvements made. The staff’s findings are documents in inspection reports that be found on
the NRC’s public Web site:
http://www.nrc.gov/NRR/OVERSIGHT/ASSESS/listofrpts_body.html.
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Throughout the years, the NRC has continued to gather lessons learned in this area. Insights
gained from the inspections were used as a basis to substantially update IP 95003 and
Inspection Manual Chapter 0305 in December 2015. For example, IP 95003 provides more
detailed guidance on conducting independent assessment of the licensee’s safety culture.
The revised procedure will be used during an inspection that will be conducted at the Pilgrim
Nuclear Power Station in late 2016 or early 2017.
12.4 Fukushima Lessons Learned
There are human factors considerations to many of the Fukushima lessons learned, including
three orders that were issued, which are described in Sections 1.3.1 and 1.3.3 of this report. For
example, for the mitigating strategies order, licensees are required to validate that their
strategies can be performed as described, including the operator actions. In addition, the
proposed Mitigation of Beyond-Design-Basis Events rulemaking includes requirements for
licensees to ensure that they can transition smoothly between the different types of response
guidelines. Furthermore, human factors are considered as part of the RFI issued by the NRC in
March 2012; specifically, the RFI required licensees to assess its emergency communications
systems and staffing levels to ensure sufficient resources are available to respond to an
accident involving all units at the site.
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ARTICLE 13. QUALITY ASSURANCE
Each Contracting Party shall take the appropriate steps to ensure that quality assurance
programmes are established and implemented with a view to providing confidence that
specified requirements for all activities important to nuclear safety are satisfied
throughout the life of a nuclear installation.
This section describes quality assurance requirements and guidance for design and
construction, operational activities, and staff licensing reviews. It also describes quality
assurance programs, and regulatory guidance.
There have been no changes to the quality assurance regulatory guidance or licensees’ quality
assurance programs as a result of the Fukushima accident. However, continued compliance
with existing programs and requirements is an important aspect of implementation of the
lessons learned from Fukushima, which are further discussed in Sections 1.3.1 and 1.3.3 of this
report.
13.1 Background
Nuclear power facilities must be designed, constructed, and operated in a manner that ensures:
(1) the prevention of accidents that could cause undue risk to public health and safety, and
(2) the mitigation of adverse consequences of such accidents if they should occur. A primary
way to achieve these objectives is to establish and effectively implement a nuclear quality
assurance program. Although a licensee may delegate aspects of the establishment or
execution of the quality assurance program to others, the licensee remains ultimately
responsible for the program’s overall effectiveness. Licensees carry out a variety of
self-assessments to validate the effectiveness of their quality assurance program
implementation. The NRC reviews descriptions of quality assurance programs and performs
onsite inspections to verify aspects of the program implementation.
13.2 Regulatory Policy and Requirements
The NRC sets forth requirements for a license to design, construct, and operate commercial
nuclear power plants in both 10 CFR Part 50, “Domestic Licensing of Production and Utilization
Facilities,” and 10 CFR Part 52, “Licenses, Certifications, and Approvals for Nuclear Power
Plants.” Specifically, 10 CFR Part 50 contains the requirements for a construction permit and a
separate operating license, and 10 CFR Part 52 includes the requirements for a single
combined license, which allows for both construction and operation of a nuclear power plant.
For either type of license, an applicant must describe its quality assurance program for all
activities affecting the safety-related functions of SSCs that prevent or mitigate the
consequences of postulated accidents that could cause undue risk to public health and safety.
High-level criteria for determining which plant SSCs are safety-related appear in 10 CFR 50.2,
“Definitions.” Based on these criteria, licensees’ engineering organizations develop plant-
specific listings of safety-related SSCs.
Under the 10 CFR Part 50 licensing process, each applicant for a construction permit must
describe its quality assurance program in its preliminary safety analysis report in accordance
with 10 CFR 50.34(a)(7). This program should apply to the design, fabrication, construction, and
testing of SSCs. In accordance with 10 CFR 50.34(b)(6)(ii), each applicant for an operating
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license under 10 CFR Part 50 must describe the managerial and administrative controls that will
be implemented during the operation of the nuclear power plant. The applicant must also
describe how it will satisfy the applicable requirements of Appendix B, “Quality Assurance
Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” to 10 CFR Part 50.
Each applicant for a combined license under 10 CFR Part 52 must describe its quality
assurance program in a safety analysis report and give a description of the managerial and
administrative controls that will be implemented during the operation of the nuclear power plant.
Like a 10 CFR Part 50 applicant, an applicant under 10 CFR Part 52 must also describe how it
will satisfy the applicable requirements of Appendix B to 10 CFR Part 50.
13.2.1 Appendix A to 10 CFR Part 50
Appendix A, “General Design Criteria for Nuclear Power Plants,” to 10 CFR Part 50 details the
general requirements for establishing quality assurance controls. General Design Criterion 1,
“Quality Standards and Records,” contains requirements that apply to the quality assurance of
items important to safety. The scope of items that are “important to safety” includes plant
equipment classified as safety-related. Appendix B to 10 CFR Part 50 (discussed in Section
13.2.2 of this report) contains quality assurance program requirements for safety-related SSCs.
Other regulatory guidance discusses quality assurance program controls that are appropriate for
some types of nonsafety-related equipment.
