[Federal Register Volume 79, Number 56 (Monday, March 24, 2014)]
[Proposed Rules]
[Pages 16106-16146]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-05562]
[[Page 16105]]
Vol. 79
Monday,
No. 56
March 24, 2014
Part II
Nuclear Regulatory Commission
-----------------------------------------------------------------------
10 CFR Parts 50 and 52
Performance-Based Emergency Core Cooling Systems Cladding Acceptance
Criteria; Proposed Rule
Federal Register / Vol. 79 , No. 56 / Monday, March 24, 2014 /
Proposed Rules
[[Page 16106]]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
10 CFR Parts 50 and 52
[NRC-2008-0332, NRC-2012-0041, NRC-2012-0042, NRC-2012-0043]
RIN 3150-AH42
Performance-Based Emergency Core Cooling Systems Cladding
Acceptance Criteria
AGENCY: Nuclear Regulatory Commission.
ACTION: Proposed rule.
-----------------------------------------------------------------------
SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to
amend its regulations to revise the acceptance criteria for the
emergency core cooling system (ECCS) for light-water nuclear power
reactors. The proposed ECCS acceptance criteria are performance-based,
and reflect recent research findings that identified new embrittlement
mechanisms for fuel rods with zirconium alloy cladding under loss-of-
coolant accident (LOCA) conditions. The proposed rule also addresses
two petitions for rulemaking (PRMs) by establishing requirements
applicable to all fuel types and cladding materials, and requiring the
consideration of crud, oxide deposits, and hydrogen content in
zirconium-based alloy fuel cladding. Further, the proposed rule
contains a provision that would allow licensees to use an alternative
risk-informed approach to evaluate the effects of debris for long-term
cooling. The NRC is also seeking public comment on three draft
regulatory guides that would support the implementation of the proposed
rule.
DATES: Submit comments on the rule and draft guidance by June 9, 2014.
To facilitate NRC review, please distinguish between comments submitted
on the proposed rule and comments submitted on the draft guidance.
Submit comments on the information collection aspects of this rule by
April 23, 2014. Comments received after these dates will be considered
if it is practical to do so, but assurance of consideration cannot be
given to comments received after these dates.
ADDRESSES: The methods for accessing information and comment
submissions, and submitting comments on the proposed rule are different
from the methods for accessing information and comment submissions, and
submitting comments on the draft regulatory guides.
Proposed Rule
You may access information and comment submissions related to this
proposed rule by searching on http://www.regulations.gov under Docket
ID NRC-2008-0332. You may submit comments on the proposed rule by any
of the following methods:
Federal rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2008-0332. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected]. For technical questions, please
contact the individuals 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.
Draft Regulatory Guides
You may access information and comment submissions related to the
draft regulatory guides (DGs) by searching on http://www.regulations.gov under Docket ID NRC-2012-0041 (DG-1261,
``Conducting Periodic Testing for Breakaway Oxidation Behavior'' (the
NRC's Agencywide Documents Access and Management System (ADAMS)
Accession No. ML12284A324)), Docket ID NRC-2012-0042 (DG-1262,
``Testing for Post Quench Ductility'' (ADAMS Accession No.
ML12284A325)), and Docket ID NRC-2012-0043 (DG-1263, ``Establishing
Analytical Limits for Zirconium-Based Alloy Cladding'' (ADAMS Accession
No. ML12284A323)), respectively. You may submit comments on the draft
regulatory guides by any of the following methods:
Federal rulemaking Web site: Go to http://www.regulations.gov and search for Docket IDs NRC-2012-0041, NRC-2012-
0042, and NRC-2012-0043, respectively. Mail comments to: Cindy Bladey,
Chief, Rules, Announcements, and Directives Branch, Office of
Administration, Mail Stop: 3WFN-06-44M, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001.
Information Collections
You may submit comments on the information collections by the
methods described in the SUPPLEMENTARY INFORMATION section of this
document, under the heading, ``Paperwork Reduction Act Statement.''
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Tara Inverso, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, telephone: 301-415-1024, email: [email protected]; or
Paul M. Clifford, Office of Nuclear Reactor Regulation, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, telephone: 301-415-
4043, email: [email protected].
SUPPLEMENTARY INFORMATION:
Table of Contents
Executive Summary.
I. Accessing Information and Submitting Comments.
A. Accessing Information.
B. Submitting Comments.
II. Background.
A. Emergency Core Cooling System: Embrittlement Research
Findings.
B. Generic Safety Issue (GSI)-191 and Long-Term Cooling.
III. Operating Plant Safety.
A. Emergency Core Cooling System: Embrittlement Research
Findings.
B. GSI-191 and Long-Term Cooling.
IV. Advance Notice of Proposed Rulemaking: Public Comments.
V. Proposed Requirements for ECCS Performance During LOCAs.
A. Applicability of Performance-Based Rule: Consideration of
PRM-50-71.
B. Performance-Based Aspects of the Proposed Rule.
1. Hydrogen-Enhanced Beta-Layer Embrittlement.
2. Oxygen Ingress From Cladding Inside Diameter.
3. Breakaway Oxidation.
4. Applicability of Ductility-Based Analytical Limits in the
Burst Region.
5. Long-Term Cooling.
6. Use of Risk-Informed Approaches To Address Debris for Long-
Term Cooling.
C. Corrective Actions and Reporting Requirements.
1. Peak Cladding Temperature and Equivalent Cladding Reacted.
2. Risk-Informed Alternative To Address Debris for Long-Term
Cooling.
D. Consideration of PRM-50-84: Thermal Effects of Crud and Oxide
Layers.
E. Implementation.
1. Staggered Implementation Schedule.
2. Compliance With Long-Term Cooling Requirements Using Risk-
Informed Approach To Address Debris Effects.
[[Page 16107]]
VI. Section-by-Section Analysis.
A. Section 50.46c--Heading.
B. Section 50.46c(a)--Applicability.
C. Section 50.46c(b)--Definitions.
D. Section 50.46c(c)--Relationship to Other NRC Regulations.
E. Section 50.46c(d)--Emergency Core Cooling System Design.
F. Section 50.46c(e)--Alternate Risk-Informed Approach for
Addressing the Effects of Debris on Long-Term Core Cooling.
G. Section 50.46c(g)--Fuel System Designs: Uranium Oxide or
Mixed Uranium-Plutonium Oxide Pellets Within Cylindrical Zirconium-
Alloy Cladding.
H. Section 50.46c(k)--Use of NRC-Approved Fuel in Reactor.
I. Section 50.46c(l)--Authority To Impose Restrictions on
Operation.
J. Section 50.46c(m)--Corrective Actions and Reporting.
K. Section 50.46c(o)--Implementation.
L. Appendix K to Part 50 of Title 10 of the Code of Federal
Regulations (10 CFR), ECCS Evaluation Models.
M. Redesignation of Venting Requirements in Sec. 50.46a.
N. Changes Throughout 10 CFR Parts 50 and 52.
VII. Specific Request for Comments on the Proposed Rule.
A. Fuel Performance Criteria.
B. Risk-Informed Alternative To Address the Effects of Debris.
C. Implementation.
D. Other Issues.
VIII. Request for Comment: Draft Regulatory Guidance.
IX. Availability of Documents.
X. Criminal Penalties.
XI. Agreement State Compatibility.
XII. Plain Writing.
XIII. Voluntary Consensus Standards.
XIV. Finding of No Significant Environmental Impact: Environmental
Assessment.
XV. Paperwork Reduction Act Statement.
XVI. Regulatory Analysis: Availability.
XVII. Regulatory Flexibility Certification.
XVIII. Backfitting and Issue Finality.
Executive Summary
Purpose of the Regulatory Action
The proposed rule would adopt performance-based regulatory
requirements for determining the acceptability of an ECCS for a nuclear
power reactor, including requirements governing the acceptability of
the cladding of fuel. (Cladding performance affects the cooling
requirements for the ECCS.) The proposed rule would expand the
applicability of the rule from uranium oxide pellets within cylindrical
zircaloy or ZIRLO\TM\ cladding to any light-water reactor (LWR),
regardless of fuel design or cladding material. The proposed rule would
also replace prescriptive requirements with performance-based
requirements. Performance-based ECCS requirements would provide more
flexibility for applicants and licensees to meet NRC requirements for
emergency core cooling systems in a manner that provides reasonable
assurance of adequate protection consistent with the requirements of
the Atomic Energy Act of 1954, as amended. The requirements of the
proposed performance-based rule also address new technical information
on fuel cladding integrity and degradation mechanisms.
The proposed rule would also address two PRMs, PRM-50-71 and PRM-
50-84. The PRM-50-71 requests that the NRC expand the applicability of
the ECCS rule beyond zircaloy and ZIRLO\TM\ cladding materials. The
PRM-50-84 requests, among other items, that the NRC require licensees
to consider the thermal effects of crud and oxide layers.
Finally, the proposed rule would allow individual nuclear power
plant licensees to resolve GSI-191, ``Assessment of Debris Accumulation
on PWR [Pressurized Water Reactor] Sump Performance,'' by using a risk-
informed approach for evaluating the effects of debris on long-term
cooling.
Summary of the Significant Changes in the Proposed Rule
The proposed rule includes several significant changes to the NRC's
existing requirements on the ECCS:
The proposed rule would replace prescriptive analytical
requirements with performance-based requirements. To demonstrate
compliance with the requirements, ECCS performance would be evaluated
using fuel-specific performance objectives and associated analytical
limits that take into consideration all known degradation mechanisms
and unique features of the particular fuel system, along with an NRC-
approved ECCS evaluation model.
The proposed rule would apply to all fuel designs and
cladding materials. The proposed rule would define two principle ECCS
performance requirements:
[ssquf] Core temperature during and following the LOCA does not
exceed the analytical limits for the fuel design used for ensuring
acceptable performance.
[ssquf] The ECCS provides sufficient coolant so that decay heat
will be removed for the extended period of time required by the long-
lived radioactivity remaining in the core.
The proposed rule would also include specific performance
requirements for fuel designs consisting of uranium oxide or mixed
uranium-plutonium oxide fuel pellets within cylindrical zirconium-alloy
cladding. New performance objectives and analytical limits may be
necessary for other fuel designs, as they are developed. These changes
address the requests of PRM-50-71.
The proposed rule would incorporate the results of recent
research findings. The current requirement to maintain the calculated
total cladding oxidation below 17 percent would be replaced with a
requirement to establish analytical limits on peak cladding temperature
(PCT) and integral time at temperature (ITT) that correspond to the
measured ductile-to-brittle transition for the zirconium-alloy cladding
material. The proposed rule would also address a newly identified
phenomenon known as breakaway oxidation by requiring that the total
accumulated time that the cladding is predicted to remain above a
temperature at which the zirconium-alloy has been shown to be
susceptible to breakaway oxidation shall not be greater than a limit
that corresponds to the measured onset of breakaway oxidation for that
cladding. The proposed rule would also add a requirement to
periodically measure breakaway oxidation. Additionally, the proposed
rule would require licensees to consider the effects of oxygen
diffusion from the cladding inside surfaces, if an oxygen source is
present on the inside surfaces at the onset of the LOCA.
The proposed rule would require that licensees evaluate
the thermal effects of crud and oxide layers that accumulate on the
fuel cladding during plant operation. Crud is defined as any foreign
substance deposited on the surface of the fuel cladding prior to
initiation of a LOCA. This addition addresses a request of PRM-50-84.
The proposed rule contains a provision that would allow
licensees to use an alternative risk-informed approach to evaluate the
effects of debris for long-term cooling. The proposed rule contains
acceptance criteria that would apply to the risk-informed approach and
its required content. Additionally, the proposed rule would add
reporting requirements that pertain to the risk-informed approach.
Costs and Benefits
The proposed rule, by requiring applicants and licensees to address
new technical matters not currently required to be addressed by the
NRC's existing ECCS requirements, would provide adequate protection to
the health and safety of the public by maintaining that level of
protection that the NRC previously thought would be achieved by the
current rule. The NRC prepared a draft regulatory analysis for this
proposed rule (ADAMS Accession No.
[[Page 16108]]
ML12283A188) to identify the benefits and costs of the particular
regulatory approach for addressing ECCS performance. The NRC notes that
adequate protection must be assured without regard to cost, but if
there is more than one way of achieving that level of protection, then
costs may be considered. The draft regulatory analysis prepared for
this rulemaking was used to help the NRC identify the most effective
way of achieving reasonable assurance of adequate protection with
respect to protection against LOCAs.
The benefits of maintaining reasonable assurance of protection with
respect to protection against LOCAs were not quantified. The NRC
estimates that the total cost of the proposed rule would be $35 million
(7 percent net present value). The benefits of the proposed rule are
several. The proposed rule would result in savings by obviating the
need for exemption requests to use additional claddings and exemption
requests stemming from the risk-informed alternative. As a more general
matter, adopting a performance-based approach to demonstrating ECCS
adequacy may afford applicants and licensees greater flexibility in
complying with the NRC's ECCS requirements. This may result in reduced
applicant and licensee costs with no adverse effect on public health
and safety.
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2008-0332, Docket ID NRC-2012-0041,
Docket ID NRC-2012-0042, or Docket ID NRC-2012-0043 when contacting the
NRC about the availability of information for this proposed rule or
draft regulatory guides, respectively. You may access information
related to this proposed rulemaking or draft regulatory guides by the
following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2008-0332 for the
proposed rule, and Docket ID NRC-2012-0041, Docket ID NRC-2012-0042, or
Docket ID NRC-2012-0043 for the draft regulatory guides.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly-available documents online in the NRC
Library 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 in this notice (if that document is
available in ADAMS) is provided the first time that a document is
referenced. In addition, for the convenience of the reader, the ADAMS
accession numbers are provided in a table in the section of this
document entitled, Availability of Documents.
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 the appropriate NRC Docket ID in the subject line of
your comment submission, in order to ensure that the NRC is able to
make your comment submission available to the public in that docket.
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 will post all comment submissions at http://www.regulations.gov as well as enter 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 submissions. 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 entering the comment submissions into ADAMS.
II. Background
A. Emergency Core Cooling System: Embrittlement Research Findings
In SECY-98-300, ``Options for Risk-Informed Revisions to 10 CFR
Part 50-`Domestic Licensing of Production and Utilization Facilities,'
'' dated December 23, 1998 (ADAMS Accession No. ML992870048), the NRC
began to explore approaches to risk-informing its regulations for
nuclear power reactors. One alternative (termed ``Option 3'') involved
making risk-informed changes to the specific requirements in the body
of 10 CFR part 50. As the NRC began to develop its approach to risk-
informing these requirements, it sought stakeholder input in public
meetings. Two of the regulations identified by industry as potentially
benefitting from risk-informed changes were Sec. Sec. 50.44 and 50.46.
Section 50.44 specifies the requirements for combustible gas control
inside reactor containment structures, and Sec. 50.46 specifies the
requirements for light-water power reactor emergency core cooling
systems. For Sec. 50.46, the potential was identified for making risk-
informed changes to requirements for both ECCS cooling performance and
ECCS analysis acceptance criteria in Sec. 50.46(b).
PRM-50-71
On March 14, 2000, as amended on April 12, 2000, the Nuclear Energy
Institute (NEI) submitted a PRM (ADAMS Accession No. ML003723791)
requesting that the NRC amend its regulations in Sec. Sec. 50.44 and
50.46 (PRM-50-71). The NEI petition noted that these two regulations
apply to only two specific zirconium-alloy fuel cladding materials
(zircaloy and ZIRLO\TM\). The NEI stated that reactor fuel vendors had
subsequently developed new cladding materials other than zircaloy and
ZIRLO\TM\ and that, in order for licensees to use these new materials
under the regulations, licensees needed to request NRC approval of
exemptions from Sec. Sec. 50.44 and 50.46.
On May 31, 2000, the NRC published a notice of receipt (65 FR
34599) and requested public comment. The public comment period ended on
August 14, 2000, and the NRC received 11 public comment letters from
public citizens and the nuclear industry. Although the majority of the
comments generally supported the requests of the PRM, one commenter
suggested that the enhanced efficiency of the proposal would be at the
expense of public health and safety. The NRC disagrees with that
commenter and notes that, while the petition's proposal would remove
specific zirconium-alloy names from the regulation, the NRC review and
approval of specific zirconium-alloys for use as reactor fuel cladding
would be required prior to their use in reactors (with the exception of
lead test assemblies permitted in technical specifications). The NRC's
detailed discussion of the public comments submitted on PRM-50-71,
including a detailed list of commenters, is contained in a separate
document, ``Section 50.46c and PRM-50-71 Comment Response Document''
(ADAMS Accession No. ML12283A213).
After evaluating the petition and public comments received, the NRC
[[Page 16109]]
decided that PRM-50-71 should be considered in the rulemaking process.
The NRC's determination was published in the Federal Register on
November 6, 2008 (73 FR 66000). Because most of the issues raised in
this PRM pertain to Sec. 50.46, the PRM is addressed in this proposed
rule.\1\
---------------------------------------------------------------------------
\1\ PRM-50-71 also requested changes to Sec. 50.44. Those
changes were addressed in a rulemaking that revised that section (68
FR 54123; September 16, 2003) to include risk-informed requirements
for combustible gas control. That regulation was also modified to be
applicable to all boiling or pressurized water reactors regardless
of type of fuel cladding material used.
---------------------------------------------------------------------------
Staff Requirements Memorandum Direction
On March 31, 2003, in response to SECY-02-0057, ``Update to SECY-
01-0133, `Fourth Status Report on Study of Risk-Informed Changes to the
Technical Requirements of 10 CFR Part 50 (Option 3) and Recommendations
on Risk-Informed Changes to 10 CFR 50.46 (ECCS Acceptance Criteria)' ''
(ADAMS Accession No. ML020660607), the Commission issued a staff
requirements memorandum (SRM) (ADAMS Accession No. ML030910476)
directing the NRC staff to move forward to risk-inform its regulations
in a number of specific areas. In addition, this SRM directed the staff
to modify the ECCS acceptance criteria to provide a more performance-
based approach to the ECCS requirements in Sec. 50.46.
Research Results
Separate from the effort to modify the regulations to provide a
more risk-informed, performance-based regulatory approach, the NRC had
also undertaken a fuel cladding research program to investigate the
behavior of high-exposure fuel cladding under accident conditions. This
research program included an extensive LOCA research and testing
program at Argonne National Laboratory (ANL), as well as jointly-funded
programs at the Kurchatov Institute (supported by the French Institute
for Radiological Protection and Nuclear Safety and the NRC) and the
Halden Reactor project (a jointly-funded program under the auspices of
the Organization for Economic Cooperative Development--Nuclear Energy
Agency, sponsored by national organizations in 18 countries), to
develop the body of technical information needed to support the new
regulations.
The effects of both alloy composition and fuel burnup (the extent
to which fuel is used in a reactor) on cladding embrittlement (e.g.,
loss of ductility) under accident conditions were studied in these
research programs. The research programs identified new cladding
embrittlement mechanisms and expanded the NRC's knowledge of previously
identified mechanisms. The research results revealed that alloy
composition has a minor effect on embrittlement, but that the cladding
corrosion that occurs as fuel burnup increases has a substantial effect
on embrittlement. One of the major findings of the NRC's research
program was that hydrogen, which is absorbed in the cladding as a
result of zirconium oxidation (e.g., corrosion) under normal operation,
has a significant influence on embrittlement during a postulated LOCA.
Increased hydrogen content increases both the solubility of oxygen in
zirconium and the rate at which it is diffused within the metal, thus
increasing the amount of oxygen in the metal during high temperature
oxidation in LOCA conditions. Further, the NRC's research program found
that oxygen from the oxide fuel pellets enters the cladding from the
inner surface if a bonding layer exists between the fuel pellet and the
cladding, in addition to the oxygen that enters from the oxide layer on
the outside of the cladding. Moreover, under some small-break LOCA
conditions (such as extended time-at-temperature around 1,000 degrees
Celsius ([deg]C) (1832 degrees Fahrenheit ([deg]F))), a phenomenon
termed breakaway oxidation can take place, allowing large amounts of
hydrogen to diffuse into the cladding, exacerbating the embrittlement
process. Breakaway oxidation is defined as the fuel cladding oxidation
phenomenon in which weight gain rate deviates from normal kinetics.
This change occurs with a rapid increase of hydrogen pickup during
prolonged exposure to a high temperature steam environment, which
promotes lack of ductility.
The research results also confirmed a previous finding that if
cladding rupture occurs during a LOCA, large amounts of hydrogen from
the steam-cladding reaction can enter the cladding inside surface near
the rupture location. These research findings have been summarized in
Research Information Letter (RIL)-0801, ``Technical Basis for Revision
of Embrittlement Criteria in 10 CFR 50.46'' (ADAMS Accession No.
ML081350225), and the detailed experimental results from the program at
ANL are contained in NUREG/CR-6967, ``Cladding Embrittlement during
Postulated Loss-of-Coolant Accidents'' (ADAMS Accession No.
ML082130389). Since the publication of NUREG/CR-6967 and RIL-0801,
additional testing was conducted related to the embrittlement
phenomenon, which has been documented in supplemental reports. Where
the additional testing relates to conclusions and recommendations in
RIL-0801, RIL-0801 has been supplemented to reference the additional
reports and incorporate findings (``Update to Research Information on
Cladding Embrittlement Criteria in 10 CFR 50.46,'' dated December 29,
2011 (ADAMS Accession No. ML113050484)).
The NRC publicly released the technical basis information in RIL-
0801 on May 30, 2008, and NUREG/CR-6967 on July 31, 2008. Also on July
31, 2008, the NRC published in the Federal Register a notice of
availability of the RIL and NUREG/CR-6967, together with a request for
comments (73 FR 44778). In that notice, the NRC stated that these
documents and comments on the documents would be discussed at a public
workshop to be scheduled in September 2008. The public workshop was
held on September 24, 2008, and included presentations and open
discussion between representatives of the NRC, international regulatory
and research agencies, domestic and international commercial power
firms, fuel vendors, and the general public. A summary of the workshop,
including a list of attendees and presentations, is available in ADAMS
under Accession No. ML083010496. The NRC has not prepared responses to
comments received on the technical basis information as a result of the
July 31, 2008, Federal Register notice (including comments received at
the September 2008 public workshop), because: (i) The public workshop
was held, in part, to discuss public comments on the technical basis
information, and (ii) further opportunity to comment is available
during this proposed rule's formal public comment period.
Based upon a preliminary safety assessment in response to the
research findings in RIL-0801, the NRC determined that immediate
regulatory action was not required, and that changes to the ECCS
acceptance criteria to account for these new findings could reasonably
be addressed through the rulemaking process. Recognizing that
finalization and implementation of the new ECCS requirements would take
several years, the NRC completed a more detailed safety assessment that
confirmed current plant safety for every operating reactor. See Section
III, ``Operating Plant Safety,'' of this document for further
information.
Since 2002, the NRC has met with the Advisory Committee on Reactor
Safeguards (ACRS) multiple times to
[[Page 16110]]
discuss the progress of the LOCA research program and rulemaking
proposals. Provided in the following table are the dates and ADAMS
accession numbers of the relevant ACRS meetings and associated
correspondence.
------------------------------------------------------------------------
Date Meeting/Letter ADAMS
------------------------------------------------------------------------
October 9, 2002............... Subcommittee Meeting. * ML023030246
October 10, 2002.............. Full Committee * ML022980190
Meeting.
October 17, 2002.............. Letter from ACRS to ML022960640
NRC staff.
December 9, 2002.............. Response letter from ML023260357
NRC staff to ACRS.
September 29, 2003............ Subcommittee Meeting. * ML032940296
July 27, 2005................. Subcommittee Meeting. * ML052230093
September 8, 2005............. Full Committee * ML052710235
Meeting.
January 19, 2007.............. Subcommittee Meeting. * ML070390301
February 2, 2007.............. Full Committee ML070430485
Meeting.
May 23, 2007.................. Letter from ACRS to ML071430639
NRC Staff.
July 11, 2007................. Response letter from ML071640115
NRC staff to ACRS.
December 2, 2008.............. Subcommittee Meeting. * ML083520501
* ML083530449
December 4, 2008.............. Full Committee * ML083540616
Meeting.
December 18, 2008............. Letter from ACRS to ML083460310
NRC staff.
January 23, 2009.............. Response letter from ML083640532
NRC staff to ACRS.
May 10, 2011.................. Subcommittee Meeting. ML111450409
June 8, 2011.................. Full Committee ML11166A181
Meeting.
June 22, 2011................. Letter from ACRS to ML11164A048
NRC staff.
June 23, 2011................. Subcommittee Meeting. ML11193A035
July 13, 2011................. Full Committee ML11221A059
Meeting.
July 21, 2011................. Response letter from ML111861706
NRC staff to ACRS.
December 15, 2011............. Subcommittee Meeting. ML120100268
January 19, 2012.............. Full Committee ML12032A048
Meeting.
January 26, 2012.............. Letter from ACRS to ML12023A089
NRC Staff.
February 17, 2012............. Response Letter from ML120260893
NRC staff to ACRS.
------------------------------------------------------------------------
* ADAMS file is a transcript of the ACRS meeting.
PRM-50-84
On March 15, 2007, Mark Leyse (the petitioner) submitted a PRM to
the NRC (ADAMS Accession No. ML070871368) requesting that all holders
of operating licenses for nuclear power plants be required to operate
such plants at operating conditions (e.g., levels of power production
and light-water coolant chemistries) necessary to effectively limit the
thickness of crud \2\ and/or oxide layers on fuel rod cladding
surfaces. The petitioner requests that the NRC conduct rulemaking in
the following three specific areas:
---------------------------------------------------------------------------
\2\ For the purpose of this discussion, the NRC defines ``crud''
as any foreign substance deposited on the surface of the fuel
cladding prior to the initiation of a LOCA. It is known that this
layer can impede the transfer of heat.
---------------------------------------------------------------------------
(1) Establish regulations that require licensees to operate light-
water power reactors under conditions that are effective in limiting
the thickness of crud and/or oxide layers on zirconium-clad fuel in
order to ensure compliance with Sec. 50.46(b) ECCS acceptance
criteria;
(2) Amend appendix K to 10 CFR part 50 to explicitly require that
steady-state temperature distribution and stored energy in the reactor
fuel at the onset of a postulated LOCA be calculated by factoring in
the role that the thermal resistance of crud deposits and/or oxide
layers plays in increasing the stored energy in the fuel (these
requirements also need to apply to any NRC-approved, best-estimate ECCS
evaluation models used in lieu of appendix K to 10 CFR part 50
calculations); and
(3) Amend Sec. 50.46 to specify a maximum allowable percentage of
hydrogen content in (fuel rod) cladding.
On May 23, 2007, the NRC published a notice of receipt for this
petition in the Federal Register (72 FR 28902) and requested public
comment. The public comment period ended on August 6, 2007. Comments in
support of PRM-50-84 were provided by the Union of Concerned
Scientists, two individuals, and the petitioner. The NEI and Strategic
Teaming and Resource Sharing organization submitted comments in
opposition to the petition. After evaluating the public comments, the
NRC resolved PRM-50-84 by deciding that each of the petitioner's issues
should be considered in the rulemaking process. The NRC's
determination, including the NRC's response to public comments received
on the petition, was published in the Federal Register on November 25,
2008 (73 FR 71564). Although there is no direct relationship between
the subject of crud and the anticipated new ECCS acceptance criteria
requirements, the petition deals with the NRC's requirements on ECCS
performance in Sec. 50.46. Given the comprehensive changes to Sec.
50.46 being addressed in this rulemaking, the NRC is considering the
petitioner's proposed changes in this rulemaking.
B. Generic Safety Issue (GSI)-191 and Long-Term Cooling
As a result of evolving staff concerns related to the adequacy of
PWR recirculation sump designs, the NRC opened Unresolved Safety Issue
(USI) A-43, ``Containment Emergency Sump Performance.'' The resolution
of USI A-43 was subsequently documented in Generic Letter (GL) 1985-
022, ``Potential for Loss of Post-LOCA Recirculation Capability Due to
Insulation Debris Blockage,'' dated December 3, 1985 (ADAMS Accession
No. ML031150731). The NRC staff found in GL 1985-022 that the 50
percent blockage assumption, identified in Regulatory Guide (RG) 1.82,
``Sumps for Emergency Core Cooling and Containment Spray Systems,''
Revision 0 (ADAMS Accession No. ML111680318), should be replaced with a
more comprehensive requirement to assess debris effects on a plant-
specific basis. Following the resolution of USI A-43, industry events
at Barsebeck and Limerick Generating Station challenged the conclusion
that no new requirements were necessary to prevent the clogging of ECCS
strainers at operating boiling water reactors (BWR).
