[Federal Register Volume 74, Number 152 (Monday, August 10, 2009)]
[Proposed Rules]
[Pages 40006-40052]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-18547]



[[Page 40005]]

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Part II





Nuclear Regulatory Commission





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10 CFR Parts 50 and 52



Risk-Informed Changes to Loss-of-Coolant Accident Technical 
Requirements; Proposed Rule

Federal Register / Vol. 74, No. 152 / Monday, August 10, 2009 / 
Proposed Rules

[[Page 40006]]


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NUCLEAR REGULATORY COMMISSION

10 CFR Parts 50 and 52

[NRC-2004-0006]
RIN 3150-AH29


Risk-Informed Changes to Loss-of-Coolant Accident Technical 
Requirements

AGENCY: Nuclear Regulatory Commission.

ACTION: Supplemental proposed rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend 
its regulations that govern domestic licensing of production and 
utilization facilities and licenses, certifications, and approvals for 
nuclear power plants to allow current and certain future power reactor 
licensees and applicants to choose to implement a risk-informed 
alternative to the current requirements for analyzing the performance 
of emergency core cooling systems (ECCS) during loss-of-coolant 
accidents (LOCAs). The proposed amendments would also establish 
procedures and acceptance criteria for evaluating certain changes in 
plant design and operation based upon the results of the new analyses 
of ECCS performance.

DATES: Submit comments on this supplemental proposed rule by September 
24, 2009. Submit comments specific to the information collections 
aspects of this supplemental proposed rule by September 9, 2009. 
Comments received after the above 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: You may submit comments by any one of the following methods. 
Comments submitted in writing or in electronic form will be made 
available for public inspection. Because your comments will not be 
edited to remove any identifying or contact information, the NRC 
cautions you against including any information in your submission that 
you do not want to be publicly disclosed. You may submit comments on 
the information collections by the methods indicated in the Paperwork 
Reduction Act Statement of this document.
    Federal e Rulemaking Portal: Go to http://www.regulations.gov and 
search for documents filed under Docket ID NRC-2004-0006. Address 
questions about NRC dockets to Carol Gallagher, (301) 415-5905; e-mail 
<a href="/cdn-cgi/l/email-protection#88cbe9fae7e4a6cfe9e4e4e9efe0edfac8e6faeba6efe7fe"><span class="__cf_email__" data-cfemail="7c3f1d0e1310523b1d10101d1b14190e3c120e1f521b130a">[email protected]</span></a>.
    Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attn: Rulemakings and Adjudications Staff.
    E-mail comments to: <a href="/cdn-cgi/l/email-protection#8bd9fee7eee6eae0e2e5eca5c8e4e6e6eee5fff8cbe5f9e8a5ece4fd"><span class="__cf_email__" data-cfemail="cd9fb8a1a8a0aca6a4a3aae38ea2a0a0a8a3b9be8da3bfaee3aaa2bb">[email protected]</span></a>. If you do not 
receive a reply e-mail confirming that we have received your comments, 
contact us directly at (301) 415-1966.
    Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland 
20852, between 7:30 a.m. and 4:15 p.m. during Federal workdays. 
(Telephone (301) 415-1966).
    Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at 
(301) 415-1101.
    You can access publicly available documents related to this 
document using the following methods:
    NRC's Public Document Room (PDR): The public may examine publicly 
available documents at the NRC's PDR, Public File Area O-F21, One White 
Flint North, 11555 Rockville Pike, Rockville, Maryland. The PDR 
reproduction contractor will copy documents for a fee.
    NRC's Agencywide Document Access and Management System (ADAMS): 
Publicly available documents created or received at the NRC are 
available electronically at the NRC's Electronic Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this page, the public can gain 
entry into ADAMS, which provides text and image files of NRC's public 
documents. If you do not have access to ADAMS or if there are problems 
in accessing the documents located in ADAMS, contact the NRC's PDR 
reference staff at 1-800-397-4209, or (301) 415-4737, or by e-mail to 
<a href="/cdn-cgi/l/email-protection#6b3b2f3945390e18041e19080e2b051908450c041d"><span class="__cf_email__" data-cfemail="8adaced8a4d8eff9e5fff8e9efcae4f8e9a4ede5fc">[email protected]</span></a>.

FOR FURTHER INFORMATION CONTACT: Richard Dudley, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone (301) 415-1116; e-mail: <a href="/cdn-cgi/l/email-protection#b1c3d8d2d9d0c3d59fd5c4d5ddd4c8f1dfc3d29fd6dec7"><span class="__cf_email__" data-cfemail="44362d272c2536206a20312028213d042a36276a232b32">[email protected]</span></a>.

SUPPLEMENTARY INFORMATION: 

Table of Contents

I. Background
II. Rulemaking Initiation
III. Description of Proposed Rule
IV. Discussion on Public Comments
    A. Comments on Selection of the TBS
    B. Comments on Seismic Considerations Related to the TBS
    C. Comments on Thermal-Hydraulic Analysis
    D. Comments Related to Probabilistic Risk Assessment
    E. Comments Related to Applicability of the Backfit Rule
    F. Comments on Topics Requested by the NRC
V. Revised Proposed Rule
    A. Overview
    B. Determination of the Transition Break Size
    C. Evaluation of the Plant-Specific Applicability of the 
Transition Break Size
    D. Alternative ECCS Analysis Requirements and Acceptance 
Criteria
    E. Risk-Informed Changes to the Facility, Technical 
Specifications, and Procedures
    F. Operational Requirements
    G. Reporting Requirements
    H. Documentation Requirements
    I. Submittal and Review of Applications
    J. Applicability to New Reactor Designs
VI. Specific Topics Identified for Public Comments
VII. Petition for Rulemaking, PRM-50-75
VIII. Section-by-Section Analysis of Changes
IX. Criminal Penalties
X. Compatibility of Agreement State Regulations
XI. Availability of Documents
XII. Plain Language
XIII. Voluntary Consensus Standards
XIV. Finding of No Significant Environmental Impact: Environmental 
Assessment
XV. Paperwork Reduction Act Statement
XVI. Regulatory Analysis
XVII. Regulatory Flexibility Certification
XVIII. Backfit Analysis

I. Background

    During the last few years, the NRC has had numerous initiatives 
underway to make improvements in its regulatory requirements that would 
reflect current knowledge about reactor risk. The overall objectives of 
risk-informed modifications to reactor regulations include:
    (1) Enhancing safety by focusing NRC and licensee resources in 
areas commensurate with their importance to health and safety;
    (2) Providing NRC with the framework to use risk information to 
take action in reactor regulatory matters, and
    (3) Allowing use of risk information to provide flexibility in 
plant operation and design, which can result in reduction of burden 
without compromising safety, improvements in safety, or both.
    The Commission published a Policy Statement on the Use of 
Probabilistic Risk Assessment (PRA) on August 16, 1995 (60 FR 42622). 
In the policy statement, the Commission stated that the use of PRA 
technology should be increased in all regulatory matters to the extent 
supported by the state-of-the-art in PRA methods and data, and in a 
manner that complements the deterministic approach and that supports 
the NRC's defense-in-depth philosophy. PRA evaluations in support of 
regulatory decisions should be as realistic as practicable and 
appropriate supporting data should be publicly available. The policy 
statement also

[[Page 40007]]

stated that, in making regulatory judgments, the Commission's safety 
goals for nuclear power reactors and subsidiary numerical objectives 
(on core damage frequency and containment performance) should be used 
with appropriate consideration of uncertainties.
    To implement the policy statement, the NRC developed guidance on 
the use of risk information for reactor license amendments and issued 
Regulatory Guide (RG) 1.174, ``An Approach for Using Probabilistic Risk 
Assessments in Risk-Informed Decisions on Plant Specific Changes to the 
Licensing Basis,'' (ADAMS Accession No. ML023240437). This RG provided 
guidance on an acceptable approach to risk-informed decision-making 
consistent with the Commission's policy, including a set of key 
principles. These principles include:
    (1) Being consistent with the defense-in-depth philosophy;
    (2) Maintaining sufficient safety margins;
    (3) Allowing only changes that 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
    (4) Incorporating monitoring and performance measurement 
strategies.
    Regulatory Guide 1.174 further clarifies that in implementing these 
principles, the NRC expects that all safety impacts of the proposed 
change are evaluated in an integrated manner as part of an overall risk 
management approach in which the licensee is using risk analysis to 
improve operational and engineering decisions broadly by identifying 
and taking advantage of opportunities to reduce risk; and not just to 
eliminate requirements that a licensee sees as burdensome or 
undesirable.

