Fuel-Cladding Issues in Postulated Spent Fuel Pool Accidents, 29761-29765 [2016-11212]
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Rules and Regulations
Federal Register
Vol. 81, No. 93
Friday, May 13, 2016
This section of the FEDERAL REGISTER
contains regulatory documents having general
applicability and legal effect, most of which
are keyed to and codified in the Code of
Federal Regulations, which is published under
50 titles pursuant to 44 U.S.C. 1510.
The Code of Federal Regulations is sold by
the Superintendent of Documents. Prices of
new books are listed in the first FEDERAL
REGISTER issue of each week.
NUCLEAR REGULATORY
COMMISSION
10 CFR Part 50
[Docket Nos. PRM–50–108; NRC–2014–
0171]
Fuel-Cladding Issues in Postulated
Spent Fuel Pool Accidents
Nuclear Regulatory
Commission.
ACTION: Petition for rulemaking; denial.
AGENCY:
The U.S. Nuclear Regulatory
Commission (NRC) is denying a petition
for rulemaking (PRM or the petition),
PRM–50–108, submitted by Mr. Mark
Edward Leyse (the petitioner). The
petitioner requested that the NRC
require power reactor licensees to
perform evaluations to determine the
potential consequences of various
postulated spent fuel pool (SFP)
accident scenarios. The evaluations
would be required to be submitted to
the NRC for informational purposes.
The NRC is denying the petition
because the NRC does not believe the
information is needed for effective NRC
regulatory decisionmaking with respect
to SFPs or for public safety,
environmental protection, or common
defense and security.
DATES: The docket for the petition,
PRM–50–108, is closed on May 13,
2016.
SUMMARY:
Please refer to Docket ID
NRC–2014–0171 when contacting the
NRC about the availability of
information for this petition. You may
obtain publicly-available information
related to this petition by any of the
following methods:
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2014–0171. Address
questions about NRC dockets to Carol
Gallagher; telephone: 301–415–3463;
email: Carol.Gallagher@nrc.gov. For
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ADDRESSES:
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technical questions, contact the
individual listed in the FOR FURTHER
INFORMATION CONTACT section of this
document.
• The NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may obtain publiclyavailable documents online in the
ADAMS Public Document collection at
https://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 pdr.resource@nrc.gov. 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
IV, ‘‘Availability of Documents,’’ of this
document.
• The NRC’s PDR: You may examine
and purchase copies of public
documents at the NRC’s PDR, O1–F21,
One White Flint North, 11555 Rockville
Pike, Rockville, Maryland 20852.
FOR FURTHER INFORMATION CONTACT:
Daniel Doyle, Office of Nuclear Reactor
Regulation; U.S. Nuclear Regulatory
Commission, Washington, DC 20555–
0001; telephone: 301–415–3748; email:
Daniel.Doyle@nrc.gov.
SUPPLEMENTARY INFORMATION:
Table of Contents
I. The Petition
II. Reasons for Denial
III. Conclusion
IV. Availability of Documents
I. The Petition
Section 2.802 of title 10 of the Code
of Federal Regulations (10 CFR),
‘‘Petition for rulemaking—requirements
for filing,’’ provides an opportunity for
any interested person to petition the
Commission to issue, amend, or rescind
any regulation. The NRC received a
petition dated June 19, 2014, from Mr.
Mark Edward Leyse and assigned it
Docket No. PRM–50–108 (ADAMS
Accession No. ML14195A388). The NRC
published a notice of docketing in the
Federal Register (FR) on October 7,
2014 (79 FR 60383). The NRC did not
request public comment on the petition
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because sufficient information was
available for the NRC staff to form a
technical opinion regarding the merits
of the petition.
The petitioner requested that the NRC
develop new regulations requiring that:
(1) SFP accident evaluation models use
data from multi-rod bundle (assembly)
severe accident experiments for
calculating the rates of energy release,
hydrogen generation, and fuel cladding
oxidation from the zirconium-steam
reaction; (2) SFP accident evaluation
models use data from multi-rod bundle
(assembly) severe accident experiments
conducted with pre-oxidized fuel
cladding for calculating the rates of
energy release (from both fuel cladding
oxidation and fuel cladding nitriding),
fuel cladding oxidation, and fuel
cladding nitriding from the zirconiumair reaction; (3) SFP accident evaluation
models be required to conservatively
model nitrogen-induced breakaway
oxidation behavior; and (4) licensees be
required to use conservative SFP
accident evaluation models to perform
annual SFP safety evaluations of:
postulated complete loss-of-coolant
accident (LOCA) scenarios, postulated
partial LOCA scenarios, and postulated
boil-off accident scenarios.
The petitioner referenced recent NRC
post-Fukushima MELCOR simulations
of boiling-water reactor Mark I SFP
accident/fire scenarios. The petitioner
stated that the conclusions from the
NRC’s MELCOR simulations are nonconservative and misleading because
their conclusions underestimate the
probabilities of large radiological
releases from SFP accidents.
The petitioner asserted that in actual
SFP fires, there would be quicker fuelcladding temperature escalations,
releasing more heat, and quicker axial
and radial propagation of zirconium (Zr)
fires than MELCOR simulations predict.
The petitioner stated that the NRC’s
philosophy of defense-in-depth requires
the application of conservative models,
and, therefore, it is necessary to improve
the performance of MELCOR and any
other computer safety models that are
intended to accurately simulate SFP
accident/fire scenarios.
The petitioner stated that the new
regulations would help improve public
and plant-worker safety. The petitioner
asserted that the first three requested
regulations, regarding zirconium fuel
cladding oxidation and nitriding, as
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well as nitrogen-induced breakaway
oxidation behavior, are intended to
improve the performance of computer
safety models that simulate postulated
SFP accident/fire scenarios. The
petitioner stated that the fourth
requested regulation would require that
licensees use conservative SFP accident
evaluation models to perform annual
SFP safety evaluations of postulated
complete LOCA scenarios, postulated
partial LOCA scenarios, and postulated
boil-off accident scenarios. The
petitioner stated that the purpose of
these evaluations would be to keep the
NRC informed of the potential
consequences of postulated SFP
accident/fire scenarios as fuel assembles
were added, removed, or reconfigured
in licensees’ SFPs. The petitioner stated
that the requested regulations are
needed because the probability of the
type of events that could lead to SFP
accidents is relatively high.
The NRC staff reviewed the petition
and, based on its understanding of the
overall argument in the petition,
identified and evaluated the following
three issues:
• Issue 1: The requested regulations
pertaining to SFP accident evaluation
models are needed because the
probability of the type of events that
could lead to SFP accidents is relatively
high.
• Issue 2: Annual licensee SFP safety
evaluations and submission of results to
the NRC is necessary so that the NRC is
aware of potential consequences of
postulated SFP accident/fire scenarios
as fuel assemblies are added, removed,
or reconfigured in licensees’ SFPs.
• Issue 3: MELCOR is not currently
sufficient to provide a conservative
evaluation of postulated SFP accident/
fire scenarios for use in the PRMproposed annual SFP evaluations.
Detailed NRC responses to the three
issues are provided in Section II,
‘‘Reasons for Denial,’’ of this document.