[[Page 16111]]
As described in NRC Bulletin 95-02, ``Unexpected Clogging of a
Residual Heat Removal (RHR) Pump Strainer While Operating in
Suppression Pool Cooling Mode,'' dated October 7, 1995 (ADAMS Accession
No. ML082490807), a safety relief valve at the Limerick Generating
Station inadvertently opened and could not be closed, the plant was
manually scrammed, and the RHR system was started in the suppression
pool cooling mode to remove the heat added by the open relief valve.
The A train of the RHR exhibited signs of pump cavitation and was
secured. The B train of the RHR was started to remove the heat from the
relief valve discharge. After the plant was stabilized, a diver
inspected the pump suction strainers and found a mat of fibers and
sludge covering them. The licensee determined that the discharge from
the relief valve did not contribute debris to the suppression pool.
As described in NRC Bulletin 96-03, ``Potential Plugging of
Emergency Core Cooling Suction Strainers by Debris in Boiling-Water
Reactors,'' dated May 6, 1996 (ADAMS Accession No. ML082401219), a
Swedish BWR, Barseback Unit 2, experienced plugging of two containment
vessel spray system (CVSS) suction strainers. The strainers were
partially plugged with mineral wool (a fibrous insulation) that was
dislodged by a steam jet from an open pilot operated relief valve. The
operators noticed an indication of high-differential pressure across
the strainers and were able to back flush them to keep the CVSS
operating.
Also described in NRC Bulletin 96-03 are two ECCS suction strainer
plugging events that occurred at the Perry Nuclear Power Plant, a BWR
located in the United States. The first event resulted from general
maintenance material and dirt in the suppression pool collecting on the
RHR suction strainers. The differential pressure caused by the debris
resulted in deformation of the suction strainers. After the suppression
pool was cleaned and the suction strainers replaced, a second event
occurred when several safety relief valves lifted. The RHR system was
used to cool the suppression pool after the steam discharge. The
suction strainers were inspected and found to be covered with fibrous
debris and corrosion products. A test of the system found that the B
train pump suction pressure dropped to zero. The fibrous debris
originated from temporary drywell cooling filter media that was
accidentally dropped into the suppression pool and not retrieved. The
fibers created a filtering bed on which particles collected, resulting
in a high-resistance debris bed.
In response to these events, the NRC issued generic communications
requesting that BWR licensees take appropriate actions to minimize the
potential for the clogging of ECCS suction strainers by debris
accumulation following a LOCA. The NRC staff concluded that all BWR
licensees have sufficiently addressed these bulletins in a memorandum,
``Completion of Staff Reviews of NRC Bulletin 96-03, `Potential
Plugging of Emergency Core Cooling Suction Strainers by Debris in
Boiling-Water Reactors,' and NRC Bulletin 95-02, `Unexpected Clogging
of a Residual Heat Removal (RHR) Pump Strainer While Operating in
Suppression Pool Cooling Mode','' dated October 18, 2001 (ADAMS
Accession No. ML012970229).
The findings regarding BWR strainers prompted the NRC to open GSI-
191, ``Assessment of Debris Accumulation on PWR Sump Performance,'' to
ensure that post-accident debris effects would not impede long-term
core cooling at PWRs. After completing its technical assessment of GSI-
191, the NRC issued Bulletin 2003-01, ``Potential Impact of Debris
Blockage on Emergency Sump Recirculation at Pressurized-Water
Reactors,'' dated June 9, 2003 (ADAMS Accession No. ML031600259). This
bulletin did not require licensees to immediately perform deterministic
evaluations for debris effects, but requested that plants take
compensatory measures to reduce risk or otherwise enhance the
capability of the ECCS and containment spray system (CSS) recirculation
functions. The bulletin also informed licensees that the staff was
preparing a generic letter that would request that plants demonstrate
through deterministic methods that long-term core cooling would not be
compromised by debris effects.
Generic Letter 2004-02, ``Potential Impact of Debris Blockage on
Emergency Recirculation During Design Basis Accidents at Pressurized-
Water Reactors,'' dated September 13, 2004 (ADAMS Accession No.
ML042360586), was issued to all operating PWRs requesting that they
perform a mechanistic evaluation of the effects of debris on the ECCS
and CSS recirculation functions. The affected plants are currently
working to address the issues identified by the generic letter. All
operating PWRs have installed larger strainers and taken other actions
toward the final resolution of the issue. Final closure of the generic
letter has been delayed to allow industry and the NRC staff to develop
appropriate methodologies for evaluation of debris related issues that
were identified after the issuance of the generic letter. The staff
generated two SECY papers on this issue to provide options and solicit
feedback from the NRC Commissioners. On December 14, 2012, the
Commission issued an SRM (ADAMS Accession No. ML12349A378) for SECY-12-
0093, ``Closure Options for Generic Safety Issue--191, Assessment of
Debris Accumulation on Pressurized-Water Reactor Sump Performance''
(ADAMS Accession No. ML121320270). In this SRM, the Commission directed
the following:
The forthcoming Sec. 50.46c proposed rulemaking should contain
a provision allowing NRC licensees on a case-by-case basis, to use
risk-informed alternatives. The license amendment process would be
used to reconstitute the long-term core cooling licensing basis.
Stakeholder comments should be solicited on the proposed provision.
Consistent with this SRM, the proposed rule includes a provision
that would allow licensees to use an alternative risk-informed approach
to evaluate the effects of debris for long-term cooling.
III. Operating Plant Safety
A. Emergency Core Cooling System: Embrittlement Research Findings
In response to the research findings in RIL-0801, the NRC performed
a preliminary safety assessment of currently operating reactors
(``Plant Safety Assessment of RIL-0801 (non-proprietary),'' dated
February 23, 2009 (ADAMS Accession No. ML090340073)). This assessment
found that, due to realistic fuel rod power history, measured cladding
performance under LOCA conditions, and current analytical
conservatisms, sufficient safety margin exists for operating reactors.
Therefore, the NRC staff determined that immediate regulatory action
was not required, and that changes to the ECCS acceptance criteria to
account for these new findings can reasonably be addressed through the
rulemaking process.
Recognizing that finalization and implementation of the new ECCS
requirements would take several years, the NRC decided that a more
detailed safety assessment was necessary. As a voluntary industry
effort, the PWR Owners Group (OG) (``Letter Report: OG-11-143 PWROG
50.46(b) Margin Assessment,'' dated April 29, 2011 (ADAMS Accession No.
ML11139A309)) and BWR OG (``BWROG-TP-11-010 (Rev. 1) Evaluation of BWR
LOCA Analyses and Margins Against High Burnup Fuel
[[Page 16112]]
Research Findings,'' dated June 2011 (ADAMS Accession No.
ML111950139)), under the auspices of NEI, submitted ECCS margin
assessment reports. After grouping plants based on similar design
features, cladding alloys, or ECCS evaluation models and defining
cladding alloy-specific analytical limits, the OG reports identified
analytical credits or performed new LOCA analyses necessary to
demonstrate that the limiting plant within each grouping had positive
margin relative to the research findings. The NRC conducted an audit of
the OG reports and supporting General Electric--Hitachi (GEH), AREVA,
and Westinghouse engineering calculations. Based on the OG reports and
supplemental information collected during the audits, the NRC was able
to confirm, for every operating reactor, current safe operation. As
documented in the audit report and safety assessment (``ECCS
Performance Safety Assessment and Audit Report,'' dated February 10,
2012 (ADAMS Accession No. ML12041A078)), the NRC intends to verify, on
an annual basis, continued safe operation until each licensee has
implemented the new ECCS requirements. See Section V.E,
``Implementation,'' of this document for the staff-recommended
implementation plan developed based on this information.
B. GSI-191 and Long-Term Core Cooling
Section II. B., ``GSI-191 and Long-Term Cooling,'' of this document
provides background information on GSI-191 and long-term cooling. That
section includes information on action taken by the NRC and licensees
to address the potential effects of debris on long-term cooling. These
actions have contributed significantly to the safety of operating
plants. The NRC staff provided information to the Commission in two
SECY papers: SECY-10-0113, ``Closure Options for Generic Safety Issue--
191, Assessment of Debris Accumulation on Pressurized Water Reactor
Sump Performance,'' dated August 26, 2010 (ADAMS Accession No.
ML101820296); and SECY-12-0093, ``Closure Options for Generic Safety
Issue--191, Assessment of Debris Accumulation on Pressurized Water
Reactor Sump Performance,'' dated July 9, 2012 (ADAMS Accession No.
ML12130270).
The Commission issued guidance for the closure of the issue in two
SRMs associated with each SECY paper. The SRM to SECY-10-0113 (``Staff
Requirements--SECY-10-0113--Closure Options for Generic Safety Issue--
191, Assessment of Debris Accumulation on Pressurized Water Reactor
Sump Performance'' (ADAMS Accession No. ML103570354)) was issued on
December 23, 2010. With respect to operating plant safety the SRM
stated:
The staff should take the time needed to consider all options to
a risk-informed, safety conscious resolution to GSI-191. While they
have not fully resolved this issue, the measures taken thus far in
response to the sump-clogging issue have contributed greatly to the
safety of U.S. nuclear power plants. Given the vastly enlarged
advanced strainers installed, compensatory measures already taken,
and the low probability of challenging pipe breaks, adequate
defense-in-depth is currently being maintained.
On December 14, 2012, the Commission issued the SRM to SECY-12-0093
(ADAMS Accession No. ML12349A378). With respect to operating plant
safety, the SRM reiterated the direction in SRM-SECY-10-0113.
As directed by the Commission, the NRC staff is currently working
with licensees to assure adequate safety by closing the issue and
updating their licensing bases to reflect full compliance on a schedule
consistent with Commission direction.
IV. Advance Notice of Proposed Rulemaking: Public Comments
On August 13, 2009, the NRC published an Advance Notice of Proposed
Rulemaking (ANPR) (74 FR 40767) to obtain stakeholder views on issues
associated with amending Sec. 50.46(b). The ANPR indicated that the
proposed scope of the rulemaking included four major objectives: (1)
Expand the applicability of Sec. 50.46 to include any light-water
reactor fuel cladding material; (2) establish performance-based
requirements and acceptance criteria specific to zirconium-based
cladding materials that reflect research findings; (3) revise the LOCA
reporting requirements; and (4) address the issues raised in PRM-50-84
that relate to crud deposits and hydrogen content in fuel cladding. The
ANPR provided interested stakeholders an opportunity to comment on the
options under consideration by the NRC during a 75-day public comment
period. In addition, the NRC asked 12 specific questions in the
following categories: Applicability Considerations, New Embrittlement
Criteria Considerations, Testing Considerations, Revised Reporting
Requirements Considerations, Crud Analysis Considerations, and Cost
Considerations. The public comment period ended on October 27, 2009.
The NRC received a total of 19 comment letters during the ANPR's
public comment period; these letters were sent from a variety of
entities, including one comment from a private citizen, 15 comments
from the nuclear industry, one comment from a non-governmental
organization, and two comments from the international community. The
NRC held a public meeting on April 28-29, 2010, to discuss, among other
things, the public comments received on the ANPR. No additional public
comments were accepted at this public meeting. The meeting summary is
available in ADAMS under Accession No. ML101300490.
As a result of comments received on the ANPR, the NRC has made a
number of changes to the proposed rule. A detailed discussion of the
public comments submitted on the ANPR, including a detailed list of
commenters, is contained in a separate document, ``Section 50.46c and
PRM-50-71 Comment Response Document'' (ADAMS Accession No.
ML12283A213). The most significant changes as the result of public
comments are:
The specific experimental technique for measuring cladding
ductility (i.e., 1.00 percent permanent strain prior to
failure during ring-compression loading at a temperature of 135 [deg]C
and a displacement rate of 0.033 millimeters per second (mm/sec)) was
removed from the rule and provided as one approved method within DG-
1262, ``Testing for Postquench Ductility'' (ADAMS Accession No.
ML12284A325).
The specific experimental technique for measuring time
until breakaway oxidation (i.e., hydrogen uptake reaches 200 weight
part per million (wppm) anywhere on a cladding segment subjected to
high-temperature steam oxidation ranging from 1200[emsp14][deg]F to
1875[emsp14][deg]F (649 [deg]C to 1024 [deg]C)) was removed from the
rule and provided as one approved method within DG-1261, ``Conducting
Periodic Testing for Breakaway Oxidation Behavior'' (ADAMS Accession
No. ML12284A324).
The proposed risk-informed change to the reporting
requirements (objective three of the ANPR) was abandoned. The majority
of public comments received on the proposed reporting criteria
suggested that the concept was complex, and might promote unnecessary
burden or misinterpretation.
The applicability of the zirconium-based alloy fuel
specific performance requirements was expanded to include uranium-
plutonium mixed oxide fuel.
The applicability of the post-quench ductility (PQD)
analytical limits in DG-1263, ``Establishing Analytical Limits for
Zirconium-Based Alloy Cladding'' (ADAMS Accession No. ML12284A323), was
expanded to
[[Page 16113]]
encompass cladding hydrogen concentration up to 800 wppm.
Many changes and improvements were made in the development
of DG-1261, DG-1262, and DG-1263.
A staged implementation plan was developed.
V. Proposed Requirements for ECCS Performance During LOCAs
The proposed rule would establish a general, performance-based rule
governing ECCS performance for LWRs, regardless of fuel design or
cladding material. This represents a significant change from the
current ECCS regulations, which apply to ``uranium oxide pellets within
cylindrical zircaloy or ZIRLO\TM\ cladding.'' Because ECCS system
requirements must be expressed independent of fuel type, and because
ECCS system performance ultimately must be based upon maintaining the
fuel in the reactor in a safe (analyzed) condition, the proposed rule
separates the ECCS system requirements from the need for the applicant/
licensee to establish the fuel system design performance criteria
constituting a safe condition.
In proposed Sec. 50.46c, the specified performance objectives of
the systems, structures, and components of the ECCS are to provide
residual heat removal during and following a postulated LOCA. As with
the current regulations, the ECCS performance is demonstrated by NRC-
approved ECCS evaluation models in proposed Sec. 50.46c. Specific
performance requirements and analytical limits have been established
for fuel designs consisting of uranium oxide or mixed uranium-plutonium
oxide pellets within zirconium cladding alloys that account for recent
research findings. New performance objectives and analytical limits may
be necessary for other fuel designs to take into consideration all
degradation mechanisms and any unique features of the particular fuel
system that the ECCS is trying to cool.
The proposed rule follows the general regulatory approach of the
existing regulations by establishing non-prescriptive, performance-
based regulatory language for demonstrating acceptable ECCS system
performance and determining the fuel's performance characteristics. The
organization and 10 CFR designations of the NRC's requirements
governing ECCS (currently in Sec. 50.46) and reactor cooling venting
systems (currently in Sec. 50.46a) are expected to change, as a result
of: (1) Ongoing rulemaking activities; (2) the proposed implementation
schedule for those activities; and (3) the need to maintain the current
requirements in place for those licensees that have not transitioned to
the new requirements (following the implementation schedule that would
be provided in the final rule). A detailed description of the
transition of 10 CFR designations is provided in Section VI, ``Section-
by-Section Analysis,'' of this document.
A. Applicability of Performance-Based Rule: Consideration of PRM-50-71
The NRC proposes to expand the applicability of the rule from
``uranium oxide pellets within cylindrical zircaloy or ZIRLO\TM\
cladding'' to any LWR, regardless of fuel design or cladding material.
The proposed rule would be applicable to applicants for and holders of
construction permits, operating licenses, combined licenses, and
standard design approvals and to applicants for certified designs and
for manufacturing licenses. The rule would not apply to any licensee
that has submitted certifications for permanent cessation of operations
and permanent removal of fuel from the reactor vessel, in accordance
with Sec. 50.82(a)(1).
Over the past 10 years, the NRC has granted exemptions from the
requirements of Sec. 50.46 (in accordance with Sec. 50.12(a)) to
licensees utilizing approved fuel designs with M5 zirconium-based alloy
cladding and, more recently, to licensees using approved fuel designs
with Optimized ZIRLO\TM\ zirconium-based alloy cladding.
The proposed rule includes general performance requirements for
future LWR fuel designs and specific performance requirements for the
current generation of LWR fuel designs with zirconium-based alloy
claddings. As such, it is anticipated that future exemption requests
would not be necessary for loading an advanced fuel design or cladding
material approved by the NRC through a rulemaking. However, the
licensee would still need to submit a license amendment. During this
approval process the NRC would determine whether, either: (1) Specified
and NRC-approved analytical limits have been established, along with an
NRC-approved ECCS evaluation model, which satisfy the specific
performance-based requirements for fuel designs consisting of uranium
oxide or mixed uranium-plutonium oxide pellets within zirconium-based
alloy cladding material; or (2) specified performance objectives and
associated analytical limits which take into consideration all
degradation mechanisms and any unique features of the particular fuel
system have been established, along with an NRC-approved ECCS
evaluation model, by which to judge the ECCS performance for new fuel
designs.
The NRC recognizes that a small number of fuel rods may experience
cladding failuare (i.e., small perforation) during normal operation due
to manufacturing defects, debris fretting, grid-to-rod fretting, etc.
The allowable number of fuel rod failures during normal operation is
not governed by ECCS performance requirements, but limited by 10 CFR
part 20, ``Standards for Protection against Radiation,'' and plant
Technical Specifications, which limit reactor coolant activity level to
maintain on-site and off-site dose during normal operation, anticipated
operational occurrences, and postulated accidents to within prescribed
limits. In addition to Technical Specifications limitations, plant
administrative limits on reactor coolant activity level further reduce
the potential number of failed fuel rods within an operating core.
Due to secondary degradation effects, the performance of these
limited failed fuel rods during a postulated LOCA may be difficult to
predict, and would most likely be outside the experimental database
used to set the NRC-approved analytical limits for coolable geometry
(i.e., cladding embrittlement for zirconium-based alloys). However, due
to their limited number relative to the total core population, any
unforeseen degradation or performance during a postulated LOCA would
not challenge the general performance requirements. As such, compliance
with ECCS performance requirements of Sec. 50.46c is not required for
this limited number of failed fuel rods.
This proposed extension to all LWR fuel types addresses PRM-50-71,
which requested that the applicable regulations be amended to allow for
the introduction of advanced zirconium-based alloy claddings, thus
eliminating the need for a licensee to pursue an exemption for alloys
which did not meet the definition of ``zircaloy or ZIRLO\TM\.'' If the
NRC adopts the proposed rule in final form, PRM-50-71 would be granted
and resolved.
B. Performance-Based Aspects of the Proposed Rule
The systems, structures, and components of the ECCS are designed to
provide residual heat removal during and following a postulated LOCA.
Failure of the ECCS to perform its intended function would result in a
loss of coolable geometry followed by core reconfiguration. While the
principal ECCS performance requirements are simple in nature (i.e.,
remove residual heat and maintain core temperatures at acceptable
levels), the system must be
[[Page 16114]]
designed to achieve specified performance objectives, taking into
consideration all degradation mechanisms and any unique features of the
particular fuel system that the ECCS is intended to cool. Sufficient
empirical data must be available for the particular fuel system to
identify all degradation mechanisms (e.g., embrittlement, loss of
structural integrity) and any unique features (e.g., eutectic or
exothermic reactions, combustible gas generation) to specify both
acceptable core temperatures and the duration for which the ECCS must
remove residual heat. In addition, fuel-specific analytical
requirements may be necessary to accurately or conservatively model
unique phenomena that impact the ECCS performance demonstration (e.g.,
fuel rod balloon and burst, cladding inside-diameter oxygen ingress).
To achieve the NRC's goal of a more performance-based rule,
significant changes in format and structure are being proposed relative
to Sec. 50.46. In place of the current prescriptive Sec. 50.46(b)
analytical limits, the proposed rule would define the following
principal ECCS performance requirements:
Core temperature during and following the LOCA event does
not exceed the analytical limits for the fuel design used for ensuring
acceptable performance. This ensures that the fuel maintains a coolable
geometry.
Sufficient cooling so that decay heat will be removed for
the extended period of time required by the long-lived radioactivity
remaining in the core so that long-term cooling is ensured.
Complying with these performance requirements provides reasonable
assurance that the overall objective of maintaining a coolable core
geometry in the event of a LOCA is met. In addition, the proposed rule
would dictate specific analytical requirements for demonstrating
compliance with the ECCS performance requirements. For instance, to
demonstrate compliance with these system performance requirements, ECCS
performance would be evaluated using fuel-specific performance
objectives and associated analytical limits that take into
consideration all degradation mechanisms and unique features of the
particular fuel system, along with an NRC-approved evaluation model.
The proposed rule includes specific performance requirements for
fuel designs consisting of uranium oxide or mixed uranium-plutonium
oxide fuel pellets within cylindrical zirconium-alloy cladding. These
performance requirements incorporate the findings of the NRC LOCA
research program. New performance objectives and analytical limits may
be necessary for other fuel designs.
For uranium oxide or mixed uranium-plutonium oxide fuel pellets
within cylindrical zirconium-alloy cladding, all known degradation
mechanisms and unique features have been identified, specific
performance objectives have been defined, and fuel design-specific
performance requirements have been established and included in the
proposed rule. For this fuel system design, the performance objective
is to maintain the coolable fuel rod bundle array. In other words, the
objective is to maintain fuel pellets within the cladding and fuel rods
within the fuel bundle lattice. Existing ECCS models and methods are
capable of accurately predicting core temperatures and demonstrating
ECCS performance, provided this core configuration is maintained. To
achieve this performance objective, the ECCS must limit core
temperatures to prevent high-temperature cladding failure, prevent
brittle cladding failure (i.e., maintain PQD and prevent breakaway
oxidation), minimize hydrogen gas generation, and provide for long-term
residual heat removal for the long-lived fission decay products
associated with uranium oxide or uranium-plutonium oxide fuel.
The following Sec. 50.46(b) requirements would remain unchanged in
the proposed Sec. 50.46c:
Peak cladding temperature. The calculated maximum fuel
element cladding temperature shall not exceed 2200[emsp14][deg]F. The
peak cladding temperature requirements currently in Sec. 50.46(b)(1)
would be moved to Sec. 50.46c(g)(1)(i).
Maximum hydrogen generation. The calculated total amount
of hydrogen generated from the chemical reaction of the cladding with
water or steam shall not exceed 0.01 times the hypothetical amount that
would be generated if all of the metal in the cladding cylinders
surrounding the fuel, excluding the cladding surrounding the plenum
volume, were to react. The maximum hydrogen generation limits currently
in Sec. 50.46(b)(3) would be moved to Sec. 50.46c(g)(1)(iv).
In the current regulations, the preservation of cladding ductility,
via compliance with regulatory criteria on peak cladding temperature
(Sec. 50.46(b)(1)) and local cladding oxidation (Sec. 50.46(b)(2)),
provides a level of assurance that fuel cladding will not experience
gross failure and that the fuel rods will remain within their coolable
lattice arrays. The recent LOCA research program identified new
cladding embrittlement mechanisms that demonstrated that the current
combination of peak cladding temperature (2200 [deg]F (1204 [deg]C))
and local cladding oxidation (17 percent equivalent cladding reacted
(ECR)) criteria may not always ensure PQD. The impact of these research
findings on cladding ductility is addressed in the following section.
1. Hydrogen-Enhanced Beta-Layer Embrittlement
As explained in Section 1.4 of NUREG/CR-6967, oxygen diffusion into
the base metal under LOCA conditions promotes a reduction in the size
(referred to as beta-layer thinning) and ductility (referred to as
beta-layer embrittlement) of the metallurgical structure within the
cladding that provides its macroscopic mechanical behavior. The
presence of hydrogen within the cladding enhances this embrittlement
process.
It is important to recognize that the embrittlement of the cladding
is the result of oxygen diffusion into the base metal and not directly
related to the rate of growth or overall thickness of a zirconium
dioxide layer on the outside cladding diameter. In combination with a
limit on peak cladding temperature, the current regulation limits
maximum local oxidation to preserve cladding ductility. Maximum local
oxidation is used as a surrogate to limit the ITT and associated oxygen
diffusion. This surrogate approach is possible because both the rate of
oxidation and rate of oxygen diffusion share strong temperature
dependence. In the recent LOCA research program, the Cathcart-Pawel
(CP) weight gain correlation was used to integrate time-at-temperature
and define the point at which ductility was lost (nil ductility).
Section 1.3 of NUREG/CR-6967 defines the following equations used to
integrate time-at-temperature:
[[Page 16115]]
[GRAPHIC] [TIFF OMITTED] TP24MR14.001
Measurements of weight gain were performed on many of the steam-
oxidized cladding samples tested in the LOCA research program. For
example, Table 22 of NUREG/CR-6967 provides both measured ECR and
calculated Cathcart-Pawel Equivalent Cladding Reacted (CP-ECR) for the
zircaloy-2 cladding samples tested. Instead of correlating measured
plastic strain or measured offset displacement with measured ECR or
measurements of the post-quench cladding microstructure (e.g., beta
layer thickness), the research findings correlate the ductile-to-
brittle transition to calculated CP-ECR (using the equations previously
stated). In this instance, calculated ECR is used to integrate time-at-
temperature and requires knowledge of measured ECR. However, an
accurate or conservative weight gain model based on measured oxidation,
which may be alloy-specific or vary significantly from CP predictions,
needs to be used for predicting rate of energy release and hydrogen
generation from the metal/water reaction in the LOCA heat balance
calculation.
In an attempt to more accurately characterize the degrading
phenomenon, the proposed rule would replace the term ``maximum local
oxidation'' with ``ITT,'' which more directly relates to the parameter
of interest (i.e., embrittlement due to oxygen diffusion). This should
clarify the need to have: (1) An accurate or conservative weight gain
correlation based on measured oxidation for estimating the rate of
energy release and hydrogen generation from the metal/water reaction,
and (2) a consistent analytical technique to integrate time-at-
temperature in both the empirical database (i.e., allowable CP-ECR) and
evaluation model (i.e., predicted CP-ECR).
During normal operation, the cladding metal absorbs some hydrogen
from the corrosion process. When that cladding is exposed to high-
temperature LOCA conditions, the elevated hydrogen levels increase the
solubility of oxygen in the beta phase and the rate of diffusion of
oxygen into the beta phase. Therefore, even for LOCA temperatures below
1204 [deg]C (2200[emsp14][deg]F), embrittlement can occur for time
periods corresponding to less than 17-percent oxidation in corroded
cladding with significant hydrogen pickup.
Figure 1 illustrates the effect of hydrogen on ring-compression
test ductility measurements. Test specimens included high-burnup (a 71-
to 74-micrometer corrosion-layer thickness) and as-fabricated (fresh)
PWR Zircaloy-4 cladding segments. Cladding samples were oxidized on two
sides at approximately 1200 [deg]C (~2200 [deg]F) and cooled at
approximately 11 [deg]C per second to 800 [deg]C (1472 [deg]F). As-
fabricated samples were quenched at 800 [deg]C, whereas the high-burnup
samples were slow-cooled from 800 [deg]C to room temperature.
Figure 1 plots ECR (a parameter correlated with oxygen pickup from
the steam) as calculated by the CP-ECR kinetics correlation vs. the
offset strain accommodated before cracking in ring compression testing.
The offset strain before cracking indicates sample ductility and an
offset strain less than 2 percent is considered brittle. Multiple ring
compression tests were conducted using rings that had been oxidized to
a range of CP-ECR levels from 0-16 percent. The results indicate that
high burnup cladding material embrittles more rapidly than fresh
material. For these tests, an ECR of 7 percent (where the high burnup
material indicated brittle behavior) corresponds to a total (integral)
oxidation time of ~155 seconds, while an ECR of 14 percent (where the
fresh material first indicated brittle behavior) corresponds to ~300
seconds.
[[Page 16116]]
[GRAPHIC] [TIFF OMITTED] TP24MR14.002
To address this phenomenon (as well as to achieve a more
performance-based rule), the NRC proposes to replace the existing
prescriptive analytical limits with a performance-based requirement
that would require licensees to establish specified and NRC-approved
analytical limits on PCT and ITT. These limits should correspond to the
measured ductile-to-brittle transition for the zirconium-based alloy
cladding based upon an NRC-approved experimental technique. If the peak
cladding temperature that preserves cladding ductility is lower than
the 2200 [deg]F limit, the licensee should use the lower temperature.
The NRC is issuing draft regulatory guide DG-1263 for comment. The
draft regulatory guide provides licensees with ``specified and NRC-
approved analytical limits on PCT and ITT,'' based upon the NRC's LOCA
research program's measured ductile-to-brittle transition for
zirconium-based alloy cladding. In addition, the NRC is issuing DG-1262
for comment, which provides licensees with ``an NRC-approved
experimental technique'' for conducting PQD measurements and developing
analytical limits. These DGs specify an approach acceptable to the NRC.
Even if the draft regulatory guides are adopted in final form,
licensees may propose alternative approaches to those described in
those regulatory guides.
