[Federal Register Volume 80, Number 223 (Thursday, November 19, 2015)]
[Proposed Rules]
[Pages 72358-72373]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-29536]
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Proposed Rules
Federal Register
________________________________________________________________________
This section of the FEDERAL REGISTER contains notices to the public of
the proposed issuance of rules and regulations. The purpose of these
notices is to give interested persons an opportunity to participate in
the rule making prior to the adoption of the final rules.
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Federal Register / Vol. 80, No. 223 / Thursday, November 19, 2015 /
Proposed Rules
[[Page 72358]]
NUCLEAR REGULATORY COMMISSION
10 CFR Parts 26, 50, 52, 73, and 140
[NRC-2015-0070]
RIN 3150-AJ59
Regulatory Improvements for Decommissioning Power Reactors
AGENCY: Nuclear Regulatory Commission.
ACTION: Advance notice of proposed rulemaking; request for comment.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is issuing this
advance notice of proposed rulemaking (ANPR) to obtain input from
stakeholders on the development of a draft regulatory basis. The draft
regulatory basis would support potential changes to the NRC's
regulations for the decommissioning of nuclear power reactors. The
NRC's goals in amending these regulations would be to provide an
efficient decommissioning process, reduce the need for exemptions from
existing regulations, and support the principles of good regulation,
including openness, clarity, and reliability. The NRC is soliciting
public comments on the contemplated action and invites stakeholders and
interested persons to participate. The NRC plans to hold a public
meeting to promote full understanding of the questions contained in
this ANPR and facilitate public comment.
DATES: Submit comments by January 4, 2016. Comments received after this
date will be considered if it is practical to do so, but the NRC is
able to ensure consideration only for comments received on or before
this date.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0070. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Email comments to: [email protected]. If you do
not receive an automatic email reply confirming receipt, then contact
us at 301-415-1677.
Fax comments to: Secretary, U.S. Nuclear Regulatory
Commission at 301-415-1101.
Mail comments to: Secretary, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, ATTN: Rulemakings and
Adjudications Staff.
Hand deliver comments to: 11555 Rockville Pike, Rockville,
Maryland 20852, between 7:30 a.m. and 4:15 p.m. (Eastern time) Federal
workdays; telephone: 301-415-1677.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Jason B. Carneal, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-1451; email: [email protected].
SUPPLEMENTARY INFORMATION:
Table of Contents
I. Obtaining Information and Submitting Comments
II. Background
A. Regulatory Actions Related to Decommissioning Power Reactors
B. Licensing Actions Related to Decommissioning Power Reactors
III. Discussion
IV. Regulatory Objectives
A. Applicability to NRC Licenses and Approvals
B. Interim Regulatory Actions
V. Specific Considerations
VI. Public Meeting
VII. Cumulative Effects of Regulation
VIII. Plain Writing
IX. Availability of Documents
X. Rulemaking Process
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2015-0070 when contacting the NRC
about the availability of information for this action. You may obtain
publicly-available information related to this action by any of the
following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0070.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in the
SUPPLEMENTARY INFORMATION section. For the convenience of the reader,
instructions about obtaining materials referenced in this document are
provided in Section IX, ``Availability of Documents,'' of this
document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2015-0070 in your comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or
[[Page 72359]]
entering the comment submissions into ADAMS.
II. Background
A. Regulatory Actions Related to Decommissioning Power Reactors
Significant regulations for the decommissioning of nuclear power
reactors were not included in NRC rules promulgated before 1988. The
NRC published a final rule in the Federal Register on June 27, 1988 (53
FR 24018), establishing decommissioning requirements for various types
of licensees. By the early 1990s, the NRC recognized a need for more
changes to the power reactor decommissioning regulations and published
a proposed rule to amend its regulations for reactor decommissioning in
1995 (60 FR 37374; July 20, 1995). In 1996, the NRC amended its
regulations for reactor decommissioning to clarify ambiguities, make
generically applicable procedures that had been used on a case-by-case
basis, and allow for greater public participation in the
decommissioning process (61 FR 39278; July 29, 1996). However, as an
increasing number of power reactor licensees began decommissioning
their reactors, it became apparent in the late 1990s that additional
rulemaking was needed on specific topics to improve the efficiency and
effectiveness of the decommissioning process.
In a series of Commission papers issued between 1997 and 2001, the
NRC staff provided options and recommendations to the Commission to
address regulatory improvements related to power reactor
decommissioning. In the Staff Requirements Memorandum (SRM) to SECY-99-
168, ``Improving Decommissioning Regulations for Nuclear Power
Plants,'' dated December 21, 1999 (ADAMS Accession No. ML003752190),
the Commission directed the NRC staff to proceed with a single,
integrated, risk-informed decommissioning rule, addressing the areas of
emergency preparedness (EP), insurance, safeguards, staffing and
training, and backfit. The objective of the rulemaking was to clarify
and remove certain regulations for decommissioning power reactors based
on the reduction in radiological risk compared to operating reactors.
At an operating reactor, the high temperature and pressure of the
reactor coolant system, as well as the inventory of relatively short-
lived radionuclides, contribute to both the risk and consequences of an
accident. With the permanent cessation of reactor operations and the
permanent removal of the fuel from the reactor core, such accidents are
no longer possible. As a result of the shutdown and removal of fuel,
the reactor, reactor coolant system, and supporting systems no longer
operate and, therefore, have no function. Hence, postulated accidents
involving failure or malfunction of the reactor, reactor coolant
system, or supporting systems are no longer applicable.
During reactor decommissioning, the principal radiological risks
are associated with the storage of spent fuel onsite. Generally, a few
months after the reactor has been permanently shut down, there are no
possible design-basis events that could result in a radiological
release exceeding the limits established by the U.S. Environmental
Protection Agency's (EPA) early- phase Protective Action Guidelines of
1 roentgen equivalent man at the exclusion area boundary. The only
accident that might lead to a significant radiological release at a
decommissioning reactor is a zirconium fire. The zirconium fire
scenario is a postulated, but highly unlikely, beyond-design-basis
accident scenario that involves a major loss of water inventory from
the spent fuel pool (SFP), resulting in a significant heat-up of the
spent fuel, and culminating in substantial zirconium cladding oxidation
and fuel damage. The analyses of spent fuel heat-up scenarios that
might result in a zirconium fire are related to the decay heat of the
irradiated fuel stored in the SFP. Therefore, the probability of a
zirconium fire scenario continues to decrease as a function of the time
that the decommissioning reactor has been permanently shut down.
On June 28, 2000, the NRC staff submitted SECY-00-0145,
``Integrated Rulemaking Plan for Nuclear Power Plant Decommissioning''
(ADAMS Accession No. ML003721626) to the Commission, proposing an
integrated decommissioning rulemaking plan. The rulemaking plan was
contingent on the completion of a zirconium fire risk study provided in
NUREG-1738, ``Technical Study of Spent Fuel Pool Accident Risk at
Decommissioning Nuclear Power Plants'' (ADAMS Accession No.
ML010430066), on the accident risks at decommissioning reactor SFPs.
The NUREG was issued on February 28, 2001.
Although NUREG-1738 could not completely rule out the possibility
of a zirconium fire after a long spent fuel decay times, it did
demonstrate that storage of spent fuel in a high-density configuration
in SFPs is safe, and that the risk of accidental release of a
significant amount of radioactive material to the environment is low.
The study used simplified and sometimes bounding assumptions and models
to characterize the likelihood and consequences of beyond-design-basis
SFP accidents. Subsequent NRC regulatory activities and studies
(described in more detail below) have reaffirmed the safety and
security of spent fuel stored in pools and shown that SFPs are
effectively designed to prevent accidents.
Because of uncertainty in the NUREG-1738 conclusions about the risk
of SFP fires, the NRC staff faced a challenge in developing a generic
decommissioning rule for EP, physical security, and insurance. To seek
additional Commission direction, on June 4, 2001, the NRC staff
submitted to the Commission SECY-01-0100, ``Policy Issues Related to
Safeguards, Insurance, and Emergency Preparedness Regulations at
Decommissioning Nuclear Power Plants Storing Fuel in Spent Fuel Pools''
(ADAMS Accession No. ML011450420). However, based on the reactor
security implications of the terrorist attacks of September 11, 2001
(9/11), and the results of NUREG-1738, the NRC redirected its
rulemaking priorities to focus on programmatic regulatory changes
related to safeguards and security. In a memorandum to the Commission,
``Status of Regulatory Exemptions for Decommissioning Plants,'' dated
August 16, 2002 (ADAMS Accession No. ML030550706), the NRC staff stated
that no additional permanent reactor shut downs were anticipated in the
foreseeable future, and that no immediate need existed to proceed with
the decommissioning regulatory improvement work that was planned.
Consequently, the NRC shifted resources allocated for reactor
decommissioning rulemaking to other activities. The NRC staff concluded
that if any additional reactors permanently shut down after the
rulemaking effort was suspended, establishment of the decommissioning
regulatory framework would continue to be addressed through the license
amendment and exemption processes.
Between 1998 and 2013, no power reactors permanently ceased
operation. Since 2013, five power reactors have permanently shut down,
defueled, and are transitioning to decommissioning. For these
decommissioning reactor licensees, the NRC has processed various
license amendments and exemptions to establish a decommissioning
regulatory framework, similar to the method used in the 1990s.
Following the 9/11 attack, the NRC took several actions to further
reduce the possibility of a SFP fire. In the wake of the attacks, the
NRC issued orders
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that required licensees to implement additional security measures,
including increased patrols, augmented security forces and
capabilities, and more restrictive site-access controls to reduce the
likelihood of an accident, including a SFP accident, resulting from a
terrorist initiated event. The NRC's regulatory actions after the
terrorist attacks of 9/11 have significantly enhanced the safety of
SFPs. A comprehensive discussion of post 9/11 activities, some of which
specifically address SFP safety and security, is provided in the
memorandum to the Commission titled, ``Documentation of Evolution of
Security Requirements at Commercial Nuclear Power Plants with Respect
to Mitigation Measures for Large Fires and Explosions,'' dated February
4, 2010 (ADAMS Accession No. ML092990438).
In addition, the NRC amended Sec. 50.55(hh)(2) of title 10 of the
Code of Federal Regulations (10 CFR) to require licensees to implement
other mitigating measures to maintain or restore SFP cooling capability
in the event of loss of large areas of the plant due to fires or
explosions, which further decreases the probability of a SFP fire (74
FR 13926, March 27, 2009). The Nuclear Energy Institute (NEI) provided
detailed guidance in ``NEI-06-12: B.5.b Phase 2 & 3 Submittal
Guideline,'' Revision 2, dated December 2006 (ADAMS Accession No.
ML070090060). The NRC endorsed this guidance on December 22, 2006 (non-
publicly available), for compliance with the Sec. 50.54(hh)(2)
requirements. Under Sec. 50.54(hh)(2), power reactor licensees are
required to implement strategies such as those provided in NEI-06-12.