13.2.2 Appendix B to 10 CFR Part 50
Appendix B to 10 CFR Part 50 outlines the quality assurance requirements that apply to
activities affecting the safety-related functions of SSCs that prevent or mitigate the
consequences of postulated accidents. Appendix B defines quality assurance as all planned
and systematic actions that are necessary for adequate confidence that SSCs will perform
satisfactorily in service. Toward that end, it specifies 18 criteria that the commitments in a
licensee’s quality assurance program must satisfy. These criteria cover such topics as
organizational independence, design control, procurement, document control, test control,
corrective action, and audits. Appendix B also stipulates that licensees establish measures to
ensure that the documents for procurement of safety-related materials, equipment, and
services, whether purchased by the licensee or its contractors or subcontractors, include or
reference the applicable regulatory requirements, design bases, and other requirements
necessary to ensure adequate quality. Consistent with the importance and complexity of the
products or services to be provided, licensees (or their designees) are responsible for
periodically verifying that suppliers’ quality assurance programs comply, as appropriate, with the
applicable criteria in Appendix B and that they are effectively implemented. Additionally, as
outlined in 10 CFR 21.41, “Inspections,” the NRC staff performs inspections at vendors that
supply basic components to the nuclear industry.
Because the requirements of Appendix B are written at a conceptual level, the NRC and the
industry needed to develop consensus standards that include acceptable ways to conform to
these requirements. The NRC then issued companion RGs, which endorsed (with conditions, if
warranted) quality assurance codes and standards.
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13.2.3 Approaches for Adopting More Widely Accepted International Quality Standards
The NRC has reviewed options for adopting more widely accepted international quality
standards, such as International Organization for Standardization Standard 9001, 2000 edition,
by considering how international standards compare with the existing framework in Appendix B
to 10 CFR Part 50. On the basis of this review, the NRC concluded that supplemental quality
requirements would be needed when implementing Standard 9001 within the existing regulatory
framework. The NRC participates in both national and international efforts associated with
quality assurance standard development and it continues to assess how various national and
international quality standards comport with NRC regulations in an ongoing effort to seek
convergence of standards.
13.3 Quality Assurance Regulatory Guidance
The NRC has developed or endorsed quality assurance guidance for use by the NRC staff,
applicants for construction permits or operating licenses, and licensees. This guidance is
applicable to the design, construction, and operational phases of a nuclear power plant.
13.3.1 Guidance for Staff Reviews for Licensing
NUREG-0800, Section 17.5, “Quality Assurance Program Description - Design Certification,
Early Site Permit and New License Applicants,” Revision 1, issued in August 2015, provides
guidance to the NRC staff for the review of applications for construction permits, operating
licenses, and combined licenses. The specific review guidance in NUREG-0800 correlates with
the 18 criteria of Appendix B to 10 CFR Part 50 and integrates a review of licensee
commitments to adopt the NRC’s quality assurance-related RGs and apply the industry’s quality
assurance codes and standards.
13.3.2 Guidance for Design and Construction Activities
Licensees may apply consensus standards developed by the American National Standards
Institute (ANSI) in its N45.2 series or by the ASME in its NQA-1 series to comply with the
requirements of Appendix B to 10 CFR Part 50. The NRC has endorsed ANSI and ASME
standards through its RGs. Through its consensus codes and standards activities, the NRC
continues to participate with ASME NQA-1 committees to revise the latest edition of the NQA-1
standard. As part of this effort, the NRC staff issued RG 1.28, “Quality Assurance Program
Requirements (Design and Construction),” Revision 4, on June 2010, to endorse NQA-1-2008
and the NQA-1a-2009 addenda.
13.3.3 Guidance for Operational Activities
The NRC has conditionally endorsed the consensus standard ANSI N18.7-1976, “Administrative
Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants,” issued in
February 1976, through RG 1.33, “Quality Assurance Program Requirements (Operations),”
Revision 2, issued in February 1978, as complying with the requirements of Appendix B to
10 CFR Part 50. The NRC staff issued RG 1.33, “Managerial, Administrative, and Quality
Assurance Controls for the Operational Phase of Nuclear Power Plants,” Revision 3, in
June 2013, endorsing ANSI/ANS 3.2-2012, “Managerial, Administrative, and Quality Assurance
Controls for Operational Phase of Nuclear Power Plants,” dated March 20, 2012.
ANSI/ANS 3.2-2012 is focused on quality assurance of plant operations because information on
quality assurance of design and construction is contained in another standard.
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13.4 Quality Assurance Programs
The NRC inspects quality assurance programs under the Reactor Oversight Process for
operating reactors and under the Construction Inspection Program (see Article 18 of this report)
for new reactors. The NRC also conducts augmented inspection activities as needed.
The baseline inspection program of the Reactor Oversight Process includes one primary
procedure related to quality assurance issues, IP 71152. Inspectors use this procedure to
assess the effectiveness of licensees’ programs to find and resolve problems through a
performance-based review of specific issues. In particular, inspectors look for cases in which a
licensee may have missed generic implications of specific problems and for the risk significance
of combinations of problems that individually may not have significance. They do not inspect
other aspects of quality assurance program implementation in the baseline inspection program
but may do so through supplemental inspections.
Some equipment in the nuclear facility may be classified as nonsafety-related and yet still be
important to safety. In specific cases, the NRC has specified that quality assurance controls are
warranted for equipment determined to be more important than commercial-grade equipment.
However, the quality assurance controls do not have to meet Appendix B requirements, which
apply only to activities affecting safety-related functions. Typically, applying quality assurance
controls to this important-to-safety, yet nonsafety-related, equipment is called “augmented
quality control.”