It is important to recognize that a consistent integration
technique should be used to quantify time at elevated temperature in
both the experiments and evaluation model. For example, the NRC-
approved analytical limits on ITT in DG-1263 were based on the NRC's
LOCA research program results, which, in turn, integrated time at
elevated temperature using the CP weight gain correlation. For
consistency with DG-1263, future LOCA analyses should integrate time at
elevated temperature using the same CP weight gain correlation when
comparing analysis results against these analytical limits. For this
case, appendix K to 10 CFR part 50 ECCS evaluation models would
continue to use the Baker-Just (BJ) weight gain correlation for
estimating the rate of energy release and hydrogen generation from the
metal/water reaction.
The NRC's LOCA research program did not investigate cladding
degradation mechanisms or develop the technical basis for performance-
based requirements beyond the existing 2200[emsp14][deg]F peak cladding
temperature criterion. Examples of degradation mechanisms beyond
cladding embrittlement (via oxygen diffusion) include excessive
exothermic metal-water reaction, alloy-specific eutectics, and loss of
fuel rod geometry due to plastic flow. As a result, the existing
2200[emsp14][deg]F limit (specified in Sec. 50.46c(g)(1)(i) of the
proposed rule) remains an absolute upper limit for zirconium-based
alloys on PCT. However, as reflected in this proposed requirement, a
lower PCT may be required to preserve ductility.
2. Oxygen Ingress From Cladding Inside Diameter
Oxygen sources may be present on the inner surface of irradiated
cladding due to gas-phase UO3 transport prior to gap
closure, fuel-cladding-bond formation (uranium dioxide in solid
solution with zirconium dioxide), and the fuel bonded to this layer.
Under LOCA conditions, this available oxygen may diffuse into the base
metal of the cladding, effectively reducing the integral time-at-
temperature to nil ductility.
To address this phenomenon, the NRC proposes to add an analytical
requirement to the ECCS evaluation model that would require licensees
to, if an oxygen source is present on the inside surfaces of the
cladding at the onset of a LOCA, consider the effects of oxygen
diffusion from the cladding inside surfaces in the ECCS evaluation
model.
The NRC recognizes that the availability of a cladding inside
diameter (ID) oxygen source and its diffusion into the base metal
during a postulated LOCA may depend on several factors (e.g., rod
design, power history). As such, applicants are responsible for
determining when the fuel-cladding bonding layer is strong enough to
allow the diffusion of oxygen from the uranium-oxide fuel to the
zirconium cladding and, therefore, must be included in the ECCS
evaluation model. It is anticipated that identifying the magnitude and
onset of oxygen ID diffusion would be part of the NRC's review and
approval of LOCA
[[Page 16117]]
evaluation models or vendor fuel designs. A conservative analytical
limit is provided in draft regulatory guide DG-1263.
3. Breakaway Oxidation
As explained in Section 1.4.5 of NUREG/CR-6967, zirconium dioxide
can exist in several crystallographic forms (allotropes). The normal
tetragonal oxide that develops under LOCA conditions is dense,
adherent, and protective with respect to hydrogen pickup. However,
there are conditions that promote a transformation to the monoclinic
phase (i.e., the phase that is grown during normal operation), which is
neither fully dense nor protective. The tetragonal-to-monoclinic
transformation is an instability that initiates at local regions of the
metal-oxide interface and grows rapidly throughout the oxide layer.
Because this transformation results in an increase in oxidation rate,
it is referred to as breakaway oxidation. Along with this increase in
oxidation rate resulting from cracks in the monoclinic oxide,
significant hydrogen pickup also occurs. Hydrogen that enters in this
manner during a LOCA transient promotes rapid embrittlement of the
cladding.
While all zirconium alloys will eventually experience breakaway
oxide phase transformation when exposed to long durations of high-
temperature steam oxidation, alloying composition and manufacturing
process (e.g., surface roughness) influence the timing of this
phenomenon.
Any fuel rod that experiences breakaway oxidation during a
postulated LOCA will rapidly become brittle and more susceptible to
gross failure and hence, is no longer in compliance with General Design
Criteria (GDC)-35 requirements for coolable core geometry. To address
this phenomenon, the NRC proposes to add a performance-based
requirement that the licensee measure the onset of breakaway oxidation
for each reload batch on manufactured cladding material and report any
changes in the onset of breakaway oxidation at least annually. This
requirement, along with a periodic test requirement, would confirm that
slight composition changes or manufacturing changes have not
inadvertently altered the cladding's susceptibility to oxidation. The
NRC is issuing DG-1261, which will provide licensees with ``an NRC
approved experimental technique'' for conducting breakaway oxidation
measurements and developing analytical limits. Even if the draft
regulatory guide is finalized, licensees may also provide an
alternative approach to that proposed in the draft regulatory guide.
4. Applicability of Ductility-Based Analytical Limits in the Burst
Region
During a postulated LOCA, a portion of the fuel rod population may
be predicted to experience fuel rod ballooning and cladding rupture as
a result of rapid depressurization of the reactor coolant system in
combination with elevated cladding temperature. The number of burst
rods depends on several variables including initial conditions (e.g.,
fuel rod design, rod internal pressure, rod power) and accident
conditions (e.g., break size, cladding temperature). This flawed
section of the fuel rod may experience degradation mechanisms beyond
oxygen diffusion embrittlement encountered in the remaining portions of
the fuel rod, including significant amounts of hydrogen uptake from
steam entering the fuel rod through the rupture.
Consistent with the technical basis of the proposed rule, DG-1262
describes an NRC-approved experimental technique for defining the
ductile-to-brittle transition. This experimental procedure involves
measuring ductility using ring compression testing performed on small,
unflawed segments of fuel rod cladding previously exposed to steam
oxidation at a defined peak cladding temperature and the integrated
time at temperature profile (expressed as CP-ECR). While this
experimental approach captures embrittlement of the zirconium metal due
to oxygen diffusion and the effects of pre-existing hydrogen on the
rate of embrittlement, it does not capture all of the degradation
mechanisms experienced in the region of the fuel rod surrounding a
cladding rupture. In addition to embrittlement due to oxygen ingress
(which is doubled in the burst region due to steam entering cladding
rupture), the burst region experiences cladding wall thinning, cladding
rupture, and increased hydrogen uptake (hydrogen absorbed from
zirconium oxidation on the cladding ID). All of these degradation
mechanisms impact the performance of the fuel rod under LOCA
conditions. As such, the ductile-to-brittle transition based on ring
compression tests of unflawed cladding segments may not fully represent
the region of the fuel rod surrounding the cladding rupture.
The rupture region contains non-uniform distributions of: (1)
Oxygen concentration within the base metal and zirconium oxide
thickness, (2) soluble hydrogen and zirconium hydrides, (3) cladding
wall thickness (due to ballooning), and (4) cladding flaws (due to
ballooning and rupture). The overall goal of preserving cladding
ductility may not apply to the rupture area that contains non-uniform
distributions of flaws, cladding thickness, hydrogen distribution, and
oxygen levels.
To investigate the mechanical behavior of ruptured fuel rods, the
NRC conducted integral LOCA testing, designed to exhibit ballooning and
burst, on as-fabricated and hydrogen-charged cladding specimens and
high-burnup fuel rod segments exposed to high-temperature steam
oxidation followed by quench. The research results and conclusions are
documented in the report ``Mechanical Behavior of Ballooned and
Ruptured Cladding'' (ADAMS Accession No. ML12048A475). The integral
LOCA testing confirms that continued exposure to a high-temperature
steam environment weakens the already flawed region of the fuel rod
surrounding the cladding rupture. Hence, limitations on PCT and ITT are
necessary to preserve an acceptable amount of mechanical strength and
fracture toughness. In addition, this research demonstrated that the
degradation in strength and fracture toughness with prolonged exposure
to steam oxidation was enhanced with pre-existing cladding hydrogen
content.
The research findings from the integral LOCA research presented the
NRC with two options for revising the fuel performance requirements:
(1) Establish a separate performance requirement within the burst
region (i.e., analytical limits that preserve sufficient fracture
toughness to ensure burst region survival), or (2) apply the ductility-
based analytical limits to the entire fuel rod.
In the absence of a credible analysis of loads, cladding stresses,
and cladding strains for a degraded LOCA core, there are no absolute
metrics to determine how much ductility or strength would be needed to
``guarantee'' that fuel-rod cladding would maintain its geometry during
and following LOCA quench. It is also not clear what impact severance
of some fuel rods into two pieces would have on core coolability.
Fragmentation of fuel rod cladding would be more detrimental to core
coolability than severance of rods into two pieces. Even minimal
ductility ensures that cladding will have high strength and toughness
and therefore, high resistance to fracturing. Brittle cladding, on the
other hand, might fail at low strength and shatter. Therefore, the
intent to maintain ductility is beneficial even without adequate
knowledge of LOCA loads. If wall thinning and double-sided oxidation
are accounted for, then it was determined that applying the hydrogen-
[[Page 16118]]
based embrittlement limit developed in previous work at ANL to limit
oxidation in the balloon region of the irradiated fuel rods tested at
Studsvik was sufficient to preserve reasonable behavior of the
ballooned and ruptured region.
The integral LOCA research concluded that application of the
hydrogen-dependent ductility-based analytical limits on PCT and ITT
(when applied within the burst region) preserve the mechanical behavior
of high-burnup rods tested to that measured for as-fabricated cladding
oxidized to 17 percent CP-ECR. Assuming highly conservative upper
bounds on thermal expansion loading during quench, the residual
mechanical behavior preserved by this limit was determined to be
adequate to demonstrate that coolable geometry is maintained. As such,
the NRC elected the second regulatory approach to apply a single
performance-based requirement to the entire fuel rod. This decision
recognizes that portions of the cladding within the burst region may
not maintain ductility. This decision is reflected in DG-1263 and
supported by the technical basis documented in the staff report, ``The
Mechanical Behavior of Ballooned and Ruptured Cladding'' (ADAMS
Accession No. ML12048A475).
5. Long-Term Cooling
The current regulation in Sec. 50.46(b)(5) requires that for long-
term cooling the calculated core temperature be maintained at an
acceptably low value following any calculated successful initial
operation of the ECCS. It also requires that decay heat be removed for
the extended period of time required by the long-lived radioactivity
remaining in the core.
The proposed rule would define a performance-based requirement to
ensure acceptable fuel performance during long-term cooling.
Specifically, the proposed rule would require that a specified and NRC-
approved analytical limit on peak cladding temperature be established
that corresponds to the measured ductile-to-brittle transition for the
zirconium-based alloy cladding material based upon an NRC-approved
experimental technique. It would also require that the calculated
maximum fuel element temperature should not exceed the established
analytical limit.
6. Use of Risk-Informed Approaches To Address Debris for Long-Term
Cooling
The proposed rule would allow all entities to use an alternative
risk-informed approach to evaluate the effects of debris for long-term
cooling. The adverse effects of debris on ECCS performance have been
documented in the NRC's actions to resolve GSI-191, ``Assessment of
Debris Accumulation on PWR Sump Performance.'' Debris may cause
increased head loss across the ECCS and CSS pump suction strainer and
restrict the flow of water to the ECCS and CSS pumps. Debris may also
pass through the strainer and cause blockage of components or the core,
or damage to components downstream of the strainer. For these reasons,
the effects of debris on long-term ECCS cooling performance must be
evaluated. However, the NRC believes that risk-informed methodologies
have progressed to the point where the NRC may allow their use in
considering the effects of debris on the adequacy of long-term ECCS
cooling performance. The entity's application and the NRC's review and
approval of the application will close that entity's required actions
under GSI-191.
For the purpose of Sec. 50.46c provisions on the risk-informed
alternative to long-term cooling, debris is material within containment
that may be transported to the suction strainer(s) for the ECCS and
CSS. Debris includes (but is not limited to) loose materials that may
transport and materials that may be damaged by a LOCA jet to the extent
that they become transportable. Debris sources of interest typically
include insulation, coatings, dust, dirt, concrete, fire barrier
material, signs and tags, and materials left in containment; however,
debris may originate from other sources. Debris may also result from
chemical interactions that cause precipitation of materials. Debris may
cause increased head loss across the strainer and restrict the flow of
water to the ECCS and CSS pumps. Debris may also pass through the
strainer and cause blockage of components or the core, or damage to
components downstream of the strainer.
The proposed Sec. 50.46c provisions allowing a risk-informed
approach for evaluating the effects of debris on long-term cooling
performance would require that the defense-in-depth philosophy and
safety margins be maintained and, as a result, defense-in-depth and
safety margins must be explicitly considered. This consideration of
defense-in-depth and safety margins is consistent with the NRC's
general guidance regarding risk-informed decisionmaking contained in RG
1.174, ``An Approach for Using Probabilistic Risk Assessment in Risk
Informed Decisions on Plant Specific Changes in the Licensing Basis,''
Revision 2, dated May 2011 (ADAMS Accession No. ML100910006). The RG
1.174 provides guidance on an acceptable approach to risk-informed
decision-making, consistent with the Commission's Policy Statement on
the Use of Probabilistic Risk Assessment (PRA) dated August 16, 1995
(60 FR 42622). The RG sets forth a set of five key principles, four of
which are relevant to the proposed rule:
Maintain the defense in depth philosophy;
Maintain sufficient safety margins;
Any changes allowed must result in no more than a small
increase in core damage frequency or risk, consistent with the intent
of the Commission's Safety Goal Policy Statement; and
Incorporate monitoring and performance measurement
strategies.
The proposed rule is consistent with the defense in depth principle
of RG 1.174. Defense-in-depth has traditionally been applied in reactor
design and operation to provide multiple means of accomplishing safety
functions and to prevent the release of radioactive material. The
applicant would need to address the intent of the general design
criteria (or similar licensing basis design criteria), national
standards, and engineering principles (e.g., single failure criterion)
in evaluating the impact of the alternative approach on defense-in-
depth. Defense-in-depth is considered sufficient if the overall
redundancy and diversity among the plant's systems and barriers,
including the containment and its support systems, is sufficient to
ensure that the risk acceptance criteria of Sec. 50.46c(e)(1)(i) are
met, and the following attributes are maintained:
Reasonable balance is preserved among prevention of core
damage, prevention of containment failure or bypass, and mitigation of
consequences of an offsite release.
There is not an over-reliance on programmatic activities
to compensate for weaknesses in plant design.
System redundancy, independence, and diversity are
preserved commensurate with the expected frequency of challenges,
consequences of failure of the system, and associated uncertainties in
determining these parameters.
Defenses against potential common cause failures are
preserved and the potential for the introduction of new common cause
failure mechanisms are assessed and addressed.
Independence of barriers is not degraded.
Defenses against human errors are preserved.
The intent of the plant's design criteria is maintained.
Regarding the maintenance of sufficient safety margins, the
applicant would need to address the impact of implementing the
alternate approach on
[[Page 16119]]
current safety margins. Consistent with RG 1.174, Revision 2,
sufficient safety margins are considered to be maintained when:
Codes and standards or their alternatives approved for use
by the NRC are met.
Safety analysis acceptance criteria in the licensing basis
are met or proposed revisions provide sufficient margin to account for
analysis and data uncertainty.
The risk-informed provisions for considering the effects of debris
on long-term cooling would also require that any potential net increase
in risk from implementation of the risk-informed approach be assessed
and that reasonable confidence is provided that this change in risk is
small. The NRC regards ``small'' changes for plants with total baseline
core damage frequencies (CDF) of 10-\4\ per year or less to
be CDF increases of up to 10-\5\ per year and plants with
total baseline CDF greater than 10-\4\ per year to be CDF
increases of up to 10-\6\ per year. However, if there is an
indication that the CDF may be considerably higher than
10-\4\ per year, the focus of the applicant should be on
finding ways to decrease rather than increase CDF and the licensee may
be required to present arguments as to why steps should not be taken to
reduce CDF in order for the alternate approach to be considered. For
plants with total baseline large early release frequency (LERF) of
10-\5\ per year or less, small LERF increases are considered
to be up to 10-\6\ per year, and for plants with total
baseline LERF greater than 10-\5\ per year, small LERF
increases are considered to be up to 10-\7\ per year.
Similar to the CDF metric, if there is an indication that the LERF may
be considerably higher than 10-\5\ per year, the focus of
the licensee should be on finding ways to decrease rather than increase
LERF and the licensee may be required to present arguments as to why
steps should not be taken to reduce LERF in order for the alternate
approach to be considered. This perspective is consistent with the
guidance in Section 2.2.4 of RG 1.174, Revision 2.
Finally, Sec. 50.46c contains requirements that would ensure that
the plant-specific PRA is of sufficient scope, level of detail, and
technical adequacy for this approach and is updated and maintained over
time and that the risk-informed approach is evaluated periodically. The
technical adequacy of the plant-specific PRA would be assessed by the
NRC taking into account appropriate standards and peer review results.
The NRC has prepared an RG (RG 1.200, ``An Approach for Determining the
Technical Adequacy of Probabilistic Risk Assessment Results for Risk-
Informed Activities,'' dated March 2009 (ADAMS Accession No.
ML090410014)) on determining the technical adequacy of PRA results for
risk-informed activities. As one step in the assurance of technical
adequacy, the PRA must have been subjected to a peer review process
assessed against a standard or set of acceptance criteria that is
endorsed by the NRC. Therefore, the NRC staff would rely on the NEI
Peer Review Process, as modified in the NRC's approval, or the American
Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS)
Peer Review Process, as modified in the NRC's approval; both processes
are documented in RG 1.200. Changes and data, including: (1)
Operational practices; (2) the facility configuration; (3) plant and
industry experience; and (4) structure, system, and component (SSC)
performance would be required to be fed back into the PRA and the Sec.
50.46c risk-informed analyses and, when appropriate, adjustments would
be made to maintain the validity of these processes. In addition, Sec.
50.46c contains requirements for corrective action and reporting, to
the NRC, conditions where the established risk-informed approach
results exceed the risk acceptance criteria. Together, these
requirements would maintain the validity of the risk-informed approach
such that the risk-informed decisionmaking principles would continue to
be satisfied over the life of the facility.
In as much as Sec. 50.46c contains requirements that would (1)
provide reasonable confidence that any net risk increase from
implementation of its requirements is small; (2) maintain defense-in-
depth; (3) maintain safety margins; and (4) require the use of
monitoring and performance measurement strategies, the proposed rule is
consistent with the Commission's policy on the use of PRA for risk-
informed decision-making and, more importantly, would maintain adequate
protection of public health and safety.
Future Development of Draft Guidance for the Risk-Informed Alternative
South Texas Project Nuclear Operating Company (STPNOC) submitted a
letter of intent to pilot a risk-informed approach for addressing GSI-
191 (ADAMS Accession No. ML103481027) in December 2010. Subsequently,
the NRC received a pilot submittal from STPNOC on January 31, 2013
(ADAMS Accession No. ML13043A013), supplemented on June 19, 2013 (ADAMS
Accession No. ML131750250). In parallel with the NRC's review of the
application, the NRC will develop draft guidance for the risk-informed
alternative to address the effects of debris on long-term cooling. That
draft guidance will be published for comment upon completion, which is
currently anticipated for early- to mid-calendar year 2015. The NRC
will then evaluate public comments received on the draft guidance, and
develop the final guidance on a timeline that ensures all guidance
(both for the risk-informed alternative and the new proposed
embrittlement criteria) is available when the NRC staff provides the
final Sec. 50.46c rule to the Commission (currently scheduled for
February 2016).
C. Corrective Actions and Reporting Requirements
1. Peak Cladding Temperature and Equivalent Cladding Reacted
The ANPR identified the third objective of the rulemaking as the
revision of the LOCA reporting requirements. Specifically, the ANPR
indicated that the NRC considered revising the reporting criteria by
redefining what constitutes a significant change or error in such a
manner as to make the reporting requirements dependent upon the margin
between the acceptance criteria limits and the calculated values of the
respective parameters (i.e., PCT or CP-ECR). After reviewing the public
comments received, the NRC recognizes that the proposed reporting
requirements specified in the ANPR were complex, and might, as a
result, promote unnecessary burden or misinterpretation. As such, the
reporting requirements of this proposed rule would not incorporate a
dependence on margin between the acceptance criteria and calculated
parameters.
The proposed rule would add a reporting requirement and definition
of significant change or error based on predicted changes in maximum
local oxidation (i.e., ECR), reformat the reporting section to clarify
existing requirements, and add a reporting requirement based on
periodic breakaway oxidation measurements. Any changes or errors that
prolong the temperature transient may further challenge the ITT
analytical limit; however, they may not significantly change the
predicted PCT. As such, this change or error would not be captured in
the reporting requirements. To improve the reporting and evaluation of
changes or errors of this type, the NRC would expand the definition of
significant change or error to include maximum local oxidation. The
[[Page 16120]]
threshold for a significant change or error, 0.4 percent ECR, would be
equivalent to a change in calculated ECR for a 50[emsp14][deg]F change
in cladding temperature.
The definition of a significant change or error (i.e.,
50[emsp14][deg]F PCT, 0.4 percent ECR) is specific to zirconium-alloy
cladding. A new definition of significant change or error may be
necessary for other cladding materials. In addition, the proposed rule
would require the use of maximum local oxidation (i.e., percent ECR) to
evaluate the impact of a change or error on the predicted ITT.
Reporting requirements with respect to any ``change to or error
discovered in an NRC-approved ECCS evaluation model or in the
application of such a model'' have been a source of confusion. Two
common misconceptions are: (1) Baseline values when estimating a
significant change or error (i.e., greater than 50[emsp14][deg]F), and
(2) 30-day reporting including ``a proposed schedule for providing a
reanalysis.'' When estimating a significant change or error, the
proposed rule provides threshold values for both PCT and local
oxidation. The baseline predictions used to assess a significant change
or error should be the PCT and maximum local oxidation values
documented in a plant's updated final safety analysis report (UFSAR).
These values should represent the latest LOCA analyses that were
submitted and reviewed by the NRC staff as part of a license amendment
request (e.g., power uprate, fuel transition) as amended by prior
annual reports. The following example illustrates the NRC's position:
In 2007, a licensee submits new LOCA analyses as part of an
extended power uprate license amendment request with a predicted PCT
of 1900[emsp14][deg]F and maximum local oxidation (MLO) of 2.4
percent ECR. The 2008 and 2009 annual reports identify no changes or
errors. In 2010, two errors in the ECCS evaluation model are
discovered and documented in the annual report with an estimated
impact on PCT of +25[emsp14][deg]F and -20[emsp14][deg]F and
estimated impact on MLO of +0.08 percent ECR and -0.01 percent ECR.
A 30-day notification was not required since the estimated impact
was below the threshold for a significant change or error. At this
point, the licensee should update the UFSAR, document the error
notification, and identify the baseline for judging future changes
or errors as 1905[emsp14][deg]F PCT and 2.5 percent ECR.
When a change to or error in an ECCS evaluation model is
discovered, the licensee would be responsible for estimating the
magnitude of changes in predicted results to: (1) Determine if
immediate steps are necessary to demonstrate compliance or bring plant
design or operation into compliance with Sec. 50.46c requirements, and
(2) identify reporting requirements. Under the proposed rule, a
licensee's obligation to report and take corrective action varies
depending upon whether the licensee's situation falls into one of three
possible scenarios, as described in this document:
1. Change, error, or operation that does not result in any
predicted response that exceeds any acceptance criteria and is itself
not significant.
The licensee must:
a. Submit an annual report documenting the change(s), error(s), or
operation along with the estimated magnitudes of changes in predicted
results.
b. Revise the UFSAR.
c. Use the UFSAR PCT/ECR predictions as a baseline for future
evaluations.
2. Change, error, or operation that does not result in any
predicted response that exceeds any acceptance criteria but is
significant.
The licensee must:
a. Submit a 30-day report documenting the change(s), error(s), or
operation, estimated magnitudes of changes in predicted results, and
the schedule for providing a new analysis of record (AOR). The NRC will
review the new AOR.
b. Revise the UFSAR to include new AOR.
c. Use the UFSAR PCT/ECR predictions as a baseline for the future
evaluations.
3. Change, error, or operation that results in any predicted
response that exceeds acceptance criteria.
The licensee must:
a. Take immediate actions to bring the plant into compliance with
acceptance criteria.
b. Report the change, error, or operation under Sec. Sec.
50.55(e), 50.72, and 50.73, as applicable.
c. Submit a 30-day report documenting the change(s), error(s), or
operation, estimated magnitudes of changes in predicted results, and
the schedule for providing a new AOR. The NRC will review the new AOR.
d. Revise the UFSAR to include new AOR.
e. Use the UFSAR PCT/ECR predictions as the baselines for future
evaluations.
The proposed reporting requirements in Sec. 50.46c(m) reflect
reformatting of the current reporting provisions in order to separately
identify these three scenarios and clarify their respective
requirements.
The proposed rule would also add the requirement to report results
of breakaway oxidation measurements to the NRC. The licensees would be
required to measure breakaway oxidation prior to each reload batch, and
report the measurements within the calendar year following the testing.
The breakaway oxidation phenomenon is explained in detail in sub-
section B.3, ``Breakaway Oxidation'' of this section, ``Proposed
Requirements for ECCS Performance During LOCAs.'' This reporting
requirement would be specific to zirconium-alloy cladding and may not
be applicable to other cladding materials.
2. Risk-Informed Alternative To Address Debris for Long-Term Cooling
Section 50.46c(e) of the proposed rule would require reasonable
confidence that any calculated increase in CDF or LERF associated with
debris is small. In the context of this paragraph, the calculated
increases in CDF and LERF represent the difference between the as-
built, as-operated plant (accounting for the effects of debris) and the
``baseline'' plant where the effects of debris are assumed to be
negligible. This approach quantifies the portions of CDF and LERF
attributable to debris and designates them as [Delta]CDF and
[Delta]LERF. These metrics inform the NRC staff's decision on whether
the effects of debris are acceptably small and consistent with the
Commission's Safety Goal Policy Statement.
Subsequent changes to the plant or the PRA model may change the
baseline CDF and LERF values as well as [Delta]CDF and [Delta]LERF.
Because the NRC staff's original decision was based in part on these
metrics, subsequent changes to their values should be assessed to
ensure that the bases for this decision are still valid. It should be
noted that the cumulative effects of operating changes (including plant
modifications, procedural changes, and SSC performance) must be
maintained within the rule's risk acceptance criteria over the life of
the plant and, therefore, the evaluation of subsequent changes needs to
address the cumulative effect of these changes.
Therefore, the proposed rule contains a corrective action and
reporting requirement that would ensure that changes and errors are
evaluated, reported to the NRC (as appropriate), and corrected in a
timely manner (as appropriate). Consistent with the NRC's integrated
approach to decisionmaking, changes that can impact risk, defense-in-
depth, or safety margins need to be evaluated and, as appropriate,
reported to the NRC. These terms, while frequently used, can have
different definitions to different stakeholders. Therefore, the NRC
intends to ensure that licensees using the risk-informed
[[Page 16121]]
approach to debris update their UFSAR to list applicable plant-specific
capabilities of defense-in-depth and safety margins with respect to the
proposed rule.
In addition, the NRC's approval under Sec. 50.46c(e)(3) would
specify the circumstances under which the entity would be required to
notify the NRC of changes or errors in the risk evaluation approach
used to address the effects of debris on long-term cooling. This
requirement would ensure that if errors in the approach are identified
subsequent to the NRC approval or if the entity seeks to change
specific aspects of their approach that were determined by the NRC to
be important to the NRC approval, such as the scope or level of detail
of the PRA, these circumstances would be clearly identified in the
NRC's approval. These requirements would ensure conditions that result
in exceeding the Sec. 50.46c(e) acceptance criteria are identified,
corrected, and reported in a timely manner, and thus, ensure the
effects of debris on long-term core cooling continue to be
appropriately addressed.
The corrective action and reporting requirements for the aspects of
the rule related to entities using the risk-informed alternative
approach of Sec. 50.46c(e) would be established in Sec. 50.46c(m)(4).
The proposed rule recognizes that there are different corrective and
reporting requirements for different entities, as depicted in Table 1,
Corrective Actions and Reporting: Risk-Informed Approach.
Table 1--Corrective Actions and Reporting: Risk-Informed Approach
----------------------------------------------------------------------------------------------------------------
Entity (and applicable proposed Requirement to Requirement to Requirement to make necessary
requirement) re[dash]evaluate? report? changes?
----------------------------------------------------------------------------------------------------------------
Design certification applicant No (But known errors Yes (Submit Yes (Changes in amended
before issuance of final design and discoveries amended application).
certification rule (covered by must be corrected). application).
Sec. 50.46c(m)(4)(i)).
Design certification applicant No.................. Yes (Only if No.
during the period of validity referenced in a
under Sec. 52.55(a) and (b)-- COL; then within
not currently referenced in any 30 days).
combined operating license
(COL) application or COL
(covered by Sec.
50.46c(m)(4)(ii)).
Design certification applicant Yes................. Yes............... No.
during the period of validity
under Sec. 52.55(a) and (b)--
once referenced in a COL
application or COL (covered by
Sec. 50.46c(m)(4)(iii)).