The NEI's guidance specifies that portable, power-independent pumping
capabilities must be able to provide at least 500 gallons per minute
(gpm) of bulk water makeup to the SFP, and at least 200 gpm of water
spray to the SFP. Recognizing that the SFP is more susceptible to a
release when the spent fuel is in a nondispersed configuration, the
guidance also specifies that the portable equipment is to be capable of
being deployed within 2 hours for a nondispersed configuration. The NRC
found the NEI guidance to be an effective means for mitigating the
potential loss of large areas due to fires or explosions.
Further, other organizations, such as Sandia National Laboratory,
have confirmed the effectiveness of the additional mitigation
strategies to maintain spent fuel cooling in the event the pool is
drained and its initial water inventory is reduced or lost entirely.
The analyses conducted by the Sandia National Laboratories
(collectively, the ``Sandia studies''), are sensitive security related
information and are not available to the public. The Sandia studies
considered spent fuel loading patterns and other aspects of a
pressurized-water reactor SFP and a boiling water reactor SFP,
including the role that the circulation of air plays in the cooling of
spent fuel. The Sandia studies indicated that there may be a
significant amount of time between the initiating event (i.e., the
event that causes the SFP water level to drop) and the spent fuel
assemblies becoming partially or completely uncovered. In addition, the
Sandia studies indicated that for those hypothetical conditions where
air cooling may not be effective in preventing a zirconium fire, there
is a significant amount of time between the spent fuel becoming
uncovered and the possible onset of such a zirconium fire, thereby
providing a substantial opportunity for both operator and system event
mitigation.
The Sandia studies, which account for relevant heat transfer and
fluid flow mechanisms, also indicated that air-cooling of spent fuel
would be sufficient to prevent SFP zirconium fires at a point much
earlier following fuel offload from the reactor than previously
considered (e.g., in NUREG-1738). Thus, the fuel is more easily cooled,
and the likelihood of an SFP fire is therefore reduced.
Additional mitigation strategies implemented subsequent to 9/11
enhance spent fuel coolability, and the potential to recover SFP water
level and cooling prior to a potential SFP zirconium fire. The Sandia
studies also confirmed the effectiveness of additional mitigation
strategies to maintain spent fuel cooling in the event the pool is
drained and its initial water inventory is reduced or lost entirely.
Based on this more recent information, and the implementation of
additional strategies following 9/11, the probability of a SFP
zirconium fire initiation is expected to be less than reported in
NUREG-1738 and previous studies.
The NUREG-2161, ``Consequence Study of a Beyond-Design-Basis
Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling
Water Reactor,'' dated September 2014 (ADAMS Accession No.
ML14255A365), evaluated the potential benefits of strategies required
in Sec. 50.54(hh)(2). The NUREG-2161 found that successful
implementation of mitigation strategies significantly reduces the
likelihood of a release from the SFP in the event of a loss of cooling
water. Additionally, NUREG-2161 found that the placement of spent fuel
in a dispersed configuration in the SFP, such as the 1 x 4 pattern,
would have a positive effect in promoting natural circulation, which
enhances air coolability and thereby reduces the likelihood of a
release from a completely drained SFP. An information notice titled,
``Potential Safety Enhancements to Spent Fuel Pool Storage,'' dated
November 14, 2014 (ADAMS Accession No. ML14218A493), was issued to all
licensees informing them of the insights from NUREG-2161. This
information notice describes the benefits of storing spent fuel in more
favorable loading patterns, placing spent fuel in dispersed patterns
immediately after core offload, and taking action to improve mitigation
strategies.
In addition, in response to the Fukushima Dai-ichi accident, the
NRC is currently implementing regulatory actions to further enhance
reactor and SFP safety. On March 12, 2012, the NRC issued Order EA-12-
051, ``Issuance of Order to Modify Licenses with Regard to Reliable
Spent Fuel Pool Instrumentation,'' (ADAMS Accession No. ML12054A679),
which requires that licensees install reliable means of remotely
monitoring wide-range SFP levels to support effective prioritization of
event mitigation and recovery actions in the event of a beyond-design-
basis external event. Although the primary purpose of the order was to
ensure that operators were not distracted by uncertainties related to
SFP conditions during the accident response, the improved monitoring
capabilities will help in the diagnosis and response to potential
losses of SFP integrity. In addition, on March 12, 2012, the NRC issued
Order EA-12-049, ``Order Modifying Licenses with Regard to Requirements
for Mitigation Strategies for Beyond-Design-Basis External Events,''
(ADAMS Accession No. ML12054A735), which requires licensees to develop,
implement, and maintain guidance and strategies to maintain or restore
SFP cooling capabilities, independent of alternating current power,
following a beyond-design-basis external event. These requirements
ensure a more reliable and robust mitigation capability is in place to
address degrading conditions in SFPs.
The NRC believes that much of the information in the SFP studies
that have been accomplished since NUREG-1738, as discussed previously,
will contribute to the development of a regulatory basis for the
current power reactor decommissioning rulemaking effort.
In the SRM to SECY-14-0118, ``Request by Duke Energy Florida, Inc.,
for Exemptions from Certain Emergency Planning Requirements,'' dated
December 30, 2014 (ADAMS Accession No. ML14364A111), the Commission
directed the NRC staff to proceed with
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rulemaking on reactor decommissioning and set an objective of early
2019 for its completion. The Commission also stated that this
rulemaking should address the following:
Issues discussed in SECY-00-0145 such as the graded
approach to emergency preparedness;
Lessons learned from the plants that have already (or are
currently) going through the decommissioning process;
The advisability of requiring a licensee's post-shutdown
decommissioning activity report (PSDAR) to be approved by the NRC;
The appropriateness of maintaining the three existing
options (DECON, SAFSTOR, and ENTOMB \1\) for decommissioning and the
timeframes associated with those options;
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\1\ These options were first identified in the 1988 Generic
Environmental Impact Statement and defined as follows:
DECON: The equipment, structures, and portions of the facility
and site that contain radioactive contaminants are promptly removed
or decontaminated to a level that permits termination of the license
shortly after cessation of operations.
SAFSTOR: The facility is placed in a safe, stable condition and
maintained in that state (safe storage) until it is subsequently
decontaminated and dismantled to levels that permit license
termination. During SAFSTOR, a facility is left intact, but the fuel
has been removed from the reactor vessel, and radioactive liquids
have been drained from systems and components and then processed.
Radioactive decay occurs during the SAFSTOR period, thus reducing
the quantity of contaminated and radioactive material that must be
disposed of during decontamination and dismantlement. The definition
of SAFSTOR also includes the decontamination and dismantlement of
the facility at the end of the storage period.
ENTOMB: Radioactive systems, structures, and components are
encased in a structurally long-lived substance, such as concrete.
The entombed structure is appropriately maintained, and continued
surveillance is carried out until the radioactivity decays to a
level that permits termination of the license.
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The appropriate role of State and local governments and
nongovernmental stakeholders in the decommissioning process; and
Any other issues deemed relevant by the NRC staff.
In SECY-15-0014, ``Anticipated Schedule and Estimated Resources for
a Power Reactor Decommissioning Rulemaking,'' dated January 30, 2015
(ADAMS Accession No. ML15082A089--redacted), the NRC staff committed to
proceed with a rulemaking on reactor decommissioning and provided an
anticipated schedule and estimate of the resources required for the
completion of a decommissioning rulemaking. In SECY-15-0127,
``Schedule, Resource Estimates, and Impacts for the Power Reactor
Decommissioning Rulemaking,'' dated October 7, 2015, (non-publicly
available), the staff provided further information to the Commission on
resource estimates and work that will be delayed or deferred in fiscal
year (FY) 2016 to enable the staff to make timely progress consistent
with Commission direction to have a final rule submitted to the
Commission by the end of FY 2019.
B. Licensing Actions Related to Decommissioning Power Reactors
In 2013, four power reactor units permanently shut down without
significant advance notice or pre-planning. These licensees and the
associated shut down reactors are: Duke Energy Florida for Crystal
River Unit 3 Nuclear Generation Plant; Dominion Energy Kewaunee for
Kewaunee Power Station; and Southern California Edison for San Onofre
Nuclear Generating Station, Units 2 and 3.
On December 29, 2014, Entergy Nuclear Operations, Inc., shut down
Vermont Yankee Nuclear Power Station (VY), and on January 12, 2015, the
licensee certified that VY had permanently ceased operation and removed
fuel from the reactor vessel. Furthermore, Exelon Generation Company,
the licensee for the Oyster Creek Nuclear Generating Station, has
indicated that it is currently planning to shut down that facility in
2019.
Both the decommissioning reactor licensees and the NRC have
expended substantial resources processing licensing actions for these
power reactors during their transition period to a decommissioning
status. Consistent with the power reactors that permanently shutdown in
the 1990s, the licensees that are currently transitioning to
decommissioning are establishing a long-term regulatory framework based
on the low risk of an offsite radiological release posed by a
decommissioning reactor. The licensees are seeking NRC approval of
exemptions and amendments, to reduce requirements no longer needed or
no longer relevant for permanently shutdown reactors.
The NRC has not identified any significant risks to public health
and safety in the current regulatory framework for decommissioning
power reactors. Consequently, the need for a power reactor
decommissioning rulemaking is not based on any identified safety-driven
or security-driven concerns. When compared to an operating reactor, the
risk of an offsite radiological release is significantly lower, and the
types of possible accidents are significantly fewer, at a nuclear power
reactor that has permanently ceased operations and removed fuel from
the reactor vessel. Although the need for a power reactor
decommissioning rulemaking is not based on safety concerns, the NRC
understands that the decommissioning process can be improved and made
more efficient and predictable by reducing its reliance on processing
licensing actions to achieve a long-term regulatory framework for
decommissioning. Therefore, the primary objective of the
decommissioning rulemaking is to implement appropriate regulatory
changes that reduce the number of licensing actions needed during
decommissioning.
The NRC anticipates that a power reactor decommissioning rulemaking
will require substantial interactions with all stakeholders. The
information developed in SECY-00-0145 provides a historical perspective
on the regulatory challenges that the NRC is facing for those licensees
currently transitioning to decommissioning. In addition, SECY-00-0145
serves as a good starting point for the current reactor decommissioning
rulemaking effort. However, as a result of the changes to operating
reactor regulations in the areas of EP and security after September 11,
2001, and the earthquake and tsunami affecting the Fukushima Dai-ichi
nuclear power station in Japan, there will likely be many differences
in the current rulemaking effort as compared to the rulemaking approach
proposed in SECY-00-0145. The proposed decommissioning rulemaking
effort needs to be carefully scoped to ensure an efficient and timely
rulemaking process. Incorporating too broad of a regulatory scope into
a single rule was one of the challenges encountered during the prior
rulemaking effort.
Until a new decommissioning rulemaking is complete, licensees that
are considering decommissioning can use recently completed
decommissioning licensing actions as a template for beginning
decommissioning activities. In addition, the NRC can use these recent
licensing action evaluations as a precedent when processing similar
decommissioning actions. The recently completed licensing actions will
also provide supporting information for the framework and context of a
power reactor decommissioning rulemaking. The NRC has also completed
interim staff guidance on processing EP license exemptions (NSIR/DPR-
ISG-02, ``Emergency Planning Exemption Requests for Decommissioning
Nuclear Power Plants,'' ADAMS Accession No. ML13304B442), and has
issued draft interim staff guidance for physical security license
exemptions (NSIR/DSP-ISG-03, ``Review of Security
[[Page 72362]]
Exemptions/License Amendment Requests for Decommissioning Nuclear Power
Plants,'' ADAMS Accession No. ML14294A170).