II. Reasons for Denial
The NRC is denying the petition
because the petitioner failed to present
any significant information or
arguments that would warrant the
requested regulations. The first three
requested regulations would establish
requirements for how the detailed
annual evaluations that would be
required by the fourth requested
regulation would be performed. It is not
necessary to require detailed annual
evaluations of the progression of SFP
severe accidents because the risk of an
SFP severe accident is low. The NRC
defines risk as the product of the
probability and the consequences of an
accident. The requested annual
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evaluations are not needed for
regulatory decisionmaking, and the
evaluations would not prevent or
mitigate an SFP accident. The petitioner
described multiple ways that an
extended loss of offsite electrical power
could occur and how this could lead to
an SFP fire. In order for an SFP fire to
occur, all SFP systems, backup systems,
and operator actions that are intended to
prevent the spent fuel in the pool from
being uncovered would have to fail. The
NRC does not agree that more detailed
accident evaluation models need to be
developed for this purpose, as requested
by the petitioner, because the requested
annual evaluations are not needed for
regulatory decisionmaking. The NRC
recognizes that the consequences of an
SFP fire could be large and that is why
there are numerous requirements in
place to prevent a situation where the
spent fuel is uncovered.
This section provides detailed NRC
responses to the three issues identified
in the petition.
Issue 1: The Requested Regulations
Pertaining to SFP Accident Evaluation
Models Are Needed Because the
Probability of the Type of Events That
Could Lead to SFP Accidents Is
Relatively High
The petitioner stated that the
requested regulations pertaining to SFP
accident evaluation models are needed
because the probability of the type of
events that could lead to SFP accidents
is relatively high. The petitioner stated
that an SFP accident could happen as a
result of a leak (rapid drain down) or
boil-off scenario. Furthermore, the
petitioner notes that in the event of a
long-term station blackout, emergency
diesel generators could run out of fuel
and SFP cooling would be lost, resulting
in a boil-off of SFP water inventory and
a subsequent release of radioactive
materials from the spent fuel. The
petitioner also provided several
examples of events that could lead to a
long-term station blackout and,
ultimately, an SFP accident, such as a
strong geomagnetic disturbance, a
nuclear device detonated in the earth’s
atmosphere, a pandemic, or a cyber or
physical attack.
NRC Response
Spent nuclear fuel offloaded from a
reactor is initially stored in an SFP. The
SFPs at all nuclear plants in the United
States are robust structures constructed
with thick, reinforced, concrete walls
and welded stainless-steel liners. They
are designed to safely contain the spent
fuel discharged from a nuclear reactor
under a variety of normal, off-normal,
and hypothetical accident conditions
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(e.g., loss of electrical power, loss of
cooling, fuel or cask drop incidents,
floods, earthquakes, or extreme weather
events). Racks fitted in the SFPs store
the fuel assemblies in a controlled
configuration so that the fuel is
maintained in a sub-critical and
coolable geometry. Redundant
monitoring, cooling, and water makeup
systems are provided. The spent fuel
assemblies are typically covered by at
least 25-feet of water, which provides
passive cooling as well as radiation
shielding. Penetrations to pools are
limited to prevent inadvertent drainage,
and the penetrations are generally
located well above spent fuel storage
elevations to prevent uncovering of fuel
from drainage.
Studies conducted over the last four
decades have consistently shown the
risk of an accident causing a zirconium
fire in an SFP to be low. The risk of an
SFP accident was examined in the
1980s as Generic Issue 82, ‘‘Beyond
Design Basis Accidents in Spent Fuel
Pools,’’ in light of increased use of highdensity storage racks and laboratory
studies that indicated the possibility of
zirconium fire propagation between
assemblies in an air-cooled environment
(Section 3 of NUREG–0933, ‘‘Resolution
of Generic Safety Issues,’’ https://
nureg.nrc.gov/sr0933/). The risk
assessment and cost-benefit analyses
developed through this effort, Section
6.2 of NUREG–1353, ‘‘Regulatory
Analysis for the Resolution of Generic
Issue 82, Beyond Design Basis
Accidents in Spent Fuel Pools’’
(ADAMS Accession No. ML082330232),
concluded that the risk of a severe
accident in the SFP was low and
appeared to meet the objectives of the
Commission’s Safety Goal Policy
Statement public health objectives (51
FR 30028; August 21, 1986) and that no
new regulatory requirements were
warranted.
The risk of an SFP accident was reassessed in the late 1990s to support a
risk-informed rulemaking for
permanently shutdown, or
decommissioned, nuclear power plants
in the United States. The study,
NUREG–1738, ‘‘Technical Study of
Spent Fuel Pool Accident Risk at
Decommissioning Nuclear Power
Plants’’ (ADAMS Accession No.
ML010430066), conservatively assumed
that if the water level in the SFP
dropped below the top of the spent fuel,
an SFP zirconium fire involving all of
the spent fuel would occur, and thereby
bounded those conditions associated
with air cooling of the fuel (including
partial-drain down scenarios) and fire
propagation. Even with this
conservative assumption, the study
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found the risk of an SFP fire to be low
and well within the Commission’s
Safety Goals.
Additional mechanisms to mitigate
the potential loss of SFP water
inventory were implemented following
the terrorist attacks of September 11,
2001, which have enhanced spent fuel
coolability and the potential to recover
SFP water level and cooling prior to a
potential SFP zirconium fire (73 FR
76204; August 8, 2008). Based on the
implementation of these additional
strategies, the probability and,
accordingly, the risk of an SFP
zirconium fire initiation has decreased
and is expected to be less than
previously analyzed in NUREG–1738
and previous studies.
Following the 2011 accident at
Fukushima Dai-ichi, the NRC took
extensive actions to ensure that portable
equipment is available to mitigate a loss
of cooling water in the SFP. 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). This order required
licensees to develop, implement, and
maintain guidance and strategies to
maintain or restore core cooling,
containment, and SFP cooling
capabilities following a beyond-designbasis external event. The NRC endorsed
the Nuclear Energy Institute (NEI)
guidance to meet the requirements of
this order.1 That guidance establishes
additional mechanisms for mitigating a
loss of SFP cooling water beyond the
requirements in 10 CFR 50.54(hh)(2),
such as installing a remote connection
for SFP makeup water that can be
accessed away from the SFP refueling
floor.
Also, in 2014, the NRC documented a
regulatory analysis in COMSECY–13–
0030, ‘‘Staff Evaluation and
Recommendation for Japan Lessons
Learned Tier 3 Issue on Expedited
Transfer of Spent Fuel’’ (ADAMS
Accession No. ML13329A918), which
considered a broad history of the NRC’s
oversight of spent fuel storage, SFP
operating experience (domestic and
international), as well as information
compiled in NUREG–2161,
‘‘Consequence Study of a BeyondDesign-Basis Earthquake Affecting the
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1 See
NEI 12–06, ‘‘Diverse and Flexible Coping
Strategies (FLEX) Implementation Guide,’’ dated
August 2012 (ADAMS Accession No.
ML12242A378), and JLD–ISG–2012–01,
‘‘Compliance with Order EA–12–049, Order
Modifying Licenses with Regard to Requirements
for Mitigation Strategies for Beyond-Design-Basis
External Events,’’ dated August 2012 (ADAMS
Accession No. ML12229A174).