II. Rulemaking Initiation

    The process described in RG 1.174 is applicable to changes to plant 
licensing bases. As NRC experience with the process and applications 
grew, the NRC recognized that further development of risk-informed 
regulation would require making changes to the regulations. In June 
1999, the Commission decided to implement risk-informed changes to the 
technical requirements of Part 50. The first risk-informed revision to 
the technical requirements of Part 50 consisted of changes to the 
combustible gas control requirements in Title 10 of the Code of Federal 
Regulations (10 CFR) Section 50.44 (68 FR 54123; September 16, 2003). 
Other risk-informed regulations promulgated by the NRC include Sec.  
50.48(c) on fire protection (69 FR 33550; June 16, 2004), Sec.  50.69 
on special treatment requirements for systems, structures, and 
components (69 FR 68047; Nov. 22, 2004), and Sec.  50.61 on fracture 
toughness requirements for protection against pressurized thermal shock 
events.
    The NRC also decided to examine the ECCS requirements for large 
break LOCAs. A number of possible changes were considered, including 
changes to General Design Criterion (GDC) 35 and changes to Sec.  50.46 
acceptance criteria, evaluation models, and functional reliability 
requirements. The NRC also proposed to refine previous estimates of 
LOCA frequency for various sizes of LOCAs to more accurately reflect 
the current state of knowledge with respect to the mechanisms and 
likelihood of primary coolant system rupture. During public meetings, 
industry representatives expressed interest in a number of possible 
changes to licensed power reactors resulting from redefinition of the 
large break LOCA. These include: containment spray system setpoint 
changes; fuel management improvements; optimization of plant 
modifications and operator actions to address postulated sump blockage 
issues; power uprates; and changes to the required number of 
accumulators, diesel start times, sequencing of equipment, and valve 
stroke times.
    The Staff Requirements Memorandum (SRM), of March 31, 2003, 
(ML030910476), on 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)' '' 
(ML020660607), approved most of the NRC staff recommendations related 
to possible changes to LOCA requirements and also directed the NRC 
staff to prepare a proposed rule that would provide a risk-informed 
alternative maximum break size. The NRC began to prepare a proposed 
rule responsive to the SRM direction. However, after holding two public 
meetings, the NRC found that there were differences between stated 
Commission and industry interests.
    To reach a common understanding about the objectives of the LOCA 
redefinition rulemaking, the NRC staff requested additional direction 
and guidance from the Commission in SECY-04-0037, ``Issues Related to 
Proposed Rulemaking to Risk-Inform Requirements Related to Large Break 
Loss-of-Coolant Accident (LOCA) Break Size and Plans for Rulemaking on 
LOCA with Coincident Loss-of-Offsite Power,'' (March 3, 2004; 
ML040490133). The Commission provided direction in a SRM dated July 1, 
2004, (ML041830412). The Commission stated that the NRC staff should 
determine an appropriate risk-informed alternative break size and that 
breaks larger than this size should be removed from the design basis 
event category. The Commission indicated that the proposed rule should 
be structured to allow operational as well as design changes and should 
include requirements for licensees to maintain capability to mitigate 
the full spectrum of LOCAs up to the double-ended guillotine break 
(DEGB) of the largest reactor coolant system (RCS) pipe. The Commission 
stated that the mitigation capabilities for beyond design-basis events 
should be controlled by NRC requirements commensurate with the safety 
significance of these capabilities. The Commission also stated that 
LOCA frequencies should be periodically reevaluated and should 
increases in frequency require licensees to restore the facility to its 
original design basis or make other compensating changes, the backfit 
rule (10 CFR 50.109) would not apply.
    On March 29, 2005, in SECY-05-0052, ``Proposed Rulemaking for 
`Risk-Informed Changes to Loss-of-Coolant Accident Technical 
Requirements,' '' the NRC staff provided a proposed rule to the 
Commission for its consideration. In an SRM on July 29, 2005, the 
Commission directed the NRC staff to publish the proposed rule for 
public comment after making certain changes. The most significant 
change requested by the Commission was to require that after 
implementing the alternative Sec.  50.46a requirements, all subsequent 
plant changes made by a licensee would be evaluated by the licensee's 
risk-informed process to ensure that they met all of the requirements 
in Sec.  50.46a. Another change requested by the Commission was to 
address the issue of seismic loading of degraded piping during very 
large earthquakes and to solicit public comments on the subject.
    On November 7, 2005, (70 FR 67598), the proposed rule was published 
in the Federal Register (FR) with a comment period of 90 days. On 
December 6, 2005, the Nuclear Energy Institute \1\ (NEI) requested that 
the comment period be extended for 30 additional days. NEI stated that 
additional time was needed to prepare high quality comments that 
reflected an industry consensus perspective. On December 20, 2005, the

[[Page 40008]]

Westinghouse Owners Group (WOG) submitted a letter endorsing the NEI 
extension request. On January 18, 2006, the NRC extended the comment 
period by 30 days to expire on March 8, 2006. As directed by the 
Commission in its SRM on SECY-05-0052, the NRC staff addressed the 
seismic issue by preparing a report entitled ``Seismic Considerations 
for the Transition Break Size'' (ML053470439). This report was posted 
on the NRC's rulemaking Web site and a notice of its availability and 
opportunity for public comment was published in the FR on December 20, 
2005, (70 FR 75501). A public workshop was held on February 16, 2006, 
to ensure that stakeholders understood the NRC's intent and 
interpretation of the proposed rule and two public meetings were held 
on June 28, 2006, and August 17, 2006, to discuss public comments 
received on the proposed rule.
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    \1\ All utilities licensed to operate commercial nuclear power 
plants in the United States are members of NEI.
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    After evaluating all written public comments and comments received 
at the public meetings, the NRC completed draft final rule language 
that addressed nearly all commenters' concerns. On October 31 and 
November 1, 2006, the NRC staff met with the Advisory Committee on 
Reactor Safeguards (ACRS) to discuss the draft final rule. In a letter 
dated November 16, 2006, (ML063190465) the ACRS provided its evaluation 
of the draft final rule. In its November 16, 2006, letter to the 
Commission, the ACRS recommended that the rule not be issued in its 
current form. The ACRS recommended numerous changes to the rule, 
primarily to increase the defense-in-depth provided for large pipe 
breaks. The NRC staff evaluated the ACRS recommendations, and in SECY-
07-0082, ``Rulemaking to Make Risk-Informed Changes to Loss-of-Coolant 
Accident Technical Requirements''; 10 CFR 50.46a ``Alternative 
Acceptance Criteria for Emergency Core Cooling Systems for Light-Water 
Nuclear Power Reactors,'' (May 16, 2007) sought additional guidance 
from the Commission on the priority of the rule and on the issues 
raised by the ACRS. In its August 10, 2007, SRM (ML072220595) in 
response to SECY-07-0082, the Commission approved NRC staff 
recommendations for a revised priority and approach for addressing the 
ACRS concerns and completing the final rule. On April 1, 2008, the NRC 
staff provided the Commission with its planned schedule (ML080370355) 
for completing the rule.
    As the NRC staff proceeded to modify the rule in response to the 
ACRS recommendations and the Commission's direction, numerous 
substantive changes were made to the requirements in the draft final 
rule. After consideration of the extent of these changes, the NRC has 
decided to provide another opportunity for public comment focusing on 
the revised proposed rule, in order to provide public stakeholders with 
another opportunity to review and comment on the new language. Because 
of the interrelated nature of the regulatory requirements, the NRC is 
republishing the entire 10 CFR 50.46a proposed rule to allow public 
comments on the changed requirements and on other closely-related 
regulatory provisions.

III. Description of November 2005 Proposed Rule

    The proposed rule published on November 7, 2005, (70 FR 67598) 
would divide the current spectrum of LOCA break sizes into two regions. 
The division between the two regions is delineated by a ``transition 
break size'' (TBS). \2\ The first region includes small size breaks up 
to and including the TBS. The second region includes breaks larger than 
the TBS up to and including the DEGB of the largest RCS pipe. Break 
area associated with the TBS is not based upon a double-ended offset 
break. Rather, it is based upon the inside area of a single-sided 
circular pipe break.
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    \2\ Different TBSs for pressurized water reactors and boiling 
water reactors would be established due to the differences in design 
and operation between those two types of reactors.
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    Pipe breaks in the smaller break size region are considered more 
likely than pipe breaks in the larger break size region. Consequently, 
each break size region is subject to different ECCS requirements, 
commensurate with likelihood of the break. LOCAs in the smaller break 
size region must be analyzed by the methods, assumptions, and criteria 
currently used for LOCA analysis; accidents in the larger break size 
region will be analyzed by less conservative assumptions based on their 
lower likelihood. Although LOCAs for break sizes larger than the 
transition break would become ``beyond design-basis accidents,'' the 
proposed rule would require licensees to maintain the ability to 
mitigate all LOCAs up to and including the DEGB of the largest RCS pipe 
during all operating configurations.
    Licensees who perform LOCA analyses using the risk-informed 
alternative requirements could find that their plant designs are no 
longer limited by certain parameters associated with previous DEGB 
analyses. Reducing the DEGB limitations could enable some licensees to 
propose a wide scope of design or operational changes up to the point 
of being limited by some other parameter associated with any of the 
required accident analyses. Potential design changes include 
modification of containment spray designs, modifying core peaking 
factors, modifying setpoints on accumulators or removing some from 
service, eliminating fast starting of one or more emergency diesel 
generators, increasing power, etc. Some of these design and operational 
changes could increase plant safety because a licensee could modify its 
systems to better mitigate the more likely small-break LOCAs. Other 
design changes, such as increasing power, could cause increases in 
plant risk. Accordingly, the risk-informed Sec.  50.46a option would 
establish risk acceptance criteria to ensure the risk acceptability of 
all subsequent facility changes. The proposed rule required that all 
future facility changes \3\ made by licensees after adopting Sec.  
50.46a be evaluated by a risk-informed integrated safety performance 
(RISP) assessment process that has been reviewed and approved by the 
NRC via the routine process for license amendments.\4\ The RISP 
assessment process would ensure that the cumulative effect of all plant 
changes involved acceptable changes in risk and was consistent with 
other criteria from RG 1.174 to ensure adequate defense-in-depth, 
safety margins and performance measurement. Licensees with an approved 
RISP assessment process could make certain facility changes without NRC 
review if they met Sec.  50.59 \5\ and Sec.  50.46a requirements, 
including the criterion that risk increases cannot exceed a ``minimal'' 
level. Licensees could make other facility changes after NRC approval 
if they met the Sec.  50.90 requirements for license amendments and the 
criteria in Sec.  50.46a, including