The NRC intends to work closely with all stakeholders to ensure
that the decommissioning rulemaking can be achieved within a reasonable
timeframe.
III. Discussion
The NRC has determined that interaction with the public and
stakeholders will help to inform the development of a regulatory basis
for the power reactor decommissioning rulemaking. This ANPR is
structured around questions intended to solicit information that: (1)
Defines the scope of stakeholder interest in a decommissioning
rulemaking, and (2) supports the development of a complete and adequate
regulatory basis. Commenters should feel free to provide feedback on
any aspect of power reactor decommissioning that would support this
ANPR's regulatory objective, whether or not in response to a question
listed in this ANPR.
IV. Regulatory Objectives
The NRC is developing a proposed rule that would amend the current
requirements for power reactors transitioning to decommissioning.
Experience has demonstrated that licensees for decommissioning power
reactors seek several exemptions and license amendments per site to
establish a long-term licensing basis for decommissioning. By issuing a
decommissioning rule, the NRC would be able to establish regulations
that would maintain safety and security at sites transitioning to
decommissioning without the need to grant specific exemptions or
license amendments in certain regulatory areas. Specifically, the
decommissioning rulemaking would have the following goals: (1) Continue
to provide reasonable assurance of adequate protection of the public
health and safety and common defense and security at decommissioning
power reactor sites; (2) Ensure that the requirements for
decommissioning power reactors are clear and appropriate; (3) Codify
those issues that are found to be generically applicable to all
decommissioning power reactors and have resulted in the need for
similarly-worded exemptions or license amendments; and (4) Identify,
define, and resolve additional areas of concern related to the
regulation of decommissioning power reactors.
A. Applicability to NRC Licenses and Approvals
The NRC would apply these updated requirements to power reactors
permanently shut down and defueled and entered into decommissioning.
Accordingly, the NRC envisions that the requirements would apply to
the following:
Nuclear power plants currently licensed under 10 CFR part
50;
Nuclear power plants currently being constructed under
construction permits issued under 10 CFR part 50, or whose construction
permits may be reinstated;
Future nuclear power plants whose construction permits and
operating licenses are issued under 10 CFR part 50; and
Current and future nuclear power plants licensed under 10
CFR part 52.
B. Interim Regulatory Actions
The NRC recognizes that it will take several years to issue a final
rule. If additional reactors begin decommissioning before
implementation of the final rule, the NRC anticipates that licensees
will continue to use existing regulatory processes (for example,
exemptions and license amendments) to establish their decommissioning
regulatory framework.
V. Specific Considerations
The NRC is seeking stakeholders' input on the following specific
areas related to power reactor decommissioning regulations. The NRC
asks that commenters provide the bases for their comments (i.e., the
underlying rationale for the position stated in the comment) to enable
the NRC to have a complete understanding of commenters' positions.
A. Questions Related to Emergency Preparedness Requirements for
Decommissioning Power Reactor Licensees
The EP requirements of 10 CFR 50.47, ``Emergency Plans,'' and
appendix E, ``Emergency Planning and Preparedness for Production and
Utilization Facilities,'' to 10 CFR part 50 continue to apply to a
nuclear power reactor after permanent cessation of operations and
removal of fuel from the reactor vessel. Currently, there are no
explicit regulatory provisions distinguishing EP requirements for a
power reactor that has been shut down from those for an operating power
reactor. The NRC is considering several changes to the EP requirements
in 10 CFR part 50, ``Domestic Licensing of Production and Utilization
Facilities,'' including Sec. 50.47, ``Emergency Plans;'' appendix E to
10 CFR part 50, ``Emergency Planning and Preparedness for Production
and Utilization Facilities''; Sec. 50.54(s), (q), and (t), and Sec.
50.72(a) and (b). These areas are discussed in more detail in this
section. The questions on EP have been listed in this document using
the acronym ``EP'' and sequential numbers.
EP-1: The NRC has previously approved exemptions from the emergency
planning regulations in Sec. 50.47 and appendix E to 10 CFR part 50 at
permanently shut down and defueled power reactor sites based on the
determination that there are no possible design-basis events at a
decommissioning licensee's facility that could result in an offsite
radiological release exceeding the limits established by the EPA's
early-phase protective action guidelines of 1 rem at the exclusion area
boundary. In addition, the possibility of the spent fuel in the SFP
reaching the point of a beyond-design-basis zirconium fire is highly
unlikely based on an analysis of the amount of time before spent fuel
could reach the zirconium ignition temperature during a SFP partial
drain-down event, assuming a reasonably conservative adiabatic heat-up
calculation. A minimum of 10 hours is the time that was used in
previously approved exemptions, which allows for onsite mitigative
actions to be taken by the licensee or actions to be taken by offsite
authorities in accordance with the comprehensive emergency management
plans (i.e., all hazards plans). For licensees that have been granted
exemptions, the EP regulations, as exempted, continue to require the
licensees to, among other things, maintain an onsite emergency plan
addressing the classification of an emergency, notification of
emergencies to licensee personnel and offsite authorities, and
coordination with designated offsite government officials following an
event declaration so that, if needed, offsite authorities may implement
protective actions using a comprehensive emergency management (all-
hazard) approach to protect public health and safety. The EP exemptions
relieve the licensee from the requirement to maintain formal offsite
radiological emergency preparedness, including the 10-mile emergency
planning zone.
a. What specific EP requirements in Sec. 50.47 and appendix E to
10 CFR part 50 should be evaluated for modification, including any EP
requirements not addressed in previously approved exemption requests
for licensees with decommissioning reactors?
b. What existing NRC EP-related guidance and other documents should
[[Page 72363]]
be revised to address implementation of changes to the EP requirements?
c. What new guidance would be necessary to support implementation
of changes to the EP requirements?
EP-2: Rulemaking may involve a tiered approach for modifying EP
requirements based on several factors, including, but not limited to,
the source term after cessation of power operations, removal of fuel
from the reactor vessel, elapsed time after permanent defueling, and
type of long-term onsite fuel storage.
a. What tiers and associated EP requirements would be appropriate
to consider for this approach?
b. What factors should be considered in establishing each tier?
c. What type of basis could be established to support each tier or
factor?
d. Should the NRC consider an alternative to a tiered approach for
modifying EP requirements? If so, provide a description of a proposed
alternative.
EP-3: Several aspects of offsite EP, such as formal offsite
radiological emergency plans, emergency planning zones, and alert and
notification systems, may not be necessary at a decommissioning site
when beyond-design-basis events--which could result in the need for
offsite protective actions--are few in number and highly unlikely to
occur.
a. Presently, licensees at decommissioning sites must maintain the
following capabilities to initiate and implement emergency response
actions: Classify and declare an emergency, assess releases of
radioactive materials, notify licensee personnel and offsite
authorities, take mitigative actions, and request offsite assistance if
needed. What other aspects of onsite EP and response capabilities may
be appropriate for licensees at decommissioning sites to maintain once
the requirements to maintain formal offsite EP are discontinued?
b. To what extent would it be appropriate for licensees at
decommissioning sites to arrange for offsite assistance to supplement
onsite response capabilities? For example, licensees at decommissioning
sites would maintain agreements with offsite authorities for fire,
medical, and law enforcement support.
c. What corresponding changes to Sec. 50.54(s)(2)(ii) and
50.54(s)(3) (about U.S. Federal Emergency Management Agency (FEMA)-
identified offsite EP deficiencies and FEMA offsite EP findings,
respectively) may be appropriate when offsite radiological emergency
plans would no longer be required?
EP-4: Under Sec. 50.54(q), nuclear power reactor licensees are
required to follow and maintain the effectiveness of emergency plans
that meet the standards in Sec. 50.47 and the requirements in appendix
E to 10 CFR part 50. These licensees must submit to the NRC, for prior
approval, changes that would reduce the effectiveness of their
emergency plans.
a. Should Sec. 50.54(q) be modified to recognize that nuclear
power reactor licensees, once they certify under Sec. 50.82,
``Termination of License,'' to have permanently ceased operation and
permanently removed fuel from the reactor vessel, would no longer be
required to meet all standards in Sec. 50.47 and all requirements in
appendix E? If so, describe how.
b. Should nuclear power reactor licensees, once they certify under
Sec. 50.82 to have permanently ceased operation and permanently
removed fuel from the reactor vessel, be allowed to make emergency plan
changes based on Sec. 50.59, ``Changes, Tests, and Experiments,''
impacting EP related equipment directly associated with power
operations? If so, describe how this might be addressed under Sec.
50.54(q).
EP-5: Under Sec. 50.54(t), nuclear power reactor licensees are
required to review all EP program elements every 12 months. Some EP
program elements may not apply to permanently shut down and defueled
sites; for example, the adequacy of interfaces with State and local
government officials when offsite radiological emergency plans may no
longer be required. Should Sec. 50.54(t) be clarified to distinguish
between EP program review requirements for operating versus permanently
shut down and defueled sites? If so, describe how.
EP-6: The Emergency Response Data System (ERDS) transmits key
operating plant data to the NRC during an emergency. Under Sec.
50.72(a)(4), nuclear power reactor licensees are required to activate
ERDS within 1 hour after declaring an emergency at an ``Alert'' or
higher emergency classification level. Much of the plant data, and
associated instrumentation for obtaining the data, would no longer be
available or needed after a reactor is permanently shut down and
defueled. Section VI.2 to appendix E of 10 CFR part 50 does not require
a nuclear power facility that is shut down permanently or indefinitely
to have ERDS. At what point(s) in the decommissioning process should
ERDS activation, ERDS equipment, and the instrumentation for obtaining
ERDS data, no longer be necessary?
EP-7: Under Sec. 50.72(a)(1)(i), nuclear power reactor licensees
are required to make an immediate notification to the NRC for the
declaration of any of the emergency classes specified in the licensee's
NRC-approved emergency plan. Notification of the lowest level of a
declared emergency at a permanently shut down and defueled reactor
facility may no longer need to be an immediate notification (e.g.,
consider changing the immediate notification category for a
Notification of Unusual Event emergency declaration to a 1-hour
notification). What changes to Sec. 50.72(a)(1)(i) should be
considered for decommissioning sites?
EP-8: Under Sec. 50.72(b)(3)(xiii), nuclear power reactor
licensees are required to make an 8-hour report of any event that
results in a major loss of emergency assessment capability, offsite
response capability, or offsite communications capability (e.g.,
significant portion of control room indication, emergency notification
system, or offsite notification system). Certain parts of this section
may not apply to a permanently shut down and defueled site (e.g., a
major loss of offsite response capability once offsite radiological
emergency plans would no longer be required). What changes to Sec.
50.72(b)(3)(xiii) should be considered for decommissioning sites?