The Construction Inspection Program provides oversight for nuclear plants licensed under
10 CFR Part 50 and 10 CFR Part 52, including quality assurance program inspection. The
quality assurance inspection program focuses on an applicant or licensee establishing and
implementing a quality assurance program in accordance with the requirements of Appendix B
to 10 CFR Part 50. The inspectors use IP 35007, “Quality Assurance Program Implementation
during Construction and Pre-Construction Activities,” dated February 26, 2015, to verify the
holder of a combined license has developed quality assurance procedures, instructions, and
other documents that are consistent with the licensee’s NRC-approved quality assurance
program description, and to verify the licensee has effectively implemented its quality assurance
program implementing documents during construction activities.
As provided in the Construction Inspection Program, the nuclear plant will transition from the
Construction Inspection Program to the Reactor Oversight Process for commercial operation
when, in accordance with 10 CFR 52.103(g), the Commission determines that all of the
inspections, tests, and analyses in the combined license have been performed, and the
associated acceptance criteria have been met.
13.5 Quality Assurance Audits Performed by Licensees
Appendix B to 10 CFR Part 50 requires licensees to verify the effectiveness of their quality
assurance program by performing internal audits of their programs. These audits are performed
in accordance with the licensee’s procedures by appropriately trained and qualified personnel
who do not have direct responsibility for performing the activities being audited. The results of
these audits are documented and given to management for review and corrective action.
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13.5.1 Audits of Vendors and Suppliers
Appendix B to 10 CFR Part 50 requires licensees that procure material, equipment, or services
from contractors or subcontractors to perform audits to ensure that suppliers implement an
effective quality assurance program, consistent with the requirements of Appendix B and the
licensee’s technical requirements.
Licensees perform these activities by using their own technical and quality assurance staff.
Industry initiatives to promote effective and efficient standardization of these audit activities
have resulted in licensees sharing their technical resources through joint audits of suppliers.
13.6 Vendor Inspection Program
The NRC interacts with manufacturers and suppliers of safety-related components through the
NRC Vendor Inspection Program that inspects compliance with quality assurance and defect
reporting requirements. Vendor inspections are conducted at vendor facilities 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, and to verify that the quality assurance
program provides controls for reporting of defects and noncompliance. Inspection Manual
Chapter 2507, “Vendor Inspections,” dated October 3, 2013, provides guidance for these
inspections.
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ARTICLE 14. ASSESSMENT AND VERIFICATION OF SAFETY
Each Contracting Party shall take the appropriate steps to ensure that:
(i)
comprehensive and systematic safety assessments are carried out before the
construction and commissioning of a nuclear installation and throughout its life.
Such assessments shall be well documented, subsequently updated in the light of
operating experience and significant new safety information, and reviewed under
the authority of the regulatory body
(ii)
verification by analysis, surveillance, testing, and inspection is carried out to
ensure that the physical state and the operation of nuclear installations continue
to be in assurance with its design, applicable national safety requirements, and
operational limits and conditions
This section explains the governing documents and process for ensuring that systematic safety
assessments are carried out during the life of the nuclear installation, including for power
uprates and the period of extended operation. It focuses on assessments performed to maintain
the licensing basis of a nuclear installation. This section explains verification of the physical
state and operation of the nuclear installation by analysis, surveillance, testing, and inspection.
Finally, this section discusses lessons learned from Fukushima and addresses the Vienna
Declaration on Nuclear Safety, which was issued in February 2015.
Other articles in this report (e.g., Articles 6, 10, 13, 18, and 19) also discuss activities to achieve
safety at nuclear installations.
14.1 Ensuring Safety Assessments throughout Plant Life
Before a nuclear facility is constructed, commissioned, and licensed, an applicant must perform
comprehensive and systematic safety assessments for NRC review and approval. Article 18 of
this report discusses these assessments and reviews.
Once a license is issued for a nuclear plant, the licensee must operate the plant in conformance
with its license and its licensing basis. The licensing basis evolves throughout the term of the
license because of the continuing regulatory activities of the NRC, as well as the activities of the
licensee. The Commission engages in a large number of regulatory activities which, when
considered together, constitute a regulatory process that provides ongoing assurance that the
licensing bases of nuclear power plants provide an acceptable level of safety. Section 14.1.5 of
this report discusses 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.
Section 18.5 of this report demonstrates how the NRC continually evaluates new information,
including lessons learned from operational experience and their potential impact on risk and
overall plant safety.
This section focuses on the assessments required throughout the life of a nuclear installation
(i.e., assessments required to maintain the licensing basis). To show conformance with the
licensing basis, a licensee must maintain records of the original design bases and any changes.
This section explains how such changes are documented, updated, and reviewed. Renewal of a
license depends on a licensee’s continuing to meet its current licensing basis; this section
explains how the license renewal process accounts for this requirement.
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14.1.1 Assessment of Safety
The Reactor Oversight Process is the NRC’s program to inspect, measure, and assess the
safety and security performance of commercial nuclear power plants. The objective of the
Reactor Oversight Process is to monitor reactor performance in three key areas (i.e., reactor
safety, radiation safety, and safeguards), which are subsequently monitored through seven
cornerstones. The Reactor Oversight Process assesses plant performance using both
inspection findings and performance indicators across the seven cornerstones. The NRC
determines its regulatory response to plant performance in accordance with an Action Matrix
that provides for a range of actions commensurate with the safety significance of the inspection
findings and performance indicators. The Action Matrix is intended to provide consistent,
predictable, and understandable agency responses to licensee performance such that the
NRC’s regulatory oversight increases as licensee performance declines.