[[Page 40009]]

the criterion that total cumulative risk increase cannot exceed a 
``small'' threshold. Potential impacts of the plant changes on facility 
security would be evaluated as part of the license amendment review 
process.
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    \3\ The scope of changes subject to the change criteria in Sec.  
50.46a(f) of the proposed rule would be greater than the changes 
currently subject to Sec.  50.59, which applies only to changes to 
``the facility as described in the FSAR.'' The change criteria in 
the proposed rule would apply to all facility and procedure changes, 
regardless of whether they are described in the Final Safety 
Analysis Report (FSAR).
    \4\ Requirements for license amendments are specified in 
Sec. Sec.  50.90, 50.91 and 50.92. They include public notice of all 
amendment requests in the Federal Register and an opportunity for 
affected persons to request a hearing. In implementing license 
amendments, the NRC typically prepares an appropriate environmental 
analysis and a detailed NRC technical evaluation to ensure that the 
facility will continue to provide adequate protection of public 
health and safety and common defense and security after the 
amendment is implemented.
    \5\ Requirements in Sec.  50.59 establish a screening process 
that licensees may use to determine whether facility changes require 
prior review and approval by the NRC. Licensees may make changes 
meeting the Sec.  50.59 requirements without requesting NRC approval 
of a license amendment under Sec.  50.90.
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    The NRC would periodically evaluate LOCA frequency information. 
Should estimated LOCA frequencies significantly increase such that the 
risk associated with pipe breaks larger than the TBS is unacceptable, 
the NRC would undertake rulemaking (or issue orders, if appropriate) to 
change the TBS. In such a case, the backfit rule (10 CFR 50.109) would 
not apply. If previous plant changes were invalidated because of a 
change to the TBS, licensees would have to modify or restore components 
or systems as necessary so that the facility would continue to comply 
with Sec.  50.46a acceptance criteria. The backfit rule (10 CFR 50.109) 
would also not apply to these licensee actions.

IV. Discussion of Public Comments

    The NRC received comments on the proposed rule from six nuclear 
power plant licensees, four nuclear industry organizations, two reactor 
vendors, and an NRC employee. The comments provided by NEI were 
specifically endorsed by the WOG, the Boiling Water Reactors Owners 
Group (BWROG), and three nuclear power plant licensees. The NRC 
considered all comments in formulating the revised proposed rule 
language. The NRC also received comments from a nuclear engineering 
professor on the expert elicitation process for determining the 
relationship between pipe break frequency and pipe size that was used 
as the baseline for selecting the transition break size. Although these 
comments were submitted for NUREG-1829 (Draft Report), ``Estimating 
Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation 
Process'' (ML051520574), they were also considered in the development 
of the Sec.  50.46a final rule.
    Comments and other publicly available documents related to this 
rulemaking may be viewed electronically on the public computers located 
at the NRC's Public Document Room (PDR), Public File Area O-F21, One 
White Flint North, 11555 Rockville Pike, Rockville, Maryland. Selected 
documents, including comments, may be viewed and downloaded 
electronically via the Federal e Rulemaking Portal. Go to http://www.regulations.gov and search for documents filed under Docket ID NRC-
2004-0006.
    Comments addressed six different general topics: selection of the 
TBS, the effect of seismic considerations on the TBS, thermal-hydraulic 
ECCS analyses, probabilistic risk analysis, applicability of the 
backfit rule, and comments on questions posed by the Commission. The 
comments are discussed below by topic area.

A. Comments on Selection of the TBS

    Comment. NEI stated that the TBS proposed for boiling water 
reactors (BWRs) is overly conservative and may unnecessarily limit or 
preclude benefits for BWRs. They suggested that the specified piping 
for the BWR TBS should be equivalent to the 16-inch schedule 80 piping 
in the shutdown cooling suction line inside containment. The BWROG 
supported a reduced TBS for BWRs consistent with the 95th percentile 
TBS noted from the expert elicitation (i.e., without additional 
conservatisms).
    NRC response. The proposed TBS for BWRs is currently based on the 
cross-sectional area of the larger of either the shutdown cooling 
residual heat removal (RHR) or feedwater pipes which are connected to 
the RCS inside containment. These pipe sizes are generally in the 18'' 
to 24'' range, and are similar in size to the 95th percentile estimates 
from the expert elicitation process results for BWRs at a 
10<SUP>-</SUP>\5\ per year frequency. (It should be noted that the NRC 
also considered uncertainties in the estimates based on analysis 
sensitivities of the expert elicitation results, such as the method of 
aggregating the individual frequency estimates. The 95th percentile 
estimate of BWR break size diameter for the geometric mean aggregation 
method is approximately 13 inches, and the corresponding break size for 
the arithmetic mean aggregation method is approximately 20 inches.) The 
actual plant pipe sizes were used as a logical selection criterion; 
because for a given size break, it is more likely that a break will be 
circumferentially oriented (i.e., a complete severance of the pipe). 
The NRC selected the TBS by considering the actual size of the attached 
piping, rather than by selecting a single break size value which would 
conservatively bound all plant configurations. For BWRs, the pipes 
connecting to the RCS, other than the largest reactor recirculation 
piping or main steam line piping, are the feedwater and RHR piping. 
Also, these pipes are large enough so that a single-ended break of one 
of them will generally bound the total cross-sectional discharge area 
for a double-sided break in smaller size feedwater or recirculation 
pipes. For these reasons, the NRC continues to believe that the TBS for 
BWRs should be based on the cross-sectional area of the larger of 
either the feedwater or RHR lines inside containment. No changes to the 
BWR TBS have been made in the revised proposed rule.
    Comment. The Nuclear Energy Institute, the Westinghouse Owners 
Group (WOG) and a reactor licensee stated that for pressurized-water 
reactors (PWRs) with large piping connected to both the hot and cold 
legs, the TBS for the hot leg should be based on the largest connecting 
hot leg pipe, and the TBS for the cold leg should be based on the 
largest connecting cold leg pipe. These are logical break sizes and 
avoid the arbitrary nature of the size of the connecting pipe on the 
hot leg being also applied to breaks on the cold leg. If no attached 
piping is connected to the cold leg, the cold leg TBS should be the 
same as the hot leg TBS. The WOG stated that the NRC and the industry 
should take the opportunity of this rule change to determine the 
appropriate transition break size and not settle for a rule that is 
needlessly conservative. Because the rulemaking cannot easily be 
changed in the future as new information becomes available, the TBS 
should be based on sound technical facts and expert opinions with some 
margin for uncertainties and unknowns that could show up in the future 
and erode margins. It is not appropriate to set the TBS on the basis of 
where the most benefit would be realized because this may change 
tomorrow and there will be no easy recourse. The WOG also said that the 
Commissioners have recommended a design basis LOCA cut-off frequency of 
10<SUP>-</SUP>\5\ per reactor year, which corresponds to a break size 
of about a three or four-inch diameter effective break (for PWRs). The 
WOG believes that selecting a TBS equal to the largest attached piping 
(8- to 12-inch diameter break) is very conservative. However, the WOG 
has conducted thermal-hydraulic and risk analyses that show that there 
are substantial potential benefits for PWR plants even with this larger 
TBS. The WOG agreed that setting the transition break size at the sizes 
of the piping attached to the RCS loop is reasonable because it will 
provide significant benefit while providing substantial margin to 
account for uncertainties or any new information that may become 
available on break size vs. frequency. The requirement that plants must 
still be able to mitigate breaks larger than the TBS provides even more 
margin.
    NRC response. In developing the basis for the PWR TBS, the NRC not 
only used the mean break frequency estimates from the expert 
elicitation but also included additional allowances for

[[Page 40010]]

various uncertainties. To address uncertainties in the elicitation 
process, the 95th percentile estimates of break size diameter were 
used. Further, the methods of aggregating the individual frequency 
estimates were evaluated for sensitivities. For PWRs, the break size at 
a 10<SUP>-</SUP>\5\ per year frequency using the geometric mean method 
is approximately 6 inches, and the corresponding break size for the 
arithmetic mean method is approximately 10 inches. This is similar in 
size to the cross-sectional area of the largest pipe attached to the 
main reactor coolant loop on which the TBS is ultimately based. The 
largest attached piping in PWRs is generally in the 12- to 14-inch 
nominal pipe size range (with inside diameters corresponding to 10.1 to 
11.2 inches), and typically corresponds to the surge line which is 
attached to the hot leg. However, on some Combustion Engineering and 
Babcock and Wilcox plants, the largest attached pipes may be the RHR, 
safety injection, or core flood lines, which may not be similarly 
attached to the hot leg. However, as stated in the statement of 
considerations for the initial proposed rule (see 70 FR at 67603-
67606), the NRC selected only one size which would uniformly apply for 
all locations in the RCS piping, because the expert elicitation did not 
provide sufficient detail to distinguish the hot leg from the cold leg 
break frequencies. The commenters did not provide additional 
information or technical data that justifies different break 
frequencies or use of a smaller TBS on the cold leg piping. Thus, no 
changes to the PWR TBS were made in the revised proposed rule.