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Spent Fuel Pool for a U.S. Mark I
Boiling Water Reactor’’ (ADAMS
Accession No. ML14255A365). In
COMSECY–13–0030, the NRC staff
concluded that SFPs are robust
structures with large safety margins and
recommended to the Commission that
assessments of possible regulatory
actions to require the expedited transfer
of spent fuel from SFPs to dry cask
storage were not warranted. The
Commission subsequently approved the
staff’s recommendation in the Staff
Requirements Memorandum to
COMSECY–13–0030 (ADAMS
Accession No. ML14143A360).
As supported by numerous
evaluations referenced in this
document, the NRC has determined that
the risk of an SFP severe accident is
low. While the risk of a severe accident
in an SFP is not negligible, the NRC
believes that the risk is low because of
the conservative design of SFPs;
operational criteria to control spent fuel
movement, monitor pertinent
parameters, and maintain cooling
capability; mitigation measures in place
if there is loss of cooling capability or
water; and emergency preparedness
measures to protect the public. The
information proposed to be provided to
the NRC is not needed for the
effectiveness of NRC’s approach for
ensuring SFP safety. The NRC notes that
the issue of long-term cooling of SFPs is
the subject of PRM–50–96, which was
accepted for consideration in the
rulemaking process (77 FR 74788;
December 18, 2012) and is being
addressed by the NRC’s rulemaking
regarding mitigation of beyond designbasis events (RIN 3150–AJ49; NRC–
2014–0240).
Issue 2: Annual Licensee SFP Safety
Evaluations and Submission of Results
to the NRC Is Necessary So That the
NRC Is Aware of Potential
Consequences of Postulated SFP
Accident/Fire Scenarios as Fuel
Assemblies Are Added, Removed, or
Reconfigured in Licensees’ SFPs
The petitioner stated that the purpose
of the proposed requirement is to keep
the NRC informed of the potential
consequences of postulated SFP
accident/fire scenarios as fuel
assemblies are added, removed, or
reconfigured in licensees’ SFPs.
NRC Response
The NRC does not agree that this is
necessary because the NRC already
evaluates SFP systems and structures
during initial licensing and license
amendment reviews. In addition,
baseline NRC inspections provide
ongoing oversight to ensure adequate
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protection. There are not sufficient
benefits that would justify the new
requirement proposed in the petition for
SFP accident evaluations. The proposed
new requirement for licensees to
perform SFP evaluations would not
prevent or mitigate an SFP accident or
provide information that is necessary for
regulatory decisionmaking. The annual
licensee SFP safety evaluations and
their results proposed to be provided to
the NRC are not needed for the
effectiveness of the NRC’s approach to
ensuring SFP safety.
The NRC issues licenses after
reviewing and approving the design and
licensing bases contained in the plant’s
safety analysis report. Licensees are
required to operate the plant, including
performing operations and surveillances
related to spent fuel, in accordance with
technical specifications and established
practices and procedures for that plant.
Any licensee changes to design,
operational or surveillance practices, or
approved spent fuel inventory limits or
configuration changes must be
evaluated using the criteria in 10 CFR
50.59, documented and retained for the
duration of the operating license, and, if
warranted, submitted to the NRC for
prior approval.
The general design criteria (GDC) in
appendix A to 10 CFR part 50 establish
general expectations that licensees must
meet through compliance with their
plant-specific licensing basis. Several
GDC apply to SFPs:
• Protecting against natural
phenomena and equipment failures
(GDC 2 and GDC 4);
• Preventing a substantial loss-ofcoolant inventory under accident
conditions (e.g., equipment failure or
loss of decay and residual heat removal)
(GDC 61);
• Preventing criticality of the spent
fuel (GDC 62); and
• Adequately monitoring the SFP
conditions for loss of decay heat
removal and radiation (GDC 63).
Additionally, emergency procedures
and mitigating strategies are in place to
address unexpected challenges to spent
fuel safety. Multiple requirements in 10
CFR part 50, as well as recent NRC
orders following the Fukushima Dai-ichi
accident, require redundant equipment
and strategies to address loss of cooling
to SFPs and protective actions for plant
personnel and the public to limit
exposure to radioactive materials.
The NRC provides oversight of the
licensee’s overall plant operations and
the SFP in several ways. The NRC
inspectors ensure that spent fuel is
stored safely by regularly inspecting
reactor and equipment vendors;
inspecting the design, construction, and
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use of equipment; and observing ‘‘dry
runs’’ of procedures. At least two NRC
resident inspectors are assigned to each
site to provide monitoring and
inspection of routine and special
activities. They are aware of, and
routinely observe, SFP activities
involving fuel manipulation. The NRC
inspectors use inspection procedures to
guide periodic inspection activities, and
the results are published in publiclyavailable inspection reports. Special
inspections may be conducted, as
necessary, to evaluate root causes and
licensee corrective actions if sitespecific events occur. Special
inspections may also evaluate generic
actions taken by some or all licensees as
a result of an NRC order or a change in
regulations.
In accordance with 10 CFR part 21,
the NRC is informed of defects and
noncompliances associated with basic
components, which include SFPs and
associated drain pipes and safety-related
systems, structures, and components for
makeup water. This information allows
the NRC to take additional regulatory
action as necessary with respect to
defects and noncompliances. The NRC
is also informed of events and
conditions at nuclear power plants, as
set forth in §§ 50.72 and 50.73.
Depending upon the nature of the event
or condition, a nuclear power plant
licensee must inform the NRC within a
specified period of time of the licensee’s
corrective action taken or planned to be
taken. These reports also facilitate
effective and timely NRC regulatory
oversight. Finally, information
identified by a nuclear power plant
applicant or licensee as having a
significant implication for public health
and safety or common defense and
security must be reported to the NRC
within 2 days of the applicant’s or
licensee’s identification of the
information.
The annual evaluations requested in
the petition would not provide
information that is necessary for
regulatory decisionmaking. The
evaluations requested in the petition
would postulate scenarios in which the
normal cooling systems, the backup
cooling methods, and the mitigation
strategies have all failed to cool the
stored fuel and would require the
calculation of the time it would take for
the stored fuel to ignite and how much
of it would ignite. Due to the robustness
of this equipment, the NRC views this
sequence of events as extremely
unlikely to occur. Since the current
regulations require that the pool be
designed to prevent the loss-of-coolant
and subsequent uncovering of the fuel,
the information that would be obtained
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from the proposed requirement in the
petition would not impact the current
design basis. Moreover, as discussed
previously, the NRC’s current regulatory
infrastructure relevant to SFPs at
nuclear power plants in the United
States already contains information
collection and reporting requirements
that support effective NRC regulatory
oversight of SFPs.
The NRC does not agree that it is
necessary to impose a new requirement
for licensees to perform annual
evaluations of their SFPs because
existing requirements and oversight are
sufficient to ensure adequate protection
of public health and safety.
Issue 3: MELCOR Is Not Currently
Sufficient To Provide a Conservative
Evaluation of Postulated SFP Accident/
Fire Scenarios
The petitioner requested that the NRC
establish requirements for SFP accident
evaluation computer models to be used
in the annual SFP evaluations requested
in Issue 2. The petitioner stated that
there are serious flaws with MELCOR,
which has been used by the NRC to
model severe accident progression in
SFPs, and, therefore, MELCOR is not
sufficient.