B. Questions Related to the Physical Security Requirements for
Decommissioning Power Reactor Licensees
Currently, the physical protection programs applied at
decommissioning reactors are managed through security plan changes
submitted to the NRC under the provisions of Sec. Sec. 50.90 and
50.54(p) and exemptions submitted to the NRC for approval under Sec.
73.5. All physical protection program requirements contained in the
current Sec. 73.55, appendix B to 10 CFR part 73, ``General Criteria
for Security Personnel,'' and appendix C to 10 CFR part 73, ``Licensee
Safeguards Contingency Plans,'' are applicable to operating reactors
and decommissioning reactors unless otherwise modified. The questions
on physical security requirements (PSR) have been listed in this
document using the acronym ``PSR'' and sequential numbers.
PSR-1: Identify any specific security requirements in Sec. 73.55
and appendices B and C to 10 CFR part 73 that should be considered for
change to reflect differences between requirements for operating
reactors and permanently shut down and defueled reactors.
[[Page 72364]]
PSR-2: The physical security requirements protecting the spent fuel
stored in the SFP from the design basis threat (DBT) for radiological
sabotage are contained in 10 CFR part 73 and would remain unchanged by
this rulemaking. However:
a. Are there any suggested changes to the physical security
requirements in 10 CFR part 73 or its appendices that would be
generically applicable to a decommissioning power reactor while spent
fuel is stored in the SFP (e.g., are there circumstances where the
minimum number of armed responders could be reduced at a
decommissioning facility)? If so, describe them.
b. Which physical security requirements in 10 CFR part 73 should be
generically applicable to spent fuel stored in a dry cask independent
spent fuel storage installation?
c. Should the DBT for radiological sabotage continue to apply to
decommissioning reactors? If it should cease to apply in the
decommissioning process, when should it end?
PSR-3: Should the NRC develop and publish additional security-
related regulatory guidance specific to decommissioning reactor
physical protection requirements, or should the NRC revise current
regulatory guidance documents? If so, describe them.
PSR-4: What clarifications should the NRC make to target sets in
Sec. 73.55(f) that addresses permanently shut down and defueled
reactors?
PSR-5: For a decommissioning power reactor, are both the central
alarm station and a secondary alarm station necessary? If not, why not?
If both alarm stations are considered necessary, could the secondary
alarm station be located offsite?
PSR-6: Under Sec. 73.54, power reactor licensees are required to
protect digital computer and communication systems and networks. These
requirements apply to licensees licensed to operate a nuclear power
plant as of November 23, 2009, including those that have subsequently
shut down and entered into decommissioning.
a. Section 73.54 clearly states that the requirements for
protection of digital computer and communications systems and networks
apply to power reactors licensed under 10 CFR part 50 that were
licensed to operate as of November 23, 2009. However, Sec. 73.54 does
not explicitly mention the applicability of these requirements to power
reactors that are no longer authorized to operate and are transitioning
to decommissioning. Are any changes necessary to Sec. 73.54 to
explicitly state that decommissioning power reactors are within the
scope of Sec. 73.54? If so, describe them.
b. Should there be reduced cyber security requirements in Sec.
73.54 for decommissioning power reactors based on the reduced risk
profile during decommissioning? If so, what would be the recommended
changes?
PSR-7: Under Sec. 73.55(p)(1)(i) and (p)(1)(ii), power reactor
licensees suspend security measures during certain emergency conditions
or during severe weather under the condition that the suspension ``must
be approved as a minimum by a licensed senior operator.'' Literal
interpretation of these regulations would require that only a licensed
senior operator could suspend certain security measures at a
decommissioning reactor facility. However, for permanently shut down
and defueled reactors, licensed operators are no longer required, and
licensees typically eliminate these positions shortly after shut down.
Decommissioning licensees create a new certified fuel handler (CFH)
position (consistent with the definition in Sec. 50.2) as the senior
non-licensed operator at the plant. These positions cannot be compared
directly, so licensees typically are unable to demonstrate that the CFH
position meets the ``as a minimum'' criteria in Sec. 73.55(p). Because
the regulation does not include a provision that authorizes a CFH to
approve the suspension of security measures for permanently shut down
and defueled reactors (similar to Sec. 50.54(y) authorizing the CFH to
approve departures from license conditions or technical
specifications), licensees have requested exemptions from Sec.
73.55(p)(1)(i) and (p)(1)(ii) to allow CFHs to have this authority.
Based on this discussion, are there any concerns about changing the
regulations to include the CFH as having the authority to suspend
certain security measures during certain emergency conditions or during
severe weather for permanently shut down and defueled reactor
facilities? If so, describe them.
PSR-8: Regulations in Sec. 73.55(j)(4)(ii) require continuous
communications capability between security alarm stations and the
control room. The intent of Sec. 73.55(j)(4)(ii) is to ensure that
effective communication between the alarm stations and operations staff
with shift command function responsibility is maintained at all times.
The control room at an operating reactor contains the controls and
instrumentation necessary to ensure safe operation of the reactor and
reactor support systems during normal, off-normal, and accident
conditions and, therefore, is the location of the shift command
function. Following certification of permanent shut down and removal of
the fuel from the reactor, operation of the reactor is no longer
permitted. Although the control room at a permanently shut down and
defueled reactor provides a central location from where the shift
command function can be conveniently performed because of existing
communication equipment, office computer equipment, and access to
reference material, the control room does not need to be the location
of the shift command function since shift command functions are not
tied to this location for safety reasons, and modern communication
systems permit continuous communication capability from anywhere on the
site.
The NRC is considering revising the requirements of Sec.
73.55(j)(4)(ii) for a permanently shut down and defueled reactor. The
revised requirements would be focused on maintaining a system of
continuous communications between the shift manager/CFH and the
security alarm stations (rather than the control room). Such a change
would provide the facility's shift manager/CFH the flexibility to leave
the control room without necessitating that other operational staff
remain in the control room to receive communications from the security
alarm stations. Personal communications systems would permit the shift
manager/CFH to perform managerial and supervisory activities throughout
the plant while maintaining the command function responsibility,
regardless of the supervisor's location.
Based on the discussion above, are there any concerns related to
changing the regulations in Sec. 73.55(j)(4)(ii) to allow another
communications system between the alarm stations and the shift manager/
CFH in lieu of the control room at permanently shut down and defueled
reactors? If so, describe them.
C. Questions Related to Fitness for Duty (FFD) Requirements for
Decommissioning Power Reactor Licensees
The NRC's regulations at Sec. 26.3 lists those licensees and other
entities that are required to comply with designated subparts of 10 CFR
part 26, ``Fitness for Duty Programs.'' Part 26 does not apply to power
reactor licensees that have certified under Sec. 50.82 to have
permanently shut down and defueled. The questions on fitness for duty
(FFD) have been listed in this document using the acronym ``FFD'' and
sequential numbers.
FFD-1: Currently, holders of power reactor licenses issued under 10
CFR part 50 or 10 CFR part 52, ``Licenses, Certifications, and
Approvals for
[[Page 72365]]
Nuclear Power Plants,'' must comply with the physical protection
requirements described in Sec. 73.55 during decommissioning. Under
Sec. 73.55, each nuclear power reactor licensee shall maintain and
implement its Commission-approved security plans as long as the
licensee has a 10 CFR part 50 or 52 license. Furthermore, Sec.
73.55(b)(9) requires the licensee to establish, maintain, and implement
an insider mitigation program (IMP) that contains elements from various
security programs, including the FFD program described in 10 CFR part
26. Each power reactor licensee has committed within its security plan
to using NEI 03-12, ``Security Plan Template,'' revision 7, as the
framework for developing its security plans to meet the requirements of
Sec. 73.55. NEI 03-12, which was endorsed by NRC Regulatory Guide (RG)
5.76, ``Physical Protection Programs at Nuclear Power Reactors
(Safeguards Information (SGI)),'' letter dated November 10, 2011,
states that the IMP is satisfied when the licensee ``implements the
elements of the IMP, utilizing the guidance provided in RG 5.77,
`Insider Mitigation Program.' '' The NRC is in the process of revising
RG 5.77 in order to clarify those FFD elements needed for the IMP.
a. Should the NRC pursue rulemaking to describe what provisions of
10 CFR part 26 apply to decommissioning reactor licensees or use
another method of establishing clear, consistent and enforceable
requirements? Describe other methods, as appropriate.
b. As an alternative to rulemaking, should the drug and alcohol
testing for decommissioning reactors be described in RG 5.77, with
appropriate reference to the applicable requirements in 10 CFR part 26?
This option would be contingent on an NEI commitment to revise NEI 03-
12 to include the most recent revision to RG 5.77 (which would include
the applicable drug and alcohol testing provisions) and an industry
commitment to update their security plans with the revised NEI 03-12.
c. Describe what drug and alcohol testing requirements in 10 CFR
part 26 are not necessary to fulfill the IMP requirements to assure
trustworthiness and reliability.
d. Should another regulatory framework be used, such as a corporate
drug testing program modelled on the U.S. Department of Health and
Human Services' Mandatory Guidelines for Federal Workplace Drug Testing
or the U.S. Department of Transportation's drug and alcohol testing
provisions in 49 CFR part 40? If this option is proposed, describe how
(i) the laboratory auditing, quality assurance, and reporting
requirements would be met by the proposal; (ii) licensees would conduct
alcohol testing; and (iii) the performance objectives of 10 CFR
26.23(a), (b), (c), and (d) would be met.
FFD-2: On March 31, 2008, the NRC published a final rule in the
Federal Register (73 FR 16966) adding subpart I, ``Managing Fatigue,''
to 10 CFR part 26. The addition of subpart I in the revised rule
provides reasonable assurance that the effects of fatigue and degraded
alertness on an individual's ability to safely and competently perform
his or her duties are managed commensurate with maintaining public
health and safety. The fatigue management provisions also reduce the
potential for worker fatigue (e.g., that associated with security
officers, maintenance personnel, control room operators, emergency
response personnel, etc.) to adversely affect the common defense and
security. The 2008 rule established clear and enforceable requirements
for operating nuclear power plant licensees and other entities for the
management of worker fatigue. Power reactor licensees that had
permanently shut down and defueled were not considered within the scope
of that rulemaking effort. This is because the scope of activities at a
facility undergoing decommissioning is much less likely to create a
public health and safety concern due to the significantly reduced risk
of a radiological event.
a. Should any of the fatigue management requirements of 10 CFR part
26, subpart I, apply to a permanently shut down and defueled reactor?
If so, which ones?
b. Based on the lower risk of an offsite radiological release from
a decommissioning reactor, compared to an operating reactor, should
only specific classes of workers, as identified in Sec. 26.4(a)
through (c), be subject to fatigue management requirements (e.g.,
security officers or certified fuel handlers)? Please provide what
classes of workers should be subject to the requirements and a
justification for their inclusion.
c. Should the fatigue management requirements of 10 CFR part 26,
subpart I, continue to apply to the specific classes of workers
identified in response to question b above, for a specified period of
time (e.g., until a specified decay heat level is reached within the
SFP, or until all fuel is in dry storage)? Please provide what period
of time workers would be subject to the requirements and the
justification for the timing.
d. Should an alternate approach to fatigue management be developed
commensurate with the plant's lower risk profile? Please provide a
discussion of the alternate approach and how the measures would
adequately manage fatigue for workers.