B. Comments on Seismic Considerations Related to the TBS

    The TBS specified by the NRC in the November 7, 2005, proposed rule 
did not include an adjustment to address the effects of seismically-
induced LOCAs. (See 70 FR at 67604.) On December 20, 2005, the NRC 
released a report discussing seismic considerations for the transition 
break size (``Seismic Considerations for the Transition Break Size'', 
December 2006; ML053470439). The NRC requested specific public comments 
on the effects of pipe degradation on seismically-induced LOCA 
frequencies and the potential for affecting the selection of the TBS. 
These public comments were considered in the final, published report 
(NUREG-1903, ``Seismic Considerations for the Transition Break Size'', 
February 2008; ML080880140).
    Comment. NEI, WOG, BWROG, and a reactor licensee all commented that 
the proposed TBS need not be further adjusted due to seismic 
considerations. NEI indicated that the NRC's December 20, 2005, report 
demonstrates that the seismically-induced LOCA frequency contribution 
is less than the 10<SUP>-5</SUP> per reactor year guideline used by the 
NRC in determining the TBS. NEI further commented that median seismic 
capacities for both the primary piping system and primary system 
components are higher than most other safety related power plant 
components within the nuclear power plant. Because of these relative 
capacities, NEI said the seismic risk from very large, low probability 
earthquakes would be controlled by consequential safety component 
failure. In addition, NEI stated that the creation of the TBS by itself 
does not produce a physical change in the plant that would result in an 
appreciable change in seismic risk. The WOG, the BWROG, and a reactor 
licensee endorsed the NEI comments. WOG included an additional comment 
which stated that the NRC's December report indicated that seismic 
loading will only have a small (10 per cent) effect on the LOCA 
frequencies estimated by the NRC expert panel (NUREG-1829, Draft 
report, June 2005) and that effect is well within the uncertainty 
bounds of the frequency estimate of the panel. Furthermore the NRC has 
already included a very substantial margin above the break size that 
would correspond to a LOCA frequency of 10<SUP>-5</SUP> per reactor 
year. Therefore, seismic effects should not change the transition break 
size.
    NRC Response. The NRC agrees with the commenters' conclusion that 
the TBS defined in the proposed rule need not be adjusted further to 
account for the effects of seismically induced LOCAs in piping greater 
than the TBS. In reaching its conclusion the NRC considered the 
comments received as well as historical information related to piping 
degradation and the potential for the presence of cracks sufficiently 
large that pipe failure would be expected under loads associated with 
rare (10<SUP>-5</SUP> per year) earthquakes.
    The NRC report NUREG-1903, ``Seismic Considerations for the 
Transition Break Size'' (February 2008; ML080880140) considered the 
potential contribution from two mechanisms: direct piping failures and 
indirect failures. Direct failures are those pipe ruptures that result 
when the combined earthquake loadings and normal stresses exceed the 
strength of the pipe. The report concluded that direct failures from 
earthquakes with return frequencies of 10<SUP>-5</SUP> per year and 
10<SUP>-6</SUP> per year would not be expected unless cracks on the 
order of 30 percent through-wall and approximately 145 degrees around 
the piping circumference were present at the time of the earthquake. 
The NRC reviewed its experience with flaws in reactor coolant system 
piping to assess whether cracks of this magnitude have ever been found 
in RCS main loop piping, or if other information suggests that cracks 
of this magnitude are likely. The NRC considered both fabrication 
induced flaws and service induced flaws. No large fabrication flaws 
have ever been reported. If large fabrication flaws were present and 
were not detected by the initial fabrication inspections and subsequent 
in-service inspections, it would be expected that some would have grown 
through-wall over time as a result of fatigue or other mechanisms and 
would have been discovered through leakage. This has not been observed 
even though most plants have been in operation for more than 20 years.
    With respect to service induced flaws, the NRC also considered the 
potential for known degradation mechanisms to induce cracks of the 
critical size. For BWRs, intergranular stress corrosion cracking 
(IGSCC) is the only mechanism that has been shown to produce large 
cracks. However, regulatory and industry programs have been in place 
for many years to specifically address this mechanism and as a result, 
IGSCC is being effectively managed. In PWRs, a number of partly 
through-wall flaws and a small number of through-wall flaws have been 
discovered and have been attributed to primary water stress corrosion 
cracking (PWSCC). To date, all flaws discovered were considerably 
smaller than flaws that would lead to failure under 10<SUP>-5</SUP> and 
10<SUP>-6</SUP> per year earthquake loadings. PWR plant owners have 
established programs to address PWSCC in susceptible reactor coolant 
system piping welds. They are inspecting these welds more frequently 
and, in most cases, are applying mitigation techniques to manage PWSCC. 
The NRC is working with the American Society of Mechanical Engineers 
(ASME) to establish a regulatory framework for improved inspection and 
mitigation of PWSCC in these welds. The NRC expects that these measures 
will ensure that PWSCC will be effectively managed. As a result of the 
above considerations, the NRC considers the likelihood of flaws large 
enough to fail under 10<SUP>-5</SUP> and 10<SUP>-6</SUP> per year 
earthquake loadings to be sufficiently low that the TBS need not be 
modified to address seismically induced direct failures.
    Indirect failures are primary system pipe ruptures that are a 
consequence of

[[Page 40011]]