Design certification renewal Yes................. Yes (as part of Yes.
applicant (covered by Sec. renewal
50.46c(m)(4)(iv)). application).
Combined license applicant No (But known errors Yes (Submit Yes (Changes in amended
(covered by Sec. and discoveries amended application).
50.46c(m)(4)(v)). must be corrected). application).
Combined license holder before No.................. Yes............... Yes.
finding under Sec. 52.103(g)
(covered by Sec.
50.46c(m)(4)(vi)).
Operating license holder or Yes................. Yes............... Yes.
combined license holder after
finding under Sec. 52.103(g)
(covered by Sec.
50.46c(m)(4)(vii)).
----------------------------------------------------------------------------------------------------------------
For design certification applicants (i.e., prior to issuance of the
final design certification rule), the proposed rule would require that,
if any errors are discovered, the applicant must submit a report to the
NRC within an amended application. That amended application would
describe any changes to the certified design and/or changes in the
analyses, evaluations, and modeling (including the debris evaluation
model and the PRA and its supporting analyses); and would demonstrate
that the acceptance criteria in Sec. 50.46c(e)(1) are met.
For design certification applicants during the period of validity
under Sec. 52.55(a) and (b) that are not currently referenced in any
COL application or COL, there would be no evaluation, reporting, or
change requirement. However, once the design certification is
referenced by a COL applicant, any information regarding compliance
with Sec. 50.46c(e)(1) must be reported in accordance with the
requirements in 10 CFR part 21.
For design certification applicants during the period of validity
under Sec. 52.55(a) and (b) that are referenced in a COL application
or COL, the proposed rule would require the design certification
applicant to evaluate and report any information concerning compliance
with the acceptance criterion of Sec. 50.46c(e)(1). However, there
would be no requirement to make changes to the analyses, evaluations,
and modeling until the time of renewal.
For design certification renewal applicants, the proposed rule
would require the applicant to re-evaluate the analyses, evaluation,
and modeling; report any changes or errors; and include in its
application any necessary changes to the certified design, debris
evaluation model, PRA, or supporting analyses to demonstrate that the
renewed certified design meets the acceptance criteria in Sec.
50.46c(e)(1).
For combined license applicants, the proposed rule would require
the applicant to report any errors that are discovered within 30 days
of the completion of that determination. The combined license
applicants would be required to report the errors and make any
necessary changes to the analyses, evaluation, or modeling within the
amended application.
For combined licenses before the finding under Sec. 52.103(g), the
proposed rule would require that any errors that are discovered be
updated in the analyses, evaluations, and modeling no later than the
scheduled date for initial fuel loading under Sec. 52.103(a). The
licensee must also confirm that the acceptance criteria of Sec.
50.46c(e)(1) continue to be met. Once this update is submitted, and
until the Commission has made the finding under Sec. 52.103(g), the
licensee shall re-perform the review to ensure the acceptance criteria
of Sec. 50.46c(e)(1) continue to be met in a timely manner; this
ensures that updating occurs if there are extended delays in the
scheduled date for initial fuel loading. If the licensee determines
that any acceptance criterion of Sec. 50.46c(e)(1) are not met, then
the licensee would be required to submit an application for amendment
of its
[[Page 16122]]
combined license and departure from a referenced design certification
rule, if applicable.
For operating licenses and combined licenses after the finding
under Sec. 52.103(g), the proposed rule would require that the
licensee re-evaluate the analysis, evaluation, and modeling by no later
than 48 months after the last review to confirm that the acceptance
criteria of Sec. 50.46c(e)(1) continue to be met. The licensee would
also be required to take action in a timely manner to bring the
licensee into compliance and report any failure to meet the acceptance
criteria of Sec. 50.46c(e)(1). Further, the amended application for
the combined license would be required to include a request for
exemption from a referenced design certification rule but would not
need to address the criteria for obtaining an exemption.
D. Consideration of PRM-50-84: Thermal Effects of Crud and Oxide Layers
Determination of PRM
This proposed rule would address issues raised in a PRM that was
submitted by Mark Leyse on March 15, 2007, and docketed as PRM-50-84.
The petition requests that the NRC conduct rulemaking in three specific
areas:
(1) Establish regulations that require licensees to operate light-
water power reactors under conditions that are effective in limiting
the thickness of crud and/or oxide layers on zirconium-clad fuel in
order to ensure compliance with Sec. 50.46(b) ECCS acceptance
criteria;
(2) Amend appendix K to 10 CFR part 50 to explicitly require that
the steady-state temperature distribution and stored energy in the
reactor fuel at the onset of the postulated LOCA be calculated by
factoring in the role that the thermal resistance of crud deposits and/
or oxide layers plays in increasing the stored energy in the fuel.
(These requirements also need to apply to any NRC-approved, best-
estimate ECCS evaluation models used in lieu of appendix K to 10 CFR
part 50 calculations); and
(3) Amend Sec. 50.46 to specify a maximum allowable percentage of
hydrogen content in [fuel rod] cladding.
On May 23, 2007 (72 FR 29802), the NRC published a notice of
receipt for this petition in the Federal Register and requested public
comment on the petition. The public comment period ended on August 6,
2007. After evaluating the public comments, the NRC decided that each
of the petitioner's issues should be considered in the rulemaking
process. On this basis, the NRC closed the docket on the petition for
rulemaking. The NRC's determination, and evaluation of public comments
received, was published in the Federal Register on November 25, 2008
(73 FR 71564).
Technical Issues in PRM-50-84
Licensees use approved fuel performance models to determine fuel
conditions at the start of a LOCA, and the impact of crud and oxidation
on fuel temperatures and pressures may be determined explicitly or
implicitly by the system of models used. With the addition of an
unambiguous regulatory requirement to address the accumulation of crud
and oxide during plant operation, the NRC believes that fuel
performance and LOCA evaluation models must include the thermal effects
of both crud and oxidation whenever their accumulation would affect the
calculated results. The NRC notes that licensees are required to
operate their facilities within the boundary conditions of the
calculated ECCS performance. During or immediately after plant
operation, if actual crud layers on reactor fuel are implicitly
determined or visually observed after shutdown to be greater than the
levels predicted by or assumed in the ECCS evaluation model, licensees
would be required to determine the effects of the increased crud on the
calculated results. In many cases, engineering judgment or simple
calculations could be used to evaluate the effects of increased crud
levels; therefore, detailed LOCA reanalysis may not be required. In
other cases, engineering judgment is used to determine that new
analyses would be performed to determine the effect the new crud
conditions have on the final calculated results. If unanticipated or
unanalyzed levels of crud are discovered, then the licensee must
determine if correct consideration of crud levels would result in a
reportable condition as provided in the relevant reporting paragraphs.
Should this proposed rule be adopted in final form, the NRC believes
this regulatory approach to address crud and oxide accumulation during
plant operation would satisfactorily address the issues raised by the
petitioner's first request.
The formation of cladding crud and oxide layers is an expected
condition at nuclear power plants. Although the thickness of these
layers is usually limited, the amount of accumulated crud and oxidation
varies from plant to plant and from one fuel cycle to another. Intended
or inadvertent changes to plant operational practices may result in
unanticipated levels of crud deposition. The NRC agrees with the
petitioner (the petitioner's second request) that crud and/or oxide
layers may directly increase the stored energy in reactor fuel by
increasing the thermal resistance of cladding-to-coolant heat transfer,
and may also indirectly increase the stored energy through an increase
in the fuel rod internal pressure. As such, to ensure that licensee
ECCS models properly account for the thermal effects of crud and/or
oxide layers that have accumulated during operations at power, the
proposed rule would add a requirement to evaluate the thermal effects
of crud and oxide layers that may have accumulated on the fuel cladding
during plant operation. If the NRC adopts the proposed rule in final
form, then the second request of PRM-50-84 would be resolved.
The petitioner's third request is for the NRC to establish a
maximum allowable percentage of hydrogen content in fuel rod cladding.
The purpose of this request is to prevent embrittlement of fuel
cladding during a LOCA. Although the NRC has decided not to propose the
specific rule language recommended by the petitioner, the proposed new
zirconium-specific requirements, if adopted in final form, would
address the petitioner's third request by considering cladding hydrogen
content in the development of analytical limits on integral time at
temperature.
The NRC believes that this proposed rule addresses each of the
three issues raised in PRM-50-84. If the NRC adopts the proposed rule
in final form, PRM-50-84 would be granted in part and resolved.
E. Implementation
The proposed rule would specify the dates for compliance with the
rule for existing operating license holders as well as holders of new
reactor construction permits, combined licenses, and applicants for
standard design certifications. The proposed rule sets forth a
staggered schedule for compliance with the final rule, depending upon
existing margin to the revised requirements with respect to
embrittlement and the anticipated level of effort to demonstrate
compliance. Apart from this staggered schedule for compliance, the rule
also allows licensees the alternative of voluntarily seeking to meet
the long-term cooling requirements of the proposed rule (and other
changes as permitted by the risk-informed alternative and noted in the
application) using a risk-informed approach, which could be
accomplished in advance of the date for compliance
[[Page 16123]]
with the rule as set forth in the staggered schedule.
1. Staggered Implementation Schedule
For existing operating nuclear power reactors, the proposed rule
includes a staged schedule for implementation. The NRC has developed
this staged implementation to improve the efficiency and effectiveness
of this migration toward the new ECCS requirements for the existing
operating fleet. As part of this plan, licensees have been divided
among three implementation tracks based upon existing margin to the
revised requirements and anticipated level of effort to demonstrate
compliance. The purpose of the staged implementation approach is to
bring licensees into compliance as quickly as possible, while
accounting for: (1) Differences between realistic and appendix K to 10
CFR part 50 LOCA models; and (2) the level of effort and scope of
analyses required for compliance. Table 2 provides an overview of the
implementation schedule for the existing fleet. Note that the
compliance schedule requirement represents the date that the licensee
submits either the letter report or license amendment request (as
opposed to the date of NRC approval). The proposed track assignments
for every operating reactor is provided in Table 1 of proposed Sec.
50.46c(o). Table 1 of proposed Sec. 50.46c(o) would be updated, as
necessary, to capture the implementation track assignments for all
operating reactors at the time the final rule is issued. Applications
for a 10 CFR part 50 operating license under review on the effective
date of the rule would be assigned an implementation track based on the
factors used in establishing the three tracks (as described in Table
1). An applicant for a new 10 CFR part 50 operating license submitted
or docketed after the effective date of the rule must comply with the
provisions of the rule. The NRC notes that Vermont Yankee Nuclear Power
Station is listed in the implementation track assignments. Although
Vermont Yankee submitted a notification of permanent cessation of power
operations under Sec. 50.82(a)(1)(i) (see ADAMS Accession No.
ML13273A204), that notification contained only an estimate of the date
of cessation. Vermont Yankee plans to supplement that letter with a
(firm) date of cessation, as required per Sec. Sec. 50.82(a)(1)(i) and
50.4(b)(8). Watts Bar, Unit 2, and Bellefonte, Units 1 and 2, have
construction permits in effect or in the process of being reinstated.
However, the ECCS margin to the proposed rule's requirements on
embrittlement for each of these plants is not yet known. (A final
safety analysis report (FSAR) has not been approved for these plants.)
The NRC will determine the appropriate track for each plant once its
ECCS margin to embrittlement is finalized. At that point, that plant
would be added to Table 1 of proposed Sec. 50.46c(o) in the
appropriate track, and the title of Table 1 would be modified
accordingly.
Table 2--Implementation Plan
--------------------------------------------------------------------------------------------------------------------------------------------------------
Number of units
Implementation track Basis Anticipated level of -------------------------- Compliance demonstration
effort BWR PWR
--------------------------------------------------------------------------------------------------------------------------------------------------------
1..................................... All plants which satisfy new Low..................... 27 37 No later than 24 months from
requirements without new effective date of rule.
analyses or model revisions.
2..................................... PWR plants using realistic Medium.................. 2 13 No later than 48 months from
large-break (LB) LOCA models effective date of rule.
requiring new analyses. BWR/
2 plants.
3..................................... PWR plants using appendix K Medium-High............. 6 15 No later than 60 months from
LB and small-break (SB) effective date of rule.
models requiring new
analyses. BWR/3 plants.
--------------------------------------------------------------------------------------------------------------------------------------------------------
To support the implementation of the proposed requirements on
individual plant dockets, fuel vendors would be encouraged to submit
for NRC review alloy-specific hydrogen uptake models and any LOCA model
updates (e.g., incorporation of CP weight gain correlation) no later
than 12 months from the effective date of the final rule. Upon
approval, these models and methods could be used to demonstrate the
ECCS performance against the new analytical limits. For Track 1 plants
that would not require new ECCS evaluations, licensees should complete
any necessary engineering calculations, update their plant UFSAR, and
provide a letter report to the NRC documenting compliance with Sec.
50.46c. The NRC recognizes that to demonstrate compliance, these plants
would need to utilize newly-approved hydrogen uptake models and
integrate time at temperature using the CP weight gain correlation (for
appendix K to 10 CFR part 50 models).
For any unit at a plant that would require a new ECCS evaluation,
including adopting a previously approved realistic evaluation model,
revising an existing evaluation model, performing a new LOCA break
spectrum analysis, performing a multiple rod survey (e.g., burnup-rod
power tradeoff), or making changes to a technical specification or core
operating limit report (COLR), licensees would need to submit the new
LOCA AOR and, where applicable, a license amendment request updating
the COLR list of approved methods.
The NRC has developed a phased implementation approach for
applicants and holders of standard design approvals, design
certifications, combined licenses, and manufacturing licenses granted
under 10 CFR part 52.
The proposed implementation plan for reactors approved under 10 CFR
part 52 would allow the applicant for a design certification, standard
design approval, or manufacturing license either submitted to, or
docketed by, the NRC prior to the effective date of the rule, to come
into compliance with the rule at the time of any application for
renewal.
An applicant for a design certification, standard design approval,
or manufacturing license submitted or docketed after the effective date
of the rule must comply with the provisions of the rule.
The holder of a combined license granted prior to the effective
date of the rule would be permitted to operate the plant for one fuel
cycle before demonstrating compliance with the rule. Doing so would
permit adequate time to submit demonstration of compliance with the
rule prior to
[[Page 16124]]
achieving fuel burnup for which the cladding limitations are imposed by
the rule. In this case the holder of the combined license would be
required to remain in compliance with the ECCS performance acceptance
criteria in place at the time the combined license was granted.
Applicants for combined licenses docketed after the effective date
of the rule must comply with the provisions of the rule.
The proposed rule reflects the NRC's determination that reactor
designs reviewed and approved under 10 CFR part 52 should have the same
constraints as the reactors operating under 10 CFR part 50 with respect
to development, submittal, and approval of ECCS performance models
necessary to demonstrate compliance with this rule. Alloy-specific
hydrogen uptake models and all ECCS performance model updates would be
expected to be submitted in a timely manner for NRC review and approval
so that demonstration of the ECCS performance with respect to the
analytical limits would not impact plant operation more than is
necessary.
The proposed rule also reflects the NRC's expectation that, for new
reactors licensed to operate prior to the effective date of the rule,
operation for at least the initial fuel cycle using fuel that has not
been analyzed under the proposed rule's provisions accounting for burn-
up effects does not present an adequate protection concern. During the
initial fuel cycle, the NRC believes that burn-up effects would not be
limiting, and the current ECCS rule's acceptance criteria are
sufficient during the initial fuel cycle to provide reasonable
assurance of adequate protection with respect to overall ECCS
performance.
2. Compliance With Long-Term Cooling Requirements Using Risk-Informed
Approach To Address Debris Effects
Implementation of the alternative approach to addressing the impact
of debris on long-term cooling is independent from implementation of
the requirements related to the embrittlement research findings. The
NRC would allow partial early implementation of the proposed
requirements of Sec. 50.46c, limited to this alternative approach. In
other words, an applicant may elect to submit its risk-informed
alternative under Sec. 50.46c(e) prior to demonstrating compliance
with the other requirements of Sec. 50.46c. In this case, the licensee
would have to receive NRC approval on both its risk-informed submittal
and the analytical limit for long-term cooling required under Sec.
50.46c(g)(1)(v) prior to using the risk-informed approach. The NRC is
proposing to allow early implementation because the NRC encourages
licensees to complete resolution of GSI-191 and this risk-informed
alternative is one way of resolving the issue.
The NRC has determined that a licensee's decision to use a risk-
informed methodology to evaluate the effects of debris on ECCS and CSS
with respect to long-term cooling following a LOCA should be reviewed
and approved by the NRC prior to implementation. The ECCS and CSS are
significant safety systems that provide necessary defense-in-depth. The
design bases for the ECCS are of high regulatory significance to the
NRC, as reflected in the detailed requirements applicable to the ECCS
(and the associated fuel system) in Sec. 50.46 and appendix K to 10
CFR part 50. In addition, the design bases for the ECCS and the CSS
affect the design bases for many other SSCs throughout the nuclear
power plant. Therefore, changes to the design assumptions for the ECCS
and CSS may have significant effects on the design bases for other SSCs
throughout the plant. These potential effects include changes in the
consequences of postulated accidents, margins of safety, and defense-
in-depth.
The NRC also determined that Sec. 50.59, properly implemented,
would not allow a change to the design bases of a plant to use a risk-
informed methodology for evaluating the effects of debris on long-term
cooling. A risk-informed methodology for addressing the effects of
debris on long-term cooling is a departure from the method of
evaluation described in the current UFSAR, as updated and used in
establishing the design bases in the safety analysis as defined in
Sec. 50.59(a)(2). Hence, under Sec. 50.59(c)(2)(viii), a licensee's
departure from the existing methodology for evaluating long-term
cooling must be reviewed and approved by the NRC as a license
amendment.
In sum, given the importance of the ECCS and CSS, the ``cascading''
effects of changes in ECCS and CSS design on the design bases of other
SSCs of a nuclear power plant, the NRC believes that a licensee's
decision to use a risk-informed methodology to evaluate the effects of
debris on ECCS with respect to long-term cooling should be reviewed and
approved by the NRC. Under the proposed rule, the NRC's review and
approval is accomplished through the license amendment process in
accordance with Sec. Sec. 50.90 through 50.92.
VI. Section-by-Section Analysis
The organization and 10 CFR designations of the NRC's requirements
governing emergency core cooling (currently in Sec. 50.46) and reactor
cooling venting systems (currently in Sec. 50.46a) are expected to
change. These changes would result from:
(1) The current schedule for Commission serial adoption of two
rulemakings: (i) The finalization of the proposed rule on risk-informed
changes to ECCS systems, currently referred to as the Sec. 50.46a
rulemaking, followed by; (ii) the finalization of this proposed rule on
performance-based changes to ECCS requirements and cladding acceptance
criteria, currently referred to as the Sec. 50.46c rulemaking;
(2) The proposed schedule for implementation of these rules; and
(3) The need to maintain current requirements in place for those
reactors that have not transitioned to the new requirements under the
implementation schedule to be specified in the final rule.
The following table shows how the organization and 10 CFR
designation of these rules will evolve, if the NRC sequentially adopts
the two final rules and licensees complete implementation of the
alternate cladding requirements. The NRC notes that, in an SRM, ``SRM-
SECY-10-0161--`Final Rule: Risk-Informed Changes to Loss-of-Coolant
Accident Technical Requirements (10 CFR 50.46a)','' dated April 26,
2012 (ADAMS Accession No. ML12117A121), the Commission approved the NRC
staff's request to withdraw SECY-10-0161, ``Risk-Informed Changes to
Loss-of-Coolant Accident Technical Requirements (10 CFR 50.46a),'' from
Commission consideration (ADAMS Accession No. ML121500380). The NRC
does not plan to publish a notice in the Federal Register withdrawing
the Sec. 50.46a proposed rule. The NRC staff plans to resubmit the
draft final rule for Commission consideration in conjunction with the
Near-Term Task Force (NTTF) Recommendation 1 activities. (For
information on NTFF Recommendation 1, see ``Recommendations for
Enhancing Reactor Safety in the 21st Century,'' dated July 12, 2011,
ADAMS Accession No. ML 112510271.) Therefore, the Sec. 50.46a
rulemaking still may be finalized before the Sec. 50.46c rulemaking,
as assumed in the following table.
[[Page 16125]]
----------------------------------------------------------------------------------------------------------------
Rulemaking and implementation activities
----------------------------------------------------------------------------
Existing NRC requirements and Adoption of final risk- Initial codification End of phased
proposed new regulations (bolded informed ECCS of final performance- implementation period for
rules are currently in effect) requirements (Sec. based fuel cladding performance-based cladding
50.46a) requirements requirements
----------------------------------------------------------------------------------------------------------------
Sec. 50.46 ECCS Acceptance Sec. 50.46 ECCS Sec. 50.46 ECCS Sec. 50.46 ECCS
Criteria. Acceptance Criteria Acceptance Criteria Acceptance Criteria (see
(unchanged). (unchanged). discussion for Sec.
50.46c under this column).
Risk-Informed ECCS Requirements Sec. 50.46a Risk- Sec. 50.46a Risk- Sec. 50.46a Risk-Informed
(currently designated in final Informed ECCS Informed ECCS ECCS Requirements.
rulemaking package as Sec. Requirements. Requirements.
50.46a).
Sec. 50.46a Reactor Coolant Redesignated as Sec. NA (Redesignation as NA (Redesignation as Sec.
Venting Systems. 50.46b. Sec. 50.46b 50.46b completed).
completed).
Performance-based ECCS and Cladding NA.................... Sec. 50.46c NA (Administrative
Requirements (currently designated Alternate Fuel rulemaking would: (i)
in draft proposed rulemaking Cladding Requirements. remove superseded fuel
package as Sec. 50.46c). cladding requirements in
Sec. 50.46, and (ii)
redesignate Sec. 50.46c
as Sec. 50.46.).
----------------------------------------------------------------------------------------------------------------
A. Section 50.46c--Heading
A new section, Sec. 50.46c, would be created in 10 CFR part 50 by
this rulemaking. The heading of Sec. 50.46c would be ``Emergency core
cooling system performance during loss-of-coolant accidents.''
B. Section 50.46c(a)--Applicability
Paragraph (a) would define the applicability of the proposed rule,
which remains limited to LWRs, but would be expanded beyond fuel
designs consisting of uranium oxide pellets within cylindrical zircaloy
or ZIRLO\TM\ cladding. The proposed rule would also be applicable to
applicants for and holders of construction permits, operating licenses,
combined licenses, and standard design approvals, and also to
applicants for standard design certifications and for manufacturing
licenses.
C. Section 50.46c(b)--Definitions
Paragraph (b) would provide definitions for terms used in this
section. The definitions of Loss-of-coolant accident and Evaluation
model would remain unchanged from those currently located in Sec.
50.46(c)(1) and (c)(2), respectively.
The definition of Breakaway oxidation and Debris evaluation model
would be added.
D. Section 50.46c(c)--Relationship to Other NRC Regulations
Paragraph (c) would describe the relationship of Sec. 50.46c to
other NRC regulations. The description in proposed paragraph (c) would
remain largely unchanged from that of the current regulation found in
Sec. 50.46(d). However, the description would be revised to make clear
that an approach approved by the NRC under Sec. 50.46c(e) may also be
used when evaluating the effects of debris to demonstrate compliance
with other requirements of this part, including GDC-35, GDC-38, and
GDC-41 (as allowed by Sec. 50.46c and requested in the application).
E. Section 50.46c(d)--Emergency Core Cooling System Design
Paragraph (d)(1) would define performance-based requirements for
the ECCS. Paragraph (d)(2) would require that ECCS performance be
demonstrated using an NRC-approved ECCS evaluation model meeting
specific requirements for a range of postulated LOCAs of different
sizes, locations, and other properties, sufficient to provide assurance
that the most severe postulated LOCA has been identified. The
provisions for a realistic ECCS model or appendix K to 10 CFR part 50
model would remain unchanged from the current regulation found in Sec.
50.46(a)(1)(i) and (ii), respectively. Similarly, the model requirement
that calculated changes in core geometry must be addressed would remain
unchanged from the current regulation found in Sec. 50.46(b)(4).
Paragraph (d)(2)(iii) would explicitly require that the ECCS evaluation
model address calculated changes in core geometry, and consider factors
that may alter localized coolant flow or inhibit delivery of coolant to
the core. Demonstration of ECCS performance in the post-accident
recovery period, or long-term cooling, is expected to consider
inhibition of core flow that can result from such factors as, but not
limited to, pump damage, piping damage, boron precipitation, and
deposition of debris and/or chemicals associated with the long-term
cooling mode of recirculation coolant collection from the reactor
building sump. Consideration of debris and/or chemical deposition is
already required by the current rule, and the proposed rule does not
alter the current efforts to address such factors under programs such
as GSI-191. Demonstration of consideration of such factors may also be
achieved through analytical models that adequately represent the
empirical data obtained regarding debris deposition. The proposed rule
would alternatively allow the use of risk-informed approaches to
evaluate the effects of debris on localized coolant flow and delivery
of coolant to the core during the long-term cooling (post-accident
recovery) period.
In addition, paragraph (d)(2)(iv) of the proposed rule would
specifically require that ECCS performance be demonstrated for both the
accident and the post-accident recovery and recirculation period.
Paragraph (d)(2)(v) would require that the ECCS model address the
fuel system modeling requirements in paragraph (g)(2) if the reactor
uses uranium oxide or mixed uranium-plutonium oxide pellets within
zirconium cladding (e.g., currently operating reactors).
Paragraph (d)(3) would provide the ECCS evaluation model
documentation requirements currently provided in appendix K, Section
II, ``Required Documentation.''
F. Section 50.46c(e)--Alternate Risk-Informed Approach for Addressing
the Effects of Debris on Long-Term Core Cooling
Paragraphs (d)(2)(iii) and (e) would allow entities to use a risk-
informed approach for addressing the effects of debris on long-term
core cooling. Paragraphs (e)(1)(i) through (e)(1)(iv) would provide the
acceptance criteria for an acceptable alternative risk-informed
approach for addressing the effects of debris on long-term core cooling
and would establish minimum requirements for the plant PRA and how it
is to be used in the alternate risk-informed approach. These proposed
requirements are intended to ensure that the implementation of the
alternate risk-informed approach to address debris
[[Page 16126]]
effects on long-term core cooling would provide reasonable confidence
that any resulting increase in CDF and LERF will be small, and that
sufficient defense-in-depth and safety margins are maintained. These
proposed requirements are consistent with the key principles of risk-
informed decisionmaking described in RG 1.174, Revision 2.
Paragraph (e)(1)(i) of the proposed rule would require that there
be reasonable confidence that any potential risk increase be small.
Paragraph (e)(1)(ii) would require that sufficient defense-in-depth and
safety margins be maintained as part of the implementation of the
alternate risk-informed approach. Further, paragraphs (e)(1)(iii) and
(iv) would contain the minimum requirements for the plant PRA and how
it is to be used in the alternate risk-informed approach.
Paragraph (e)(2) would require those applicants seeking to use the
alternative risk-informed approach under paragraph (e)(1) to submit an
application that contains the information provided in paragraphs
(e)(2)(i) through (e)(2)(v).
Paragraph (e)(2)(i) would require applicants to follow established
regulatory guidance that the NRC expects to finalize concurrent with
the final rule. If an applicant wishes to use a different approach, the
submittal must provide a sufficient description of how the alternative
risk-informed approach would be conducted and why it is acceptable.
Paragraph (e)(2)(ii) would require that initiating events from
sources both internal and external to the plant and for all modes of
operation, including low power and shutdown modes, be considered when
evaluating the effects of debris on long-term core cooling using the
alternate approach. This aspect of the rule recognizes that the minimum
PRA that would be required by paragraph (e)(1)(iv) may not address all
sources of initiating events and modes of operations, and as such,
other approaches may be used. Therefore, the application would need to
describe the measures taken to assure the scope, level of detail, and
technical adequacy of all the analyses performed to address severe
accidents are sufficient for this application and address the full
spectrum of initiating events and modes of operation.
Paragraph (e)(2)(iii) would specifically address the need to
provide the results of the PRA review process. This aspect includes
such items as any peer reviews performed, any actions taken to address
peer review findings that are important to the application, and any
efforts to compare the plant-specific PRA to the ASME/ANS PRA standard,
as endorsed by the NRC in RG 1.200.
In paragraph (e)(2)(iv), the applicant would be required to include
information about the evaluations they conduct to provide reasonable
confidence that any potential increase in risk would be small. The
applicant would be required to provide sufficient information to the
NRC, describing the evaluations and the basis for their acceptability
as appropriately representing the potential increase in risk from
implementation of the requirements in this rule.
In paragraph (e)(2)(v), the applicant would be required to provide
a description of the analytical limit on long-term peak cladding
temperature established in accordance with paragraph (g)(1)(v).