D. Questions Related to Training Requirements of Certified Fuel
Handlers for Decommissioning Power Reactor Licensees
Reactor operators are licensed under 10 CFR part 55 to manipulate
the controls of operating power reactors. The regulations at Sec. 55.4
define ``controls'' to mean, ``when used with respect to a nuclear
reactor . . . apparatus and mechanisms the manipulation of which
directly affects the reactivity or power level of the reactor.''
``Controls'' are not relevant at decommissioning reactors because the
reactors are permanently shutdown and defueled and no longer authorized
to load fuel into the reactor vessel. Consequently, without fuel in the
reactor vessel, decommissioning reactors are in a configuration in
which the reactivity or power level of the reactor is no longer
meaningful and there are no conditions where the manipulation of
apparatus or mechanisms can affect the reactivity or power level of the
reactor. Therefore, licensed operators are not required at
decommissioning reactors. The NRC regulations do not explicitly state
the staffing alternative for licensed operators after a reactor has
permanently shutdown and defueled under Sec. 50.82(a)(1). When
licensees permanently shut down their reactors, they must continue to
meet minimum staffing requirements in technical specifications and
regulatory required programs (e.g., emergency response organizations,
fire brigade, security, etc.). Given the reduced risk of a radiological
incident once the certifications of permanent cessation of operation
and permanent removal of fuel from the reactor vessel have been
submitted, licensees typically transition their operating staff to a
decommissioning organization. This transition includes replacing
licensed operators with CFHs as the on-shift management representative
responsible for supervising and directing the monitoring, storage,
handling, and cooling of irradiated nuclear fuel in a manner consistent
with ensuring the health and safety of the public. Regulations in Sec.
50.2 define a CFH for a nuclear power reactor as a non-licensed
operator who has qualified in accordance with a fuel handler training
program approved by the Commission. The transition to the use of CFHs
from licensed operators at decommissioning reactors occurs following
the NRC's
[[Page 72366]]
approval of a licensee's CFH training program and an amendment to the
administrative and organization section of the licensee's defueled
technical specifications.
However, the NRC regulations do not contain criteria for an
acceptable CFH training program. Because of the reduced risks and
relative simplicity of the systems needed for safe storage of the spent
fuel, the Commission stated in the 1996 decommissioning final rule that
``[t]he degree of regulatory oversight required for a nuclear power
reactor during its decommissioning stage is considerably less than that
required for the facility during its operating stage'' (61 FR 39278).
In the proposed rule, the Commission also provided insights as to the
responsibilities of the CFH position. Specifically, the CFHs are needed
at decommissioning reactors to ensure that emergency action decisions
necessary to protect the public health and safety are made by an
individual who has both the requisite knowledge and plant experience
(60 FR 37374, 37379).
In previous evaluations of licensee CFH training programs (ADAMS
Accession Nos. ML14104A046, ML13268A165), the NRC has determined that
an acceptable CFH training program should ensure that the trained
individual has requisite knowledge and experience in spent fuel
handling and storage and reactor decommissioning, and is capable of
evaluating plant conditions and exercising prudent judgment for
emergency action decisions. In addition, since the CFH is defined as a
non-licensed operator, the NRC staff has also evaluated the CFH
training program in accordance with Sec. 50.120, which includes a
requirement in Sec. 50.120(b)(2) that the training program must be
derived from a systems approach to training as defined in Sec. 55.4.
However, as previously noted, the specific training requirements
for the CFH program are not in the regulations. In addition, Sec.
50.120 specifies the training and qualification requirements for non-
licensed reactor personnel but does not address the CFH staffing
position. Because the regulations are silent on the training attributes
of the CFH, regulatory uncertainty regarding the CFH training program
exists. In addition, because the NRC's regulations do not address the
replacement of licensed operators by CFHs, licensees also have
questions regarding the transition from licensed operator training
programs to CFHs' training programs. The questions on CFH have been
listed in this document using the acronym ``CFH'' and sequential
numbers.
CFH-1: Based on the NRC's experience with the review of the CFH
training/retraining programs submitted by licensees that have recently
permanently shutdown, the following questions are focused on areas that
may need additional clarity. Specifically:
a. When should licensees that are planning to enter decommissioning
submit requests for approval of CFH training/retraining programs?
b. What training and qualifications should be required for
operations staff at power reactors that decommission earlier than
expected and that do not have an approved CFH training/retraining
program?
c. Should the NRC issue new requirements that prohibit licensees
from surrendering operators' licenses before implementation of an
approved CFH training/retraining program, or should other incentives or
deterrents be considered? If so, what factors must be included?
d. Should the contents of a CFH training/retraining program be
standardized throughout the industry? If so, how should this be
implemented?
e. Should a process be implemented that requires decommissioning
power reactor licensees to independently manage the specific content of
their CFH training/retraining program based on the systems and
processes actually used at each particular plant instead of
standardization? If so, how should this work?
f. Is there any existing or developing document or program (from
the Institute of Nuclear Power Operations, NEI, NRC, or other related
sources) that provides relevant guidance on the content and format of a
CFH training/retraining program that could be made applicable to CFH
training?
g. Should the requirements for CFH training programs be
incorporated into an overall decommissioning rule, or addressed using
other regulatory vehicles such as associated NUREGs, regulatory guides,
standard review plan chapters or sections, and inspection procedures?
E. Questions Related to the Current Regulatory Approach for
Decommissioning Power Reactor Licensees
In the SRM to SECY-15-0014, the Commission directed the staff to
determine the appropriateness of (1) maintaining the three existing
options for decommissioning and the timeframes associated with those
options, and (2) address the appropriate role of State and local
governments and non-governmental stakeholders in the decommissioning
process. Based on the Commission's direction, the NRC staff is seeking
additional information on the need for any regulatory changes
concerning the use of decommissioning options, the timeframe to
complete decommissioning, and the role of external stakeholders in the
decommissioning process. The questions on regulatory approach (REG)
have been listed in this document using the acronym ``REG'' and
sequential numbers.
REG-1: The NRC has evaluated the environmental impacts of three
general methods for decommissioning power reactor facilities, DECON,
SAFSTOR, or ENTOMB, as described in Section II.A, footnote 1 of this
document. The choice of the decommissioning method is left entirely to
the licensee, provided that the decommissioning method can be performed
in accordance with NRC's regulations. The NRC would require the
licensee to re-evaluate its decision on the method of the
decommissioning process that it chose if it (1) could not be completed
as described, (2) could not be completed within 60 years of the
permanent cessation of plant operations, (3) included activities that
would endanger the health and safety of the public by being outside of
the NRC's health and safety regulations, or (4) would result in a
significant impact to the environment. The licensee's choice is
communicated to the NRC and the public in the PSDAR. To date, most
utilities have used DECON or SAFSTOR to decommission reactors. Several
sites have performed some incremental decontamination and dismantlement
during the storage period of SAFSTOR, a combination of SAFSTOR and
DECON as personnel, money, or other factors become available. No
utilities have used the ENTOMB option for a commercial nuclear power
reactor.
a. Should the current options for decommissioning--DECON, SAFSTOR,
and ENTOMB--be explicitly addressed and defined in the regulations
instead of solely in guidance documents, and how so?
b. Should other options for decommissioning be explored? If so,
what other technical or programmatic options are reasonable and what
type of supporting documents would be most effective for providing
guidance on these new options or requirements?
c. The NRC regulations state that decommissioning must be completed
within 60 years of permanent cessation of operations. A duration of 60
years was chosen because it roughly corresponds to 10 half-lives for
cobalt-60, one of the predominant isotopes remaining in the facility.
By 60 years, the initial short-lived isotopes,
[[Page 72367]]
including cobalt-60, will have decayed to background levels. In
addition, the 60-year period appears to be reasonable from the
standpoint of expecting institutional controls to be maintained.
Completion of decommissioning beyond 60 years will be approved by the
NRC only when necessary to protect public health and safety. Should the
requirements be changed so that the timeframe for decommissioning is
something other than the current 60-year limit? Would this change be
dependent on the method of decommissioning chosen, site specific
characteristics, or some other combination of factors? If so, please
describe.
REG-2: In support of decommissioning planning for a permanently
shut down and defueled power reactor, the licensee submits to the NRC a
PSDAR that: (1) Informs the public of the licensee's planned
decommissioning activities; (2) assists in the scheduling of NRC
resources necessary for the appropriate oversight activities; (3)
ensures that the licensee has considered the costs of the planned
decommissioning activities and has funding for the decommissioning
process; and (4) ensures that the environmental impacts of the planned
decommissioning activities are bounded by those considered in existing
environmental impact statements. After receiving a PSDAR, the NRC
publishes a notice of receipt, makes the PSDAR available for public
review and comment, and holds a public meeting in the vicinity of the
plant to discuss the licensee's plans and address the public's
comments. Although the NRC will determine if the information is
consistent with the regulations, NRC approval of the PSDAR is not
required. However, should the NRC determine that the informational
requirements of the regulations are not met in the PSDAR, the NRC will
inform the licensee, in writing, of the deficiencies and require that
they be addressed before the licensee initiates any major
decommissioning activities. Any decommissioning activities that could
preclude release of the site for possible unrestricted use, impact a
reasonable assurance finding that adequate funds will be available for
decommissioning, or potentially result in a significant environmental
impact not previously reviewed, must receive prior NRC approval.
Specifically, the licensee is required to submit a license amendment
request for NRC review and approval, which provides an opportunity for
public comment and/or a public hearing. Unless the NRC staff approves
the license amendment request, the licensee is not to conduct the
requested activity. Consistent with Commission direction, the NRC staff
is seeking comment on the appropriate role for the NRC in reviewing and
approving the licensee's proposed decommissioning strategy and
associated planning activities.
a. Is the content and level of detail currently required for the
licensee's PSDAR, adequate? If not, what should be added or removed to
enhance the document?
b. Should the regulations be amended to require NRC review and
approval of the PSDAR before allowing any ``major decommissioning
activity,'' as that term is defined in Sec. 50.2, to commence? What
value would this add to the decommissioning process?
REG-3: The NRC's regulations currently offer the public
opportunities to review and provide comments on the decommissioning
process. Specifically, under the NRC's regulations in Sec. 50.82, the
NRC is required to publish a notice of the receipt of the licensee's
PSDAR, make the PSDAR available for public comment, schedule separate
meetings in the vicinity of the location of the licensed facility to
discuss the PSDAR within 60 days of receipt, and publish a notice of
the meetings in the Federal Register and another forum readily
accessible to individuals in the vicinity of the site. For many years,
the NRC has strongly recommended that licensees involved in
decommissioning activities form a community committee to obtain local
citizen views and concerns regarding the decommissioning process and
spent fuel storage issues. It has been the NRC's view that those
licensees who actively engage the community maintain better relations
with the local citizens. The NRC's guidance related to creating a site-
specific community advisory board can be found in NUREG-1757,
``Consolidated Decommissioning Guidance,'' Appendix M, ``Overview of
the Restricted Use and Alternate Criteria Provisions of 10 CFR part 20,
subpart E,'' Section M.6 (ADAMS Accession No. ML063000243). Appendix M
does not require licensees to create a community advisory board, but
only provides recommendations for methods of soliciting public advice.