failures in non-primary system components or structural support 
failures (such as a steam generator support). Structural support 
failures could then cause displacements in components that stress the 
piping and result in pipe failure. The NRC performed studies on two 
plants to estimate the conditional pipe failure probability due to 
structural support failure given a low return frequency earthquake 
(10<SUP>-5</SUP> to 10<SUP>-6</SUP> per year). The results indicated 
that the conditional failure probability was on the order of 0.1. These 
studies used seismic hazard curves from NUREG-1488, ``Revised Livermore 
Seismic Hazard Estimates for Sixty-Nine Nuclear Power Plant Sites East 
of the Rocky Mountains,'' (April 1994; ML052640591). More recent 
indirect failure studies were completed by the Electric Power Research 
Institute (EPRI) on three plants using updated seismic hazard 
estimates. The updated seismic hazard increases the peak ground 
acceleration at some sites. The highest pipe failure probability 
calculated for the three plants in the industry analyses was 6 x 
10<SUP>-6</SUP> per year. Although the EPRI failure probability was 
higher than either of the two cases calculated by the NRC, the result 
is still lower than the TBS selection guideline of 10<SUP>-5</SUP> per 
reactor year. The NRC noted in its report that indirect failure 
analyses are highly plant-specific. Therefore it is possible that 
example plants assessed in the NRC and EPRI analyses are not limiting 
for all plants.
    The NRC has considered the importance of indirect failures on the 
selection of the TBS. For the cases considered in both the EPRI and NRC 
studies, the likelihood of indirectly induced piping failures resulting 
from major component support failures is less than 10<SUP>-5</SUP> per 
reactor year, the frequency criterion used to select the TBS. Also, as 
noted in the public comments, the median seismic capacities for both 
the primary piping system and primary system components are typically 
higher than other safety related components within the nuclear power 
plant. Because of these relative capacities, it is expected that a 
seismic event of sufficient magnitude to cause consequential failure 
within the primary system would also induce failure of components in 
multiple trains of mitigation systems, or even induce multiple RCS pipe 
breaks. Consequently, the risk contribution from seismically induced 
indirect failures is expected to depend more heavily on the relative 
fragilities of plant components and systems than the size of the TBS. 
Therefore, adjustment to the TBS for seismically induced indirect LOCAs 
is also not warranted.
    Comment. In the proposed rule, the NRC stated that the final rule 
might include requirements for licensees to perform plant-specific 
assessments of seismically-induced pipe breaks and, if necessary, 
implement augmented in-service inspection plans before implementing the 
alternative ECCS requirements. NEI, WOG, BWROG, and a reactor licensee 
all commented that plant specific assessments should not be required to 
demonstrate that the seismically induced pipe breaks do not 
significantly affect the likelihood of pipe breaks larger than the TBS. 
NEI indicated that the NRC's December 20, 2005 report, ``Seismic 
Considerations for the Transition Break Size'' demonstrates that the 
seismically induced LOCA frequency contribution is less than the 
10<SUP>-5</SUP> per reactor year guideline limit used by the NRC in 
determining the TBS. NEI further commented that indirect LOCA seismic 
studies had been performed by EPRI for a limited number of plants using 
more recent seismic hazard estimates than those used in the NRC's 
December study. The EPRI study estimated that the indirect LOCA 
probability was less than 10<SUP>-5</SUP> per year for the plants 
examined. The EPRI study found that although the latest seismic hazard 
has increased for some parts of the central and eastern United States, 
there are several mitigating phenomena that have been established 
within the new plant seismic program which tend to counter much of that 
increase. NEI also stated that for a risk informed application, the 
change in risk should be the primary metric for decision making. The 
change in risk relative to seismic events is estimated to be negligible 
based upon the fact that the TBS threshold does not directly impact 
either the seismic hazard or the plant seismic fragilities. The WOG, 
the BWROG, and a licensee all endorsed the NEI comments. WOG included 
an additional comment which stated that the NRC's December report 
indicated that seismic loading will only have a small (~10 percent) 
effect on the LOCA frequencies estimated by the NRC expert panel 
(NUREG-1829 Draft Report, June 2005) and that effect is well within the 
uncertainty bounds of the frequency estimate of the panel. A reactor 
licensee had an additional comment that plant specific assessments to 
determine the frequency of seismically induced pipe breaks would be 
very difficult to complete. The licensee said that because pipe 
inspection and repair are such an integral part of plant operations, 
after a plant seismic assessment was completed, its conclusions would 
then be prejudiced by implementation of piping inspection and repair 
programs. The commenter did not explain in detail how the results would 
be prejudiced. The commenter also suggested that more technically valid 
piping failure probabilities might be obtainable through an extensive 
research program, but noted it is questionable whether this would 
provide additional risk insights.
    NRC response. The NRC disagrees with the commenters that plant 
specific assessments of seismically induced pipe breaks are not 
necessary before implementing the alternative ECCS requirements. As 
discussed in the previous comment, although seismic considerations do 
not significantly affect TBS selection, the generic nature of the 
seismic risk studies requires an applicant to demonstrate that these 
studies are applicable to its plant and site.
    The NUREG-1903 study did generically conclude (based on operating 
experience, probabilistic risk assessment insights, experimental 
testing, and analysis) that the likelihood of seismic-induced unflawed 
piping failure was much less than 10<SUP>-5</SUP> per year. However, a 
general conclusion about the likelihood of seismic-induced flawed 
piping failure could not be reached for all plants. Twenty-six plant-
specific calculations were conducted in NUREG-1903 using available 
seismic hazard assessments for plants east of the Rocky Mountains 
(i.e., from NUREG-1488; April, 1994) and piping stress and material 
information obtained from historical leak-before-break applications. 
These calculations indicated that extremely large circumferential flaws 
(i.e., greater than 30 percent of the piping wall thickness for a flaw 
approximately 145 degrees around the piping circumference) would be 
required before failure would occur due to earthquakes with a return 
frequency of 10<SUP>-5</SUP> or 10<SUP>-6</SUP> per year. However, the 
plant-specific conditions used in the calculations were not chosen to 
bound conditions at all nuclear power plants. Additionally, some plants 
may have updated seismic hazard, piping stress, material property, or 
other information used in the flawed piping evaluation. Thus, the 
NUREG-1903 results may not be applicable to every plant.
    The ACRS, in its letter dated November 16, 2006 (ML063190465), also 
noted that seismic hazards are very plant specific. The ACRS further 
recommended that licensees who adopt Sec.  50.46a should demonstrate 
that the results developed by the NRC bound the

[[Page 40012]]

likelihood of seismically induced failure at their plants. The 
Committee further stated that licensees may have to perform additional 
calculations to demonstrate a comparable robustness of flawed piping. 
The ACRS recommendations are consistent with the limitations of the 
NUREG-1903 study as noted above.
    It would also be inconsistent with the Commission's intent to allow 
the relaxation of ECCS requirements at a plant with a seismically 
induced large break LOCA frequency greater than the 10<SUP>-5</SUP> per 
reactor year criteria used for selecting the TBS in the proposed rule. 
Because seismic analyses and, in particular, indirect failure estimates 
are highly plant and site specific (as noted in NUREG-1903 and in ACRS 
comments), the NRC believes that it is necessary for a licensee to 
demonstrate that its seismic LOCA frequency is sufficiently low before 
implementation of the alternative ECCS requirements. Depending upon the 
results of the plant specific assessment, it may be necessary to 
implement augmented in-service inspection plans. As discussed below in 
Section V.C. of this document, the NRC is currently preparing guidance 
for conducting these plant-specific assessments (``Plant-Specific 
Applicability of 10 CFR 50.46 Technical Basis,'' February 2009; 
ML090350757).

C. Comments on Thermal-Hydraulic Analysis

    Comment. Both NEI and WOG recommended that the proposed new 
reporting requirement in Sec.  50.46a(g)(1)(i) of a 0.4 percent change 
in oxidation as the threshold for reporting a change, or the sum of 
changes, in calculated clad oxidation be changed from 0.4 percent to 
2.0 percent. WOG noted that the rationale for selecting 0.4 percent is 
that it is the same, on a percentage basis, as the existing peak 
cladding temperature (PCT) change reporting requirement. WOG also 
stated that this rationale is only true if one considers the range of 
interest of PCT as 0 to 2200 degrees Fahrenheit ([deg]F) [(50 [deg]F/
2200 [deg]F) x (17 percent) = 0.4 percent]. If instead, one considers 
the range of interest of PCT as 1700-2200 [deg]F or 1800-2200 [deg]F, 
from the perspective of transient oxide build-up, this same rationale 
gives a significance threshold of 1.7 or 2.1 percent. On this basis, 
WOG recommended that the significance threshold for changes in 
oxidation be revised to 2.0 percent.
    WOG also noted that changes in oxidation are much more difficult to 
estimate than changes in peak cladding temperature because oxidation is 
an integrated parameter based on the temperature transient versus time, 
whereas PCT is a point value. If the significance threshold for 
oxidation is not adjusted as recommended above, it is anticipated that 
the new oxidation reporting requirement will require more frequent re-
analyses than the current regulations require, with no commensurate 
benefit to the public health and safety.
    NRC response. The basis for the 0.4 per year oxidation change is 
that the ratio of the reporting threshold value to the change in 
oxidation from a ``normal'' operating level of 4 percent (based on a 
twice-burned oxidation thickness of 65 [mu] for Zircalloy-4) to a 
maximum level of 17 percent should be the same as the ratio of the 
reporting threshold value to the change from the normal operating 
cladding temperature of 600 [deg]F to the allowed PCT of 2200 [deg]F. 
On that basis the oxidation change of 0.4 percent was chosen. The trend 
toward thinner cladding material raises the initial oxidation 
percentage even closer to the maximum local oxidation limit and reduces 
the margin for change in predicted oxidation.
    Additionally, the NRC agrees with the WOG comment that calculating 
oxidation is more time-consuming than calculating PCT. However, the NRC 
believes WOG is incorrect in stating that not reducing the significance 
threshold for reporting changes in calculated oxidation will cause the 
need for performing additional oxidation calculations. The significance 
threshold for reporting to the NRC only affects the frequency of 
reporting and has no effect on the need to do reanalysis. Reanalysis is 
necessary when licensees discover errors or make changes to analytical 
codes.
    The Commission has directed the NRC staff to revise the ECCS 
acceptance criteria in Sec.  50.46(b) to account for new experimental 
data on cladding ductility and to allow for the use of advanced 
cladding alloys. The NRC will soon issue an Advance Notice of Proposed 
Rulemaking (ANPR) seeking public comments on a planned regulatory 
approach. The NRC expects that this rulemaking (Docket ID NRC-2008-
0332) will establish new cladding embrittlement acceptance criteria in 
Sec.  50.46(b) for design basis LOCAs. As these new acceptance criteria 
are being established, the NRC will also make conforming changes to 
Sec.  50.46a as necessary for both below and above TBS breaks. As a 
consequence, the NRC now believes that the need for a reporting 
requirement in Sec.  50.46a associated with errors or changes in ECCS 
analysis methodology would be more appropriately addressed in the 
ongoing Sec.  50.46(b) proceeding. Accordingly, the changes to the 
oxidation reporting requirements in the initial proposed rule have been 
removed from the revised proposed rule.
    Comment. Framatome commented that the analysis or case requirements 
in Sec.  50.46a(e)(2) for beyond the transition break size evaluations 
are excessive. The desire for this portion of the regulation is to 
establish in a reasonable way that the plant remains able to mitigate a 
large break LOCA. It is unnecessary and inconsistent to elevate the 
consideration of break size effects beyond that of other portions or 
aspects of the evaluation that are to be treated as reasonable values. 
Under the proposed rule language, a full Sec.  50.46 evaluation will be 
required for breaks of area less than the TBS. The results for these 
analyses can be extended to the smaller break sizes in the greater than 
TBS spectrum with assurance. Combining a reasonable selection of 
discharge coefficient (0.6) with the use of the 1994 ANS decay heat 
standard would roughly equate a 14-inch schedule 160 pipe area (0.7 ft 
\2\), treated as below the TBS, with a 1.4 ft \2\ break, treated as a 
beyond TBS break. Similarly, at the upper end of the break spectrum, 
what used to be considered as an 8 to 9 ft \2\ break of the cold leg 
will be the equivalent of a historical 5 ft \2\ break. The requirement 
to perform sensitivity studies to identify a worst case break between 
these two limits seems unwarranted. It would be reasonable to just 
perform the full double area break or at most that break and one 
intermediate break. The only sensitivity required should be relative to 
break location. Historically, break location can have a substantial 
influence on the calculated results. This should be resolved prior to 
the greater than TBS calculation either by sensitivity studies or by 
reference to appropriate historical analyses. The concern can be 
allayed by either altering the rule so that the identification of the 
most severe break size is not required or by inserting the concept of 
reasonable confidence that breaks within the beyond TBS spectrum will 
not pose consequences substantially more severe than those of the 
calculations performed.
    The WOG stated that for NRC-approved best-estimate or Appendix K 
evaluation models, the requirement for analyzing a spectrum of break 
sizes is unwarranted. The BWROG said that the requirement to re-
validate over 30 years of experience with performing large break LOCA 
analysis to confirm ``for a number of postulated LOCAs of different 
sizes and locations * * * that