Paragraph (e)(3) would provide that the NRC may approve an
application to implement the alternative risk-informed approach if it
determines that the proposed approach satisfies the requirements of
paragraph (e)(1) and establishes an acceptable long-term peak cladding
temperature limit. The NRC staff would review the description of the
alternative risk-informed approach set forth in the application, and
the associated evaluations, to confirm that it contains the elements
required by the rule. The NRC staff would also review the information
provided about the plant-specific PRA and other systematic evaluations
used to evaluate severe accidents in support of the application to
assure that the scope, level of detail, and technical adequacy of the
analyses are commensurate with the reliance on the risk information.
This aspect of the review would involve the NRC assessment of the
information provided about: 1) the peer review process to which the
plant-specific PRA was subjected, 2) the reliance on other systematic
evaluations to address areas not covered by the plant-specific PRA, and
3) the approach for maintaining sufficient defense-in-depth and safety
margins. The NRC staff intends to use review guidance for this purpose.
The NRC's approval of the use of the risk-informed approach to address
long-term cooling would specify the circumstances under which the
entity would be required to notify the NRC of changes or errors in the
risk evaluation approach used to address the effects of debris on long-
term cooling. Depending upon the nature of the underlying application
(e.g., license, design certification rule, or design approval), the
approval and notification requirement will be implemented through a
license condition, a provision in the design certification rule, or a
condition of the design approval, as applicable.
Paragraph (f) would be added to reserve rulemaking space for future
amendments to Sec. 50.46c.
G. Section 50.46c(g)--Fuel System Designs: Uranium Oxide or Mixed
Uranium-Plutonium Oxide Pellets Within Cylindrical Zirconium-Alloy
Cladding
This section would be added to set forth fuel design specific
analytical limits and performance-based requirements by which to judge
the overall ECCS performance in accordance with paragraph (d)(1) for
LWRs using uranium oxide or mixed uranium-plutonium oxide pellets
within cylindrical zirconium alloy cladding. The fuel performance
criteria in paragraph (g)(1) and fuel system modeling requirements in
paragraph (g)(2) are based on the established degradation mechanisms
and performance objectives for this specific fuel type.
Paragraph (g)(1)(i) would establish an analytical limit on peak
cladding temperature to avoid cladding embrittlement, high temperature
failure modes, and run-away exothermic oxidation. Except as calculated
in paragraph (g)(1)(ii), the calculated maximum fuel element cladding
temperature should not exceed 2200[emsp14][deg]F. This requirement
remains unchanged from the current requirement at Sec. 50.46(b)(1).
Paragraph (g)(1)(ii) would require that the zirconium alloy
cladding maintains sufficient post-quench ductility in order to avoid
gross failure. This requirement replaces the current prescriptive
analytical limit, 17 percent ECR, in Sec. 50.46(b)(2).
Paragraph (g)(1)(iii) would be added to establish a performance-
based requirement to preclude breakaway oxidation in order to avoid
cladding embrittlement and gross failure. Breakaway oxidation is a new
requirement relative to Sec. 50.46(b).
Paragraph (g)(1)(iv) would establish an analytical limit on maximum
hydrogen generation to avoid an explosive concentration of hydrogen
gas. This requirement would be the same as that of the current
regulation in Sec. 50.46(b)(3).
Paragraph (g)(1)(v) would be added to establish a performance-based
requirement to ensure acceptable fuel performance during long-term
cooling. This performance requirement is consistent with the current
requirement to ``maintain the calculated core
[[Page 16127]]
temperature at an acceptably low value'' located in Sec. 50.46(b)(5).
Paragraph (g)(2) would establish fuel design specific modeling
requirements that are needed in addition to the generic ECCS evaluation
model requirements in paragraph (d)(2). Paragraph (g)(2)(i) would
require consideration of oxygen diffusion from the cladding inside
surface. This would be a new ECCS evaluation model requirement.
Paragraph (g)(2)(ii) would be added to include a requirement to
evaluate the thermal effects of crud and oxide layers that may have
accumulated on the fuel cladding during plant operation.
Paragraphs (h) through (j) would be added to reserve rulemaking
space for future amendments to Sec. 50.46c, including any changes that
stem from using newly designed fuel and cladding materials.
H. Section 50.46c(k)--Use of NRC-Approved Fuel in Reactor
Paragraph (k) would prohibit licensees from loading fuel into a
reactor, or operating the reactor, unless the licensee either
determines that the fuel meets the requirements in paragraph (d), or
complies with technical specifications governing lead test assemblies
in its license.
I. Section 50.46c(l)--Authority To Impose Restrictions on Operation
Paragraph (l) would provide that the Director of the Office of
Nuclear Reactor Regulation or the Director of the Office of New
Reactors may impose restrictions on reactor operation if it is found
that the evaluations of ECCS cooling performance submitted are not
consistent with the requirements of this section. The authority to
impose restrictions would be expanded, relative to the authority
currently granted in Sec. 50.46(a)(2), to address licenses issued
under 10 CFR part 52.
J. Section 50.46c(m)--Corrective Actions and Reporting
Paragraph (m) would provide reporting requirements applicable to
the ECCS evaluation model and reporting requirements applicable to
entities that elect to use the risk-informed alternative to address the
effects of debris on long-term cooling. Paragraphs (m)(1) through
(m)(3) would apply to all entities subject to Sec. 50.46c; paragraphs
(m)(4) would apply to those entities demonstrating acceptable long-term
core cooling under the provisions of paragraph (e).
Paragraph (m)(1) would establish required action and reporting
requirements if an entity identifies any change to, or error in, an
ECCS evaluation model or the application of such a model, or any
operation inconsistent with the evaluation model. For clarity, this
paragraph was divided into three categories of changes or errors, each
with its own proposed actions and reporting. These requirements are
unchanged from the current Sec. 50.46(a)(3), with the exception of
conforming to analytical limits established in the proposed rule.
Paragraph (m)(1)(i) would establish required action and reporting
requirements if an entity identifies any change to, or error in, an
ECCS evaluation model or the application of such a model, or any
operation inconsistent with the evaluation model, that does not result
in any predicted response that exceeds any acceptance criteria and is
itself not significant.
Paragraph (m)(1)(ii) would establish required action and reporting
requirements if a licensee identifies any change to, or error in, an
ECCS evaluation model or the application of such a model, or any
operation inconsistent with the evaluation model, that does not result
in any predicted response that exceeds any acceptance criteria but is
significant (as defined in paragraph (m)(2)).
Paragraph (m)(1)(iii) would establish required action and reporting
requirements for an entity who identifies any change to, or error in,
an ECCS evaluation model.
Paragraph (m)(1)(iv) would require an amendment to a design
certification application reflecting any reanalysis required by
paragraph (m)(1)(ii) to be submitted by the applicant in concert with
the reanalysis.
Paragraph (m)(2) would be added to provide the definition of a
significant change or error. The definition would be expanded, relative
to the 50[emsp14][deg]F change in calculated peak cladding temperature
in Sec. 50.46(a)(3)(i), to include a 0.4 percent ECR change in
calculated cladding oxidation.
Paragraph (m)(3) would require the onset of breakaway oxidation to
be measured for each reload batch, and would require any changes in the
time to the onset of breakaway oxidation to be assessed against the
integral time and to be reported annually. This would be a new
reporting requirement.
Paragraph (m)(4) would establish required action and reporting
requirements for entities choosing to implement the alternative risk-
informed approach for addressing the effects of debris on long-term
core cooling. Paragraph (m)(4) would specify the evaluation, reporting,
and change requirements for the various categories of entities that may
elect to use the risk-informed approach.
Paragraph (n) would be added to reserve rulemaking space for future
amendments to Sec. 50.46c.
K. Section 50.46(o)--Implementation
This section would establish the implementation requirements and
schedule for the existing fleet and for new reactors. Paragraph (o)(1)
would require construction permits under 10 CFR part 50 issued after
the effective date of the rule to comply with the requirements of Sec.
50.46c.
Paragraph (o)(2) would require operating licenses under 10 CFR part
50 based upon construction permits (including deferred and reinstated
construction permits) to comply with the requirements of Sec. 50.46c
by no later than the time frame established for operating reactors in
the implementation table. Until that point, the construction permits
identified by this paragraph must comply with Sec. 50.46.
Paragraph (o)(3) would require operating licenses under 10 CFR part
50 issued after the effective date of the rule to comply with the
requirements of Sec. 50.46c.
Paragraph (o)(4) would require operating licenses under 10 CFR part
50 (as of the effective date of the rule) to comply with the
requirements of Sec. 50.46c by no later than the applicable date set
forth in the implementation table for operating reactors.
Paragraph (o)(5) would require standard design certifications,
standard design approvals, and manufacturing licenses under 10 CFR part
52, whose applications (including applications for amendment) are
docketed after the effective date of the rule (including branches of
these certifications whose applications are docketed after the
effective date of the rule), to comply with the provisions of the rule.
Applicants submitting after the rule has been adopted should have had
ample time to develop and receive approval for the analysis methods
necessary to comply with the provisions of the rule.
Paragraph (o)(6) would require standard design certifications under
10 CFR part 52 issued before the effective date of the rule to comply
no later than the time of renewal of certification. Similar to the
requirements of paragraph (o)(5), such applicants will have had ample
time necessary to comply with the provisions of the rule.
Paragraph (o)(7) would require standard design certifications,
standard design approvals, and manufacturing licenses, along with new
branches of certifications under 10 CFR part 52
[[Page 16128]]
whose applications are pending as of the effective date of the rule to
comply with Sec. 50.46c no later than the time of renewal. Those
entities that are in the approval process at the time the rule becomes
effective will be required to comply at time of renewal. This will
provide ample time to develop and receive approval for the
methodologies necessary to comply with the rule. Paragraph (o)(8) would
require combined license applications under 10 CFR part 52 that are
docketed after the effective date of the rule to comply with the
provisions of the rule.
Paragraph (o)(9) would require applications for combined licenses
under 10 CFR part 52 that are docketed or issued after the effective
date of the rule to comply with Sec. 50.46c no later than completion
of the first refueling outage after the initial fuel load. Those
entities that are issued combined licenses prior to the effective date
of the rule must comply with the rule no later than the first refueling
outage after initial fuel load. This affords those entities ample time
to develop and submit the necessary methodologies.
Entities that elect to use the voluntary alternative to the long-
term cooling requirements of the proposed rule using a risk-informed
approach can do so in advance of the date for compliance with the rule.
In this case, the entity would have to receive NRC approval on both its
risk-informed submittal and the analytical limit for long-term cooling
required under Sec. 50.46c(g)(1)(v) prior to using the risk-informed
approach.
L. Appendix K to Part 50 of Title 10 of the Code of Federal Regulations
(10 CFR) ECCS Evaluation Models
In appendix K, a new paragraph II.6 would be added to clarify that,
for those entities that have implemented Sec. 50.46c, the requirements
for documentation are located within Sec. 50.46c(d)(3).
M. Redesignation of Venting Requirements in Sec. 50.46a
This proposed rule would redesignate the current Sec. 50.46a,
``Acceptance criteria for reactor coolant system venting systems,'' as
proposed Sec. 50.46b. A new Sec. 50.46a would be added and reserved
for future use as the rulemaking to provide a risk-informed alternative
to the LOCA technical requirements.
N. Changes Throughout 10 CFR Parts 50 and 52
Several administrative changes would be made throughout 10 CFR
parts 50 and 52 in order to conform with the proposed rule and proposed
redesignation of the venting requirements in current Sec. 50.46a.
Section 50.8 would be amended to add the proposed rule to the list of
approved information collections. Where Sec. Sec. 50.34(a)(4),
50.34(b)(4), 52.47(a)(4), 52.79(a)(5), 52.137(a)(4), and 52.157(f)(1)
refer to Sec. 50.46, the proposed rule would add ``and Sec. 50.46c,
as applicable.'' Where Sec. Sec. 50.34(a)(4), 52.47(a)(4),
52.79(a)(5), 52.137(a)(4), and 52.157(f)(1) refer to Sec. 50.46a, the
proposed rule would instead refer to Sec. 50.46b.
Changes are also made to GDC-35, GDC-38, and GDC-41 in appendix A
to 10 CFR part 50 to promulgate the acceptability of using a risk-
informed alternative for long-term cooling when demonstrating
compliance with these regulations, as allowed by Sec. 50.46c and
requested in the application.
VII. Specific Request for Comments on the Proposed Rule
In addition to the request for general comments on the proposed
rule, the NRC also requests specific comments on the following topics:
A. Fuel Performance Criteria
NRC Question 1. Performance-Based Peak Cladding Temperature Limit.
The NRC is proposing, in Sec. 50.46c(g)(1)(i), to maintain the
existing prescriptive criterion on PCT for zirconium alloy cladding.
Limits on cladding temperature are necessary to protect against a loss
of coolable geometry resulting from brittle failure upon quench, to
protect against high-temperature ductile failure, and to prevent
reaching the point at which the zirconium-water reaction would become
autocatalytic. In the original Sec. 50.46 rulemaking, the
2200[emsp14][deg]F limit on PCT was based on cladding embrittlement
(i.e., protection against brittle failure upon quench), which was
determined to be more limiting than either high temperature ductile
failure or autocatalytic oxidation. The NRC's LOCA research program did
not investigate cladding degradation mechanisms or develop the
technical basis for performance-based requirements beyond the existing
2200[emsp14][deg]F PCT criterion. Since the cladding embrittlement
mechanism, oxygen diffusion, is strongly dependent on temperature,
there exists an upper temperature at which the allowable time duration
to nil ductility approaches zero (i.e., PCT [deg]limit). As described
in Section V.B.1 of this document, recent research has confirmed that
2200[emsp14][deg]F remains an appropriate upper limit to protect
against cladding embrittlement since nil ductility is achieved rapidly
at higher temperature. As such, the proposed Sec. 50.46c maintains the
2200[emsp14][deg]F prescriptive PCT criterion.
The NRC requests comment on the proposed rule's retention of the
prescriptive PCT criterion, specifically:
a. In place of the prescriptive PCT criterion, should the NRC adopt
performance-based requirements for zirconium alloy cladding to protect
against high temperature ductile failure and autocatalytic oxidation?
b. Do established testing procedures already exist for
demonstrating acceptable high temperature cladding performance and
defining acceptance criteria to meet these new performance-based
requirements?
NRC Question 2. Periodic Breakaway Testing. To address the
breakaway oxidation phenomenon, the NRC proposes to add a performance-
based requirement in Sec. 50.46c(m)(3) that the licensee measure the
onset of breakaway oxidation periodically on manufactured cladding
material and report any changes in the onset of breakaway oxidation at
least annually. This requirement, along with a periodic test
requirement (defined as each reload batch in the proposed rule
language), would confirm that slight composition changes or
manufacturing changes have not inadvertently altered the cladding's
susceptibility to breakaway oxidation. The NRC is considering adopting,
as a final rule, a requirement that each licensee measure breakaway
oxidation behavior for each re-load batch. The NRC requests specific
comment on the type of data reported and the proposed frequency of
required testing. The objective of periodic testing is to prevent
affected fuel from being loaded into a reactor. At the same time, the
objective is to do so without adding ineffective requirements and
unnecessary burden. Other sampling approaches may be more effective.
For example, should the licensee be required to report data relevant
solely to their reload fuel batch or should the licensee be able to
report representative data based on periodic testing (e.g., test every
10,000 rods, tubing lot, or ingot) of the same zirconium-based alloy
cladding compiled during the period from the last report?
NRC Question 3. Analytical Long-Term Peak Cladding Temperature
Limit. Section 50.46c(g)(1)(v) of the proposed rule would require that
a specified and NRC-approved limit on long-term peak cladding
temperature be established which preserves a measure of cladding
ductility throughout the period of long-term demonstration (e.g., 30
days). The current regulation at Sec. 50.46(b)(5) stipulates that
long-term temperature be maintained ``at an acceptably low
[[Page 16129]]
value.'' The proposed rule would define the performance-based metric to
judge an acceptably low temperature. The overall goal of preserving
ductility would provide reasonable assurance that the fuel rods will
maintain their coolable bundle array. The NRC is requesting input
regarding this performance objective to determine if this is the most
suitable performance-based metric to demonstrate long-term cladding
performance.
Alternatively, the proposed rule could establish an analytical
limit of long-term fuel rod cladding temperature related to observed
corrosion behavior. For example, the Pressurized Water Reactor Owners
Group (PWROG) has applied as a long-term core cooling acceptance
criterion that the cladding temperature be maintained below
800[emsp14][deg]F (see Topical Report (TR) Westinghouse Commercial
Atomic Power (WCAP)-16793-NP, Revision 2, ``Evaluation of Long-Term
Cooling Considering Particulate, Fibrous and Chemical Debris in the
Recirculating Fluid,'' Appendix A (ADAMS Accession No. ML11292A021)).
Doing so will ensure that additional corrosion and hydrogen pickup over
a 30-day period will not significantly affect cladding properties. The
NRC seeks comment on the acceptance criterion for long-term cooling and
whether there is justification for a different temperature limit (other
than the 800[emsp14][deg]F provided in the WCAP).
B. Risk-Informed Alternative To Address the Effects of Debris
NRC Question 4. Acceptance Criteria for Risk-Informed Alternative.
Section 50.46c(e) of the proposed rule contains the high-level
acceptance criteria for an alternative that would allow entities to
use, on a case-by-case basis, a risk-informed approach to address the
effects of debris on long-term core cooling. In addition, the NRC will
develop draft regulatory guidance for this provision concurrent with
the staff's review of the STPNOC's pilot application for a risk-
informed approach to address the closely related topic of GSI-191. The
NRC seeks comment on whether the detailed acceptance criteria should be
set forth in Sec. 50.46c, or in the associated regulatory guidance.
NRC Question 5. Regulatory Approach for Risk-Informed Regulation.
The NRC seeks comment on whether the risk-informed alternative offered
by this regulation should require meeting numeric-risk acceptance
criteria as a matter of compliance (similar to Sec. 50.48c) or whether
other risk-informed approaches that use risk-importance insights to
establish measurable criteria or performance objectives, such as those
in use by Sec. Sec. 50.62, 50.63, and 50.65, or approaches using both
risk importance and numeric-risk acceptance criteria, such as those in
use by Sec. 50.69, would be preferable.
NRC Question 6. Operational Modes Considered in Risk-Informed
Alternative. Deterministic evaluations of GSI-191 are currently
required only for those modes of operation where both recirculation
from the sump is relied upon and the plant accident can cause high
pressure jets that can result in generation and transport of debris to
the sump. By contrast, probabilistic evaluations generally consider all
modes of operation. The NRC seeks comment on whether the risk-informed
approach provided in Sec. 50.46(e) could generically exclude any plant
operational modes (e.g., low power or shutdown) from consideration. If
so, what are the bases for excluding these operational modes from
consideration?
NRC Question 7. Reporting Criteria for the Risk-Informed
Alternative. The NRC is proposing in Sec. 50.46c(m) corrective actions
and reporting criteria specific to the risk-informed approach for
addressing the effects of debris on long-term cooling. These criteria
are performance-based and similar in concept to the reporting criteria
in Sec. 50.69. Per proposed Sec. 50.46c(m), the NRC's approval of the
entity's risk-informed application would specify the circumstances
under which the licensee or design certification applicant shall notify
the NRC of changes or errors in the risk evaluation approach. In
addition, the proposed rule would require entities to review the
analyses, evaluations, and modeling for changes and errors and
incorporate changes to the design, plant, operational practices, and
operation experience. The entity would then be required to update the
debris evaluation model and the PRA and its supporting analyses, and
re-perform the evaluations of risk, defense-in-depth, and safety
margins to confirm the acceptance criteria for the risk-informed
approach continue to be met. The NRC seeks specific comment on the
reporting criteria for the risk-informed approach.
Alternatively, the NRC seeks comment on whether the reporting
criteria for the risk-informed approach should be more prescriptive and
establish requirements similar to those for the ECCS model (i.e., Sec.
50.46c(m)(1) through (m)(3)). For instance, should the rule establish
values for changes in [Delta] CDF, [Delta] LERF, defense-in-depth, and
safety margins that would trigger specific reporting actions? If so,
what values should reporting criteria establish as reporting triggers
and what are the bases for selecting those values?
NRC Question 8. Exemptions Needed to Implement the Risk-Informed
Alternative. One objective of the proposed rule is to allow entities to
submit a risk-informed alternative to address the effects of debris on
long-term core cooling without the need to submit an exemption request.
The NRC identified that, in order to eliminate the need for an
exemption, changes may be necessary in GDCs 35, 38, and 41, as provided
in the proposed rule. The NRC seeks input on whether conforming changes
to other regulations would be necessary or desirable. Such conforming
changes may avoid the need for entities wishing to use the risk-
informed alternative to request exemptions from those regulations in
order to effectively implement the risk-informed alternative. If you
believe it is necessary or desirable to provide a conforming change to
a regulation in order to avoid an exemption from that regulation, then
please identify the specific regulation (and specific regulatory
provisions, if applicable) for which a conforming change would be made,
either the language of the change or a description of the conforming
change's objective, and the reason(s) why an exemption would otherwise
be needed if the NRC did not make a conforming change to that
regulation.
C. Implementation
NRC Question 9. Staged Implementation. The NRC is proposing, in
Sec. 50.46c(o), a staged implementation plan for the proposed rule. As
part of this plan, licensees have been divided among three
implementation tracks based upon existing margin to the revised
requirements and anticipated level of effort to demonstrate compliance.
The NRC requests specific comment on the staged implementation plan,
track assignments, or alternative means to implement the requirements
of the proposed rule.
NRC Question 10. New Reactor Implementation. The NRC is proposing,
in Sec. 50.46c(o)(5) through (9), an implementation approach that
takes into account design certifications, standard design approvals,
manufacturing licenses, and combined licenses and their status in
relation to the effective date of the rule. The proposed implementation
plan for new reactors would allow applicants for a design
certification, standard design approval, and manufacturing license
under review at the time of the effective date of the rule to come into
compliance with the rule at time of renewal. The holder of a combined
license issued prior to the
[[Page 16130]]
effective date of the rule would be permitted to operate the plant for
one fuel cycle before coming into compliance with the rule. Therefore,
the NRC is proposing to recognize that new reactors may operate for the
initial fuel cycle with fuel for which the burnup effects being
accounted for in the rule would not be a consideration. Applications
for design certifications, standard design approvals, manufacturing
licenses and combined licenses submitted after the effective date of
the rule would be expected to be in compliance with the rule at the
time of approval.
The NRC is requesting input regarding this implementation proposal,
including suggestions for alternate approaches.
D. Other Issues
NRC Question 11. Re-structuring 10 CFR Chapter I with respect to
ECCS Regulations. The NRC is considering restructuring its ECCS
regulations as part of the finalization of this rulemaking due to: (1)
Commission direction to include in the proposed rule a provision
allowing licensees to use a risk-informed submittal to address the
effects of debris during the long-term recovery period; and (2) the
potential benefit and efficiency of collocating all ECCS-related
requirements within the CFR. As such, the NRC seeks comment on the
following potential administrative changes:
Codify the performance-based ECCS and cladding
requirements (as proposed in this document) as a new section, Sec.
50.181.
Reserve Sec. 50.183 for the potential future risk-
informed ECCS requirements rule (currently referred to as the draft
final Sec. 50.46a rule).
Codify the requirements for the risk-informed submittals
(proposed as Sec. 50.46c(e) in this proposed rule) to address the
effects of debris in the long-term recovery period as a new section,
Sec. 50.185.
Duplicate the content of appendix K to 10 CFR part 50,
ECCS evaluation models, and add the content as a new section, Sec.
50.187. (The NRC notes that appendix K to 10 CFR part 50 will remain in
place until all licensees have implemented the proposed requirements
(i.e., until completion of the proposed staged implementation period).)
If this restructure is pursued, following the completion
of the proposed staged implementation period, the NRC would make the
following administrative changes:
[cir] Remove the current Sec. 50.46, ECCS acceptance criteria, in
its entirety.
[cir] Remove the current appendix K to 10 CFR part 50, in its
entirety. (The content will exist as Sec. 50.187.)
[cir] Redesignate the current Sec. 50.46a, ``Acceptance criteria
for reactor coolant system venting systems,'' as Sec. 50.46.
The tables that follow depict the described potential changes:
----------------------------------------------------------------------------------------------------------------
Rulemaking and implementation activities
---------------------------------------------------------------------------
Existing NRC requirements and End of phased Finalization of
proposed new regulations (bolded Initial codification of implementation period risk[dash]informed ECCS
rules are currently in effect) final performance-based for performance-based requirements (currently
fuel cladding fuel cladding referred to as draft
requirements requirements final Sec. 50.46a)
----------------------------------------------------------------------------------------------------------------
Sec. 50.46 ECCS Acceptance Sec. 50.46 ECCS Removed from 10 CFR Removed from 10 CFR
Criteria. acceptance criteria Chapter I in its Chapter I in its
(no change). entirety. entirety.
Sec. 50.46a Reactor Coolant NO CHANGE.............. Sec. 50.46.......... Sec. 50.46.
Venting Systems.
Draft final rule: Sec. 50.46a Risk- See Note 1............. See Note 1............ Sec. 50.183 Risk-
Informed ECCS Requirements. informed emergency core
cooling system
requirements.
Performance-based ECCS and cladding Sec. 50.181 Emergency Sec. 50.181......... Sec. 50.181.
requirements (currently designated core cooling system
in draft proposed rulemaking performance during
package as Sec. 50.46c). loss[dash]of[dash]cool
ant accidents.
Requirements for risk-informed Sec. 50.185 Sec. 50.185 Sec. 50.185.
submittals to address effects of Requirements for risk- Requirements for risk-
debris in the long-term post-quench informed submittals to informed submittals
cooling period (currently address effects of to address effects of
designated in draft proposed debris in the long- debris in the long-
rulemaking package as Sec. term post-quench term post-quench
50.184). cooling period. cooling period.
Appendix K to 10 CFR part 50: ECCS Appendix K to 10 CFR Sec. 50.187 ECCS Sec. 50.187.
Evaluation Models. part 50: ECCS evaluation models.
Evaluation Models.
And....................
Sec. 50.187 ECCS
evaluation models.
See Note 2.............
----------------------------------------------------------------------------------------------------------------
Note 1: The staff plans to submit the draft final Sec. 50.46a rulemaking package to the Commission following
completion of NTTF Recommendation 1 activities. At this time, it is uncertain whether finalization of the
draft final Sec. 50.46a rule would occur before the finalization of the proposed Sec. 50.46c rule.
Note 2: Until all licensees have implemented the proposed requirements (i.e., the proposed staged implementation
is complete), appendix K to 10 CFR part 50, ``ECCS Evaluation Models,'' and Sec. 50.187, ``ECCS Evaluation
Models,'' would coexist.
Should this restructure be pursued, the following table depicts the
structure of 10 CFR part 50 after finalization of the Sec. 50.46a
Risk-Informed ECCS Requirements and after the proposed staged
implementation of the Sec. 50.46c Performance-based ECCS and Cladding
Requirements rulemaking is complete:
------------------------------------------------------------------------
Section Title
------------------------------------------------------------------------
Sec. 50.46......................... Reactor coolant venting systems.
Sec. 50.181........................ Emergency core cooling system
performance during loss-of-
coolant accidents (Sec.
50.46c).
Sec. 50.183........................ Risk-informed emergency core
cooling system requirements
(Sec. 50.46a).
Sec. 50.185........................ Requirements for risk-informed
submittals to address effects of
debris in the long-term post-
quench cooling period.
[[Page 16131]]
Sec. 50.187........................ ECCS evaluation models (appendix
K to 10 CFR part 50).
------------------------------------------------------------------------
The NRC acknowledges that such changes could have a large impact on
licensees and vendors with regard to procedures, plans, programs,
topical reports, and engineering calculations that reference appendix K
to 10 CFR part 50 and the current ECCS regulations. In your comments,
please include the estimated cost for conforming changes to topical
reports, licensing amendments, and other technical documents. Please
also comment on whether the anticipated benefits and efficiencies would
outweigh the administrative burden, costs, and complexities.
NRC Question 12. Cumulative Effects of Regulation. The cumulative
effects of regulation (CER) consist of the challenges licensees face in
addressing the implementation of new regulatory positions, programs,
and requirements (e.g., rulemaking, guidance, generic letters,
backfits, inspections). The CER is manifested in several ways,
including the total burden imposed on licensees by the NRC from
simultaneous or consecutive regulatory actions that can adversely
affect the licensee's capability to implement those requirements while
continuing to operate or construct its facility in a safe and secure
manner. Consistent with SECY-11-0032, ``Consideration of the Cumulative
Effects of Regulation in the Rulemaking Process,'' dated March 2, 2011
(ADAMS Accession No. ML110190027), the NRC is requesting comments on
CER with respect to this proposed rulemaking. The NRC's consideration
of CER will be based, in part, on the NRC's confirmation of the safe
operation for each operating reactor, as described in Section III,
``Operating Plant Safety,'' of this document.