NRC Response
The NRC does not agree that it is
necessary to establish requirements for
SFP accident evaluation computer
models because the annual SFP
evaluations requested in Issue 2 are not
necessary for regulatory
decisionmaking. Therefore, it is not
necessary for the NRC to establish
requirements for how such an
evaluation should be conducted.
Furthermore, the NRC disagrees with
the petitioner’s statements that
MELCOR is flawed.
There are inherent uncertainties in
the progression of severe accidents.
There are many interrelated phenomena
that need to be properly understood;
otherwise, conservatism in one area may
lead to overall non-conservative results.
Conservatism can be meaningfully
introduced into the relevant analysis
after the best estimate analysis is done
and uncertainties are properly taken
into account.
The important question for a severe
accident analysis is whether the
uncertainties are appropriately
considered in the analysis results. For
example, Section 9 of the SFP study
(NUREG–2161) is devoted to discussing
the major uncertainties that can affect
the radiological releases (e.g., hydrogen
combustion, core concrete interaction,
multi-unit or concurrent accident, or
fuel loading). In addition, the regulatory
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analysis in COMSECY–13–0030 only
relied on SFP study insights for the
boiling-water reactors with Mark I and
II containments, and, even then, the
results were conservatively biased
towards higher radiological releases. For
other designs, the release fractions were
based on previous studies (i.e., NUREG–
1738) that used bounding or
conservative estimates.
The MELCOR computer code is the
NRC’s best estimate tool for severe
accident analysis. It has been validated
against experimental data, and it
represents the current state of the art in
severe accident analysis. In NUREG–
2161, the NRC stated that ‘‘MELCOR has
been developed through the NRC and
international research performed since
the accident at Three Mile Island in
1979. MELCOR is a fully integrated,
engineering-level computer code and
includes a broad spectrum of severe
accident phenomena with capabilities to
model core heatup and degradation,
fission product release and transport
within the primary system and
containment, core relocation to the
vessel lower head, and ex-vessel core
concrete interaction.’’ Furthermore,
MELCOR has been benchmarked against
many experiments, including separate
and integral effects tests for a wide
range of phenomena. Therefore, the
NRC has determined that MELCOR is
acceptable for its intended use.
Additional information about the
capabilities of the MELCOR code to
model SFP accidents can be found in
the NRC response to stakeholder
comments in Appendix E to NUREG–
2161. The NRC also addressed questions
regarding MELCOR in Appendix D to
NUREG–2157, Volume 2, ‘‘Generic
Environmental Impact Statement for
Continued Storage of Spent Nuclear
Fuel’’ (ADAMS Accession No.
ML14196A107).
III. Conclusion
For the reasons described in Section
II, ‘‘Reasons for Denial,’’ of this
document, the NRC is denying the
petition under 10 CFR 2.803. The
petitioner failed to present any
information or arguments that would
warrant the requested amendments. The
NRC does not believe that the
information that would be reported to
the NRC as requested by the petitioner
is necessary for effective NRC regulatory
decisionmaking with respect to SFPs.
The NRC continues to conclude that the
current design and licensing
requirements for SFPs provide adequate
protection of public health and safety.
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IV. Availability of Documents
interested persons as indicated. For
more information on accessing ADAMS,
The documents identified in the
following table are available to
see the ADDRESSES section of this
document.
ADAMS accession number/
Federal Register citation
Date
Document
August 21, 1986 .................................
Safety Goals for the Operations of Nuclear Power Plants; Policy Statement; Republication.
NUREG–1353, ‘‘Regulatory Analysis for the Resolution of Generic Issue
82, Beyond Design Basis Accidents in Spent Fuel Pools’’.
NUREG–1738, ‘‘Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants’’.
EA–12–049, ‘‘Order Modifying Licenses with Regard to Requirements for
Mitigation Strategies for Beyond-Design-Basis External Events’’.
NEI 12–06, ‘‘Diverse and Flexible Coping Strategies (FLEX) Implementation Guide’’.
JLD–ISG–2012–01, ‘‘Compliance with Order EA–12–049, Order Modifying
Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events’’.
Long-Term Cooling and Unattended Water Makeup of Spent Fuel Pools ...
COMSECY–13–0030, ‘‘Staff Evaluation and Recommendation for Japan
Lessons Learned Tier 3 Issue on Expedited Transfer of Spent Fuel’’.
SRM–COMSECY–13–0030, ‘‘Staff Requirements—COMSECY–13–0030—
Staff Evaluation and Recommendation for Japan Lessons-Learned Tier
3 Issue on Expedited Transfer of Spent Fuel’’.
Incoming Petition (PRM–50–108) from Mr. Mark Edward Leyse .................
NUREG–2157, ‘‘Generic Environmental Impact Statement for Continued
Storage of Spent Nuclear Fuel,’’ Volume 2.
NUREG–2161, ‘‘Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling-Water Reactor’’.
Notice of Docketing for PRM–50–108 ...........................................................
April 1989 ...........................................
February 2001 ....................................
March 12, 2012 ..................................
August 2012 .......................................
August 2012 .......................................
December 18, 2012 ............................
November 12, 2013 ............................
May 23, 2014 .....................................
June 19, 2014 ....................................
September 2014 .................................
September 2014 .................................
October 7, 2014 .................................
Dated at Rockville, Maryland, this 5th day
of May, 2016.
For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 2016–11212 Filed 5–12–16; 8:45 am]
BILLING CODE 7590–01–P
PENSION BENEFIT GUARANTY
CORPORATION
29 CFR Parts 4010, 4041, 4071, and
4302
RIN 1212–AB33
Adjustment of Civil Penalties
Pension Benefit Guaranty
Corporation.
ACTION: Interim final rule.
AGENCY:
The Pension Benefit Guaranty
Corporation is amending its regulations
to adjust the penalties provided for in
sections 4071 and 4302 of the Employee
Retirement Income Security Act of 1974.
This action is being taken in accordance
with the Federal Civil Penalties
Inflation Adjustment Act Improvements
Act of 2015 and Office of Management
and Budget memorandum M–16–06.
The regulations being amended are
those on Penalties for Failure to Provide
Certain Notices or Other Material
Information (29 CFR part 4071) and
ehiers on DSK5VPTVN1PROD with RULES
SUMMARY:
VerDate Sep<11>2014
16:59 May 12, 2016
Jkt 238001
Penalties for Failure to Provide Certain
Multiemployer Plan Notices (29 CFR
part 4302). Conforming amendments are
also being made to the regulations on
Annual Financial and Actuarial
Information Reporting (29 CFR part
4010) and Termination of SingleEmployer Plans (29 CFR part 4041).
DATES: The amendments are effective
August 1, 2016. Also see Applicability,
below.
FOR FURTHER INFORMATION CONTACT:
Deborah C. Murphy, Deputy Assistant
General Counsel for Regulatory Affairs
(murphy.deborah@pbgc.gov), Office of
the General Counsel, Pension Benefit
Guaranty Corporation, 1200 K Street
NW., Washington, DC 20005–4026; 202–
326–4400 extension 3451. (TTY and
TDD users may call the Federal relay
service toll-free at 800–877–8339 and
ask to be connected to 202–326–4400
extension 3451.)