[[Page 40013]]

the most severe postulated LOCAs * * * are analyzed'' is unnecessarily 
burdensome and appears to serve no specific technical need. Current 
best-estimate large break LOCA models, which are benchmarked to testing 
data, have yielded no insights that would invalidate the previous 
analytical experience and knowledge. WOG concluded that this provision 
in the rule language should be removed.
    NRC response. The NRC disagrees with the commenters on the need for 
analyzing a spectrum of break sizes. The proposed rule language was 
selected because there are two peak cladding temperatures, one that 
occurs below the TBS and one that occurs above the TBS. The peak above 
the TBS may not occur for the DEGB, but rather, for a break area in the 
range of 0.6 to 0.8 times the DEGB area. Because there can be a fairly 
large temperature difference between that break and the DEGB, use of 
the DEGB could be non-conservative. The NRC also believes that the 
language of the rule provides considerable flexibility in 
implementation (relative to the stated comments) because the 
requirement is to analyze a ``number of postulated LOCAs * * * 
sufficient to provide assurance that the most severe LOCAs * * * are 
analyzed''. The use of historical analyses is not precluded. No changes 
were made in the revised proposed rule.
    Comment. NEI commented that in Sec.  50.46a(e)(2) on ECCS analysis 
methods, one requirement is that ``comparisons to applicable 
experimental data must be made.'' NEI stated that other approaches such 
as comparison of results to accepted analysis techniques or to textbook 
approaches are also appropriate and suggested that the requirement be 
reworded to state that ``sufficient justification'' must be provided.
    NRC response. The NRC disagrees with this commenter. Computer code-
to-code comparisons are not adequate because all codes have uncertainty 
in their results. Only code-to-data comparisons can be used to 
accurately assess code uncertainties. Similarly, computer code results 
cannot be validated by comparison to ``textbook approaches'' because no 
simple textbook approaches exist for modeling the highly complex 
thermal-hydraulic phenomena associated with pipe break analyses. No 
changes were made in the revised proposed rule.
    Comment. WOG submitted four options for how to perform ECCS 
analysis in the beyond-TBS region to assist the NRC staff in developing 
the regulatory guide for implementing the Sec.  50.46a rule.
    NRC Response. The NRC will evaluate the WOG ECCS analysis options 
and will provide additional implementation guidance in the associated 
regulatory guide.
    Comment. The BWROG stated that it supports applying the 
requirements of Sec.  50.46a(b)(1) to reactors with MOX [mixed oxide] 
fuel.
    NRC response. The proposed Sec.  50.46a is intended to be an 
alternative to the current ECCS requirements in Sec.  50.46. Because 
Sec.  50.46 does not address the use of mixed oxide fuel, the NRC 
believes that the commenter's proposal is beyond the scope of this 
rulemaking. The NRC did not make changes in the revised proposed rule 
to address MOX fuel.
    Comment. Proposed Sec.  50.46a(e)(2): The following sentence should 
be moved from its current location to just in front of the sentence 
beginning, ``These calculations * * *'': ``The evaluation must be 
performed for a number of postulated LOCAs of different sizes and 
locations sufficient to provide assurance that the most severe 
postulated LOCAs larger than the TBS up to the double-ended rupture of 
the largest pipe in the reactor coolant system are analyzed.'' This 
relocated sentence should begin a new paragraph. These changes will 
properly group the more detailed analysis requirements.
    NRC response. The NRC agrees that movement of the noted sentence 
improves the rule presentation. In the revised proposed rule, this 
sentence has been relocated as the commenter suggested, but the 
structure of Sec.  50.46a(e)(2) was not modified.
    Comment. In proposed Sec.  50.46a(e)(2), the NRC should clarify the 
requirements for licensee documentation to be maintained onsite versus 
generic documentation in or supporting a licensing topical report.
    NRC response. In the revised proposed rule, the NRC modified Sec.  
50.46a(e) to require that analysis methods for all LOCAs ``must be 
approved for use by the NRC. Appendix K, Part II, to 10 CFR Part 50, 
sets forth the documentation requirements for evaluation models.'' 
Thus, the documentation requirements for analysis methods used for 
breaks larger than the TBS are the same as for analysis methods used 
for breaks smaller than the TBS. The purpose of this change is to 
increase confidence in the ability to mitigate breaks greater than the 
TBS, as recommended by the Advisory Committee on Reactor Safeguards.
    Comment. In proposed Sec.  50.46a(e)(2), the NRC states that these 
calculations [for breaks larger than the TBS] may take credit for the 
availability of offsite power and do not require the assumption of a 
single failure. It should also be noted that availability of equipment 
is not limited to safety-related equipment.
    NRC response. The NRC agrees that the suggested language is more 
descriptive and has incorporated the change into that last sentence of 
Sec.  50.46a(e)(2).
    Comment. For PWR LOCAs below and above the TBS, the mitigating 
systems and equipment are the same for the full spectrum of LOCAs. 
Although non-safety LOCA mitigation systems/components may be 
applicable in the context of BWR LOCA analysis, this is not the case 
for PWRs. If this element of the proposed regulation (allowing the use 
of non-safety grade systems) is intended to address a situation that is 
only applicable to BWRs, then it should not be required for PWRs.
    NRC response. The element of the proposed regulation--allowing the 
use of non-safety grade systems--noted by the commenter is not intended 
to address a situation that is only applicable to BWRs. Although PWR 
plants may not currently have non-safety systems that could be credited 
for LOCA mitigation (for breaks larger than the TBS), modifications 
could be made in the future that facilitate use of non-safety systems. 
The revised proposed rule would relax existing Sec.  50.46 requirements 
to allow ECCS analyses of breaks larger than the TBS to take credit for 
both safety-grade and non-safety-grade equipment if such equipment 
exists, is maintained available and reliable, and is capable of being 
powered by an on-site source of electrical power.
    Comment. The WOG commented that the rule should not contain a 
requirement for licensees to submit beyond TBS thermal-hydraulic 
analyses to the NRC for approval. One reactor licensee commented that 
the proposed rule states that licensees will not be required to submit 
their beyond-TBS analysis method or application to the NRC for review 
and approval; instead, the NRC intends to maintain regulatory oversight 
of these analyses by inspection. That licensee said that although not 
requiring NRC review and approval has the appearance of a benefit to 
the licensees, it actually introduces a risk of a regulatory crisis 
should an inspection identify a deficiency in the beyond-TBS analysis 
method following implementation. Such an identified deficiency could 
result in a consequence such as the regulator imposing restrictions on 
reactor operation. This risk is greater than for

[[Page 40014]]

the current situation where LOCA evaluation models and applications are 
pre-approved by the NRC. It would be preferable that NRC review and 
approval of Sec.  50.46a applications be obtained prior to 
implementation to avoid such a regulatory crisis. This commenter 
proposed that the NRC agree to perform a pre-approval of a licensee's 
beyond-TBS analysis method and application if requested by a licensee.
    NRC response. The NRC has changed the proposed rule to require NRC 
review and approval of analysis methods used to evaluate plant response 
to LOCAs larger than the transition break size. The purpose of this 
change is to increase confidence in the ability to mitigate breaks 
greater than the TBS, as recommended by the ACRS.
    Comment. NEI, a reactor vendor, and a reactor licensee requested 
that M5 cladding (M5) be specified as an approved fuel cladding 
material in existing Sec.  50.46(a) and in proposed Sec.  50.46a(b)(1) 
to avoid the need for requesting an exemption to allow its use. The 
reactor vendor stated that because M5 is currently being used in 11 
nuclear power reactors of varying designs across the United States, it 
is obvious that M5 is an acceptable and desirable cladding material. 
The BWROG stated that Sec.  50.46a should be made available to reactors 
with alternate cladding materials.
    NRC response. As previously discussed, the Commission directed the 
NRC staff to initiate a separate rulemaking effort to amend Sec.  
50.46(b) to address the use of advanced cladding alloys. The NRC is 
considering cladding specific issues in that proceeding and will also 
incorporate appropriate conforming changes to Sec.  50.46a. The NRC is 
working to revise the ECCS acceptance criteria in Sec.  50.46(b) to 
account for new experimental data on cladding ductility and to 
facilitate the licensing review of advanced cladding alloys such as M5. 
The NRC plans to issue an ANPR during the summer of 2009 to solicit 
public comments on a planned regulatory approach. In the interim, the 
NRC will continue to evaluate the use of cladding materials other than 
Zircalloy or ZIRLO on a case-by-case basis.