Section 6.3.2 of this report discusses the Reactor Oversight Process and results of the
regulatory assessment in greater detail.
The Construction Reactor Oversight Process monitors and assesses the construction of
commercial nuclear power plants in a similar manner to that employed by the Reactor Oversight
Process. The NRC monitors plant construction in three key areas (i.e., construction reactor
safety, operational readiness, and safeguards programs) and assesses construction using
inspection findings across six cornerstones. The NRC determines its regulatory response to
licensee construction performance in accordance with the Construction Action Matrix.
14.1.2 Maintaining the Licensing Basis
The NRC carries out regulatory programs to give reasonable assurance that plants continue to
conform to the licensing basis. Article 6 of this report discusses these programs.
This section explains the governing documents and process used to maintain the licensing
basis, as required by 10 CFR, Section 50.90, “Application for Amendment of License, or
Construction Permit, or Early Site Permit,” 10 CFR 50.59, “Changes, Tests and Experiments,”
and 10 CFR 50.71, “Maintenance of Records, Making of Reports.”
14.1.2.1 Governing Documents and Process
A licensee is to operate its facility in accordance with the license and as described in its final
safety analysis report. To change its license or reactor facility, a licensee must follow the review
and approval processes established in the regulations. For license amendments, including
changes to technical specifications, the licensee must ask for NRC approval in accordance with
10 CFR 50.90. However, 10 CFR 50.59 contains requirements for the process by which, under
certain conditions, licensees may make changes to their facilities and procedures as described
in the safety analysis report without prior NRC approval.
10 CFR 50.59. In 10 CFR 50.59, the NRC establishes the conditions under which licensees may
make changes to the facility or procedures and conduct tests or experiments without prior NRC
approval. The NRC must review and approve proposed changes, tests, and experiments that
satisfy the definitions and one or more of the criteria in the rule before implementation. Thus, the
rule provides a threshold for regulatory review, not the final determination of safety, for
proposed activities. After determining that a proposed activity is safe and effective through
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appropriate engineering and technical evaluations, the 10 CFR 50.59 process is applied to
determine if a license amendment will be required before implementation. The process involves
three basic steps:
(1) applicability and screening to determine if a 10 CFR 50.59 evaluation is
required, (2) an evaluation that applies the eight evaluation criteria of 10 CFR 50.59(c)(2) to
determine if a license amendment must be obtained from the NRC, and (3) documentation and
reporting to the NRC of activities implemented under 10 CFR 50.59.
A licensee shall obtain a license amendment in accordance with 10 CFR 50.90 before
implementing a proposed change, test, or experiment if the change, test, or experiment would
do any of the following:
x
Result in more than a minimal increase in the frequency of occurrence of a previously
evaluated accident.
x
Result in more than a minimal increase in the likelihood of occurrence of a malfunction of
an SSC important to safety.
x
Result in more than a minimal increase in the consequences of a previously evaluated
accident.
x
Result in more than a minimal increase in the consequences of a malfunction of an SSC
important to safety.
x
Create a possibility for an accident of a different type than any previously evaluated.
x
Create a possibility for a malfunction of an SSC important to safety with a different result
than any previously evaluated.
x
Result in exceeding or altering a design-basis limit for a fission product barrier.
x
Result in a departure from a method of evaluation used in establishing the design bases
or in the safety analyses.
RG 1.187, “Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments,”
dated November 2000, which endorses industry guidance document NEI 96-07, Revision 1,
“Guidelines for 10 CFR 50.59 Evaluations,” dated February 2000, provides methods that are
acceptable to the NRC staff for complying with the provisions of 10 CFR 50.59.
On March 6, 2015, the NRC staff issued a report, “Review of Lessons Learned from the San
Onofre Steam Generator Tube Degradation Event,” along with an accompanying White Paper,
“10 CFR 50.59; the Process, Application to Substantial Modifications to Licensee Facilities, and
NRC Staff Assessment of Licensee Implementation,” dated February 25, 2015. The San Onofre
Nuclear Generating Station lessons learned report highlights important aspects of the guidance
in NEI 96-07, Revision 1, related to issues with the San Onofre 10 CFR 50.59 screening and
evaluation for the replacement steam generators. This was followed by the issuance of RIS
2016-03, “10 CFR 50.59 Issues Identified In NRC’s San Onofre Steam Generator Tube
Degradation Lessons Learned Report,” issued in April 13, 2016.
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10 CFR 50.90. According to 10 CFR 50.90, whenever a holder of a license, including a
construction permit and operating license under 10 CFR Part 50, “Domestic Licensing of
Production and Utilization Facilities,” or an early site permit, combined license, or manufacturing
license under 10 CFR Part 52, “Licenses, Certifications, and Approvals for Nuclear Power
Plants,” wants to amend the license or permit, it must file an application for an amendment with
the Commission, as specified in 10 CFR 50.4, “Written Communications,” or 10 CFR 52.3,
“Written Communications,” fully describing the changes desired, and following, as far as
applicable, the form prescribed for original applications. The NRC performs and documents a
safety evaluation in these instances before it authorizes the change.
10 CFR 50.71. Section (e) of 10 CFR 50.71 requires licensees to update their final safety
analysis reports periodically to incorporate the information and analyses that they submitted to
the Commission or prepared in accordance with Commission requirements. Revisions to the
updated final safety analysis reports are to include the effects of changes that occur in the
vicinity of the plant, changes made in the facility or procedures described in the report, safety
evaluations for approved license amendments and for changes made under 10 CFR 50.59, and
safety analyses conducted at the request of the Commission to address new safety issues.