SUPPLEMENTARY INFORMATION:
Executive Summary
Purpose of the Regulatory Action
This rule is needed to carry out the
requirements of the Federal Civil
Penalties Inflation Adjustment Act
Improvements Act of 2015. The rule
adjusts the maximum civil penalties
that PBGC may assess for failure to
provide certain notices or other material
information.
PO 00000
Frm 00005
Fmt 4700
Sfmt 4700
29765
51 FR 30028.
ML082330232.
ML010430066.
ML12054A735.
ML12242A378.
ML12229A174.
77 FR 74788.
ML13329A918.
ML14143A360.
ML14195A388.
ML14196A107.
ML14255A365.
79 FR 60383.
PBGC’s legal authority for this action
comes from the Federal Civil Penalties
Inflation Adjustment Act of 1990 as
amended by the Federal Civil Penalties
Inflation Adjustment Act Improvements
Act of 2015 and from sections
4002(b)(3), 4071, and 4302 of the
Employee Retirement Income Security
Act of 1974.
Major Provisions of the Regulatory
Action
This rule adjusts the maximum civil
penalties that PBGC may assess under
sections 4071 and 4302 of ERISA. The
new maximum amounts are $2,063 for
section 4071 penalties and $275 for
section 4302 penalties.
Background
The Pension Benefit Guaranty
Corporation (PBGC) administers title IV
of the Employee Retirement Income
Security Act of 1974 (ERISA). Title IV
has two provisions that authorize PBGC
to assess civil monetary penalties.1
Section 4302, added to ERISA by the
Multiemployer Pension Plan
1 Under the Federal Civil Penalties Inflation
Adjustment Act of 1990, a penalty is a civil
monetary penalty if (among other things) it is for
a specific monetary amount or has a maximum
amount specified by Federal law. Title IV also
provides (in section 4007) for penalties for late
payment of premiums, but those penalties are
neither in a specified amount nor subject to a
specified maximum amount.
E:\FR\FM\13MYR1.SGM
13MYR1
]]>Agencies
[Federal Register Volume 81, Number 93 (Friday, May 13, 2016)]
[Rules and Regulations]
[Pages 29761-29765]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-11212]
========================================================================
Rules and Regulations
Federal Register
________________________________________________________________________
This section of the FEDERAL REGISTER contains regulatory documents
having general applicability and legal effect, most of which are keyed
to and codified in the Code of Federal Regulations, which is published
under 50 titles pursuant to 44 U.S.C. 1510.
The Code of Federal Regulations is sold by the Superintendent of Documents.
Prices of new books are listed in the first FEDERAL REGISTER issue of each
week.
========================================================================
Federal Register / Vol. 81, No. 93 / Friday, May 13, 2016 / Rules and
Regulations
[[Page 29761]]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
[Docket Nos. PRM-50-108; NRC-2014-0171]
Fuel-Cladding Issues in Postulated Spent Fuel Pool Accidents
AGENCY: Nuclear Regulatory Commission.
ACTION: Petition for rulemaking; denial.
-----------------------------------------------------------------------
SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is denying a
petition for rulemaking (PRM or the petition), PRM-50-108, submitted by
Mr. Mark Edward Leyse (the petitioner). The petitioner requested that
the NRC require power reactor licensees to perform evaluations to
determine the potential consequences of various postulated spent fuel
pool (SFP) accident scenarios. The evaluations would be required to be
submitted to the NRC for informational purposes. The NRC is denying the
petition because the NRC does not believe the information is needed for
effective NRC regulatory decisionmaking with respect to SFPs or for
public safety, environmental protection, or common defense and
security.
DATES: The docket for the petition, PRM-50-108, is closed on May 13,
2016.
ADDRESSES: Please refer to Docket ID NRC-2014-0171 when contacting the
NRC about the availability of information for this petition. You may
obtain publicly-available information related to this petition by any
of the following methods:
Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2014-0171. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
The NRC's Agencywide Documents Access and Management
System (ADAMS): You may obtain publicly-available documents online in
the ADAMS Public Document collection at https://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 pdr.resource@nrc.gov. 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 IV, ``Availability of Documents,'' of this
document.
The NRC's PDR: You may examine and purchase copies of
public documents at the NRC's PDR, O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
FOR FURTHER INFORMATION CONTACT: Daniel Doyle, Office of Nuclear
Reactor Regulation; U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-3748; email: Daniel.Doyle@nrc.gov.
SUPPLEMENTARY INFORMATION:
Table of Contents
I. The Petition
II. Reasons for Denial
III. Conclusion
IV. Availability of Documents
I. The Petition
Section 2.802 of title 10 of the Code of Federal Regulations (10
CFR), ``Petition for rulemaking--requirements for filing,'' provides an
opportunity for any interested person to petition the Commission to
issue, amend, or rescind any regulation. The NRC received a petition
dated June 19, 2014, from Mr. Mark Edward Leyse and assigned it Docket
No. PRM-50-108 (ADAMS Accession No. ML14195A388). The NRC published a
notice of docketing in the Federal Register (FR) on October 7, 2014 (79
FR 60383). The NRC did not request public comment on the petition
because sufficient information was available for the NRC staff to form
a technical opinion regarding the merits of the petition.
The petitioner requested that the NRC develop new regulations
requiring that: (1) SFP accident evaluation models use data from multi-
rod bundle (assembly) severe accident experiments for calculating the
rates of energy release, hydrogen generation, and fuel cladding
oxidation from the zirconium-steam reaction; (2) SFP accident
evaluation models use data from multi-rod bundle (assembly) severe
accident experiments conducted with pre-oxidized fuel cladding for
calculating the rates of energy release (from both fuel cladding
oxidation and fuel cladding nitriding), fuel cladding oxidation, and
fuel cladding nitriding from the zirconium-air reaction; (3) SFP
accident evaluation models be required to conservatively model
nitrogen-induced breakaway oxidation behavior; and (4) licensees be
required to use conservative SFP accident evaluation models to perform
annual SFP safety evaluations of: postulated complete loss-of-coolant
accident (LOCA) scenarios, postulated partial LOCA scenarios, and
postulated boil-off accident scenarios.
The petitioner referenced recent NRC post-Fukushima MELCOR
simulations of boiling-water reactor Mark I SFP accident/fire
scenarios. The petitioner stated that the conclusions from the NRC's
MELCOR simulations are non-conservative and misleading because their
conclusions underestimate the probabilities of large radiological
releases from SFP accidents.
The petitioner asserted that in actual SFP fires, there would be
quicker fuel-cladding temperature escalations, releasing more heat, and
quicker axial and radial propagation of zirconium (Zr) fires than
MELCOR simulations predict. The petitioner stated that the NRC's
philosophy of defense-in-depth requires the application of conservative
models, and, therefore, it is necessary to improve the performance of
MELCOR and any other computer safety models that are intended to
accurately simulate SFP accident/fire scenarios.