During the development of this proposed rulemaking, the NRC engaged
external stakeholders through multiple public meetings, an ANPR, and
solicitation of public comments. Additionally, the proposed rule would
establish a staged implementation plan, which would reduce the overall
implementation burden on licensees.
With regard to CER, the NRC requests specific comment on the
proposed rule's implementation schedule in light of any existing CER
challenges, specifically:
a. Do the proposed rule's effective date, compliance date, and
submittal dates provide sufficient time to implement the new proposed
requirements, including changes to programs, procedures, and the
facility, in light of any ongoing CER challenges?
b. If there are ongoing CER challenges, what do you suggest as a
means to address this situation (e.g., if more time is required for
implementation of the new requirements, what time period is
sufficient)?
c. Are there unintended consequences (e.g., does the proposed rule
create conditions that would be contrary to the proposed rule's purpose
and objectives)? If so, what are the unintended consequences?
d. Please comment on the NRC's cost and benefit estimates in the
proposed rule regulatory analysis (ADAMS Accession No. ML12283A188).
Specifically, please comment on the vendor hydrogen uptake and LOCA
model costs, costs of PQD and breakaway testing, and licensee analysis
costs.
VIII. Request for Comment: Draft Regulatory Guidance
The NRC is seeking public comment on three regulatory guides: DG-
1261, ``Conducting Periodic Testing for Breakaway Oxidation Behavior''
(ADAMS Accession No. ML12284A324); DG-1262, ``Testing for Post Quench
Ductility'' (ADAMS Accession No. ML12284A325); and DG-1263,
``Establishing Analytical Limits for Zirconium-Based Alloy Cladding''
(ADAMS Accession No. ML12284A323). You can access these documents as
described in Section IX, ``Availability of Documents,'' of this
document, or online at http://www.nrc.gov/reading-rm/doc-collections/.
The proposed rule would add the requirement (see Sec.
50.46c(g)(1)(iii)) to measure the onset of breakaway oxidation for a
zirconium cladding alloy based on an acceptable experimental technique.
The proposed rule also calls for the evaluation of the measurement
relative to emergency core cooling system performance (see Sec.
50.46c(g)(1)(iii)), and periodic testing and reporting of the values
measured (see Sec. 50.46c(m)(3)). The DG-1261 describes an
experimental technique acceptable to the NRC staff to measure the onset
of breakaway oxidation in order to support a specified and acceptable
limit on the total accumulated time that a cladding may remain at high
temperature, as well as a method acceptable to the NRC to implement the
periodic testing and reporting requirements in the proposed rule.
The proposed rule would also require licensees to establish
analytical limits on peak cladding temperature and time at elevated
temperature corresponding to the measured ductile-to-brittle transition
for the zirconium-alloy cladding material (see Sec. 50.46c(g)(1)(i)
and (ii)). The DG-1262 describes an experimental technique that is
acceptable to the NRC for measuring the ductile-to-brittle transition
for a zirconium-based cladding alloy. The DG-1263 provides a method of
using experimental data to establish regulatory limits.
You may submit comments on the draft regulatory guides as indicated
in the ADDRESSES section of this document.
IX. Availability of Documents
The NRC is making the documents identified in the following table
available to interested persons through one or more of the methods
provided in the ADDRESSES section of this document:
----------------------------------------------------------------------------------------------------------------
Document PDR ADAMS Web
----------------------------------------------------------------------------------------------------------------
SECY-98-300 ``Options for Risk-Informed Revisions to 10 CFR part X ML992870048 ............
50--Domestic Licensing of Production and Utilization
Facilities,'' dated December 23, 1998...........................
Petition for Rulemaking submitted by David J. Modeen on behalf of X ML003723791 ............
the Nuclear Energy Institute requesting amendment of 10 CFR
50.44 and 50.46.................................................
Federal Register Notice (65 FR 34599), ``Petition for Rulemaking X ML081780439 X
filed by David J. Modeen, Nuclear Energy Institute;
Consideration of Petition in the Rulemaking Process''...........
SRM-SECY-02-0057, ``Update to SECY-01-0133, `Fourth Status Report X ML030910476 X
on Study of Risk-Informed Changes to the Technical Requirements
of 10 CFR part 50 (Option 3) and Recommendations on Risk-
Informed Changes to 10 CFR 50.46 (ECCS Acceptance Criteria),'''
dated March 31, 2003............................................
[[Page 16132]]
Petition for Rulemaking submitted by Mark Edward Leyse re X ML070871368 X
addressing corrosion of fuel cladding surfaces and a change in
the calculations for a loss-of-coolant accident.................
Federal Register Notice (72 FR 28902), ``Mark Edward Leyse; X ML071290466 X
Receipt of Petition for Rulemaking''............................
Federal Register Notice (73 FR 71564), ``Mark Edward Leyse; X ML082240164 X
Consideration of Petition in Rulemaking Process''...............
NUREG/CR-6967, ``Cladding Embrittlement During Postulated Loss-of- X ML082130389 X
Coolant Accidents''.............................................
Research Information Letter (RIL)-0801, ``Technical Basis for X ML081350225 X
Revision of Embrittlement Criteria in 10 CFR 50.46''............
Summary of September 24, 2008, Public Workshop on Technical Basis X ML083010496 ............
GL-1985-022, ``Potential for Loss of Post-LOCA Recirculation X ML031150731 ............
Capability Due to Insulation Debris Blockage,'' dated December
3, 1985.........................................................
RG 1.82, ``Sumps for Emergency Core Cooling and Containment Spray X ML111680318 ............
Systems, Revision 0,'' dated June 1974..........................
Bulletin 95-02, ``Unexpected Clogging of a Residual Heat Removal X ML082490807 ............
Pump Strainer While Operating in Suppression Pool Cooling
Mode,'' dated October 7, 1995...................................
Bulletin 96-03, ``Potential Plugging of Emergency Core Cooling X ML082401219 ............
Suction Strainers by Debris in Boiling Water Reactors,'' dated
May 6, 1996.....................................................
Completion of Staff Reviews of NRC Bulletin 96-03, ``Potential X ML012970229 ............
Plugging of Emergency Core Cooling Suction Strainers by Debris
in Boiling-Water Reactors,'' and NRC Bulletin 95-02,
``Unexpected Clogging of a Residual Heat Removal (RHR) Pump
Strainer While Operating in Suppression Pool Cooling Mode,''
dated October 18, 2001..........................................
Bulletin 2003-01, ``Potential Impact of Debris Blockage on X ML031600259 ............
Emergency Sump Recirculation at Pressurized Water Reactors,''
dated June 9, 2003..............................................
GL 2004-02, ``Potential Impact of Debris Blockage on Emergency X ML042360586 ............
Recirculation During Design Basis Accidents at Pressurized Water
Reactors,'' dated September 13, 2004............................
SECY-10-0113, ``Closure Options for Generic Safety Issue--191, X ML101820296 ............
Assessment of Debris Accumulation on Pressurized Water Reactor
Sump Performance,'' dated August 26, 2010.......................
SRM-SECY-10-0113, dated December 23, 2010........................ X ML103570354 ............
SECY-12-0093, ``Closure Options for Generic Safety Issue--191, X ML121320270 ............
Assessment of Debris Accumulation on Pressurized Water Reactor
Sump Performance,'' dated July 9, 2012..........................
SRM-SECY-12-0093, dated December 14, 2012........................ X ML12349A378 ............
RG 1.174, Revision 2, ``An Approach for Using Probabilistic Risk X ML100910006 ............
Assessment in Risk[dash]Informed Decisions on
Plant[dash]Specific Changes in the Licensing basis,'' dated May
2011............................................................
RG 1.200, ``An Approach for Determining the Technical Adequacy of X ML090410014 ............
Probabilistic Risk Assessment Results for Risk[dash]Informed
Activities,'' dated March 2009..................................
Plant Safety Assessment of RIL 0801.............................. X ML090340073 ............
Federal Register Notice (73 FR 44778), ``Notice of Availability ............ ................. X
and Solicitation of Public Comments on Documents Under
Consideration to Establish the Technical Basis for New
Performance-Based Emergency Core Cooling System Requirements''..
Supplemental research material--additional PQD tests............. X ML090690711 ............
Supplemental research material--additional breakaway testing..... X ML090700193 ............
Draft proposed procedure for Conducting Oxidation and Post-Quench X ML090900841 X
Ductility Tests with Zirconium-Based Alloys.....................
Draft proposed procedure for Conducting Breakaway Oxidation Tests X ML090840258 X
with Zirconium-based cladding alloys............................
Update on Breakaway Oxidation of Westinghouse ZIRLOTM Cladding... X ML091330334 X
Impact of Speciment Preparation of Breakaway Oxidation of X ML091350581 X
Westinghouse ZIRLOTM Cladding...................................
Advance Notice of Proposed Rulemaking, published on August 13, X ML091250132 X
2009 (74 FR 40765)..............................................
Summary of April 28-29, 2010, Public Meeting on ANPR............. X ML101300490 ............
SRM-SECY-12-0034, ``Proposed Rulemaking--10 CFR 50.46c: Emergency X ML13007A478 X
Core Cooling System Performance During Loss of Coolant Accidents
(RIN 3150-AH42)''...............................................
TR WCAP 16793-NP, Revision 2, ``Evaluation of Long-Term Cooling X ML11292A021 ............
Considering Particulate, Fibrous, and Chemical Debris in the
Recirculating Fluid,'' Appendix A...............................
PWROG ECCS Analysis Report....................................... X ML11139A309 ............
BWROG ECCS Analysis Report....................................... X ML111950139 ............
ECCS Audit Report................................................ X ML12041A078 ............
Supplement to RIL-0801, ``Technical Basis for Revision of X ML113050484 ............
Embrittlement Criteria in 10 CFR 50.46''........................
NUREG-2119, ``Mechanical Behavior of Ballooned and Ruptured X ML12048A475 X
Cladding''......................................................
Sec. 50.46c and PRM-50-71 Comment Response Document............ X ML12283A213 ............
Regulatory Analysis.............................................. X ML12283A188 ............
Proposed Rule Information Collection Analysis.................... X ML112520328 ............
Draft Regulatory Guide 1261, ``Conducting Periodic Testing for X ML12284A324 ............
Breakaway Oxidation Behavior''..................................
Draft Regulatory Guide 1262, ``Testing for Post Quench X ML12284A325 ............
Ductility''.....................................................
Draft Regulatory Guide 1263, ``Establishing Analytical Limits for X ML12284A323 ............
Zirconium[dash]Based Alloy Cladding''...........................
Request to Withdraw 50.46a from Commission Consideration......... X ML121500380 ............
Staff Requirements--SECY-10-0161--Final Rule: Risk-Informed X ML12117A121 ............
Changes to Loss-of-Coolant Accident Technical Requirements (10
CFR 50.46a) (RIN 3150-AH29).....................................
----------------------------------------------------------------------------------------------------------------
[[Page 16133]]
X. Criminal Penalties
For the purposes of Section 223 of the Atomic Energy Act of 1954,
as amended (AEA), the NRC is issuing the proposed rule to amend
Sec. Sec. 50.8, 50.34, 50.46a, 50.46c, appendix A to 10 CFR part 50,
appendix K to 10 CFR part 50, and Sec. Sec. 52.47, 52.79, 52.137, and
52.157 under one or more sections of 161b, 161i, or 161o of the AEA.
Willful violations of the rule would be subject to criminal
enforcement. Criminal penalties, as they apply to regulations in 10 CFR
part 50, are discussed in Sec. 50.111.
XI. Agreement State Compatibility
Under the Policy Statement on Adequacy and Compatibility of
Agreement States Programs, approved by the Commission on June 20, 1997,
and published in the Federal Register (62 FR 46517; September 3, 1997),
this rule is classified as compatibility category ``NRC.''
Compatibility is not required for Category ``NRC'' regulations. The NRC
program elements in this category are those that relate directly to
areas of regulation reserved to the NRC by the AEA or the provisions of
Title 10 of the CFR, and although an Agreement State may not adopt
program elements reserved to the NRC, it may wish to inform its
licensees of certain requirements via a mechanism that is consistent
with the particular State's administrative procedure laws, but does not
confer regulatory authority on the State.
XII. Plain Writing
The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal
agencies to write documents in a clear, concise, well-organized manner
that also follows other best practices appropriate to the subject or
field and the intended audience. Although regulations are exempt under
the act, the NRC is applying the same principles to its rulemaking
documents. Therefore, the NRC has written this document, including the
proposed new and amended rule language, to be consistent with the Plain
Writing Act. In addition, where existing rule language must be changed,
the NRC has rewritten that language to improve its organization and
readability. The NRC requests comment on the proposed rule specifically
with respect to the clarity and effectiveness of the language used.
Comments should be sent to the NRC as explained in the ADDRESSES
section of this document.
XIII. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995,
Public Law 104-113, requires that Federal agencies use technical
standards that are developed or adopted by voluntary consensus
standards bodies unless using such a standard is inconsistent with
applicable law or is otherwise impractical. The NRC is not aware of any
voluntary consensus standard that could be used as an alternative to
the proposed Government-unique standard in the proposed rule, in order
to determine the acceptability of emergency core cooling systems and
fuel assemblies for nuclear power reactors. The NRC will consider using
a voluntary consensus standard if an appropriate standard is
identified.
XIV. Finding of No Significant Environmental Impact: Environmental
Assessment
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in
subpart A of 10 CFR part 51, that this rule, if adopted, would not be a
major Federal action significantly affecting the quality of the human
environment and, therefore, an environmental impact statement is not
required. Further, initial implementation of these proposed amendments
would require licensees, in some cases, to submit an additional license
amendment. The NRC's consideration of these license amendments would
each contain an environmental assessment of the proposed licensee-
specific action. The basis for this determination is as follows:
Identification of the Action
The proposed action is the amendment of 10 CFR part 50 by adding a
new Sec. 50.46c which would contain the NRC's requirements for ECCSs
for LWRs (that are currently contained in Sec. 50.46). The proposed
amendment would establish performance-based requirements and also
account for the new research information, as discussed in Section II,
``Background,'' of this document. This research identified previously
unknown embrittlement mechanisms. The research indicated that the
current combination of peak cladding temperature (2200[emsp14][deg]F
(1204 [deg]C)) and local cladding oxidation criteria do not always
ensure PQD. Further, the proposed amendment would expand the
applicability of Sec. 50.46 to all fuel design and fuel cladding
materials. In addition, this proposed rule would address the issues
raised in two PRMs (docketed as PRM-50-71 and PRM-50-84). The proposed
rule would also contain a provision that would allow licensees to use
an alternative risk-informed approach to evaluate the effects of debris
for long-term cooling.
The Need for Action
The proposed action is needed in response to recent research into
the behavior of fuel cladding under LOCA conditions. This research, as
discussed in Section II, ``Background,'' of this document, indicated
that the current combination of peak cladding temperature
(2200[emsp14][deg]F (1204 [deg]C)) and local cladding oxidation
criteria do not always ensure PQD. The research also identified
previously unknown embrittlement mechanisms. The proposed action would
replace the limits on peak cladding temperature and local oxidation
with specific cladding performance requirements and acceptance criteria
that ensure that an adequate level of cladding ductility is maintained
throughout the postulated LOCA.
The proposal to expand applicability to all light-water nuclear
power reactors, regardless of fuel design or cladding material used,
will allow for the development and use of cladding materials other than
zircaloy and ZIRLO\TM\. Under the current Sec. 50.46, licensees that
use different types of cladding material are required to request NRC
approval for an exemption from the rule, in accordance with Sec.
50.12.
The proposed rule would require licensees to take into account the
deposition of crud on the fuel cladding during plant operation. This
change addresses PRM-50-84.
The NRC identified the need for an approach that would allow
entities to address the effects of debris on long-term cooling in a
manner that would be more timely and cost-effective than the current
use of deterministic methods.
Environmental Impacts of the Proposed Action
This environmental assessment focuses on those aspects of the
proposed rulemaking through which the revised requirements could
potentially affect the environment. The NRC has concluded that there
will be no significant radiological environmental impacts associated
with the implementation of the proposed rule requirements for the
following reasons:
(1) The proposed amendments to the ECCS requirements of Sec. 50.46
are unrelated to the integrity of reactor coolant system piping whose
sudden failure would initiate a LOCA. Therefore, the proposed rule does
not affect the probability of an accident.
(2) The proposed amendments to the 10 CFR part 50 ECCS requirements
are
[[Page 16134]]
unrelated to the physical make-up of the systems, structures, and
components that mitigate the consequences of a LOCA. These proposed
amendments, if approved, would revise and expand the performance
requirements for which the ECCS response is judged. With these
enhancements, the reactor core would remain coolable because, by
addressing previously unknown degradation mechanisms, cladding
ductility would be preserved following a postulated LOCA. Therefore,
the consequences of a postulated LOCA are not adversely changed by the
proposed rule.
(3) The proposed amendments to the 10 CFR part 50 ECCS requirements
would not impact a facility's release of radiological effluents during
and following a postulated LOCA. Therefore, the rule does not affect
the amount of effluent released as a result of a possible accident.
(4) The proposed rule would allow entities to address the effects
of debris on long-term cooling using a risk-informed approach. The
effects of debris are currently addressed using deterministic methods.
Any change in CDF and LERF allowed by a risk-informed approach would be
small and within criteria already established in RG 1.174, Revision 2,
for making risk-informed changes to plant licensing bases.
This proposed rulemaking would amend calculated ECCS evaluation
models used to assess the emergency core cooling system's response to a
postulated LOCA. The rulemaking would not affect any other procedures
used to operate the plant, nor alter the plant's geometry or
construction. Further, the proposed amendments would ensure post quench
ductility and core coolability following a postulated LOCA, and as
such, would not affect the dose to any plant workers following
postulated accidents. Similarly, dose to any individual member of the
public would not be affected.
For the reasons discussed, the action will not significantly
increase the probability or consequences of accidents, nor result in
changes being made in the types of any effluents that may be released
off-site, and there would be no increase in occupational or public
radiation exposure.
With regard to potential nonradiological impacts, the proposed rule
would have no significant impact on the environment. The proposed rule
to revise and expand the ECCS performance requirements would be applied
by an NRC nuclear reactor power plant licensee to the restricted area
of its facility only, and in many cases would not result in any
physical changes to the plant. Restricted areas of nuclear power plants
are industrial portions of the facility constructed upon previously
disturbed land, to which access is limited to authorized personnel. As
such, it is extremely unlikely that the proposed amendments, if
approved, would create any significant impact on any aquatic or
terrestrial habitat in the vicinity of the plant, or to any threatened,
endangered, or protected species under the Endangered Species Act, or
have any impacts to essential fish habitat covered by the Magnuson-
Stevens Act. Similarly, it is extremely unlikely that there will be any
impacts to socioeconomic, or to historic properties and cultural
resources. Therefore, there would be no significant nonradiological
environmental impacts associated with the proposed action.
Licensee compliance with the proposed amendments would require an
additional license amendment. A National Environmental Policy Act
analysis would be conducted for each licensee-specific license
amendment review.
Alternatives to the Proposed Action
As an alternative to the rulemakings previously described, the NRC
considered not taking the action (i.e., the ``no-action'' alternative).
Not revising the ECCS cladding acceptance criteria could result in
instances, following a LOCA, in which cladding ductility is not
guaranteed to be maintained. Under the no action alternative, licensees
will continue to submit exemption requests for NRC approval of fuel
cladding other than zircaloy or ZIRLO\TM\.
The NRC does not find this alternative acceptable to preserving
public health and safety. The revised requirements are necessary
because recent research has indicated that the current PCT and
oxidation restrictions do not take into consideration newly discovered
cladding embrittlement mechanisms, and that the current restrictions
may not always be adequate to ensure post quench ductility of fuel
cladding. The revised requirements ensure post quench ductility and
core coolability following a postulated LOCA.
The proposed rule would allow entities to use a risk-informed
approach to address the effects of debris for long-term cooling. An
alternative to addressing debris using this risk-informed approach is
to continue to address the effects of debris using deterministic
methods and approved models, as described in SECY-12-0093, ``Closure
Options for Generic Safety Issue--191, Assessment of Debris
Accumulation on Pressurized-Water Reactor Sump Performance,'' dated
July 9, 2012 (ADAMS Accession No. ML121310648). However, the NRC has
added the alternative approach to provide entities the additional
flexibility to address the effects of debris on long-term cooling using
risk-informed methodologies, which may be implemented in a more timely
and cost-efficient manner.
Alternative Use of Resources
This action would not involve the use of any resources not
previously considered by the NRC in its past environmental statements
for issuance of operating licenses for the facilities that would be
affected by this action.
Agencies and Persons Consulted
The NRC staff developed the proposed rule and this environmental
assessment. In accordance with its stated policy, the NRC provided a
copy of the proposed rule and the environmental assessment to
designated State Liaison Officers and requested their comments. No
other agencies were consulted.
There appears to be no significant impact to human health or the
environment from implementation of the proposed action. However, the
general public should note that the NRC is seeking public
participation. Comments on any aspect of the environmental assessment
may be submitted to the NRC via email to [email protected] or
via mail to Secretary, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
XV. Paperwork Reduction Act Statement
This proposed rule contains new or amended information collection
requirements that are subject to the Paperwork Reduction Act of 1995
(44 U.S.C. 3501 et seq.). This rule has been submitted to the Office of
Management and Budget for review and approval of the information
collection requirements.
Type of submission, new or revision: Revision.
The title of the information collection: 10 CFR 50.46c, Emergency
Core Cooling System Performance During Loss-of-Coolant Accidents.
The form number if applicable: Not applicable.
How often the collection is required: LOCA model updates, Licensee
Amendment Requests, and compliance letters will be submitted one time
during implementation; significant errors will be reported on occasion
[[Page 16135]]
(within 30 days); other errors or changes in analysis will be reported
annually.
Who will be required or asked to report: Fuel design vendors, all
operating reactors, all applicants for or holders of construction
permits, each applicant for an operating license, each applicant for or
holder of a combined license, each applicant for a standard design
certification, each applicant for a standard design approval, and each
applicant for a manufacturing license.
An estimate of the number of annual responses: 290.
The estimated number of annual respondents: 70 during the first 3
years of implementation; a total of 111 will be impacted by the rule.
An estimate of the total number of hours needed annually to
complete the requirement or request: 61,131 hours (an increase of
61,891 hours reporting and a decrease of 760 hours recordkeeping
resulting from eliminating the need for exemptions).
Abstract: The NRC is proposing to amend its regulations to revise
the acceptance criteria for the emergency core cooling system for
light-water nuclear power reactors as currently required by 10 CFR part
50. The rule would establish a 5-year staged implementation approach to
improve the efficiency and effectiveness of the migration to the new
ECCS requirements. The vendors would also propose post-quench ductility
limits by either selecting analytical limits provided in Figure 2 of
draft regulatory guide DG-1263, ``Establishing Analytical Limits for
Zirconium-Based Alloy Cladding,'' using an NRC-approved experimental
approach to obtain the post-quench ductility limits, or using an
experimental approach developed by the vendor to obtain the post-quench
ductility limits. Those ductility limits which are developed via an
experimental method would be submitted to the NRC via a topical report
for NRC approval. The DG-1262, ``Testing for Post Quench Ductility,''
provides guidance on an acceptable testing approach for developing
post-quench ductility. The DG-1263 provides a methodology for using
test results, generated from DG-1262 or an alternate NRC-approved
experimental approach, to establish and support a new cladding-specific
analytical limit. The vendors would also obtain post-quench ductility
analytical methods by either selecting analytical limits provided in a
regulatory guide, using an NRC-approved experimental approach, or using
an experimental approach developed by the vendor. Those PQD limits
developed via an experimental method would be submitted to the NRC via
a topical report. The vendors would also perform long-term cooling
tests to determine the long-term cooling limits for each of the nine
cladding alloys. In addition, vendors would perform initial breakaway
testing. The licensees would report the initial breakaway results to
the NRC via their license amendment request. Those licensees that meet
the new requirements without new analyses or model revisions would
complete any necessary engineering calculations, update their plant
UFSAR, and provide a letter report to the NRC documenting compliance.
Those licensees that would require new analyses or model revisions to
demonstrate compliance would be required to submit a new LOCA analysis
of record. The rule would also require licensees to conduct periodic
breakaway testing, and include those results in the yearly ECCS report.
Lastly, the rule would add a requirement to report errors in ECR to the
NRC. This would be submitted within the same yearly ECCS report.
The rule would include a provision allowing entities to use an
alternative risk-informed approach to evaluate the effects of debris
for long-term cooling. If an entity voluntarily chooses to use this
approach, they would need to submit an application for NRC review and
approval, report all errors and changes in their plant-specific PRA,
and conduct periodic updates to their PRA.
The NRC is seeking public comment on the potential impact of the
information collections contained in this proposed rule and on the
following issues:
1. Is the proposed information collection necessary for the proper
performance of the functions of the NRC, including whether the
information will have practical utility?
2. Is the estimate of burden accurate?
3. Is there a way to enhance the quality, utility, and clarity of
the information to be collected?
4. How can the burden of the information collection be minimized,
including the use of automated collection techniques?
The public may examine and have copied, for a fee, publicly-
available documents, including the draft supporting statement, at the
NRC's Public Document Room, One White Flint North, 11555 Rockville
Pike, Room O-1 F21, Rockville, Maryland 20852. The OMB clearance
requests are available on the NRC's Web site: http://www.nrc.gov/public-involve/doc-comment/omb/index.html. The document will be
available on the NRC's Web site for 30 days after the signature date of
this document.
Send comments on any aspect of these proposed information
collections, including suggestions for reducing the burden and on the
above issues, by May 23, 2014 to the FOIA, Privacy, and Information
Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, or by email to [email protected]
and to the Desk Officer, Chad Whiteman, Office of Information and
Regulatory Affairs, NEOB-10202, (3150-0011), Office of Management and
Budget, Washington, DC 20503. Comments received after this date will be
considered if it is practical to do so, but assurance of consideration
cannot be given to comments received after this date. Comments can also
be emailed to [email protected] or submitted by telephone
at 202-395-4718.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a request for information or an information collection
requirement unless the requesting document displays a currently valid
OMB control number.
XVI. Regulatory Analysis: Availability
The NRC has prepared a draft regulatory analysis on this proposed
regulation (ADAMS Accession No. ML12283A188). The analysis examines the
costs and benefits of the alternatives considered by the Commission.
The NRC requests public comments on the draft regulatory analysis.
Availability of the draft regulatory analysis is indicated in
Section IX of this document. Comments on the draft regulatory analysis
may be submitted to the NRC by any method provided in the ADDRESSES
section of this document.
XVII. Regulatory Flexibility Certification
Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the
Commission certifies that this rule would not, if promulgated, have a
significant economic impact on a substantial number of small entities.
This proposed rule affects light water nuclear power reactors. None of
the companies that own and operate these facilities falls within the
scope of the definition of ``small entities'' set forth in the
Regulatory Flexibility Act or the size standards established by the NRC
(Sec. 2.810).
[[Page 16136]]
XVIII. Backfitting and Issue Finality
Proposed Sec. 50.46c Rule
The proposed rule would be applicable to all existing and future
nuclear power plant designs, regardless of fuel design or cladding
material, but the time by which compliance must be achieved would vary
as described in the proposed rule. The proposed rule, if finalized,
would replace existing ECCS requirements in Sec. 50.46. The proposed
rule would provide an option (``voluntary alternative'') to address
consideration of the effects of debris on long-term cooling (following
a LOCA) using a risk-informed approach, and to use the same risk-
informed approach for consideration of debris with respect to long-term
cooling to demonstrate compliance with GDC-35, GDC-38, and GDC-41 in
appendix A to 10 CFR part 50. The proposed rule, if finalized, would
apply to and be imposed on (``apply to'') all current nuclear power
plant licensees (including holders of renewed licenses and combined
licenses under 10 CFR part 52). The proposed rule, if finalized, would
also apply to current and future applicants for combined licenses under
10 CFR part 52, including those applicants referencing one of the
existing standard design certification rules in appendices A through D
to 10 CFR part 52. The proposed rule would also apply to all current
and future applicants for LWR standard design certification rules under
10 CFR part 52. The proposed rule, if finalized, would not apply to the
existing four design certifications in appendices A through D to 10 CFR
part 52 until their renewal. Finally, the proposed rule would apply to
all future applicants for manufacturing licenses under 10 CFR part 52
(there are no current applicants or holders of manufacturing licenses).
Each of these classes of licenses and regulatory approvals is
discussed in the following sections.
Operating Licenses
With respect to current nuclear power plant licensees, the NRC
assumes that imposition of the proposed rule would constitute
backfitting as defined in Sec. 50.109(a)(1). However, the NRC believes
that the proposed rule must be imposed upon current nuclear power plant
licensees in order to ensure adequate protection to the public health
and safety. The proposed rule will ensure that the level of protection
intended to be achieved by the current rule is maintained. Therefore,
the NRC has determined that the proposed rule is necessary to ensure
that the facility provides adequate protection to the health and safety
of the public, and that a backfit analysis as described in Sec.