14.1.3 Power Uprates
This section explains the NRC power uprate licensing process, including the governing
documents, regulatory process, recent experience, and relevant examples.
14.1.3.1 Governing Documents and Process
Background. The NRC regulates the maximum power level at which a commercial nuclear
power plant may operate. This power level is used, with other data, in many of the licensing
analyses that demonstrate plant safety. This power level is included in the license and technical
specifications for the plant. NRC approval is required to make changes to the license and
technical specifications for a plant. Thus, a licensee must receive NRC approval, through the
license amendment process, before it can operate at a higher power level.
Categories of Power Uprates. The NRC has specified three categories of power uprates:
x
Measurement Uncertainty Recapture Power Uprates - measurement uncertainty
recapture power uprates are power increases of less than 2 percent and are achieved by
implementing enhanced techniques for calculating reactor power. This involves the use
of state-of-the-art devices to more precisely measure feedwater flow that is used to
calculate reactor power. More precise measurements reduce the degree of uncertainty
in the power level, which analysts use to predict the ability of the reactor to be safely
shut down under postulated accident conditions.
x
Stretch Power Uprates - stretch power uprates typically are on the order of up to
7 percent and are within the design capacity of the plant. The actual value for
percentage increase in power a plant can achieve and stay within the stretch power
uprate category is plant-specific and depends on the operating margins included in the
design of a particular plant. Stretch power uprates usually involve changes to
instrumentation setpoints but do not involve major plant modifications.
x
Extended Power Uprates - extended power uprates are greater than stretch power
uprates and have been approved for increases as high as 20 percent. Extended power
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uprates usually require significant modifications to major balance-of-plant equipment
such as the high pressure turbines, condensate pumps and motors, main generators, or
transformers.
Review Process, Regulatory Requirements, and Guidance Documents. Because uprates affect
a reactor’s licensed power level, licensees apply for NRC permission to amend their operating
license to implement a power uprate. The process for requesting and approving a change to a
plant’s power level is governed by 10 CFR 50.90 through 10 CFR 50.92, “Issuance of
Amendment.” The applications and reviews are often complex and involve many areas of
expertise in the NRC’s Office of Nuclear Reactor Regulation and Office of the General Counsel.
Some reviews also may involve the Office of Nuclear Regulatory Research, Office of New
Reactors, and the Advisory Committee on Reactor Safeguards. In evaluating a power uprate
request, the NRC reviews data and accident analyses that a licensee submits to confirm that the
plant can operate safely at the higher power level.
The NRC uses RS-001, “Review Standard for Extended Power Uprates,” issued in
December 2003, for evaluating extended power uprates and stretch power uprates. The
Advisory Committee on Reactor Safeguards has endorsed this standard, which provides a
comprehensive process and technical guidance for reviews by the NRC staff, and useful
information to licensees considering applying for an extended power uprate. RIS 2002-03,
“Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications,”
issued in January 2002, discusses the scope and detail of the information that should be
provided to the NRC for reviewing measurement uncertainty recapture uprate applications.
Additionally, the staff uses NUREG-0800, “Standard Review Plan for the Review of Safety
Analysis Reports for Nuclear Power Plants: LWR Edition,” where appropriate, when
conducting power uprate regulatory reviews.
After a licensee submits an uprate application, the NRC issues a Federal Register notice to alert
the public that the agency is considering the application. The public has 30 days to comment on
the licensee’s request and 60 days to request a hearing where the application could be
contested. The NRC thoroughly reviews the application and any public comments, while the
Atomic Safety and Licensing Board considers any requests for hearings. The NRC documents
its review in a safety evaluation, and, if acceptable, the NRC will issue a license amendment
approving the power uprate. The NRC will issue another Federal Register notice to inform the
public if the amendment is issued. After the approval, the NRC performs inspections of the
power uprate implementation using IP 71004, “Power Uprates,” dated May 21, 2015, to review
plant modifications and operator readiness.
If the Atomic Safety and Licensing Board determines that a hearing is required, a separate legal
process takes place, and NRC staff provides technical information, if needed. The safety
evaluation and any hearing rulings form the basis for the NRC’s final decision on the uprate
request. However, the staff can authorize an uprate before the hearing is completed. The NRC
issues a press release for any approved uprate.
The NRC’s current schedule is to complete power uprate reviews within 18 months of
application review acceptance for extended power uprates, within 12 months of application
review acceptance for stretch power uprates, and within 9 months of application review
acceptance for measurement uncertainty recapture uprates. The application acceptance
process is intended to provide the NRC staff an opportunity to ensure that application quality is
sufficient for regulatory review such that these schedules can be met.
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14.1.3.2 Experience
The NRC issued the first power uprate amendment for the Calvert Cliffs nuclear power plant in
1977. As of May 2016, the NRC had approved 157 uprates, resulting in a gain of approximately
22,034 MWt (megawatts thermal) or 7,346 MWe (megawatts electric), at existing plants. The
NRC is currently reviewing three power uprate applications that would authorize an additional
1,482 MWt. In addition, licensees plan to submit 10 measurement uncertainty recapture power
uprate applications in the next 2 years. If these expected applications are approved, the
resulting uprates would authorize an additional 571 MWt (190 MWe).