The petitioner stated that the new regulations would help improve
public and plant-worker safety. The petitioner asserted that the first
three requested regulations, regarding zirconium fuel cladding
oxidation and nitriding, as
[[Page 29762]]
well as nitrogen-induced breakaway oxidation behavior, are intended to
improve the performance of computer safety models that simulate
postulated SFP accident/fire scenarios. The petitioner stated that the
fourth requested regulation would require that licensees use
conservative SFP accident evaluation models to perform annual SFP
safety evaluations of postulated complete LOCA scenarios, postulated
partial LOCA scenarios, and postulated boil-off accident scenarios. The
petitioner stated that the purpose of these evaluations would be to
keep the NRC informed of the potential consequences of postulated SFP
accident/fire scenarios as fuel assembles were added, removed, or
reconfigured in licensees' SFPs. The petitioner stated that the
requested regulations are needed because the probability of the type of
events that could lead to SFP accidents is relatively high.
The NRC staff reviewed the petition and, based on its understanding
of the overall argument in the petition, identified and evaluated the
following three issues:
Issue 1: The requested regulations pertaining to SFP
accident evaluation models are needed because the probability of the
type of events that could lead to SFP accidents is relatively high.
Issue 2: Annual licensee SFP safety evaluations and
submission of results to the NRC is necessary so that the NRC is aware
of potential consequences of postulated SFP accident/fire scenarios as
fuel assemblies are added, removed, or reconfigured in licensees' SFPs.
Issue 3: MELCOR is not currently sufficient to provide a
conservative evaluation of postulated SFP accident/fire scenarios for
use in the PRM-proposed annual SFP evaluations.
Detailed NRC responses to the three issues are provided in Section
II, ``Reasons for Denial,'' of this document.
II. Reasons for Denial
The NRC is denying the petition because the petitioner failed to
present any significant information or arguments that would warrant the
requested regulations. The first three requested regulations would
establish requirements for how the detailed annual evaluations that
would be required by the fourth requested regulation would be
performed. It is not necessary to require detailed annual evaluations
of the progression of SFP severe accidents because the risk of an SFP
severe accident is low. The NRC defines risk as the product of the
probability and the consequences of an accident. The requested annual
evaluations are not needed for regulatory decisionmaking, and the
evaluations would not prevent or mitigate an SFP accident. The
petitioner described multiple ways that an extended loss of offsite
electrical power could occur and how this could lead to an SFP fire. In
order for an SFP fire to occur, all SFP systems, backup systems, and
operator actions that are intended to prevent the spent fuel in the
pool from being uncovered would have to fail. The NRC does not agree
that more detailed accident evaluation models need to be developed for
this purpose, as requested by the petitioner, because the requested
annual evaluations are not needed for regulatory decisionmaking. The
NRC recognizes that the consequences of an SFP fire could be large and
that is why there are numerous requirements in place to prevent a
situation where the spent fuel is uncovered.
This section provides detailed NRC responses to the three issues
identified in the petition.
Issue 1: The Requested Regulations Pertaining to SFP Accident
Evaluation Models Are Needed Because the Probability of the Type of
Events That Could Lead to SFP Accidents Is Relatively High
The petitioner stated that the requested regulations pertaining to
SFP accident evaluation models are needed because the probability of
the type of events that could lead to SFP accidents is relatively high.
The petitioner stated that an SFP accident could happen as a result of
a leak (rapid drain down) or boil-off scenario. Furthermore, the
petitioner notes that in the event of a long-term station blackout,
emergency diesel generators could run out of fuel and SFP cooling would
be lost, resulting in a boil-off of SFP water inventory and a
subsequent release of radioactive materials from the spent fuel. The
petitioner also provided several examples of events that could lead to
a long-term station blackout and, ultimately, an SFP accident, such as
a strong geomagnetic disturbance, a nuclear device detonated in the
earth's atmosphere, a pandemic, or a cyber or physical attack.
NRC Response
Spent nuclear fuel offloaded from a reactor is initially stored in
an SFP. The SFPs at all nuclear plants in the United States are robust
structures constructed with thick, reinforced, concrete walls and
welded stainless-steel liners. They are designed to safely contain the
spent fuel discharged from a nuclear reactor under a variety of normal,
off-normal, and hypothetical accident conditions (e.g., loss of
electrical power, loss of cooling, fuel or cask drop incidents, floods,
earthquakes, or extreme weather events). Racks fitted in the SFPs store
the fuel assemblies in a controlled configuration so that the fuel is
maintained in a sub-critical and coolable geometry. Redundant
monitoring, cooling, and water makeup systems are provided. The spent
fuel assemblies are typically covered by at least 25-feet of water,
which provides passive cooling as well as radiation shielding.
Penetrations to pools are limited to prevent inadvertent drainage, and
the penetrations are generally located well above spent fuel storage
elevations to prevent uncovering of fuel from drainage.
Studies conducted over the last four decades have consistently
shown the risk of an accident causing a zirconium fire in an SFP to be
low. The risk of an SFP accident was examined in the 1980s as Generic
Issue 82, ``Beyond Design Basis Accidents in Spent Fuel Pools,'' in
light of increased use of high-density storage racks and laboratory
studies that indicated the possibility of zirconium fire propagation
between assemblies in an air-cooled environment (Section 3 of NUREG-
0933, ``Resolution of Generic Safety Issues,'' https://nureg.nrc.gov/sr0933/). The risk assessment and cost-benefit analyses developed
through this effort, Section 6.2 of NUREG-1353, ``Regulatory Analysis
for the Resolution of Generic Issue 82, Beyond Design Basis Accidents
in Spent Fuel Pools'' (ADAMS Accession No. ML082330232), concluded that
the risk of a severe accident in the SFP was low and appeared to meet
the objectives of the Commission's Safety Goal Policy Statement public
health objectives (51 FR 30028; August 21, 1986) and that no new
regulatory requirements were warranted.
The risk of an SFP accident was re-assessed in the late 1990s to
support a risk-informed rulemaking for permanently shutdown, or
decommissioned, nuclear power plants in the United States. The study,
NUREG-1738, ``Technical Study of Spent Fuel Pool Accident Risk at
Decommissioning Nuclear Power Plants'' (ADAMS Accession No.
ML010430066), conservatively assumed that if the water level in the SFP
dropped below the top of the spent fuel, an SFP zirconium fire
involving all of the spent fuel would occur, and thereby bounded those
conditions associated with air cooling of the fuel (including partial-
drain down scenarios) and fire propagation. Even with this conservative
assumption, the study
[[Page 29763]]
found the risk of an SFP fire to be low and well within the
Commission's Safety Goals.
Additional mechanisms to mitigate the potential loss of SFP water
inventory were implemented following the terrorist attacks of September
11, 2001, which have enhanced spent fuel coolability and the potential
to recover SFP water level and cooling prior to a potential SFP
zirconium fire (73 FR 76204; August 8, 2008). Based on the
implementation of these additional strategies, the probability and,
accordingly, the risk of an SFP zirconium fire initiation has decreased
and is expected to be less than previously analyzed in NUREG-1738 and
previous studies.