D. Comments Related to Probabilistic Risk Assessment

1. Summary
    The initial proposed rule required that all future facility changes 
\6\ made by licensees after adopting Sec.  50.46a be evaluated by a 
risk-informed integrated safety performance (RISP) assessment process 
that has been reviewed and approved by the NRC via the routine process 
for license amendments.\7\ (See 70 FR 67612-67615.) Most of the 
commenters on the proposed rule stated that current regulatory 
processes that control changes to the facility are adequate and 
therefore, there is no need for the RISP change control process. In 
comments generally supported by all nuclear industry commenters, NEI 
argued that the controls on the existing licensing basis make it 
virtually impossible to make significant adverse changes to the risk 
profile of the plant without being required to submit a license 
amendment request for prior NRC review and approval. NEI concluded that 
the only item that might be missing from the current framework that 
would provide additional assurance that the licensee is appropriately 
maintaining the risk profile of the facility after adoption of Sec.  
50.46a would be a requirement that the licensee periodically assess the 
cumulative impact of facility changes to the risk profile.
---------------------------------------------------------------------------

    \6\ The scope of changes subject to the change criteria in Sec.  
50.46a(f) of the proposed rule would be greater than the changes 
currently subject to Sec.  50.59, which applies only to changes to 
``the facility as described in the FSAR.'' The change criteria in 
the proposed rule would apply to all facility and procedure changes, 
regardless of whether they are described in the FSAR.
    \7\ Requirements for license amendments are specified in 
Sec. Sec.  50.90, 50.91 and 50.92. They include public notice of all 
amendment requests in the Federal Register and an opportunity for 
affected persons to request a hearing. In implementing license 
amendments, the NRC typically prepares an appropriate environmental 
analysis and a detailed NRC technical evaluation to ensure that the 
facility will continue to provide adequate protection of public 
health and safety and common defense and security after the 
amendment is implemented.
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    Industry commenters also considered the proposed rule's unbounded 
scope of the facility changes requiring a RISP assessment to be an 
unnecessary burden and some argued that this requirement is potentially 
adverse to safety. In this regard, the commenters said that because 
most facility changes have no material safety significance, requiring a 
RISP assessment of facility changes beyond even the criteria 
established in current regulations, such as in Sec.  50.59, would add a 
wide range of activities and components to the licensing basis that 
were never reviewed or ever intended to be reviewed by the NRC. Thus, 
licensees would be forced to divert valuable resources from monitoring 
plant safety to tracking a multitude of items that have no safety or 
risk significance. A few commenters recognized that most facility 
changes could be dispositioned with a qualitative RISP assessment but 
argued that there would still be cost associated with the performance 
and documentation of the assessment.
    All commenters stated that the rule should not include the 
operational restriction that all allowable at-power configurations be 
demonstrated to meet the ECCS acceptance criteria. The suggested 
alternatives ranged from reducing the restrictions and placing them 
under licensee control to eliminating them entirely. The commenters 
argued that:
    (1) Existing plant configuration control programs, including 
technical specifications and implementation of the maintenance rule, 
provide sufficient controls to ensure that implementation of Sec.  
50.46a will not lead to plant operation in high risk configurations;
    (2) Because of the low frequency of breaks greater than the TBS 
there should be a minimum of associated operating restrictions;
    (3) Any operating restrictions for breaks larger than the TBS need 
to be commensurate with risk contribution of these larger break sizes; 
and
    (4) Operating restrictions would remove or reduce any potential 
benefit that licensees might gain from the adoption of Sec.  50.46a.
    NRC summary response. The NRC believes that a risk-informed change 
process is a necessary component of this rule because this rule would 
permit changes to facility design bases that would not be allowed under 
current regulations. The current regulatory processes that control 
facility changes are not adequate to control risk-informed plant 
changes that would be allowed under Sec.  50.46a. However, the NRC has 
modified the risk-informed change process considerably by reducing the 
scope of facility changes for which a risk assessment is required. The 
NRC considered requiring all facility changes to be evaluated as risk 
informed changes and permitting licensees to make all facility changes, 
with some exceptions, that satisfy the criteria in Sec.  50.59 or other 
NRC regulations without prior NRC review and approval. The ACRS 
commented that requiring the change in risk from all facility changes 
to be compared to the acceptable risk increase criteria was a 
significant departure from RG 1.174 guidance and other past risk-
informed applications. The ACRS recommended that this proposal be 
reviewed for its implications.
    Instead of requiring risk assessment of all future facility 
changes, the revised proposed rule would require risk assessments for 
only those facility changes enabled by the new ECCS requirements for 
pipe breaks greater than the TBS. This change would

[[Page 40015]]

reduce unnecessary burden and bring the change control process into 
conformance with RG 1.174 and other risk-informed rules and licensing 
actions. Two previous risk-informed regulations promulgated by the NRC 
(i.e., Sec. Sec.  50.69 and 50.48(c)) have included similar 
requirements related to the use of PRA and risk-informed principles to 
demonstrate the acceptability of facility changes enabled by new, risk-
informed regulations before being implemented by licensees.
    The revised proposed rule defines facility changes enabled by Sec.  
50.46a as changes to the facility, technical specifications, and 
procedures that satisfy the revised ECCS analysis requirements in Sec.  
50.46a but do not satisfy the ECCS analysis requirements in Sec.  
50.46. A risk-informed analysis, consistent with that described in RG 
1.174, shall be applied to facility changes enabled by the rule. The 
risk-informed framework established in RG 1.174 permits licensees to 
propose several individual changes to a facility's licensing basis that 
have been evaluated and will be implemented in an integrated fashion. 
Some facility changes proposed by licensees may not be enabled by the 
rule but may lead to a risk decrease. RG 1.174 permits integrated 
(bundled) changes in risk to be compared to the acceptance guidelines 
from RG 1.174 in order to encourage changes that reduce risk. The NRC 
has retained this guidance in Sec.  50.46a(f)(2)(iv) which would permit 
the change in risk from changes enabled by the rule to be combined with 
the change in risk from other plant changes unrelated to the rule for 
the purpose of demonstrating that the change in risk from all changes 
made under the rule meets the acceptance criteria.
    In addition to reducing the scope of facility changes to which the 
risk-informed change process must be applied, the NRC has discarded the 
acronym ``RISP'' in favor of the simpler ``risk-informed'' label 
because the elements and processes described by the RISP are the 
elements and processes that make up a risk-informed evaluation.
    The NRC considered whether to simplify the risk-assessment process 
further by relying primarily on current regulations to identify which 
facility changes a licensee must submit for prior NRC review and 
approval. The ACRS commented that the NRC should use risk criteria to 
determine whether a licensee should submit a change enabled by the rule 
for review and approval. Subsequently, the NRC retained the criteria 
specifying the maximum risk increase for a change that a licensee may 
make without prior NRC review and approval. This requirement frees 
licensees and the NRC from the burden of evaluating and accounting for 
the many individual facility changes that do not have a significant 
impact on risk while retaining NRC review and approval for changes that 
might pose a safety concern.
    In response to comments received on the operational restrictions in 
the proposed rule, the NRC has decided that restrictions must remain on 
plant operation in configurations where it has not been demonstrated 
that breaks larger than the TBS can be mitigated, but the restrictions 
will be modified. The proposed rule prohibited at-power operation in 
any configuration without the demonstrated ability to mitigate a LOCA 
larger than the TBS. The revised proposed rule would restrict at-power 
operation in such a configuration to not exceed a total of fourteen 
days in any 12 month period. Rather than requiring licensees to use 
risk methods to determine how long such operation would be permitted, 
what actions would be required, and how the controls would be 
implemented, in the republished proposed rule the NRC is specifying a 
time limit that simplifies implementation without sacrificing 
flexibility and introducing unnecessary burden. The NRC believes it is 
unlikely that licensees would experience circumstances when they would 
consider operating in such a condition for more than fourteen days but 
feels that maintaining the restriction is necessary.
    Although the LOCA frequencies on which the TBS are founded indicate 
that the expected frequency of breaks larger than the TBS is low, these 
frequencies are estimates derived from an expert elicitation process. 
The NRC has addressed the associated uncertainty, in part, by 
incorporating other elements into the selection of the TBS while 
recognizing that facility changes permitted by the rule could reduce 
the capability to mitigate some accidents that would currently be 
mitigated. The NRC concluded that the consequence of a challenge to the 
facility from an unmitigated break larger than the TBS is severe enough 
to warrant some confidence that the break could be mitigated.
    Although the NRC currently has no guidance explicitly applicable to 
determine an acceptable time interval for operation without mitigation 
capability for a beyond-TBS LOCA, some related guidance is available. 
Previously, the NRC determined that events having at least a 
10<SUP>-7</SUP> probability per year should generally be taken into 
consideration in facility design. This approach is reflected in NUREG-
0800, ``Standard Review Plan for the Review of Safety Analysis Reports 
for Nuclear Power Plants.'' Events taken into consideration in facility 
design are design basis events and must meet the regulations specifying 
the required ability to mitigate the event. This guideline indicates 
that events with a frequency less than 10<SUP>-7</SUP> per year need 
not be considered in facility design. Applying this criterion to 
develop an acceptable time interval during which a beyond-TBS LOCA 
might not be successfully mitigated yields about 4 days per year. 
Regulatory Guide 1.177, ``An Approach for Plant-Specific Risk-Informed 
Decisionmaking; Technical Specifications,'' provides risk guidelines 
that are routinely used to judge the acceptability of time intervals 
that safety-related equipment can be unavailable. Applying the RG 1.177 
criterion yields about 18 days. Neither of these guidelines is fully 
applicable to this configuration. The 10<SUP>-7</SUP> annual 
probability was developed to identify events external to the plant that 
need not be included in the design basis and is not specifically 
applicable to internal events such as LOCAs. Regulatory Guide 1.177 
guidelines are normally applied to an operating configuration when 
mitigation capability would still be available although a single 
failure might fail that capability. Nevertheless, they provide an 
indication that an acceptable period of time should be measured in 
days.
    The NRC chose fourteen days as the appropriate limit on how long a 
plant can operate in a configuration not demonstrated to meet the ECCS 
acceptance criteria for LOCA break sizes larger than the TBS. The NRC 
believes that fourteen days should be sufficient to allow completion of 
on-line maintenance activities relied on to ensure high reliability for 
safety systems while providing adequate protection of public health and 
safety, consistent with the low frequency of these LOCAs. The NRC 
believes that a longer time period for operating in such a plant 
condition would not be consistent with its stated goal of retaining the 
ability to successfully mitigate the full spectrum of LOCAs and would 
not adequately address uncertainties in the evaluation used to select 
the TBS. Conversely, a shorter time period could lead to significant 
burden to the industry with no clear safety benefits and, if 
maintenance activities were adversely affected, a possible reduction in 
safety. Therefore, the NRC will limit the allowed time period for 
operation in an