Peach Bottom, Units 2 and 3, Extended Power Uprate
On August 25, 2014, the NRC approved a 12.4 percent extended power uprate for Peach
Bottom, Units 2 and 3. Stretch and measurement uncertainty power uprates were previously
approved for Peach Bottom. As such, the extended power uprate represents a power level
equivalent to 120 percent of the original licensed thermal power level. As part of the plant
modifications associated with the extended power uprate, the steam dryer in each unit was
replaced.
The NRC approval of the extended power uprate was based, in part, on the capability for the
licensee to monitor, evaluate, and take prompt action in response to potential adverse flow
effects as a result of extended power uprate operation on plant SSCs, including verifying the
continued structural integrity of the replacement steam dryer. A license condition was added to
the facility operating license for each unit, as part of the extended power uprate amendment, to
provide the necessary requirements associated with potential adverse flow effects.
For Peach Bottom Unit 2, the licensee completed the plant modifications needed to implement
the extended power uprate during the fall 2014 refueling outage, including installation of the
replacement steam dryer. During the power ascension following the refueling outage, data
collected at about the 89 percent power level (equivalent to the preuprate 100 percent power
level) identified strain responses on the replacement steam dryer in the low frequency range
that were not previously predicted by the extended power uprate approved methodology. As a
result, the licensee needed to change the steam dryer stress analysis methodology to better
account for the low frequency loads. After NRC review and approval of the methodology, Peach
Bottom, Unit 2 reached the new 100 percent extended power uprate power level on May 15,
2015.
For Peach Bottom, Unit 3, the licensee completed the plant modifications during the fall 2015
refueling outage to implement the extended power uprate.
Monticello Nuclear Generating Plant Extended Power Uprate
On December 9, 2013, the NRC approved the extended power uprate for the Monticello Nuclear
Generating Plant. The extended power uprate application was submitted in a letter dated
November 5, 2008. The amendment authorized an increase of the maximum core thermal
power level by approximately 13 percent, from the previously licensed thermal power level of
1,775 MWt to 2,004 MWt.
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The extended power uprate represents a power level equivalent to 120 percent of the original
licensed thermal power level. The Atomic Energy Commission (predecessor of the NRC)
originally issued an operating license to Monticello for a thermal power level of 1,670 MWt. In
September 1998, the NRC approved a 6.3 percent power uprate to increase the power output to
1,775 MWt.
As part of the plant modifications associated with the extended power uprate, the steam dryer in
each unit was replaced. The original Monticello steam dryer was a parallel vane bank, square
hood design by General Electric, which does not have perforated plates at the inlet and outlet
sides of the vane banks. In 2011, the licensee replaced its original dryer with a Westinghouse
steam dryer that consists of three parallel vane banks of octagonal shape and a cylindrical skirt.
The NRC approval of the extended power uprate was based, in part, on the capability for the
licensee to monitor, evaluate, and take prompt action in response to potential adverse flow
effects as a result of extended power uprate operation on plant SSCs, including verifying the
continued structural integrity of the replacement steam dryer. A license condition was added to
the facility operating license as part of the extended power uprate amendment to provide the
necessary requirements associated with potential adverse flow effects. During the power
ascension, data was collected at various power plateaus. Monticello reached the new
100 percent extended power uprate power level on July 1, 2015.
14.1.4 License Renewal
This section explains license renewal, including the governing documents, regulatory process,
recent experience, and relevant examples.
14.1.4.1 Governing Documents and Process
Background. The Atomic Energy Act and NRC regulations limit commercial power reactor
licenses to 40 years but permit such licenses to be renewed. The original 40-year term was
selected on the basis of economic and antitrust considerations, not technical limitations. The
decision to seek license renewal rests entirely with the nuclear power plant owners and typically
is based on the plant’s economic situation and whether it can meet NRC requirements.
The NRC has established a license renewal process with requirements to ensure safe plant
operation for up to 20 additional years at a time. The NRC’s current schedule is to complete the
review of a license renewal application within 30 months of receipt of the application if a hearing
is conducted and within 22 months if a hearing is not conducted. As of August 2016, eight
license renewal applications, spanning 12 units, are under NRC review.
NRC’s final rule on “Continued Storage of Spent Nuclear Fuel,” supports the agency’s licensing
decisions, in particular new reactor licensing and reactor license renewal by allowing the NRC to
proceed with environmental reviews of new reactors or license renewal applications without
considering the site-specific effects of spent fuel storage after the end of the reactor’s licensed
life for operation in the environmental analysis. See Section 1.3.3 of this report for further details
on Continued Storage.
Research has concluded that aging phenomena are readily manageable and do not pose
technical issues that would prevent life extension for nuclear power plants. Studies have also
found that facilities deal adequately with many aging effects during the initial license period and
that credit should be given for these existing programs, particularly those under the NRC’s
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Maintenance Rule, 10 CFR 50.65, “Requirements for Monitoring the Effectiveness of
Maintenance at Nuclear Power Plants,” which provides requirements for maintenance and
monitoring of active and passive SSCs.
The license renewal process proceeds along two tracks: one for the review of safety issues
and another for environmental issues. An applicant must give the NRC an evaluation that
addresses the technical aspects of plant aging and describes the ways it will manage those
effects. It must also prepare an evaluation of the potential impact on the environment if the plant
operates for up to 20 more years. The NRC reviews the application and verifies the safety and
environmental issues through onsite audits and inspections. The NRC documents its findings in
a safety evaluation report and an environmental impact statement.