Following the 2011 accident at Fukushima Dai-ichi, the NRC took
extensive actions to ensure that portable equipment is available to
mitigate a loss of cooling water in the SFP. 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). This order required
licensees to develop, implement, and maintain guidance and strategies
to maintain or restore core cooling, containment, and SFP cooling
capabilities following a beyond-design-basis external event. The NRC
endorsed the Nuclear Energy Institute (NEI) guidance to meet the
requirements of this order.\1\ That guidance establishes additional
mechanisms for mitigating a loss of SFP cooling water beyond the
requirements in 10 CFR 50.54(hh)(2), such as installing a remote
connection for SFP makeup water that can be accessed away from the SFP
refueling floor.
---------------------------------------------------------------------------
\1\ See NEI 12-06, ``Diverse and Flexible Coping Strategies
(FLEX) Implementation Guide,'' dated August 2012 (ADAMS Accession
No. ML12242A378), and JLD-ISG-2012-01, ``Compliance with Order EA-
12-049, Order Modifying Licenses with Regard to Requirements for
Mitigation Strategies for Beyond-Design-Basis External Events,''
dated August 2012 (ADAMS Accession No. ML12229A174).
---------------------------------------------------------------------------
Also, in 2014, the NRC documented a regulatory analysis in COMSECY-
13-0030, ``Staff Evaluation and Recommendation for Japan Lessons
Learned Tier 3 Issue on Expedited Transfer of Spent Fuel'' (ADAMS
Accession No. ML13329A918), which considered a broad history of the
NRC's oversight of spent fuel storage, SFP operating experience
(domestic and international), as well as information compiled in NUREG-
2161, ``Consequence Study of a Beyond-Design-Basis Earthquake Affecting
the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor'' (ADAMS
Accession No. ML14255A365). In COMSECY-13-0030, the NRC staff concluded
that SFPs are robust structures with large safety margins and
recommended to the Commission that assessments of possible regulatory
actions to require the expedited transfer of spent fuel from SFPs to
dry cask storage were not warranted. The Commission subsequently
approved the staff's recommendation in the Staff Requirements
Memorandum to COMSECY-13-0030 (ADAMS Accession No. ML14143A360).
As supported by numerous evaluations referenced in this document,
the NRC has determined that the risk of an SFP severe accident is low.
While the risk of a severe accident in an SFP is not negligible, the
NRC believes that the risk is low because of the conservative design of
SFPs; operational criteria to control spent fuel movement, monitor
pertinent parameters, and maintain cooling capability; mitigation
measures in place if there is loss of cooling capability or water; and
emergency preparedness measures to protect the public. The information
proposed to be provided to the NRC is not needed for the effectiveness
of NRC's approach for ensuring SFP safety. The NRC notes that the issue
of long-term cooling of SFPs is the subject of PRM-50-96, which was
accepted for consideration in the rulemaking process (77 FR 74788;
December 18, 2012) and is being addressed by the NRC's rulemaking
regarding mitigation of beyond design-basis events (RIN 3150-AJ49; NRC-
2014-0240).
Issue 2: Annual Licensee SFP Safety Evaluations and Submission of
Results to the NRC Is Necessary So That the NRC Is Aware of Potential
Consequences of Postulated SFP Accident/Fire Scenarios as Fuel
Assemblies Are Added, Removed, or Reconfigured in Licensees' SFPs
The petitioner stated that the purpose of the proposed requirement
is to keep the NRC informed of the potential consequences of postulated
SFP accident/fire scenarios as fuel assemblies are added, removed, or
reconfigured in licensees' SFPs.
NRC Response
The NRC does not agree that this is necessary because the NRC
already evaluates SFP systems and structures during initial licensing
and license amendment reviews. In addition, baseline NRC inspections
provide ongoing oversight to ensure adequate protection. There are not
sufficient benefits that would justify the new requirement proposed in
the petition for SFP accident evaluations. The proposed new requirement
for licensees to perform SFP evaluations would not prevent or mitigate
an SFP accident or provide information that is necessary for regulatory
decisionmaking. The annual licensee SFP safety evaluations and their
results proposed to be provided to the NRC are not needed for the
effectiveness of the NRC's approach to ensuring SFP safety.
The NRC issues licenses after reviewing and approving the design
and licensing bases contained in the plant's safety analysis report.
Licensees are required to operate the plant, including performing
operations and surveillances related to spent fuel, in accordance with
technical specifications and established practices and procedures for
that plant. Any licensee changes to design, operational or surveillance
practices, or approved spent fuel inventory limits or configuration
changes must be evaluated using the criteria in 10 CFR 50.59,
documented and retained for the duration of the operating license, and,
if warranted, submitted to the NRC for prior approval.
The general design criteria (GDC) in appendix A to 10 CFR part 50
establish general expectations that licensees must meet through
compliance with their plant-specific licensing basis. Several GDC apply
to SFPs:
Protecting against natural phenomena and equipment
failures (GDC 2 and GDC 4);
Preventing a substantial loss-of-coolant inventory under
accident conditions (e.g., equipment failure or loss of decay and
residual heat removal) (GDC 61);
Preventing criticality of the spent fuel (GDC 62); and
Adequately monitoring the SFP conditions for loss of decay
heat removal and radiation (GDC 63).
Additionally, emergency procedures and mitigating strategies are in
place to address unexpected challenges to spent fuel safety. Multiple
requirements in 10 CFR part 50, as well as recent NRC orders following
the Fukushima Dai-ichi accident, require redundant equipment and
strategies to address loss of cooling to SFPs and protective actions
for plant personnel and the public to limit exposure to radioactive
materials.
The NRC provides oversight of the licensee's overall plant
operations and the SFP in several ways. The NRC inspectors ensure that
spent fuel is stored safely by regularly inspecting reactor and
equipment vendors; inspecting the design, construction, and
[[Page 29764]]
use of equipment; and observing ``dry runs'' of procedures. At least
two NRC resident inspectors are assigned to each site to provide
monitoring and inspection of routine and special activities. They are
aware of, and routinely observe, SFP activities involving fuel
manipulation. The NRC inspectors use inspection procedures to guide
periodic inspection activities, and the results are published in
publicly-available inspection reports. Special inspections may be
conducted, as necessary, to evaluate root causes and licensee
corrective actions if site-specific events occur. Special inspections
may also evaluate generic actions taken by some or all licensees as a
result of an NRC order or a change in regulations.
In accordance with 10 CFR part 21, the NRC is informed of defects
and noncompliances associated with basic components, which include SFPs
and associated drain pipes and safety-related systems, structures, and
components for makeup water. This information allows the NRC to take
additional regulatory action as necessary with respect to defects and
noncompliances. The NRC is also informed of events and conditions at
nuclear power plants, as set forth in Sec. Sec. 50.72 and 50.73.
Depending upon the nature of the event or condition, a nuclear power
plant licensee must inform the NRC within a specified period of time of
the licensee's corrective action taken or planned to be taken. These
reports also facilitate effective and timely NRC regulatory oversight.
Finally, information identified by a nuclear power plant applicant or
licensee as having a significant implication for public health and
safety or common defense and security must be reported to the NRC
within 2 days of the applicant's or licensee's identification of the
information.
