[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
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 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

[[Page 72360]]

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

[[Page 72361]]

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 
docket folder [NRC-2015Y-0070]; (2) click the ``Sign up for Email 
Alerts'' link; and (3) enter your email address and select how 
frequently you would like to receive emails (daily, weekly, or 
monthly).

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]
 BILLING CODE 7590-01-P