50.109(a)(3) and (b) need not be prepared, under the exception in Sec.
50.109(a)(4)(ii).
Imposing the redefinition of fuel cladding acceptance criteria on
current nuclear power plant licensees is justified under the provisions
of Sec. 50.109(a)(4)(ii) as the requirements of the proposed rule are
necessary to ensure adequate protection to the public health and safety
by maintaining that level of protection (i.e., reasonable assurance of
adequate protection) which the NRC previously thought would be achieved
(throughout the entire term of licensed operation) by the current rule.
Information developed through the NRC's high burnup fuel research
program has identified that the current criterion for preventing fuel
cladding embrittlement may not be adequate in the future to ensure the
health and safety of the public. As discussed in Sections II and V of
this document, zirconium-based alloy fuel cladding materials may be
subject to embrittlement at a lower combination of temperature and
level of oxygen absorption (17 percent) than currently allowed under
Sec. 50.46(b)(1) due to absorption of hydrogen during normal
operation. The proposed rule would correct those limits initially
established to prevent embrittlement of zirconium-based alloy cladding
material based on the new research information. In addition, the
research work has identified new phenomena, such as breakaway oxidation
and oxygen diffusion from the cladding inside surfaces, which are
believed to further adversely affect the fuel cladding embrittlement
process. Therefore, PQD (which is necessary to ensure coolable core
geometry) \3\ is not guaranteed following a postulated LOCA. The
proposed rule would establish new requirements for zirconium-based
alloys to prevent breakaway oxidation and account for oxygen diffusion
from the oxide fuel pellet during the operating life of the fuel. In
sum, the NRC believes that imposing the requirements of the proposed
rule is necessary to prevent embrittlement of fuel cladding and to
ensure that the rule maintains reasonable assurance of adequate
protection to public health and safety.
---------------------------------------------------------------------------
\3\ The Commission concluded, as part of the 1973 Emergency Core
Cooling System rulemaking, that retention of ductility in the
zircaloy cladding material was determined to be the best guarantee
of its remaining intact during the hypothetical loss-of-coolant
accident, thereby maintaining a coolable core geometry. See
Acceptance Criteria for Emergency Core Cooling Systems for Light-
Water-Cooled Nuclear Power Reactors, CLI-73-39, at page 1098
(December 28, 1973).
---------------------------------------------------------------------------
The proposed rule includes the option of allowing an applicant or
licensee to address the effects of debris on long-term cooling with
respect to ECCS performance requirements in Sec. 50.46c and GDC-35
using a risk-informed approach. Inasmuch as this is a voluntary
alternative to existing requirements as well the proposed requirements
on ECCS, the inclusion of this option in the proposed rule is not
backfitting or inconsistent with issue finality provisions in 10 CFR
part 52. The proposed rule would also allow applicants and licensees
who select the option of using the risk-informed approach for
addressing the effects of debris on long-term cooling, to also use the
same approach in demonstrating compliance with GDC-38 and GDC-41.
Because this is a voluntary alternative with respect to a portion of
the existing requirements in GDC-38 and GDC-41, inclusion of this
option in the proposed rule is not backfitting as defined in Sec.
50.109(a)(1).
Combined License Holders as of the Date of a Final Sec. 50.46c Rule
Currently, there are two holders of combined licenses for the
Vogtle and Summer facilities, each referencing the AP1000 standard
design certification rule. In addition, there may be other combined
licenses issued referencing one or more of the standard design
certification rules approved in the appendices to 10 CFR part 52, by
the time that a final Sec. 50.46c rule is issued by the NRC. Imposing
the requirements of the proposed rule on current holders of combined
licenses as of the date of a final Sec. 50.46c rule would represent an
inconsistency with the general issue finality provision applicable to
standard design certifications in Sec. 52.63, the issue finality
provision included in each design certification rule at Section VI,
``Issue Resolution,'' of this document, and the issue finality
provisions applicable to combined licenses in Sec. Sec. 52.83 and
52.98.
Therefore, the NRC has addressed the criteria in those provisions
that would allow imposition of the proposed rule on current holders of
combined licenses despite the issue finality accorded to the combined
license holders. The NRC believes that the proposed rule may be imposed
as a change needed to provide reasonable assurance of adequate
protection. The key differences between the existing ECCS requirements
and the proposed rules are in the areas of embrittlement. The bases for
this adequate protection determination are presented in this document
in Section
[[Page 16137]]
II, ``Background;'' Section III, ``Operating Plant Safety;'' and
Section V, ``Proposed Requirements for ECCS Performance during LOCAs.''
Therefore, the NRC believes that the NRC has met the requirements in
the applicable issue finality provisions for not according issue
finality to the subject of ECCS performance under Sec. 50.46 and GDC-
35.
The proposed rule includes the option of allowing a combined
license holder (such as the holders of the Vogtle and Summer combined
licenses) to address the effects of debris on long-term cooling with
respect to ECCS performance requirements in Sec. 50.46c and GDC-35
using a risk-informed approach. Inasmuch as this is a voluntary
alternative to existing requirements as well as the proposed
requirements on ECCS, the inclusion of this option in the proposed rule
is not backfitting or inconsistent with issue finality provisions in 10
CFR part 52. The proposed rule would also allow combined license
applicants and holders who select the option of using the risk-informed
approach for addressing the effects of debris on long-term cooling, to
also use the same approach in demonstrating compliance with GDC-38 and
GDC-41. Because this is a voluntary alternative with respect to a
portion of the existing requirements in GDC-38 and GDC-41, inclusion of
this option in the proposed rule is not backfitting or inconsistent
with the issue finality provisions in 10 CFR part 52.
Combined License Applicants
Imposing the requirements of the proposed rule on current and
future applicants for combined licenses under subpart C of 10 CFR part
52 would not constitute backfitting. Neither the Backfit Rule nor the
finality provisions for combined licenses in Sec. Sec. 52.83 or 52.98
protect either a current or prospective applicant for a combined
license from changes in the NRC rules and regulations. The NRC has long
adopted the position that the Backfit Rule does not protect current or
prospective applicants from changes in NRC requirements or guidance
because the policies underlying the Backfit Rule are largely
inapplicable in the context of a current or future application. This
position also applies to each of the issue finality provisions in 10
CFR part 52.
The proposed rule includes the option of allowing a combined
license applicant to address the effects of debris on long-term cooling
with respect to ECCS performance requirements in Sec. 50.46c and GDC-
35 using a risk-informed approach. Inasmuch as this is a voluntary
alternative to existing requirements as well as the proposed
requirements on ECCS, the inclusion of this option in the proposed rule
is not inconsistent with any applicable issue finality provision in 10
CFR part 52. The proposed rule would also allow combined license
applicants who select the option of using the risk-informed approach
for addressing the effects of debris on long-term cooling, to also use
the same approach in demonstrating compliance with GDC-38 and GDC-41.
Because this is a voluntary alternative with respect to a portion of
the existing requirements in GDC-38 and GDC-41, inclusion of this
option in the proposed rule is not inconsistent with any applicable
issue finality provision in 10 CFR part 52.
Standard Design Certifications
The requirements of the proposed rule would not apply to any of the
four existing standard design certification rules in appendices A
through D to 10 CFR part 52 during the period in which they may be
referenced. However, inasmuch as the proposed rule would also require
any combined license applicant and holder referencing a design
certification to comply with the Sec. 50.46c rule, this would
effectively constitute an inconsistency with the general issue finality
provision applicable to standard design certifications in Sec. 52.63,
and the issue finality provision included in each design certification
rule at Section VI, ``Issue Resolution,'' of this document. Therefore,
the NRC has addressed the criteria in those provisions that would allow
imposition of the proposed rule on entities referencing the standard
design certification rule despite the issue finality accorded by Sec.
52.63 and Section VI of this document of each of the four existing
standard design certification rules.
The NRC believes that the proposed rule may be imposed as a change
needed to provide reasonable assurance of adequate protection. The key
differences between the existing ECCS requirements and the proposed
rules are in the areas of embrittlement. The bases for this adequate
protection determination are presented in this document in Section II,
``Background;'' Section III, ``Operating Plant Safety;'' and Section V,
``Proposed Requirements for ECCS Performance during LOCAs.'' Therefore,
the NRC believes that the NRC has met the requirements in the
applicable issue finality provisions for not according issue finality
to the subject of ECCS performance under Sec. 50.46 and GDC-35.
The requirements of the proposed rule would apply to the four
existing standard design certification rules in 10 CFR part 52,
appendices A through D at the time of their renewal. The NRC believes
that the proposed rule may be imposed as a change needed to provide
reasonable assurance of adequate protection. The bases for this
adequate protection determination are presented in this document in
Section II, ``Background;'' Section III, ``Operating Plant Safety;''
and Section V, ``Proposed Requirements for ECCS Performance during
LOCAs.'' Therefore, the new requirements may be imposed at renewal in
accordance with Sec. 51.51(b)(1).
The proposed rule includes the option of allowing a design
certification applicant (including applicants after the NRC has issued
a final design certification rule) to address the effects of debris on
long-term cooling with respect to ECCS performance requirements in
Sec. 50.46c and GDC-35 using a risk-informed approach. Inasmuch as
this is a voluntary alternative to existing requirements as well as the
proposed requirements on ECCS, the inclusion of this option in the
proposed rule is not inconsistent with any applicable issue finality
provisions. The proposed rule would also allow a design certification
applicant who selects the option of using the risk-informed approach
for addressing the effects of debris on long-term cooling, to also use
the same approach in demonstrating compliance with GDC-38 and GDC-41.
Because this is a voluntary alternative with respect to a portion of
the existing requirements in GDC-38 and GDC-41, inclusion of this
option in the proposed rule is not inconsistent with any applicable
issue finality provision.
Imposing the requirements of the proposed rule on current and
future applicants for standard design certification rules would not
constitute backfitting. Neither the Backfit Rule nor the finality
provisions for final design certification rules in Sec. 52.63 protect
either a current or prospective applicant for a standard design
certification rule from changes in the NRC rules and regulations.
Manufacturing Licenses
Imposing the requirements of the proposed rule on future applicants
for manufacturing licenses would not constitute backfitting. The NRC
has not issued any manufacturing licenses under 10 CFR part 52, and
neither the Backfit Rule nor the finality provisions for manufacturing
licenses in Sec. 52.171 protect a prospective manufacturing applicant
from changes in the NRC rules and regulations.
[[Page 16138]]
The proposed rule includes the option of allowing a manufacturing
license applicant or holder to address the effects of debris on long-
term cooling with respect to ECCS performance requirements in Sec.
50.46c and GDC-35 using a risk-informed approach. Inasmuch as this is a
voluntary alternative to existing requirements as well as the proposed
requirements on ECCS, the inclusion of this option in the proposed rule
is not inconsistent with Sec. 52.171. The proposed rule would also
allow combined license applicants and holders who select the option of
using the risk-informed approach for addressing the effects of debris
on long-term cooling, to also use the same approach in demonstrating
compliance with GDC-38 and GDC-41. Because this is a voluntary
alternative with respect to a portion of the existing requirements in
GDC-38 and GDC-41, inclusion of this option in the proposed rule is not
inconsistent with Sec. 52.171.
Draft Regulatory Guides
The NRC is issuing, for public comment, three draft regulatory
guides that would support implementation of Sec. 50.46c. These draft
regulatory guides are DG-1261, ``Conducting Periodic Testing for
Breakaway Oxidation Behavior'' (ADAMS Accession No. ML12284A324); DG-
1262, ``Testing for Post Quench Ductility'' (ADAMS Accession No.
ML12284A325); and DG-1263, ``Establishing Analytical Limits for
Zirconium-Based Alloy Cladding'' (ADAMS Accession No. ML12284A323). The
draft regulatory guides provide guidance on compliance with those
proposed new requirements for ECCS not contained in the current ECCS
rule, Sec. 50.46.
The NRC also plans to issue regulatory guidance on the voluntary
alternative for addressing the effects of debris on long-term cooling
using a risk-informed approach. The NRC currently intends to issue the
guidance in the form of one or more regulatory guides, and that the
regulatory guides would be published in draft form for public comment
before being issued in final form as part of a final Sec. 50.46c rule.
The first issuance of new guidance on a new rule provision \4\ does
not constitute backfitting, inasmuch as: i) The guidance on the new
rule provision must be consistent with the regulatory requirements in
the new rule provision; and ii) the backfittiing basis for the new rule
provision should also be applicable to the issuance of guidance on that
new rule provision. Therefore, the first issuance of new guidance
addressing new provisions of Sec. 50.46c does not constitute issuance
of ``changed'' or ``new'' guidance within the meaning of the definition
of ``backfitting'' in Sec. 50.109(a)(1), or constitute an action
inconsistent with any of the issue finality provisions in 10 CFR part
52. Accordingly, no further consideration of backfitting is needed to
support issuance of the new regulatory guides on Sec. 50.46c in final
form.
---------------------------------------------------------------------------
\4\ The NRC notes that while the proposed Sec. 50.46c includes
both ``amended'' requirements and ``new'' requirements, the three
draft regulatory guides only provide ``new'' guidance on ``new''
Sec. 50.46c requirements. By ``new'' requirements, the NRC means
that these requirements have no analogue in the current ECCS rule.
For example, the proposed Sec. 50.46c(g)(1)((iii) criterion on
breakaway oxidation is a ``new'' requirement because there is no
provision in current Sec. 50.46 requiring consideration of that
phenomenon. By contrast, ``amended,'' means that the proposed rule
contains several requirements that have analogues to requirements in
the existing rule but are being addressed differently. An example of
an ``amended'' requirement would be proposed Sec. 50.46c(d)(1),
because that provision: i) Addresses, in language that differs from
the current rule's language, matters that are addressed in the
current rule, including Sec. 50.46(a)(1)(i); and ii) contains
substantively different (proposed) requirements when compared to the
current rule, but the proposed requirements are directed at
technical matters already addressed in the current ECCS rule. For
example, the proposed Sec. 50.46c(g)(1)((iii) criterion on
breakaway oxidation is a ``new'' requirement because there is no
provision in current Sec. 50.46 requiring consideration of that
phenomenon. By contrast, ``amended'' means that the proposed rule
contains several requirements which have analogues to requirements
in the existing rule but are being addressed differently. An example
of an ``amended'' requirement would be proposed Sec. 50.46c(d)(1),
because that provision: i) Addresses, in language that differs from
the current rule's language, matters that are addressed in the
current rule, including Sec. 50.46(a)(1)(i); and ii) contains
substantively different (proposed) requirements when compared to the
current rule, but the proposed requirements are directed at
technical matters already addressed in the current rule.
---------------------------------------------------------------------------
List of Subjects
10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Intergovernmental relations, Nuclear power plants and
reactors, Radiation protection, Reactor siting criteria, Reporting and
recordkeeping requirements.
10 CFR Part 52
Administrative practice and procedure, Antitrust, Backfitting,
Combined license, Early site permit, Emergency planning, Fees,
Inspection, Limited work authorization, Nuclear power plants and
reactors, Probabilistic risk assessment, Prototype, Reactor siting
criteria, Redress of site, Reporting and recordkeeping requirements,
Standard design, Standard design certification.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974; and 5 U.S.C. 553, the NRC is proposing to adopt the
following amendments to 10 CFR parts 50 and 52.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
0
1. Revise the authority citation for part 50 to read as follows:
Authority: Atomic Energy Act secs. 102, 103, 104, 105, 147,
149, 161, 181, 182, 183, 186, 189, 223, 234 (42 U.S.C. 2132, 2133,
2134, 2135, 2167, 2169, 2201, 2231, 2232, 2233, 2236, 2239, 2273,
2282); Energy Reorganization Act secs. 201, 202, 206 (42 U.S.C.
5841, 5842, 5846); Nuclear Waste Policy Act sec. 306 (42 U.S.C.
10226); Government Paperwork Elimination Act sec. 1704 (44 U.S.C.
3504 note); Energy Policy Act of 2005, Pub. L. 109-58, 119 Stat. 594
(2005). Section 50.7 also issued under Pub. L. 95-601, sec. 10, as
amended by Pub. L. 102-486, sec 2902 (42 U.S.C. 5851). Section 50.10
also issued under Atomic Energy Act secs. 101, 185 (42 U.S.C. 2131,
2235); National Environmental Protection Act sec. 102 (42 U.S.C.
4332). Sections 50.13, 50.54(dd), and 50.103 also issued under
Atomic Energy Act sec. 108 (42 U.S.C. 2138).
Sections 50.23, 50.35, 50.55, and 50.56 also issued under Atomic
Energy Act sec. 185 (42 U.S.C. 2235). Appendix Q also issued under
National Environmental Protection Act sec. 102 (42 U.S.C. 4332).
Sections 50.34 and 50.54 also issued under sec. 204 (42 U.S.C.
5844). Sections 50.58, 50.91, and 50.92 also issued under Pub. L.
97-415 (42 U.S.C. 2239). Section 50.78 also issued under Atomic
Energy Act sec. 122 (42 U.S.C. 2152). Sections 50.80-50.81 also
issued under Atomic Energy Act sec. 184 (42 U.S.C. 2234).
0
2. In Sec. 50.8, paragraph (b) is revised to read as follows:
Sec. 50.8 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 50.30, 50.33, 50.34, 50.34a, 50.35,
50.36, 50.36a, 50.36b, 50.44, 50.46, 50.46c, 50.47, 50.48, 50.49,
50.54, 50.55, 50.55a, 50.59, 50.60, 50.61, 50.61a, 50.62, 50.63, 50.64,
50.65, 50.66, 50.68, 50.69, 50.70, 50.71, 50.72, 50.74, 50.75, 50.80,
50.82, 50.90, 50.91, 50.120, 50.150, and appendices A, B, E, G, H, I,
J, K, M, N, O, Q, R, and S to this part.
* * * * *
0
3. In Sec. 50.34, paragraphs (a)(4) and (b)(4) are revised to read as
follows:
Sec. 50.34 Contents of applications; technical information.
(a) * * *
(4) A preliminary analysis and evaluation of the design and
[[Page 16139]]
performance of structures, systems, and components of the facility with
the objective of assessing the risk to public health and safety
resulting from operation of the facility and including determination of
the margins of safety during normal operations and transient conditions
anticipated during the life of the facility, and the adequacy of
structures, systems, and components provided for the prevention of
accidents and the mitigation of the consequences of accidents. Analysis
and evaluation of ECCS cooling performance and the need for high point
vents following postulated loss-of-coolant accidents must be performed
in accordance with the requirements of Sec. Sec. 50.46, 50.46b, and
50.46c, as applicable, for facilities for which construction permits
may be issued after December 28, 1974.
* * * * *
(b) * * *
(4) A final analysis and evaluation of the design and performance
of structures, systems, and components with the objective stated in
paragraph (a)(4) of this section and taking into account any pertinent
information developed since the submittal of the preliminary safety
analysis report. Analysis and evaluation of ECCS cooling performance
following postulated loss-of-coolant accidents shall be performed in
accordance with the requirements of Sec. Sec. 50.46 and 50.46c, as
applicable, for facilities for which a license to operate may be issued
after December 28, 1974.
* * * * *
Sec. 50.46a [Added and Reserved]
0
4. Section 50.46a is redesignated as Sec. 50.46b, and a new Sec.
50.46a is added and reserved.
0
5. A new Sec. 50.46c is added to read as follows:
Sec. 50.46c Emergency core cooling system performance during loss-of-
coolant accidents (LOCA).
(a) Applicability. The requirements of this section apply to the
design of a light water nuclear power reactor (LWR) and to the
following entities who design, construct or operate an LWR: Each
applicant for or holder of a construction permit under this part, each
applicant for or holder of an operating license under this part (until
the licensee has submitted the certification required under Sec.
50.82(a)(1) to the NRC), each applicant for or holder of a combined
license under part 52 of this chapter, each applicant for a standard
design certification (including the applicant for that design
certification after the NRC has adopted a final design certification
rule), each applicant for a standard design approval under part 52 of
this chapter, and each applicant for or holder of a manufacturing
license under part 52 of this chapter.
(b) Definitions. As used in this section:
Breakaway oxidation, for zirconium-alloy cladding material, means
the fuel cladding oxidation phenomenon in which weight gain rate
deviates from normal kinetics. This change occurs with a rapid increase
of hydrogen pickup during prolonged exposure to a high-temperature
steam environment, which promotes loss of cladding ductility.
ECCS evaluation model means the calculational framework for
evaluating the behavior of the reactor system (including fuel) during a
postulated LOCA. It includes one or more computer programs and all
other information necessary for application of the calculational
framework to a specific LOCA, such as mathematical models used,
assumptions included in the programs, procedure for treating the
program input and output information, specification of those portions
of analysis not included in computer programs, values of parameters,
and all other information necessary to specify the calculational
procedure.
Debris evaluation model means the calculational framework used to
quantify the impact of debris generation, transport, sump head loss,
in-vessel effects, chemical precipitation, and other phenomena
important to long-term cooling. It includes one or more computer
programs and other information necessary for application of the
calculational framework to a set of initiating events, the mitigation
of which requires long term cooling via recirculation. It also includes
mathematical models used, assumptions used by the programs, procedures
for treating the program input and output information, specifications
of those portions of analysis not included in computer programs, values
of parameters, and all other information necessary to specify the
calculational procedure. The debris evaluation model is used, along
with the probabilistic risk assessment (PRA), to quantify the portion
of core damage frequency and large early release frequency attributable
to debris.
Loss-of-coolant accident (LOCA) means a hypothetical accident that
would result from the loss of reactor coolant, at a rate in excess of
the capability of the reactor coolant makeup system, from breaks in
pipes in the reactor coolant pressure boundary up to and including a
break equivalent in size to the double-ended rupture of the largest
pipe in the reactor coolant system.
(c) Relationship to other NRC regulations. The requirements of this
section are in addition to any other requirements applicable to an
emergency core cooling system (ECCS) set forth in this part, except as
noted in this paragraph. The analytical limits established in
accordance with this section, with cooling performance calculated in
accordance with an NRC approved ECCS evaluation model, are in
implementation of the general requirements with respect to ECCS cooling
performance design set forth in this part, including in particular
Criterion 35 of appendix A to this part. If the effects of debris on
long-term cooling are evaluated using a risk-informed method as
described in paragraph (e) of this section, then this method and
results can be relied upon to demonstrate compliance with other
requirements of this part as allowed by this section and requested in
the application.
(d) Emergency core cooling system design.
(1) ECCS performance criteria. Each LWR must be provided with an
ECCS designed to satisfy the following performance requirements in the
event of, and following, a postulated LOCA. The demonstration of ECCS
performance must comply with paragraph (d)(2) of this section:
(i) Core temperature during and following the LOCA event does not
exceed the analytical limits for the fuel design used for ensuring
acceptable performance as defined in this section.
(ii) The ECCS provides sufficient coolant so that decay heat will
be removed for the extended period of time required by the long-lived
radioactivity remaining in the core.
(2) ECCS performance demonstration. ECCS performance must be
demonstrated using an ECCS evaluation model meeting the requirements of
paragraph (d)(2)(i) or (d)(2)(ii) of this section, and satisfy the
analytical requirements in paragraphs (d)(2)(iii), (d)(2)(iv), and
(d)(2)(v) of this section. Paragraph (e) of this section may be used
for consideration of debris as described in paragraph (d)(2)(iii) of
this section. The ECCS evaluation model must be reviewed and approved
by the NRC.
(i) Realistic ECCS model. A realistic model must include sufficient
supporting justification to show that the analytical technique
realistically describes the behavior of the reactor system during a
loss-of-coolant accident. Comparisons to applicable experimental data
must be made and
[[Page 16140]]
uncertainties in the analysis method and inputs must be identified and
assessed so that the uncertainty in the calculated results can be
estimated. This uncertainty must be accounted for, so that when the
calculated ECCS cooling performance is compared to the applicable
specified and NRC-approved analytical limits, there is a high level of
probability that the limits would not be exceeded.
(ii) Appendix K model. Alternatively, an ECCS evaluation model may
be developed in conformance with the required and acceptable features
of appendix K to this part, ECCS Evaluation Models.
(iii) Core geometry and coolant flow. The ECCS evaluation model
must address calculated changes in core geometry and must consider
those factors, including debris, that may alter localized coolant flow
in the core or inhibit delivery of coolant to the core. A licensee may
evaluate effects of debris using a risk-informed approach to
demonstrate long-term ECCS performance, as specified in paragraph (e)
of this section.
(iv) LOCA analytical requirements. ECCS performance must be
demonstrated for a range of postulated loss-of-coolant accidents of
different sizes, locations, and other properties, sufficient to provide
assurance that the most severe postulated loss-of-coolant accidents
have been identified. ECCS performance must be demonstrated for the
accident, and the post-accident recovery and recirculation period.
(v) Modeling requirements for fuel designs: Uranium oxide or mixed
uranium-plutonium oxide pellets within zirconium-alloy cladding. If the
reactor is fueled with uranium oxide or mixed uranium-plutonium oxide
pellets within cylindrical zirconium-alloy cladding, then the ECCS
evaluation model must address the fuel system modeling requirements in
paragraph (g)(2) of this section.
(3) Required documentation. Upon implementation of this section in
accordance with paragraph (o) of this section, the documentation
requirements of this paragraph apply and supersede the requirements in
appendix K to this part, section II, ``Required Documentation.''
(i)(A) A description of each ECCS evaluation model must be
furnished. The description must be sufficiently complete to permit
technical review of the analytical approach, including the equations
used, their approximations in difference form, the assumptions made,
and the values of all parameters or the procedure for their selection,
as for example, in accordance with a specified physical law or
empirical correlation.
(B) A complete listing of each computer program, in the same form
as used in the ECCS evaluation model, must be furnished to the NRC upon
request.
(ii) For each computer program, solution convergence must be
demonstrated by studies of system modeling or noding and calculational
time steps.
(iii) Appropriate sensitivity studies must be performed for each
ECCS evaluation model, to evaluate the effect on the calculated results
of variations in noding, phenomena assumed in the calculation to
predominate, including pump operation or locking, and values of
parameters over their applicable ranges. For items to which results are
shown to be sensitive, the choices made must be justified.
(iv) To the extent practicable, predictions of the ECCS evaluation
model, or portions thereof, must be compared with applicable
experimental information.
(v) Elements of ECCS evaluation models reviewed will include
technical adequacy of the calculational methods, including: For models
covered by paragraph (d)(2)(ii) of this section, compliance with
required features of section I of appendix K to this part; and, for
models covered by paragraph (d)(2)(i) of this section, assurance of a
high level of probability that the performance criteria of paragraph
(d)(1) of this section would not be exceeded.
(vi) For operating licenses issued under this part as of [EFFECTIVE
DATE OF RULE], required documentation of Table 1 in paragraph (o) of
this section must be submitted to demonstrate compliance by the date
specified in Table 1 in paragraph (o) of this section.
(e) Alternate risk-informed approach for addressing the effects of
debris on long-term core cooling.
(1) Risk-informed approach acceptance criteria. An entity may
request the NRC to approve a risk-informed approach for addressing the
effects of debris on long-term core cooling to demonstrate compliance
with the requirements in paragraph (d)(1)(ii) of this section. The
risk-informed approach must:
(i) Provide reasonable confidence that any increase in core damage
frequency and large early release frequency resulting from implementing
the alternative risk-informed approach will be small;
(ii) Maintain sufficient defense-in-depth and safety margins;
(iii) Consider results and insights from the probabilistic risk
assessment (PRA); and
(iv) Utilize a PRA that, at a minimum, models severe accident
scenarios resulting from internal events occurring at full power
operation and reasonably reflects the current plant configuration and
operating practices, and applicable plant and industry operational
experience, is of sufficient scope, level of detail, and technical
adequacy to support the alternative process, and is subjected to a peer
review process that assesses the PRA against a standard or set of
acceptance criteria that is endorsed by the NRC.
(2) Contents of application. An entity seeking to use the risk-
informed approach under paragraph (e)(1) of this section, must submit
an application with the following information:
(i) A description of the alternative risk-informed approach;
(ii) A description of the measures taken to assure that the scope,
level of detail and technical adequacy of the systematic processes that
evaluate the plant for internal and external events initiated during
full power, low power, and shutdown operation (including the PRA,
margins-type approaches, or other systematic evaluation techniques used
to evaluate severe accidents) are commensurate with the reliance on
risk information;
(iii) Results of the PRA review process conducted to satisfy the
requirements of paragraphs (e)(1)(iii) and (iv) of this section;
(iv) A description of, and basis for acceptability of, the
evaluations conducted to demonstrate compliance with paragraphs
(e)(1)(i) and (ii) of this section; and
(v) The analytical limit on long-term peak cooling temperature as
established in paragraph (g)(1)(v) of this section.