Public participation is an important part of the license renewal process. Members of the public
have opportunities to comment on the environmental review and question how aging will be
managed during the period of extended operation. All information related to the review and
approval of a renewal application is publicly available. Significant safety and environmental
concerns also may be litigated in an adjudicatory hearing if any party that would be adversely
affected asks for a hearing.
10 CFR Part 54. Known as the License Renewal Rule, 10 CFR Part 54, “Requirements for
Renewal of Operating Licenses for Nuclear Power Plants,” establishes the technical and
procedural requirements for renewing operating licenses. License renewal requirements for
power reactors are based on two key principles:
(1)
When continued into the extended period of operation, the regulatory process, which
assesses and verifies safety, is adequate to ensure that the licensing basis of all
currently operating plants provides an acceptable level of safety. The possible exception
is detrimental effects of aging on certain SSCs, and possibly a few other issues applying
to safety only during the period of extended operation.
(2)
Each plant must maintain its licensing basis throughout the renewal term.
Guidance that applies to license renewal includes RG 1.188, “Standard Format and Content for
Applications to Renew Nuclear Power Plant Operating Licenses,” Revision 1, issued in
September 2005, to help applicants apply to renew a license; and NUREG-1800, “Standard
Review Plan for Review of License Renewal of Applications for Nuclear Power Plants,”
Revision 2, issued in December 2010, which guides the staff in reviewing applications. The
standard review plan for license renewal incorporates by reference NUREG-1801, “Generic
Aging Lessons Learned (GALL) Report,” Revision 2, issued in December 2010, which
generically documents the basis for determining when existing programs are adequate for
license renewal and when they should be augmented. As lessons are learned from the review of
renewal applications or generic technical issues are resolved, the NRC issues improved
guidance for interim use by applicants until the guidance is incorporated into the next formal
update of the documents. The staff is developing the Standard Review Plan for license renewal
and the Generic Aging Lessons Learned Report for plants intending to operate beyond 60
years. Additional information on plant operation beyond 60 years can be found in Section
14.1.4.3 of this report.
10 CFR Part 51. The NRC’s environmental protection regulation, 10 CFR Part 51,
“Environmental Protection Regulations for Domestic Licensing and Related Regulatory
Functions,” also applies to license renewal of nuclear power plants. The license renewal
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environmental review requirements under 10 CFR Part 51 are founded on the conclusion that
certain environmental issues can be resolved generically and do not need to be evaluated in
each plant-specific review. These issues are listed in Table B-1 of Appendix B, “Environmental
Effect of Renewing the Operating License of a Nuclear Power Plant,” to Subpart A, “National
Environmental Policy Act—Regulations Implementing Section 102(2),” of 10 CFR Part 51.
In June 2013, the agency amended Part 51 and its technical basis documented in
NUREG-1437, “Generic Environmental Impact Statement for License Renewal of Nuclear
Plants,” to incorporate lessons learned and knowledge gained from previous license renewal
environmental reviews conducted since 1996. During the development of the revised rule, the
NRC added new environmental impact issues and consolidated similar ones. No environmental
issues were deleted. The NRC’s revised Table B-1 and updated NUREG-1437 identify
78 environmental issues; of these, 59 issues are considered generic or applicable to all nuclear
power plants, 17 issues require a plant-specific analysis, and 2 issues require further
information and remain uncategorized. The NRC conducts independent reviews of these
environmental impacts to determine whether the effects are significant enough to preclude
license renewal as an option for energy-planning decisionmakers. In June 2013, the NRC also
updated its associated guidance documentation for license renewal applicants and its technical
guidance for use by NRC staff. RG 4.2, “Preparation of Environmental Reports for Nuclear
Power Plants License Renewal Applications,” Supplement 1, Revision 1, provides guidance to
applicants preparing environmental reports to be included as part of license renewal
applications. NUREG-1555, Supplement 1, “Standard Review Plans for Environmental Reviews
of Nuclear Power Plants: Environmental Standard Review Plan for Operating License
Renewal,” Revision 1, guides the NRC staff’s review of the environmental issues associated
with license renewal.
14.1.4.2 Experience
The NRC issued the first renewed licenses for the Calvert Cliffs Nuclear Power Plant and the
Oconee Nuclear Station in 2000. As of August 2016, 83 reactors, including Kewaunee, Vermont
Yankee, Pilgrim, Unit 1, and FitzPatrick,14 have received renewed licenses. Thirty-nine of the 81
reactors have completed 40 years of operation and are operating in the extended period. One
reactor entered the period of extended operation in 2015, and six reactor units are expected to
enter the period of extended operation in 2016. On the basis of industry statements, the NRC
expects that almost all remaining plants that have yet to tender license renewal application will
apply for license renewal. Please refer to
that are expected to apply for license renewal.
14.1.4.3 Operating beyond 60 Years
The provisions of 10 CFR Part 54 allow a previously renewed operating license to be
subsequently renewed with no additional requirements imposed and no limit on the number of
times a license can be subsequently renewed, provided that it is justified and that safety is
ensured. The earliest that a licensee can submit a license renewal application is 20 years before
the expiration of its current license; therefore, a licensee is eligible to apply for a subsequent
14 Subsequent to receiving their renewed licenses, Kewaunee, Vermont Yankee, Pilgrim Unit 1, and Fitzpatrick
announced that they would cease operations.
