[Federal Register Volume 79, Number 56 (Monday, March 24, 2014)]
[Proposed Rules]
[Pages 16105-16146]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2014-05562]



[[Page 16105]]

Vol. 79

Monday,

No. 56

March 24, 2014

Part II





 Nuclear Regulatory Commission





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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]]


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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.

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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DATES: Submit comments on the rule and draft guidance by June 9, 2014.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on the proposed rule and comments submitted on the draft guidance.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: Carol.Gallagher@nrc.gov. For technical questions, please 
contact the individuals listed in the FOR FURTHER INFORMATION CONTACT 
section of this document.
     Email comments to: Rulemaking.Comments@nrc.gov. 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: Tara.Inverso@nrc.gov; or 
Paul M. Clifford, Office of Nuclear Reactor Regulation, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, telephone: 301-415-
4043, email: Paul.Clifford@nrc.gov.

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 PDR.Resource@nrc.gov. 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\
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    \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.
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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 Rulemaking.Comments@nrc.gov 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 INFOCOLLECTS.RESOURCE@NRC.GOV 
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 Chad_S_Whiteman@omb.eop.gov 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