The annual evaluations requested in the petition would not provide
information that is necessary for regulatory decisionmaking. The
evaluations requested in the petition would postulate scenarios in
which the normal cooling systems, the backup cooling methods, and the
mitigation strategies have all failed to cool the stored fuel and would
require the calculation of the time it would take for the stored fuel
to ignite and how much of it would ignite. Due to the robustness of
this equipment, the NRC views this sequence of events as extremely
unlikely to occur. Since the current regulations require that the pool
be designed to prevent the loss-of-coolant and subsequent uncovering of
the fuel, the information that would be obtained from the proposed
requirement in the petition would not impact the current design basis.
Moreover, as discussed previously, the NRC's current regulatory
infrastructure relevant to SFPs at nuclear power plants in the United
States already contains information collection and reporting
requirements that support effective NRC regulatory oversight of SFPs.
The NRC does not agree that it is necessary to impose a new
requirement for licensees to perform annual evaluations of their SFPs
because existing requirements and oversight are sufficient to ensure
adequate protection of public health and safety.
Issue 3: MELCOR Is Not Currently Sufficient To Provide a Conservative
Evaluation of Postulated SFP Accident/Fire Scenarios
The petitioner requested that the NRC establish requirements for
SFP accident evaluation computer models to be used in the annual SFP
evaluations requested in Issue 2. The petitioner stated that there are
serious flaws with MELCOR, which has been used by the NRC to model
severe accident progression in SFPs, and, therefore, MELCOR is not
sufficient.
NRC Response
The NRC does not agree that it is necessary to establish
requirements for SFP accident evaluation computer models because the
annual SFP evaluations requested in Issue 2 are not necessary for
regulatory decisionmaking. Therefore, it is not necessary for the NRC
to establish requirements for how such an evaluation should be
conducted. Furthermore, the NRC disagrees with the petitioner's
statements that MELCOR is flawed.
There are inherent uncertainties in the progression of severe
accidents. There are many interrelated phenomena that need to be
properly understood; otherwise, conservatism in one area may lead to
overall non-conservative results. Conservatism can be meaningfully
introduced into the relevant analysis after the best estimate analysis
is done and uncertainties are properly taken into account.
The important question for a severe accident analysis is whether
the uncertainties are appropriately considered in the analysis results.
For example, Section 9 of the SFP study (NUREG-2161) is devoted to
discussing the major uncertainties that can affect the radiological
releases (e.g., hydrogen combustion, core concrete interaction, multi-
unit or concurrent accident, or fuel loading). In addition, the
regulatory analysis in COMSECY-13-0030 only relied on SFP study
insights for the boiling-water reactors with Mark I and II
containments, and, even then, the results were conservatively biased
towards higher radiological releases. For other designs, the release
fractions were based on previous studies (i.e., NUREG-1738) that used
bounding or conservative estimates.
The MELCOR computer code is the NRC's best estimate tool for severe
accident analysis. It has been validated against experimental data, and
it represents the current state of the art in severe accident analysis.
In NUREG-2161, the NRC stated that ``MELCOR has been developed through
the NRC and international research performed since the accident at
Three Mile Island in 1979. MELCOR is a fully integrated, engineering-
level computer code and includes a broad spectrum of severe accident
phenomena with capabilities to model core heatup and degradation,
fission product release and transport within the primary system and
containment, core relocation to the vessel lower head, and ex-vessel
core concrete interaction.'' Furthermore, MELCOR has been benchmarked
against many experiments, including separate and integral effects tests
for a wide range of phenomena. Therefore, the NRC has determined that
MELCOR is acceptable for its intended use.
Additional information about the capabilities of the MELCOR code to
model SFP accidents can be found in the NRC response to stakeholder
comments in Appendix E to NUREG-2161. The NRC also addressed questions
regarding MELCOR in Appendix D to NUREG-2157, Volume 2, ``Generic
Environmental Impact Statement for Continued Storage of Spent Nuclear
Fuel'' (ADAMS Accession No. ML14196A107).
III. Conclusion
For the reasons described in Section II, ``Reasons for Denial,'' of
this document, the NRC is denying the petition under 10 CFR 2.803. The
petitioner failed to present any information or arguments that would
warrant the requested amendments. The NRC does not believe that the
information that would be reported to the NRC as requested by the
petitioner is necessary for effective NRC regulatory decisionmaking
with respect to SFPs. The NRC continues to conclude that the current
design and licensing requirements for SFPs provide adequate protection
of public health and safety.
[[Page 29765]]
IV. Availability of Documents
The documents identified in the following table are available to
interested persons as indicated. For more information on accessing
ADAMS, see the ADDRESSES section of this document.
----------------------------------------------------------------------------------------------------------------
ADAMS accession number/Federal Register
Date Document citation
----------------------------------------------------------------------------------------------------------------
August 21, 1986...................... Safety Goals for the 51 FR 30028.
Operations of Nuclear Power
Plants; Policy Statement;
Republication.
April 1989........................... NUREG-1353, ``Regulatory ML082330232.
Analysis for the Resolution
of Generic Issue 82, Beyond
Design Basis Accidents in
Spent Fuel Pools''.
February 2001........................ NUREG-1738, ``Technical Study ML010430066.
of Spent Fuel Pool Accident
Risk at Decommissioning
Nuclear Power Plants''.
March 12, 2012....................... EA-12-049, ``Order Modifying ML12054A735.
Licenses with Regard to
Requirements for Mitigation
Strategies for Beyond-Design-
Basis External Events''.
August 2012.......................... NEI 12-06, ``Diverse and ML12242A378.
Flexible Coping Strategies
(FLEX) Implementation
Guide''.
August 2012.......................... JLD-ISG-2012-01, ``Compliance ML12229A174.
with Order EA-12-049, Order
Modifying Licenses with
Regard to Requirements for
Mitigation Strategies for
Beyond-Design-Basis External
Events''.
December 18, 2012.................... Long-Term Cooling and 77 FR 74788.
Unattended Water Makeup of
Spent Fuel Pools.
November 12, 2013.................... COMSECY-13-0030, ``Staff ML13329A918.
Evaluation and
Recommendation for Japan
Lessons Learned Tier 3 Issue
on Expedited Transfer of
Spent Fuel''.
May 23, 2014......................... SRM-COMSECY-13-0030, ``Staff ML14143A360.
Requirements--COMSECY-13-003
0--Staff Evaluation and
Recommendation for Japan
Lessons-Learned Tier 3 Issue
on Expedited Transfer of
Spent Fuel''.
June 19, 2014........................ Incoming Petition (PRM-50- ML14195A388.
108) from Mr. Mark Edward
Leyse.
September 2014....................... NUREG-2157, ``Generic ML14196A107.
Environmental Impact
Statement for Continued
Storage of Spent Nuclear
Fuel,'' Volume 2.
September 2014....................... NUREG-2161, ``Consequence ML14255A365.
Study of a Beyond-Design-
Basis Earthquake Affecting
the Spent Fuel Pool for a
U.S. Mark I Boiling-Water
Reactor''.
October 7, 2014...................... Notice of Docketing for PRM- 79 FR 60383.
50-108.
----------------------------------------------------------------------------------------------------------------
Dated at Rockville, Maryland, this 5th day of May, 2016.
For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 2016-11212 Filed 5-12-16; 8:45 am]
BILLING CODE 7590-01-P