Nonetheless, Section M.6 contains useful guidance and suggestions for
effective public involvement in the decommissioning process that could
be adopted by any licensee.
a. Should the current role of the States, members of the public, or
other stakeholders in the decommissioning process be expanded or
enhanced, and how so?
b. Should the current role of the States, members of the public, or
other stakeholders in the decommissioning process for non-radiological
areas be expanded or enhanced, and how so? Currently, for all non-
radiological effluents created during the decommissioning process,
licensees are required to comply with EPA or State regulations related
to liquid effluent discharges to bodies of water.
c. For most decommissioning sites, the State and local governments
are involved in an advisory capacity, often as part of a Community
Engagement Panel or other organization aimed at fostering communication
and information exchange between the licensee and the public. Should
the NRC's regulations mandate the formation of these advisory panels?
F. Questions Related to the Application of Backfitting Protection to
Decommissioning Power Reactor Licensees
In the SRM to SECY-98-253, ``Applicability of Plant-Specific
Backfit Requirements to Plants Undergoing Decommissioning,'' dated
February 12, 1999 (ADAMS Accession No. ML12311A689), the Commission
approved development of a Backfit Rule for plants undergoing
decommissioning. The Commission directed the staff to continue to apply
the then-current Backfit Rule to plants undergoing decommissioning
until the final rule was issued. The Commission ordered the development
of a rulemaking plan, which became SECY-00-0145. In SECY-00-0145, the
staff proposed amendments to Sec. 50.109 to clearly show that the
Backfit Rule applies during decommissioning and to remove factors that
are not applicable to nuclear power plants in decommissioning. As
explained in section II.A of this document, that rulemaking never
occurred, but the Commission, in SRM-SECY-14-0118, directed the staff
to proceed with a rulemaking that addresses, among other things, the
issues discussed in SECY-00-0145.
The questions on backfitting protection (BFP) have been listed in
this document using the acronym ``BFP'' and sequential numbers.
BFP-1: The protections provided by the backfitting and issue
finality provisions in 10 CFR parts 50 and 52, respectively, can apply
to a holder of a nuclear power reactor license when the reactor is in
decommissioning. Backfitting and issue finality during decommissioning
can be divided into two areas:
a. When a licensee's licensing basis for operations continues to
apply during
[[Page 72368]]
decommissioning until: (1) The licensee changes the licensing basis,
(2) the NRC's regulations set forth generic criteria delineating when
changes can be made to the licensing basis, or (3) the NRC takes a
facility-specific action that changes the licensee's licensing basis.
Why would backfitting protection apply in this area?
b. When a licensee engages in an activity during decommissioning
for which no prior NRC approval was provided. The activity could be
required by an NRC regulation or new NRC approval (through an order or
licensing action). Why would backfitting protection apply in this area?
BFP-2: Should the NRC propose amendments to Sec. 50.109 consistent
with the preliminary amendments proposed in SECY-00-0145 that would
have created a two-section Backfit Rule: one section that would apply
to nuclear power plants undergoing decommissioning and the other
section that would apply to operating reactors?
G. Questions Related to Decommissioning Trust Funds
The questions on decommissioning trust fund (DTF) have been listed
in this document using the acronym ``DTF'' and sequential numbers.
DTF-1: The Commission's regulation at Sec. 50.75 includes the
reporting requirements for providing reasonable assurance that
sufficient funds will be available for the decommissioning process. The
regulation at Sec. 50.82 contains, in part, requirements on the use of
decommissioning funds. Every 2 years each operating power reactor
licensee must report to the NRC the status of the licensee's
decommissioning funding to provide assurance to the NRC that the
licensee will have sufficient financial resources to accomplish
radiological decommissioning. After decommissioning has begun,
licensees must annually submit a financial assurance status report to
the NRC.
The NRC's authority is limited to assuring that licensees
adequately decommission their facilities with respect to cleanup and
removal of radioactive material prior to license termination.
Activities that go beyond the scope of decommissioning, as defined in
Sec. 50.2, such as waste generated during operations or demolition
costs for greenfield restoration, are not appropriate costs for
inclusion in the decommissioning cost estimate. The collection of funds
for spent fuel management is addressed in Sec. 50.54(bb) where it
indicates that licensees need to have a plan, including financing, for
spent fuel management.
The NRC has not precluded the commingling of the funds in a single
trust fund account to address radiological decommissioning, spent fuel
management, and site restoration, as long as the licensee is able to
identify and account for these specific funds. In the 1996
decommissioning rule, the Commission indicated that the rule ``does not
prohibit licensees from having separate subaccounts for other
activities in the decommissioning trust fund if minimum amounts
specified in the rule are maintained for radiological
decommissioning.'' Similarly, in the 2002 Decommissioning Trust
Provisions Rule, the Commission stated that it ``appreciates the
benefits that some licensees may derive from their use of a single
trust fund for all of their decommissioning costs, both radiological
and not; but, as stated above, a licensee must be able to identify the
individual amounts contained within its single trust. Therefore, where
a licensee has not separately identified and accounted for expenses
related to non-radiological decommissioning in its DTF, licensees are
required to request exemptions from Sec. 50.82(a)(8)(i)(A) and either
Sec. 50.75(h)(1)(iv) or Sec. 50.75(h)(2), to gain access to monies in
the decommissioning trust fund for purposes other than decommissioning
(e.g., spent fuel management). The NRC has approved exemptions from the
requirements of Sec. Sec. 50.82 and 50.75 allowing withdrawals to be
made from decommissioning trust funds for spent fuel management in
instances where the level of funding needed to complete decommissioning
is not adversely affected. In each instance, the NRC found, pursuant to
Sec. 50.12, the exemptions were authorized by law, presented no undue
risk to public health and safety, and were consistent with the common
defense and security, and found that the application of the rules was
unnecessary to achieve the underlying purpose of the rules.
In some cases, a licensee will not need an exemption. Those cases
exist when a licensee can clearly show that (1) its decommissioning
trust includes State-required funds and (2) the amount of radiological
decommissioning funds in the trust exceeds the amount of money
estimated to be needed for radiological decommissioning in the
licensee's site specific decommissioning cost estimate (or if the
licensee does not have a site specific decommissioning cost estimate
yet, then the minimum amount necessary to provide financial assurance
under Sec. 50.75). If the licensee meets these criteria, then
reasonable assurance of adequate radiological decommissioning funding
still exists after removal of the State-required funds, and the
licensee does not need an exemption to use those State-required funds.
The NRC issued Regulatory Issue Summary (RIS) 2001-07, Revision 1,
``10 CFR 50.75 Reporting and Recordkeeping for Decommissioning
Planning,'' on January 8, 2009 (ADAMS Accession No. ML083440158), to
clarify the need for licensees to preserve the distinction in their
decommissioning trust accounts between the radiological decommissioning
fund balance and amounts accumulated for other purposes, such as paying
for spent fuel management and site restoration, when using the trust
for commingled funds. However, based on NRC experience with the power
reactors that have recently and permanently shut down and entered into
decommissioning, licensees continue to report funds they have
accumulated to address spent fuel management and site restoration as
part of the amount of funds reported for radiological decommissioning.
Should the regulations in Sec. Sec. 50.75 and 50.82 be revised to
clarify the collection, reporting, and accounting of commingled funds
in the decommissioning trust fund, that is in excess of the amount
required for radiological decommissioning and that has been designated
for other purposes, in order to preclude the need to obtain exemptions
for access to the excess monies?
DTF-2: The regulation at Sec. 50.82(a)(8)(i)(A) states that
decommissioning trust funds may only be used by licensees if their
withdrawals ``are for expenses for legitimate decommissioning
activities consistent with the definition of decommissioning in Sec.
50.2.'' In accordance with Sec. 50.2, decommission means to remove a
nuclear facility or site safely from service and reduce residual
radioactivity to a level that permits: (1) Release of the property for
unrestricted use and termination of the license; or (2) release of the
property under restricted conditions and termination of the NRC
license. Thus, ``legitimate decommissioning activities'' include only
those activities whose expenses are related to removing a nuclear
facility or site safely from service and reducing residual
radioactivity to a level that permits license termination and release
of the property for restricted or unrestricted use.
While the regulations are silent with regards to what specific
expenses are related to legitimate decommissioning
[[Page 72369]]
activities, the NRC's guidance documents identify some specific
expenses that may or may not be paid from the decommissioning trust
fund. For example, Regulatory Guide (RG) 1.184, Revision 1,
``Decommissioning of Nuclear Power Reactors'' (ADAMS Accession No.
ML13144A840), states that the amount set aside for radiological
decommissioning as required by Sec. 50.75 ``should not be used for:
(1) The maintenance and storage of spent fuel in the spent fuel pool,
(2) the design, construction, or decommissioning of spent fuel dry
storage facilities directly related to permanent disposal, (3) other
activities not directly related to radiological decontamination or
dismantlement of the facility or site.'' Similarly, other NRC guidance
explain that the NRC's definition of decommissioning does not include
other activities related to facility deactivation and site closure,
including operation of the spent fuel storage pool, construction and/or
operation of an ISFSI, demolition of decontaminated structures, and/or
site restoration activities after residual radioactivity has been
removed. The NRC also has additional guidance that states that removing
uncontaminated material, such as soil or a wall, to gain access to
contamination to be removed would be a legitimate decommissioning cost.
Finally, guidance also exists that provides examples of activities
outside the scope of decommissioning including, ``(1) the maintenance
and storage of spent fuel, (2) the design and/or construction of a
spent fuel dry storage facility, (3) activities that are not directly
related to supporting long-term storage of the facility, or (4) any
other activities not directly related to radiological decontamination
of the site.''
a. What changes should be considered for Sec. Sec. 50.2 and
50.82(a)(8) to clarify what constitutes a legitimate decommissioning
activity?
b. Regulations in Sec. 50.82(8)(ii) states that 3 percent of the
decommissioning funds may be used during the initial stages of
decommissioning for decommissioning planning activities. What should be
included or specifically excluded in the definition of
``decommissioning planning activities?''
H. Questions Related to Offsite Liability Protection Insurance
Requirements for Decommissioning Power Reactor Licensees
The questions on offsite liability protection insurance (LPI) have
been listed in this document using the acronym ``LPI'' and sequential
numbers.
LPI-1: The Price Anderson Act of 1957 (PAA) requires that nuclear
power reactor licensees have insurance to compensate the public for
damages arising from a nuclear incident, including such expenses as
those for personal injury, property damage, or the legal cost
associated with lawsuits. Regulations in 10 CFR part 140, ``Amounts of
Financial Protection for Certain Reactors,'' set forth the amounts of
insurance each power reactor licensee must have. Specifically, Sec.