(3) NRC approval. If the NRC determines that the application
demonstrates that the requirements of paragraph (e)(1) of this section
are met, and the application establishes an acceptable long-term peak
cladding temperature limit, then it may approve the use of the risk-
informed approach for addressing debris effects on long-term cooling
when issuing the license, regulatory approval or amendments thereto.
The NRC's approval must specify the circumstances under which the
licensee or design certification applicant, as applicable, shall notify
the NRC of changes or errors in the risk evaluation approach utilized
to address the effects of debris on long-term cooling.
(f) [Reserved]
[[Page 16141]]
(g) Fuel system designs: Uranium oxide or mixed uranium-plutonium
oxide pellets within cylindrical zirconium-alloy cladding.
(1) Fuel performance criteria. Fuel consisting of uranium oxide or
mixed uranium-plutonium oxide pellets within cylindrical zirconium-
alloy cladding must be designed to meet the following requirements:
(i) Peak cladding temperature. Except as provided in paragraph
(g)(1)(ii) of this section, the calculated maximum fuel element
cladding temperature shall not exceed 2200[emsp14][deg]F.
(ii) Cladding embrittlement. Analytical limits on peak cladding
temperature and integral time at temperature shall be established that
correspond to the measured ductile-to-brittle transition for the
zirconium-alloy cladding material based on an NRC-approved experimental
technique. The calculated maximum fuel element temperature and time at
elevated temperature shall not exceed the established analytical
limits. The analytical limits must be approved by the NRC. If the peak
cladding temperature, in conjunction with the integral time at
temperature analytical limit, established to preserve cladding
ductility is lower than the 2200 [deg]F limit specified in paragraph
(g)(1)(i) of this section, then the lower temperature shall be used in
place of the 2200[emsp14][deg]F limit.
(iii) Breakaway oxidation. The total accumulated time that the
cladding is predicted to remain above a temperature at which the
zirconium-alloy has been shown to be susceptible to breakaway oxidation
shall not be greater than a limit that corresponds to the measured
onset of breakaway oxidation for the zirconium-alloy cladding material
based on an NRC-approved experimental technique. The limit must be
approved by the NRC.
(iv) Maximum hydrogen generation. The calculated total amount of
hydrogen generated from any chemical reaction of the fuel cladding with
water or steam shall not exceed 0.01 times the hypothetical amount that
would be generated if all of the metal in the cladding cylinders
surrounding the fuel, excluding the cladding surrounding the plenum
volume, were to react.
(v) Long-term cooling. An analytical limit on long-term peak
cladding temperature shall be established that corresponds to the
ductile-to-brittle transition for the zirconium-alloy cladding material
determined using an NRC-approved experimental technique. The analytical
limit must be approved by the NRC.
(2) Fuel system modeling requirements. The ECCS evaluation model
required by paragraph (d)(2) of this section must model the fuel system
in accordance with the following requirement:
(i) If an oxygen source is present on the inside surfaces of the
cladding at the onset of the LOCA, then the effects of oxygen diffusion
from the cladding inside surfaces must be considered in the ECCS
evaluation model.
(ii) The thermal effects of crud and oxide layers that accumulate
on the fuel cladding during plant operation must be evaluated. For the
purposes of this paragraph, crud means any foreign substance deposited
on the surface of fuel cladding prior to initiation of a LOCA.
(h) [Reserved]
(i) [Reserved]
(j) [Reserved]
(k) Use of NRC-approved fuel in reactor. A licensee may not load
fuel into a reactor, or operate the reactor, unless the licensee either
determines that the fuel meets the requirements of paragraph (d) of
this section, or complies with technical specifications governing lead
test assemblies in its license.
(l) Authority to impose restrictions on operation. The Director of
the Office of Nuclear Reactor Regulation or the Director of the Office
of New Reactors may impose restrictions on reactor operation if it is
found that the evaluations of ECCS cooling performance submitted are
not consistent with the requirements of this section.
(m) Corrective actions and reporting. Each entity subject to the
requirements of this section must comply with paragraphs (m)(1) through
(3) of this section. Each entity demonstrating acceptable long-term
core cooling under the provisions of paragraph (e) of this section
shall also comply with the requirements of paragraph (m)(4) of this
section.
(1) Categories of changes, errors, or operation inconsistent with
the ECCS evaluation model.
(i) If an entity identifies any change to, or error in, an ECCS
evaluation model or the application of such a model, or any operation
inconsistent with the ECCS evaluation model or resulting noncompliance
with the acceptance criteria in this section, that does not result in
any predicted response that exceeds any acceptance criteria specified
in this section and is itself not significant, then a report describing
each such change, error, or operation and a demonstration that the
error, change, or operation is not significant must be submitted to the
NRC no later than 12 months after the change or discovery of the error,
or operation.
(ii) If an entity identifies a change, error, or operation
inconsistent with the ECCS evaluation model that does not result in any
predicted response that exceeds any of the acceptance criteria but is
significant, then a report describing each such change, error, or
operation, and a schedule for submitting a reanalysis and
implementation of corrective actions must be submitted within 30 days
of the change, discovery of the error, or operation.
(iii) If a licensee of a facility licensed to operate identifies a
change, error, or operation inconsistent with the ECCS evaluation model
that results in any of the acceptance criteria specified in this
section to be exceeded at the facility, then the licensee shall report
the change, error, or operation under Sec. Sec. 50.55(e), 50.72, and
50.73, as applicable, and submit a report describing each such change,
error, or operation and a schedule for submitting a reanalysis and
implementation of corrective actions within 30 days of the change,
discovery of the error, or operation. In addition, the licensee (in the
case of a combined license under part 52 of this chapter, after the
Commission has made the finding under Sec. 52.103(g) shall take
immediate action to bring the facility into compliance with the
acceptance criteria.
(iv) If a design certification applicant is required by paragraphs
(m)(1)(ii) of this section to submit a reanalysis, or identifies a
change, error, or operation that results in any predicted response that
exceeds any of the acceptance criteria specified in this section, then
the applicant must submit a reanalysis, accompanied by either a
revision to its design certification application under review, or an
application to amend the design certification application, as
applicable, reflecting the reanalysis.
(2) Significant change or error in the ECCS evaluation model. For
the purposes of paragraph (m)(1) of this section, a significant change
or error in an ECCS evaluation model is one that results in a
calculated-
(i) Peak fuel cladding temperature different by more than
50[emsp14][deg]F from the temperature calculated for the limiting
transient using the last NRC-approved ECCS evaluation model, or is a
cumulation of changes and errors such that the sum of the absolute
magnitudes of the respective temperature changes is greater than
50[emsp14][deg]F; or
(ii) Integral time at temperature different by more than 0.4
percent ECR from the oxidation calculated for the
[[Page 16142]]
limiting transient using the last NRC-approved ECCS evaluation model,
or is a cumulation of changes and errors such that the sum of the
absolute magnitudes of the respective oxidation changes is greater than
0.4 percent ECR.
(3) Breakaway oxidation. Each holder of an operating license or
combined license shall measure breakaway oxidation for each reload
batch. The holder must report the results to the NRC annually (i.e.,
anytime within each calendar year), in accordance with Sec. 50.4 or
Sec. 52.3 of this chapter, and evaluate the results to determine if
there is a failure to conform or a defect that must be reported in
accordance with the requirements of 10 CFR part 21.
(4) Updates to risk-informed consideration of debris in long-term
cooling.
(i) Design certification before issuance of final design
certification rule. If a design certification applicant, after
performing the evaluation under paragraph (e) of this section and
including the information in its application, determines that any
acceptance criterion of paragraph (e)(1) of this section is not met,
then the applicant shall submit a report describing its determination.
Thereafter, the applicant shall submit, in a timely manner, an
amendment to its pending design certification application. The
amendment application must describe any changes to the certified design
and/or changes in the analyses, evaluations, and modeling (including
the debris evaluation model and the PRA and its supporting analyses)
needed to demonstrate that the certified design meets the acceptance
criteria in paragraph (e)(1) of this section.
(ii) Design certification during the period of validity under Sec.
52.55(a) and (b) of this chapter--not currently referenced in any COL
application or COL. The design certification applicant need not report
any information concerning compliance with the acceptance criterion of
paragraph (e)(1) of this section in accordance with the requirements of
part 21 of this chapter until 30 days after the design certification is
referenced by a COL applicant.
(iii) Design certification during the period of validity under
Sec. 52.55(a) and (b) of this chapter--once referenced in a COL
application or COL. The design certification applicant shall evaluate
and report any information concerning compliance with the acceptance
criterion of paragraph (e)(1) of this section in accordance with the
requirements of part 21 of this chapter.
(iv) Design certification--renewal. The applicant for renewal of a
design certification shall update the debris evaluation model and the
PRA and its supporting analyses, taking into account all known
applicable industry operational experience. The applicant shall re-
perform the evaluations of risk, defense-in-depth, and safety margins
using the updated model. If any of the acceptance criteria in paragraph
(e)(1) of this section are not met, then applicant shall include
necessary changes to the certified design, debris evaluation model, PRA
or supporting analyses to demonstrate that the renewed certified design
meets the acceptance criteria in paragraph (e)(1) of this section.
(v) Combined license application. If a combined license applicant,
after performing the evaluation required by paragraph (e) of this
section and including the information in its application, determines
that any acceptance criterion of paragraph (e)(1) of this section is
not met, then the applicant shall submit a report describing its
determination within 30 days of completion of the determination.
Thereafter, the applicant shall submit, in a timely manner, an
amendment to its pending combined license application. The amendment
application must describe any changes to the design of the facility
and/or changes in the analyses, evaluations, and modeling (including
the debris evaluation model and the PRA and its supporting analyses)
needed to demonstrate that the design of the facility meets the
acceptance criteria in paragraph (e)(1) of this section, any necessary
changes to previously-submitted inspections, tests, analyses and
acceptance criteria, and either the bases for any change to the
inspections, tests, analyses, and acceptance criteria (ITAAC) or why no
changes to the ITAAC are needed.
(vi) Combined licenses before finding under Sec. 52.103(g)of this
chapter. Each holder of a combined license must, no later than the
scheduled date for initial loading of fuel under Sec. 52.103(a) of
this chapter, update the analyses, evaluations, and modeling performed
under paragraph (e) of this section. The updating must correct
identified errors, and incorporate licensee-adopted changes to the
plant design, the licensee's proposed operational practices, and any
applicable industry operational experience known to the licensee. As
appropriate, the licensee shall update the debris evaluation model and
the PRA and its supporting analyses, and re-perform the evaluations of
risk, defense-in-depth, and safety margins to confirm that the
acceptance criteria identified in paragraph (e)(1) of this section
continue to be met. After submitting the update under this paragraph
and until the Commission has made the finding under Sec. 52.103(g) of
this chapter, the licensee shall re-perform this evaluation in a timely
manner if the licensee identifies a change or error in the analyses,
evaluations, and modeling, makes a change in the plant design or the
plant's proposed operational practices, or identifies applicable
industry operational experience. The licensee shall re-perform the
evaluation, even if no changes or errors are identified, by no later
than 48 months after the last review. If the licensee determines that
any acceptance criterion of paragraph (e)(1) of this section is not
met, then the licensee shall submit, in a timely fashion, an
application for amendment of its combined license (and departure from a
referenced design certification rule, if applicable), including
necessary changes to its updated final safety analysis report and any
necessary changes to the ITAAC. The amendment application must
demonstrate that the acceptance criteria of paragraph (e)(1) of this
section are met, and must describe any changes to the analyses,
evaluations and modeling needed to support that conclusion. The
application must explain either the bases for any change to ITAAC or
why no changes to ITAAC are needed. The application must, if
applicable, include a request for exemption from a referenced design
certification rule, but need not address the criteria for obtaining an
exemption. The licensee shall also submit any report required by Sec.
52.99 of this chapter. The NRC need not address the issue finality
criteria in Sec. Sec. 52.63, 52.83, and 52.98 of this chapter when
acting on this amendment, and shall--as part of any approved
amendment--issue any necessary exemption upon a finding that the
exemption is authorized by law and will not endanger life or property
or the common defense and security and are otherwise in the public
interest.
(vii) Operating licenses and combined licenses after finding under
Sec. 52.103(g) of this chapter--updating and corrections. The licensee
shall review the analyses, evaluations, and modeling performed under
paragraph (e) of this section for changes and errors and incorporate
changes to the design, plant, operational practices, and applicable
plant and industry operational experience. As appropriate, the licensee
shall update the debris evaluation model and the PRA and its supporting
analyses, and re-perform the evaluations of risk, defense-in-depth, and
safety margins to confirm that the acceptance criteria identified in
paragraph (e)(1) of
[[Page 16143]]
this section continue to be met. The licensee shall perform this review
in a timely manner after a change or error is identified in the
analyses, evaluations, and modeling or a change is identified in the
design, plant, operational practices, or applicable plant and industry
operational experience. The licensee shall perform this review even if
no changes or errors are identified, by no later than 48 months after
the last review. If the licensee, at any time, determines that any
acceptance criterion of paragraph (e)(1) of this section is not met,
then the licensee shall take action in a timely manner to bring the
facility into compliance with the acceptance criteria of paragraph
(e)(1) of this section. The licensee shall also report the failure to
meet the long-term cooling acceptance criterion in paragraph (e)(1) of
this section. The report must be prepared and submitted in accordance
with, Sec. Sec. 50.72, and 50.73, as applicable. Thereafter, the
licensee shall submit, in a timely fashion, an application for
amendment of its license, including necessary changes to its updated
final safety analysis report. The amendment application must
demonstrate that the acceptance criteria of paragraph (e)(1) of this
section are met, and must describe any changes to the analyses,
evaluations and modeling needed to support that conclusion. The
amendment application for a combined license must, if applicable,
include a request for exemption from a referenced design certification
rule, but need not address the criteria for obtaining an exemption. The
NRC need not address either the backfitting criteria in Sec. 50.109 or
the issue finality criteria in Sec. Sec. 52.63, 52.83, and 52.98 of
this chapter when acting on this amendment and shall, as part of any
approved amendment, issue any necessary exemption upon a finding that
the exemption is authorized by law and will not endanger life or
property or the common defense and security and are otherwise in the
public interest.
(n) [Reserved]
(o) Implementation.
(1) Construction permits issued under this part after [EFFECTIVE
DATE OF RULE] must comply with the requirements of this section at
their issuance.
(2) Operating licenses issued under this part that are based upon
construction permits in effect as of [EFFECTIVE DATE OF RULE]
(including deferred and reinstated construction permits) must comply
with the requirements of this section by no later than the applicable
date set forth in Table 1 in paragraph (o) of this section. Until such
compliance is achieved, the requirements of Sec. 50.46 continue to
apply.
(3) Operating licenses issued under this part after [EFFECTIVE DATE
OF RULE] must comply with the requirements of this section.
(4) Operating licenses issued under this part as of [EFFECTIVE DATE
OF RULE] must comply with the requirements of this section by no later
than the applicable date set forth in Table 1 in paragraph (o) of this
section. Until such compliance is achieved, the requirements of Sec.
50.46 continue to apply.
(5) Standard design certifications, standard design approvals, and
manufacturing licenses under part 52 of this chapter, whose
applications (including applications for amendment) are docketed after
[EFFECTIVE DATE OF RULE], and new branches of these certifications
whose applications are docketed after [EFFECTIVE DATE OF RULE] must
comply with this section at their issuance.
(6) Standard design certifications under part 52 of this chapter
issued before [EFFECTIVE DATE OF RULE] must comply with this section by
the time of renewal.
(7) Standard design certifications, standard design approvals, and
manufacturing licenses under part 52 of this chapter issued after
[EFFECTIVE DATE OF RULE] whose applications were pending as of
[EFFECTIVE DATE OF RULE] and new branches of certifications issued
after [EFFECTIVE DATE OF RULE] whose applications were pending as of
[EFFECTIVE DATE OF RULE] must comply with this section by the time of
renewal.
(8) Combined license applications under part 52 of this chapter
whose applications are docketed after [EFFECTIVE DATE OF RULE] must
comply with this section.
(9) Combined licenses issued under part 52 of this chapter, before
[EFFECTIVE DATE OF RULE] and combined licenses issued after the
[EFFECTIVE DATE OF RULE] whose applications were docketed before
[EFFECTIVE DATE OF RULE] must comply with this section no later than
completion of the first refueling outage after initial fuel load. Until
such compliance is achieved, the requirements in Sec. 50.46 continue
to apply.
Table 1: Implementation Dates for Nuclear Power Plants with
Operating Licenses as of [EFFECTIVE DATE OF RULE].
----------------------------------------------------------------------------------------------------------------
Track Reactor type Plant name Compliance demonstration
----------------------------------------------------------------------------------------------------------------
1........................ PWR...................... Arkansas Nuclear One--Unit 1 No later than 24 months from
Braidwood Station--Unit 1... effective date of rule.
Byron Station--Unit 1.......
Calvert Cliffs Nuclear Power
Plant--Unit 1.
Calvert Cliffs Nuclear Power
Plant--Unit 2.
Comanche Peak Nuclear Power
Plant--Unit 1.
Comanche Peak Nuclear Power
Plant--Unit 2.
Davis-Besse Nuclear Power
Station--Unit 1.
Diablo Canyon Power Plant--
Unit 2.
Fort Calhoun Station--Unit 1
H.B. Robinson Steam Electric
Plant--Unit 2.
Indian Point Nuclear
Generating Station--Unit 2.
J.M. Farley Nuclear Plant--
Unit 1.
J.M. Farley Nuclear Plant--
Unit 2.
Millstone Power Station--
Unit 2.
Millstone Power Station--
Unit 3.
North Anna Power Station--
Unit 1.
North Anna Power Station--
Unit 2.
Oconee Nuclear Station--Unit
1.
Oconee Nuclear Station--Unit
2.
Oconee Nuclear Station--Unit
3.
Palisades Nuclear Plant.....
Point Beach Nuclear Plant--
Unit 1.
[[Page 16144]]
Point Beach Nuclear Plant--
Unit 2.
Prairie Island Nuclear
Generating Plant--Unit 1.
Prairie Island Nuclear
Generating Plant--Unit 2.
R.E. Ginna Nuclear Power
Plant.
Saint Lucie Plant--Unit 1...
Seabrook Station--Unit 1....
Sequoyah Nuclear Plant--Unit
1.
Sequoyah Nuclear Plant--Unit
2.
Three Mile Island--Unit 1...
Turkey Point Nuclear
Generating Station--Unit 3.
Turkey Point Nuclear
Generating Station--Unit 4.
Vogtle Electric Generating
Plant--Unit 1.
Vogtle Electric Generating
Plant--Unit 2.
Wolf Creek Generating
Station--Unit 1.
BWR...................... Browns Ferry Nuclear Plant--
Unit 1.
Browns Ferry Nuclear Plant--
Unit 2.
Browns Ferry Nuclear Plant--
Unit 3.
Brunswick Steam Electric
Plant--Unit 1.
Brunswick Steam Electric
Plant--Unit 2.
Clinton Power Station--Unit
1.
Columbia Generating Station.
Cooper Nuclear Station......
Duane Arnold Energy Center..
E.I. Hatch Nuclear Plant--
Unit 1.
E.I. Hatch Nuclear Plant--
Unit 2.
Fermi--Unit 2...............
Hope Creek Generating
Station--Unit 1.
Grand Gulf Nuclear Station--
Unit 1.
J.A. Fitzpatrick Nuclear
Power Plant.
LaSalle County Station--Unit
1.
LaSalle County Station--Unit
2.
Limerick Generating Station--
Unit 1.
Limerick Generating Station--
Unit 2.
Nine Mile Point Nuclear
Station--Unit 2.
Peach Bottom Atomic Power
Station--Unit 2.
Peach Bottom Atomic Power
Station--Unit 3.
Perry Nuclear Power Plant--
Unit 1.
River Bend Station--Unit 1..
Susquehanna Steam Electric
Station--Unit 1.
Susquehanna Steam Electric
Station--Unit 2.
Vermont Yankee Nuclear Power
Station.
2........................ PWR...................... Beaver Valley Power Station-- No later than 48 months from
Unit 1. effective date of rule.
Beaver Valley Power Station--
Unit 2..
Braidwood Station--Unit 2...
Byron Station--Unit 2.......
Catawba Nuclear Station--
Unit 1.
Catawba Nuclear Station--
Unit 2.
D.C. Cook Nuclear Plant--
Unit 1.
D.C. Cook Nuclear Plant--
Unit 2.
Diablo Canyon Power Plant--
Unit 1.
Indian Point Nuclear
Generating Station--Unit 3.
McGuire Nuclear Station--
Unit 1.
McGuire Nuclear Station--
Unit 2.
Watts Bar Nuclear Plant--
Unit 1.
BWR...................... Nine Mile Point Nuclear
Station--Unit 1.
Oyster Creek Nuclear
Generating Station.
3........................ PWR...................... Arkansas Nuclear One--Unit 2 No later than 60 months from
Callaway Plant--Unit 1...... effective date of rule.
Palo Verde Nuclear
Generating Station--Unit 1..
Palo Verde Nuclear
Generating Station--Unit 2.
Palo Verde Nuclear
Generating Station--Unit 3.
Saint Lucie Plant--Unit 2...
Salem Nuclear Generating
Station--Unit 1.
Salem Nuclear Generating
Station--Unit 2.
Shearon Harris Nuclear Power
Plant--Unit 1.
South Texas Project--Unit 1.
South Texas Project--Unit 2.
Surry Power Plant--Unit 1...
Surry Power Plant--Unit 2...
V.C. Summer Nuclear Station--
Unit 1.
Waterford Steam Electric
Station--Unit 3.
BWR...................... Dresden Nuclear Power
Station--Unit 2.
Dresden Nuclear Power
Station--Unit 3.
Monticello Nuclear
Generating Plant--Unit 1.
[[Page 16145]]
Pilgrim Nuclear Power
Station.
Quad Cities Nuclear Power
Station--Unit 1.
Quad Cities Nuclear Power
Station--Unit 2.
----------------------------------------------------------------------------------------------------------------
* * * * *
0
6. In appendix A to part 50, under the heading, ``Criteria,'' criteria
35, 38, and 41 are revised to read as follows:
Appendix A to Part 50--General Design Criteria for Nuclear Power Plants
* * * * *
Criterion 35--Emergency core cooling. A system to provide
abundant emergency core cooling shall be provided. The system safety
function shall be to transfer heat from the reactor core following
any loss of reactor coolant at a rate such that 1) fuel and clad
damage that could interfere with continued effective core cooling is
prevented and 2) clad metal-water reaction is limited to negligible
amounts.
Suitable redundancy in components and features, and suitable
interconnections, leak detection, isolation, and containment
capabilities shall be provided to assure that for onsite electric
power operation (assuming offsite power is not available) and for
offsite electric power system operation (assuming onsite power is
not available) the system safety function can be accomplished,
assuming a single failure.
The effects of debris on system safety function with respect to
long-term cooling may be evaluated in accordance with all
requirements applicable to the risk-informed approach in Sec.
50.46c.
* * * * *
Criterion 38--Containment heat removal system. A system to
remove heat from the reactor containment shall be provided. The
system safety function shall be to reduce rapidly, consistent with
the functioning of other associated systems, the containment
pressure and temperature following any loss-of-coolant accident and
maintain them at acceptably low levels.
Suitable redundancy in components and features, and suitable
interconnections, leak detection, isolation, and containment
capabilities shall be provided to assure that for onsite electric
power system operation (assuming offsite power is not available) and
for offsite electric power system operation (assuming onsite power
is not available) the system safety function can be accomplished,
assuming a single failure.
The effects of debris on safety system function with respect to
the maintenance of containment pressure and temperature may be
evaluated in accordance with all requirements applicable to the
risk-informed approach in Sec. 50.46c.
* * * * *
Criterion 41--Containment atmosphere cleanup. Systems to control
fission products, hydrogen, oxygen, and other substances which may
be released into the reactor containment shall be provided as
necessary to reduce, consistent with the functioning of other
associated systems, the concentration and quality of fission
products released to the environment following postulated accidents,
and to control the concentration of hydrogen or oxygen and other
substances in the containment atmosphere following postulated
accidents to assure that containment integrity is maintained.
Each system shall have suitable redundancy in components and
features, and suitable interconnections, leak detection, isolation,
and containment capabilities to assure that for onsite electric
power system operation (assuming offsite power is not available) and
for offsite electric power system operation (assuming onsite power
is not available) its safety function can be accomplished, assuming
a single failure.
The effects of debris on system safety function following
occurrence of the postulated accidents may be evaluated in
accordance with all requirements applicable to the risk-informed
approach in Sec. 50.46c.
* * * * *
0
7. In appendix K to part 50, a new paragraph II.6 is added to read as
follows:
Appendix K to Part 50--ECCS Evaluation Models
* * * * *
II. * * *
6. Upon implementation of Sec. 50.46c in accordance with Sec.
50.46c(o), the documentation requirements in Sec. 50.46c(d)(3)
apply and supersede the requirements of section II of this appendix.
PART 52--LICENSES, CERTIFICATIONS AND APPROVALS FOR NUCLEAR POWER
PLANTS
0
8. The authority citation for part 52 continues to read as follows:
Authority: Secs. 103, 104, 147, 149, 161, 181, 182, 183, 185,
186, 189, 223, 234 (42 U.S.C. 2133, 2167, 2169, 2201, 2232, 2233,
2235, 2236, 2239, 2282); Energy Reorganization Act secs. 201, 202,
206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Government Paperwork
Elimination Act sec. 1704 (44 U.S.C. 3504 note); Energy Policy Act
of 2005, Pub. L. 109-58, 119 Stat. 594 (2005).
0
9. In Sec. 52.47, paragraph (a)(4) is revised to read as follows:
Sec. 52.47 Contents of applications; technical information
* * * * *
(a) * * *
(4) An analysis and evaluation of the design and performance of
structures, systems, and components with the objective of assessing the
risk to public health and safety resulting from operation of the
facility and including determination of the margins of safety during
normal operations and transient conditions anticipated during the life
of the facility, and the adequacy of structures, systems, and
components provided for the prevention of accidents and the mitigation
of the consequences of accidents. Analysis and evaluation of emergency
core cooling system (ECCS) cooling performance and the need for high-
point vents following postulated loss-of-coolant accidents shall be
performed in accordance with the requirements of Sec. Sec. 50.46,
50.46b and 50.46c of this chapter, as applicable;
* * * * *
0
10. In Sec. 52.79, paragraph (a)(5) is revised to read as follows:
Sec. 52.79 Contents of applications; technical information in final
safety analysis report.
(a) * * *
(5) An analysis and evaluation of the design and performance of
structures, systems, and components with the objective of assessing the
risk to public health and safety resulting from operation of the
facility and including determination of the margins of safety during
normal operations and transient conditions anticipated during the life
of the facility, and the adequacy of structures, systems, and
components provided for the prevention of accidents and the mitigation
of the consequences of accidents. Analysis and evaluation of ECCS
cooling performance and the need for high-point vents following
postulated loss-of-coolant accidents shall be performed in accordance
with the requirements of Sec. Sec. 50.46, 50.46b and 50.46c of this
chapter, as applicable;
* * * * *
0
11. In Sec. 52.137, paragraph (a)(4) is revised to read as follows:
Sec. 52.137 Contents of applications; technical information.
* * * * *
(a) * * *
(4) An analysis and evaluation of the design and performance of
SSCs with the objective of assessing the risk to public health and
safety resulting from operation of the facility and including
determination of the margins of safety during normal operations and
transient conditions anticipated during the life of the facility, and
the adequacy of SSCs provided for the prevention of accidents and the
mitigation of the consequences of accidents. Analysis and evaluation of
ECCS cooling performance and the need
[[Page 16146]]
for high-point vents following postulated loss-of-coolant accidents
shall be performed in accordance with the requirements of Sec. Sec.
50.46, 50.46b, and 50.46c of this chapter, as applicable;
* * * * *
0
12. In Sec. 52.157, paragraph (f)(1) is revised to read as follows:
Sec. 52.157 Contents of applications; technical information in the
final safety analysis report.
* * * * *
(f) * * *
(1) An analysis and evaluation of the design and performance of
structures, systems, and components with the objective of assessing the
risk to public health and safety resulting from operation of the
facility and including determination of the margins of safety during
normal operations and transient conditions anticipated during the life
of the facility, and the adequacy of structures, systems, and
components provided for the prevention of accidents and the mitigation
of the consequences of accidents. Analysis and evaluation of ECCS
cooling performance and the need for high-point vents following
postulated loss-of-coolant accidents shall be performed in accordance
with the requirements of Sec. Sec. 50.46, 50.46b, and 50.46c of this
chapter, as applicable;
* * * * *
Dated at Rockville, Maryland, this 6th day of March, 2013.
For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 2014-05562 Filed 3-21-14; 8:45 am]
BILLING CODE 7590-01-P