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license renewal once it enters the initial period of extended operation (i.e., the 20-year renewal
period beyond its initial 40-year license period). Several industry representatives have
expressed an interest in operating nuclear power plants beyond 60 years and on November 5,
2015, the Commission received a formal letter of intent to pursue such a renewal for Surry
Power Station, Units 1 and 2 in 2019. The renewed operating licenses for Surry Power Station,
Units 1 and 2 were issued on March 20, 2003, and will expire on May 25, 2032, and January 29,
2033, respectively. On June 7, 2016, the Commission received a formal letter of intent to pursue
a second license renewal for Peach Bottom Atomic Power Station, Units 2 and 3, in 2018. The
renewed operating licenses for Peach Bottom Atomic Power Station, Units 2 and 3, were issued
on May 7, 2013, and will expire on August 8, 2033, and July 2, 2034, respectively.
To prepare for the review of subsequent license renewal applications, on January 31, 2014, the
NRC staff submitted to the Commission SECY-14-0016, “Ongoing Staff Activities to Assess
Regulatory Considerations for Power Reactor Subsequent License Renewal.” In
SRM-SECY-14-0016, dated August 29, 2014, the Commission affirmed that the current
regulatory framework for the first license renewal (i.e., operation from 40 years to 60 years) is
sufficient to support the review of subsequent license renewal. In addition, in
SRM-SECY-14-0016 the Commission directed the staff to:
x
Continue to update the license renewal guidance, as needed, to provide additional clarity
on the implementation of the license renewal regulatory framework.
x
Address Option 2 and Option 3 as presented in SECY-14-0016 through alternative
vehicles (e.g., issuance of generic communications, voluntary industry initiatives, or
updates to the Generic Aging Lessons Learned Report). Option 2 recommended minor
editorial changes to 10 CFR Part 54 to add alternate fracture toughness requirements
and clarify how existing recordkeeping requirements apply to newly identified systems,
structures, and components. Option 3, which includes Option 2, recommended an
expansion in scope of 10 CFR Part 54 to include equipment associated with
10 CFR 50.54(hh)(2) and adds a provision to address timely renewal so that a licensee
must implement aging-management activities before the expiration of its current license.
x
Submit an information paper to the Commission reporting on the progress of the
implementation of the inspection enhancements described in the Reactor Oversight
Process Enhancement Project related to aging management and the Inspection
Procedure Operating Experience Update Process.
x
Keep the Commission informed on the progress of technical issues resolution, research
activities, the staff’s readiness to receive and evaluate the acceptability of an application
for subsequent license renewal, and any further need for regulatory process changes, or
rulemaking, related to subsequent license renewal.
x
Emphasize in communications with industry the need to strive for satisfactory resolution
of these issues before the NRC begins a review of any subsequent license renewal
application.
To address the unique aspects of material aging and degradation that apply to subsequent
license renewal, the NRC is collaborating on research activities with both domestic and
international partners to ensure that important research topics are being addressed and to
effectively leverage both resources and knowledge.
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The NRC, in cooperation with the U.S. DOE Light Water Reactor Sustainability Program,
completed NUREG/CR-7153, “Expanded Materials Degradation Assessment (EMDA),”
Volumes 1 through 5, dated October 2014, to identify the most significant technical issues for
nuclear power reactor operations beyond 60 years. The expanded materials degradation
assessment ranked the significance, current knowledge, and uncertainty associated with aging-
related degradation issues that could affect systems, structures, and components over 80 years
of operation. As outlined SRM-SECY-14-0016, the major technical issue areas are:
x
reactor pressure vessel neutron embrittlement at high fluence
x
irradiation assisted stress corrosion cracking of reactor internals and primary system
components
x
concrete and containment degradation
x
electrical cable qualification and condition assessment
Licensees and applicants have the primary responsibility for providing the technical basis to
support their safety analysis and application for a subsequent license renewal. NRC staff
conducts confirmatory research to independently verify licensee data, determine safety margins,
and explore uncertainties. The NRC continues to track industry’s work in this area, evaluate
areas for research, gather data to help assess the effectiveness of licensee’s aging
management programs, and provide confirmatory research on the results of industry’s work.
Results from NRC’s research will be used, in part, to confirm the adequacy of industry’s
technical basis for a subsequent license renewal and the associated aging management
programs. The aging management programs are cornerstones for managing materials
degradation in safety-significant components during a subsequent license renewal. In addition,
the NRC research will support and increase the efficiency of staff review of subsequent license
renewal applications.
To support the review of the first subsequent license renewal application in 2019, the NRC staff
is developing guidance documents to address the unique aging management needs for
subsequent license renewal using as a starting point the existing license renewal guidance
documents, NUREG-1801, Revision 2, and NUREG-1800, Revision 2. The NRC staff plans to
issue the Generic Aging Lessons Learned Report and the Standard Review Plan for operations
beyond 60 years by mid-2017.
14.1.5 The United States and Periodic Safety Reviews
To a large extent, the international community conducts periodic safety reviews (typically carried
out every 10 years) to assess the cumulative effects of plant aging, plant modifications,
operating experience, technical developments, and plant siting aspects. The reviews include an
assessment of plant design and operation against current safety standards and practices, with
the objective of ensuring a high level of safety throughout the plant’s operating lifetime.
Some countries use routine comprehensive safety assessment programs that deal with specific
safety issues, significant events, and changes in safety standards and practices as they arise.
These programs, if applied with appropriate scope, frequency, depth, and rigor, achieve the
same review standards and objectives as a periodic safety review. Some countries also use
periodic safety reviews to support the decisionmaking process for long-term operation or license
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