140.11(a)(4) requires a reactor licensee to maintain $375 million in
offsite liability insurance coverage. In addition, the primary
insurance is supplemented by a secondary insurance tier. In the event
of an accident causing offsite damages in excess of $375 million, each
licensee would be assessed a prorated share of the excess damages, up
to $121.3 million per reactor, for a total of approximately $13
billion.
Regulations in Sec. 140.11(a)(4) do not distinguish between a
reactor that is authorized to operate and a reactor that has
permanently shut down and defueled. Most of the accident scenarios
postulated for operating power reactors involve failures or
malfunctions of systems that could affect the fuel in the reactor core,
which in the most severe postulated accidents, would involve the
release of large quantities of fission products. With the permanent
cessation of reactor operations and the permanent removal of the fuel
from the reactor core, such reactor accidents are no longer possible
with a decommissioning reactor.
The PAA requires licensees of facilities with a rated capacity of
100,000 electrical kilowatts or more to have the primary and secondary
insurance coverage described above, which the NRC establishes in 10 CFR
part 140. Typically, the NRC will issue a decommissioning licensee a
license amendment to remove the rated capacity of the reactor from the
license. This has the effect of removing the reactor licensee from the
category of licensees that are required to maintain the primary and
secondary insurance amounts under the PAA and 10 CFR part 140.
Most permanently shut down and defueled power reactor licensees
have requested exemptions from Sec. 140.11(a)(4) to reduce the
required amount of primary offsite liability insurance coverage from
$375 million to $100 million and to withdraw from the secondary
insurance pool. As noted above, these licensees are no longer within
the category of licensees that are legally required under the PAA to
have these amounts of offsite liability insurance. The technical
criteria for granting these exemptions are based on the determination
that there are no possible design-basis events at a licensee's facility
that could result in an offsite radiological release exceeding the
limits established by the EPA's early-phase Protective Action
Guidelines of 1 rem at the exclusion area boundary. In addition, the
exemptions are predicated on the licensee demonstrating that the heat
generated by the spent fuel in the SFP has decayed to the point where
the possibility of a zirconium fire is highly unlikely. Specifically,
if all coolant were drained from the SFP as the result of a highly
unlikely beyond design-basis accident, the fuel assemblies would remain
below a temperature of incipient cladding oxidation for zirconium based
on air-cooling alone. For a postulated situation where the cooling
configuration of a highly unlikely beyond design basis accident results
in an unknown cooling configuration of the spent fuel, analysis should
demonstrate that even with no cooling of any kind (conduction,
convection, or radiative heat transfer), the spent fuel stored in the
SFP would not reach the zirconium ignition temperature in fewer than 10
hours starting from the time at which the accident was initiated. The
NRC has considered 10 hours sufficient time to take mitigative actions
to cool the spent fuel. Based on this discussion:
a. Should the NRC codify the current conservative exemption
criteria (i.e., 10 hours to take mitigative actions) that have been
used in granting decommissioning reactor licensees exemptions to Sec.
140.11(a)(4)?
b. As an alternative to codifying the current conservative
exemption criteria (i.e., 10 hours to take mitigative actions), should
the NRC codify a requirement to allow decommissioning reactor licensees
to generate site specific criteria (i.e., time period to take
mitigative actions) based upon a site specific analysis?
c. The use of $100 million for primary liability insurance level is
based on Commission policy and precedent from the early 1990s. The
amount established was a qualitative value to bound the claims from the
Three Mile Island accident. Should this number be adjusted?
d. What other factors should be considered in establishing an
appropriate primary insurance liability level (based on the potential
for damage claims) for a decommissioning plant once the risk of any
kind of offsite radiological release is highly unlikely?
[[Page 72370]]
I. Questions Related to Onsite Damage Protection Insurance Requirements
for Decommissioning Power Reactor Licensees
The questions on onsite damage protection insurance (ODI) have been
listed in this document using the acronym ``ODI'' and sequential
numbers.
ODI-1: The requirements of Sec. 50.54(w)(1) call for each power
reactor licensee to have insurance to provide minimum coverage for each
reactor site of $1.06 billion or whatever amount of insurance is
generally available from private sources, whichever is less. The
insurance would be used, in the event of an accident at the licensee's
reactor, to provide financial resources to stabilize the reactor and
decontaminate the reactor site, if needed.
The requirements in Sec. 50.54(w)(1) do not distinguish between a
reactor authorized to operate and a reactor that has permanently shut
down and defueled. With the permanent cessation of reactor operations
and the permanent removal of the fuel from the reactor core, operating
reactor accidents are no longer possible. Therefore, the need for
onsite insurance at a decommissioning reactor to stabilize accident
conditions or decontaminate the site following an accident, should be
significantly lower compared to the need for insurance at an operating
reactor.
Based on NRC policy and precedent, permanently shut down and
defueled reactor licensees have requested exemptions from Sec.
50.54(w)(1). The exemption granted to a permanently shut down reactor
licensee permits the licensee to reduce the required level of onsite
property damage insurance from the amount established in Sec.
50.54(w)(1) to $50 million. The NRC has previously determined that $50
million bounds the worst radioactive waste contamination event (caused
by a liquid radioactive waste storage tank rupture) once the heat
generated by the spent fuel in the SFP has decayed to the point where
the possibility of a zirconium fire in any beyond design-basis accident
is highly unlikely, and in any case, there is sufficient time to take
mitigative actions. The technical criteria used in assessing the
possibility of a zirconium fire, as discussed in question LPI-1 above,
is also used for exemptions from Sec. 50.54(w)(1). Based on this
discussion:
a. Should the NRC codify the current exemption criteria that have
been used in granting decommissioning reactor licensees exemptions from
Sec. 50.54(w)(1)? If so, describe why.
b. The use of $50 million insurance level for bounding onsite
radiological damages is based on a postulated liquid radioactive waste
storage tank rupture using analyses from the early 1990s. Should this
number be adjusted? If so, describe
c. Is the postulated rupture of a liquid radioactive waste storage
tank an appropriate bounding postulated accident at a decommissioning
reactor site once the possibility of a zirconium fire has been
determined to be highly unlikely?
J. General Questions Related to Decommissioning Power Reactor
Regulations
The general (GEN) questions related to decommissioning power
reactor regulations have been listed in this document using the acronym
``GEN'' and sequential numbers.
GEN-1: Section 50.51, ``Continuation of License,'' states in
paragraph (b)(1) that all permanently shut down and defueled reactor
licensees shall continue to take actions to maintain the facility, and
the storage and control and maintenance of spent fuel, in a safe
condition beyond the license expiration date until the Commission
notifies the licensee in writing that the license is terminated. The
NRC has recently focused on the licensee's maintenance of long lived,
passive structures and components at decommissioning reactors. The NRC
expects that many long-lived, passive structures and components may
generally not have performance and condition characteristics that can
be readily monitored, or could be considered inherently reliable by
licensees and do not need to be monitored under Sec. 50.65(a)(1).
There may be few, if any, actual maintenance activities (e.g.,
inspection or condition monitoring) that a licensee conducts for such
structures and components. Treatment of long-lived, passive structures
and components under the maintenance rule is likely to involve minimal
preventive maintenance or monitoring to maintain functionality of such
structures and components in the original licensing period. The NRC is
interested in the need to provide reasonable assurance that certain
long-lived, passive structures and components (e.g., neutron absorbing
materials, SFP liner) are maintained and monitored during the
decommissioning period while spent fuel is in the SFP.
Based on the discussion above, what regulatory changes should be
considered that address the performance or condition of certain long-
lived, passive structures and components needed to provide reasonable
assurance that they will remain capable of fulfilling their intended
functions during the decommissioning period?
GEN-2: Section 50.54(m) of the NRC's regulations for operating
reactors specifies the minimum licensed operator staffing levels (e.g.,
minimum staffing per shift for licensed operators and senior operators)
for power reactors authorized to operate. The regulations define the
duties of licensed operators as either the manipulation of controls or
supervising the manipulation of controls that directly affect the
reactor reactivity or power level of the reactor. A decommissioning
plant is clearly not operating and no manipulation of controls that
affect reactor reactivity or power can occur at a permanently defueled
reactor. Therefore, the requirements in Sec. 50.54(m) concerning
licensed operator staffing levels for operating reactors are not
applicable to a decommissioning plant. For a decommissioning power
reactor, the senior on-shift management representative is a certified
fuel handler who, as stated in Sec. 50.2, is a non-licensed operator
that has qualified in accordance with a fuel handler training program
approved by the Commission. However, there are no regulatory provisions
similar to Sec. 50.54(m) concerning operator staffing levels for a
power reactor licensee once it has certified that it is permanently
shut down and defueled under Sec. 50.82(a)(1). Because the
decommissioning regulations are silent regarding staffing levels,
licensees have sought amendments in their defueled technical
specifications to specify minimum non-licensed operator staffing. Based
on precedent used at most previous permanently shut down reactors, and
considering the demonstrated safety performance of reactor
decommissioning sites over many years, the NRC has found that an
operations staff crew complement consisting of one certified fuel
handler and one non-certified operator is an acceptable minimum
staffing level.
Considering the discussion above, should minimum operations shift
staffing at a permanently shutdown and defueled reactor be codified by
regulation?
GEN-3: Related to the decommissioning plant operator staffing
levels is the requirement for and the use of a control room during
decommissioning. Section 50.54(m) specifies the control room staffing
requirements for licensed operators at an operating reactor with a
fueled reactor vessel. No such requirements exist for the location of
operations staff at a permanently shutdown and defueled reactor. The
control room at an
[[Page 72371]]
operating reactor contains the controls and instrumentation necessary
for complete supervision and response needed to ensure safe operation
and shutdown of the reactor and support systems during normal, off-
normal, and accident conditions and, therefore, is the location of the
shift command function. Following permanent shutdown and removal of
fuel from the reactor, operation of the reactor is no longer permitted
and the control room no longer performs all of the functions that were
required for an operating reactor. There are no longer any activities
at a permanently shutdown and defueled reactor that require a quick
decision and response by operations staff in the control room. For most
decommissioning reactors, the NRC has approved license amendments to
the technical specifications that require at least one non-licensed
operator to remain in a control room. This technical specification
change is primarily based on precedent. However, the NRC has noted in
the license amendment safety evaluations that the primary functions of
the control room at a permanently shutdown reactor are monitoring,
response, communications, and coordination. Specifically, the control
room at a decommissioning reactor is where many plant systems and
equipment parameters are monitored (for operating status and
conditions, radiation levels, electrical anomalies, or fire alarms for
example). Control room personnel assess plant conditions; evaluate the
magnitude and potential consequences of abnormal conditions; determine
preventative, mitigating and corrective actions; and perform
notifications. The control room provides a central location from where
the shift command function can be conveniently performed because of the
availability of existing monitoring and assessment instrumentation,
communication systems and equipment, office computer equipment, and
ready access to reference material. The control room also provides a
central location from which emergency response activities are
coordinated. When activated, the emergency response organization
reports to the control room.
During reactor decommissioning, the control room may be subject to
extensive changes, which are evaluated by the licensee for safety
implications under the Sec. 50.59 process. There is precedent among
some previous decommissioning reactor licensees to design and construct
a decommissioning control room that is independent of the original
operating control room. Most decommissioning reactors can probably
demonstrate that the command, communications, and monitoring functions
performed in the control room could be readily performed at an
alternate onsite location, based on the site-specific needs of a
licensee during its decommissioning process. Consequently, several
decommissioning licensees have questioned the meaning of the control
room as it relates to decommissioning nuclear power plants.
Based on the discussion above, what regulatory changes should be
considered for a permanently shutdown and defueled reactor to prevent
ambiguities concerning the meaning of the control room for
decommissioning reactors and should minimum staffing levels be
specified for the control room?
GEN-4: Are there any other changes to 10 CFR Chapter I, ``Nuclear
Regulatory Commission,'' that could be clarified or amended to improve
the efficiency and effectiveness of the reactor decommissioning
process?
GEN-5: The NRC is attempting to gather information on the costs and
benefits of the changes in the regulatory areas discussed in this
document as early as possible in the rulemaking process. Given the
topics discussed, please provide estimated costs and benefits of
potential changes in these areas from either the perspective of a
licensee or from the perspective of an external stakeholder.
a. From your perspective, which areas discussed are the most
beneficial or detrimental?
b. From your perspective, assuming you believe changes are needed
to the NRC's reactor decommissioning regulatory infrastructure, what
are the factors that drive the need for changes in these regulatory
areas? If at all possible, please provide specific examples (e.g.,
expected savings, expectations for efficiency, anticipated effects on
safety, etc.) about how these changes will affect you.
c. Are there areas that are of particular interest to you, and for
what reason?
d. Please provide any suggested changes that would further enhance
benefits or reduce risks that may not have been addressed in this ANPR.
VI. Public Meeting
The NRC will conduct a public meeting to discuss the contents of
this ANPR and to answer questions from the public regarding the
contents of this ANPR. The NRC will publish a notice of the location,
time, and agenda of the meeting on the NRC's public meeting Web site at
least 10 calendar days before the meeting. Stakeholders should monitor
the NRC's public meeting Web site for information about the public
meeting at: http://www.nrc.gov/public-involve/public-meetings/index.cfm. In addition, the meeting information will be posted on
www.regulations.gov under Docket ID NRC-2015-0070. For instructions on
how to receive alerts when changes or additions occur in a docket
folder, see Section IX of this document.
VII. Cumulative Effects of Regulation
The NRC has implemented a program to address the possible
Cumulative Effects of Regulation (CER), in the development of
regulatory bases for rulemakings. The CER describes the challenges that
licensees, or other impacted entities (such as State partners) may face
while implementing new regulatory positions, programs, and requirements
(e.g., rules, generic letters, backfits, inspections). The CER is an
organizational effectiveness challenge that results from a licensee or
impacted entity implementing a number of complex positions, programs or
requirements within a limited implementation period and with available
resources (which may include limited available expertise to address a
specific issue). The NRC is specifically requesting comment on the
cumulative effects that may result from this potential rulemaking. In
developing comments on the development of the regulatory basis for
revisions to the requirements for decommissioning power reactor
licensees relative to CER, consider the following questions:
(1) In light of any current or projected CER challenges, what
should be a reasonable effective date, compliance date, or submittal
date(s) from the time the final rule is published to the actual
implementation of any new proposed requirements including changes to
programs, procedures, or the facility?
(2) If current or projected CER challenges exist, what should be
done to address this situation (e.g., if more time is required to
implement the new requirements, what period of time would be
sufficient, and why such a time frame is necessary)?
(3) Do other (NRC or other agency) regulatory actions (e.g.,
orders, generic communications, license amendment requests, and
inspection findings of a generic nature) influence the implementation
of the potential proposed requirements?
(4) Are there unintended consequences? Does the potential proposed
action create conditions that would be contrary to the potential
proposed action's purpose and objectives? If so, what are the
[[Page 72372]]
consequences and how should they be addressed?
(5) Please provide information on the costs and benefits of the
potential proposed action. This information will be used to support any
regulatory analysis performed by the NRC.
VIII. Plain Writing
The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal
agencies to write documents in a clear, concise, and well-organized
manner. The NRC has written this document to be consistent with the
Plain Writing Act as well as the Presidential Memorandum, ``Plain
Language in Government Writing,'' published June 10, 1998 (63 FR
31883). The NRC requests comment on this document with respect to the
clarity and effectiveness of the language used.
IX. Availability of Documents
The documents identified in the following table are available to
interested persons through one or more of the following methods, as
indicated.
------------------------------------------------------------------------
ADAMS Accession
No./ Federal
Date Document Register
citation
------------------------------------------------------------------------
May 10, 1993.................. SECY-93-127, ML12257A628.
``Financial
Protection Required
of Licensees of Large
Nuclear Power Plants
during
Decommissioning''.
July 20, 1995................. Proposed Rule: 60 FR 37374.
Decommissioning of
Nuclear Power
Reactors.
July 29, 1996................. Final Rule: 61 FR 39278.
Decommissioning of
Nuclear Power
Reactors.
December 17, 1996............. SECY-96-256, ``Changes ML15062A483.
to Financial
Protection
Requirements for
Permanently Shutdown
Nuclear Power
Reactors, 10 CFR
50.54(w)(1) and
140.11''.
June 30, 1998................. SRM to SECY-98-075, ML003752383.
``DSI-24
Implementation: Risk-
Informed, Performance-
Based Concepts
Applied to
Decommissioning''.
November 4, 1998.............. SECY-98-258, ``DSI-24 ML992870144.
Implementation:
Decommissioning
Licensing Actions and
Priorities and
Milestones for
Addressing Rulemaking
and Guidance
Development''.
February 24, 1999............. SRM to SECY-98-258.... ML003753861.
June 30, 1999................. SECY-99-168, ML992800087.
``Improving
Decommissioning
Regulations for
Nuclear Power
Plants''.
December 21, 1999............. SRM to SECY-99-168.... ML003752190.
June 28, 2000................. SECY-00-0145, ML003721626.
``Integrated
Rulemaking Plan for
Nuclear Power Plant
Decommissioning''.
September 27, 2000............ SRM to SECY-00-0145... ML003754381.
February 2001................. NUREG-1738, ML010430066.
``Technical Study of
Spent Fuel Pool
Accident Risk at
Decommissioning
Nuclear Power
Plants''.
June 4, 2001.................. SECY-01-0100, ``Policy ML011450420.
Issues Related to
Safeguards,
Insurance, and
Emergency
Preparedness
Regulations at
Decommissioning
Nuclear Power Plants
Storing Fuel in Spent
Fuel Pools''.
August 16, 2002............... Memorandum to the ML030550706.
Commission: Status of
Regulatory Exemptions
for Decommissioning
Plants.
September 18, 2002............ SECY-02-0169, ``Annual ML022120432.
Update Status of
Decommissioning
Program''.
February 4, 2010.............. Memorandum to the ML092990438.
Commission,
``Documentation of
Evolution of Security
Requirements at
Commercial Nuclear
Power Plants with
Respect to Mitigation
Measures for Large
Fires and
Explosions''.
December 2006................. NEI-06-12, ``B.5.b. ML070090060.
Phase 2 & 3 Submittal
Guideline, Revision
2''.
December 22, 2006............. Response to December Non-publicly
14, 2006 request to available.
endorse NEI 06-12,
``B.5.b Phase 2& 3
Submittal Guideline''.
August 8, 2008................ The Attorney General 73 FR 46204.
of Commonwealth of
Massachusetts, the
Attorney General of
California; Denial of
Petitions for
Rulemaking.
November 12, 2013............. COMSECY-13-0030, ML13329A918.
``Staff Evaluation
and Recommendation
for Japan Lessons-
Learned Tier 3 Issue
on Expedited Transfer
of Fuel''.
September 2014................ NUREG-2161, ML14255A365.
``Consequence Study
of a Beyond-Design-
Basis Earthquake
Affecting the Spent
Fuel Pool for a U.S.
Mark I Boiling Water
Reactor''.
November 14, 2014............. IN-2014-14, ML14218A493.
``Potential Safety
Enhancements to Spent
Fuel Storage''.
December 30, 2014............. SRM to SECY-14-0118, ML14364A111.
``Request by Duke
Energy Florida, Inc.,
for Exemptions from
Certain Emergency
Planning
Requirements''.
January 30, 2015.............. SECY-15-0014, ML15082A089.
``Anticipated
Schedule and
Estimated Resources
for a Power Reactor
Decommissioning
Rulemaking''.
December 23, 2013............. NSIR/DPR-ISG-02, ML13304B442.
``Emergency Planning
Exemption Requests
for Decommissioning
Nuclear Power
Plants''.
November 25, 2014............. NSIR/DSP-ISG-03, ML14294A170.
``Review of Security
Exemptions/License
Amendment Requests
for Decommissioning
Nuclear Power
Plants''.
November 10, 2011............. Letter Endorsing NEI ML112800379.
03-12, Revision 7.
March 2009.................... RG 5.77, ``Insider Non-publicly
Mitigation Program''. available.
March 31, 2008................ Final Rule: ``Fitness 73 FR 16966.
for Duty Programs''.
March 12, 2012................ Order EA-12-051, ML12054A679.
``Issuance of Order
to Modify Licenses
with Regard to
Reliable Spent Fuel
Pool
Instrumentation''.
March 12, 2012................ Order EA-12-049, ML12054A734.
``Issuance of Order
to Modify Licenses
with Regard to
Requirements for
Mitigation Strategies
for Beyond-Design-
Basis External
Events''.
[[Page 72373]]
October 7, 2015............... SECY-15-0127, Non-publicly
``Schedule, Resource available.
Estimates, and
Impacts for the Power
Reactor
Decommissioning
Rulemaking''.
------------------------------------------------------------------------
The NRC may post additional materials to the Federal rulemaking Web
site at www.regulations.gov, under Docket NRC-2015-0070. The Federal
rulemaking Web site allows you to receive alerts when changes or
additions occur in a docket folder. To subscribe: (1) Navigate to the
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Alerts'' link; and (3) enter your email address and select how
frequently you would like to receive emails (daily, weekly, or
monthly).
X. Rulemaking Process
The NRC does not intend to provide detailed comment responses for
information provided in response to this ANPR. The NRC will consider
comments on this ANPR in the rule development process. If the NRC
develops a regulatory basis sufficient to support a proposed rule,
there will be an opportunity for additional public comment when the
draft regulatory basis and the proposed rule are published. If
supporting guidance is developed for the proposed rule, stakeholders
will have an opportunity to provide feedback on the guidance as well.
Alternatively, if the regulatory basis does not provide sufficient
support for a proposed rule, the NRC will publish a Federal Register
notice withdrawing this ANPR and summarizing the public comments
received on this ANPR.
Dated at Rockville, Maryland, this 6th day of November 2015.
For the U.S. Nuclear Regulatory Commission.
Frederick D. Brown,
Acting Executive Director for Operations.
[FR Doc. 2015-29536 Filed 11-18-15; 8:45 am]
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