[[Page 40016]]

unanalyzed condition to fourteen days to ensure that mitigation 
capability is maintained except for occasional, brief periods necessary 
to perform online maintenance of mitigation structures, systems and 
components.
    The NRC concludes that the fourteen day operational restriction 
would protect public health and safety, provide adequate time for 
licensees to perform beneficial maintenance activities, be commensurate 
with the safety significance of LOCAs with a break size larger than the 
TBS and be consistent with the Commission's intent that mitigation 
capability be retained for the full spectrum of LOCA events 
``commensurate with the safety significance of these capabilities.''
    The NRC agrees with commenters that operational restrictions could 
reduce the benefits that may be derived from adopting Sec.  50.46a, but 
the NRC believes that this reduction in benefits is necessary and 
prudent to ensure that some capability to successfully mitigate LOCAs 
larger than the TBS is retained consistent with the risk of these 
events.
    As an example, because the new Sec.  50.46a ECCS analysis 
requirements provide relief from the single failure criterion for pipe 
breaks larger than the TBS, they could permit a facility to increase 
power to the extent that flow from both low pressure safety injection 
trains would be required to fully mitigate beyond-TBS breaks. However, 
the operational restriction in the re-noticed proposed rule would 
require that such a facility reduce power to a level where injection 
from one train is sufficient to mitigate beyond-TBS breaks if the 
second train is inoperable or is removed from service for preventative 
maintenance for longer than fourteen days. The plant would be permitted 
to operate at the increased power level at all other times.
2. Discussion of Specific Comments
    Comment. The RISP process would be an extreme regulatory burden on 
licensees and the NRC to implement. Five reactor licensees said they 
would not implement the proposed rule because of excessive burden.
    NRC response. The NRC disagrees with the commenters that the burden 
to develop and implement a risk-informed evaluation process as 
described in the initial proposed rule is an extreme regulatory burden 
because many elements of a risk-informed evaluation process should 
already exist at power reactors. However, as discussed above, the NRC 
has substantially reduced the scope of facility changes requiring a 
risk-informed evaluation. The revised proposed rule now would require a 
risk-informed evaluation as described in RG 1.174 which is consistent 
with the risk-informed evaluations required by other risk-informed 
applications and regulations. The NRC believes that the burden 
associated with implementing a risk-informed evaluation program would 
be offset by the flexibility provided by the new ECCS analysis 
requirements that will permit facility changes that were not permitted 
by the previous ECCS analysis requirements.
    Comment. The risk-informed evaluation process emphasizes 
insignificant facility changes. The proposed change control 
requirements would require the NRC to be in the business of 
individually reviewing a myriad of insignificant facility changes. The 
risk acceptance criteria for allowing minimal risk changes appear to be 
contrary to the stated goal of enhancing safety. It seems illogical to 
adopt more restrictive requirements on safeguards for beyond design 
basis events than exist for design basis events.
    NRC response. The NRC disagrees that the proposed rule's 
requirements would lead to the NRC individually reviewing insignificant 
facility changes. Facility changes that are enabled by the new ECCS 
requirements may lead to a wide range of estimated increases in risk, 
from immeasurably small to very large. The NRC has established an 
acceptance criterion that specifies the total amount of risk increase 
that would be considered acceptable from changes made under this rule. 
The revised proposed rule also includes a provision that prior NRC 
review is not required for individual facility changes that cause no 
more than a minimal increase in risk when compared to the overall plant 
risk profile. As discussed below, the NRC would consider any increase 
that is less than ten percent of the total acceptable risk increase to 
be minimal. The revised proposed rule includes these criteria to 
prevent NRC review of insignificant changes while retaining the 
capability to review facility changes that might pose a safety concern 
before implementation.
    Comment. The scope of the required PRA is excessive. One commenter 
stated that the PRA scope requirements of Sec.  50.46a(f)(4)(i) in the 
proposed rule appear excessive and should instead use text from NRC 
policy regarding PRA scope requirements relative to an application, 
i.e. ``* * * the PRA scope is such that all operational modes and 
initiating events that could change the regulatory decision 
substantially are included in the model quantitatively.'' Another 
commenter stated that requirements for PRA should not be prescribed in 
the rule. Standards and processes exist to establish requirements for 
PRA technical adequacy (e.g., RG 1.174, RG 1.200, ASME PRA standard). A 
peer-reviewed internal events PRA that meets RG 1.200 should be 
sufficient for Sec.  50.46a implementation. A final commenter stated 
that a requirement for shutdown PRAs is not appropriate because of the 
low risk associated with shutdown configurations at BWRs. Requirements 
for seismic PRAs are also inappropriate because these constitute a 
typically small fraction of the overall risk for most plants.
    NRC response. The NRC does not agree with commenters that the scope 
of the required PRA is excessive and has made no changes to the PRA 
requirements in the revised proposed rule. Further, the NRC believes 
that the proposed rule language regarding PRA scope requirements 
provided by one of the above commenters is consistent with the language 
in both the proposed and the revised proposed rules. Thus, the 
commenter's text was not incorporated into the revised proposed rule.
    The required overall characteristics of the PRA (and the non-PRA 
risk assessment) are included in the rule because these characteristics 
have been determined to be necessary to support decision making and 
inclusion of the characteristics in the rule provides clarity and 
predictability. The revised proposed rule does not prescribe how it 
will be determined whether a licensee's risk-assessment complies with 
these characteristics. The process to evaluate the suitability of each 
licensees' risk assessment will be described in the regulatory guide 
associated with this rule. This process will include staff-endorsed 
industry standards and the peer review process currently used by the 
NRC to evaluate the technical adequacy of PRAs supporting license 
amendment requests.
    Comment. The requirement to update the PRA at a frequency no less 
often than once every two refueling cycles is potentially burdensome. 
An alternative would be to require that after every second refueling 
cycle, that the need for a PRA update is assessed and that appropriate 
action be initiated.
    NRC response. The commenter's suggestion that the need for a PRA 
update be first assessed and appropriate action then be taken is 
consistent with the revised proposed rule. Section 50.46a(f)(2)(iv) 
would require that the PRA reasonably represent the current 
configuration of the plant. If a PRA continues to reasonably represent 
the configuration of the plant after a periodic review, the update 
requirement could be satisfied with a simple conclusion that changes to 
the PRA are not needed. The NRC believes that an

[[Page 40017]]

update interval no longer than two operating cycles is not unduly 
burdensome; thus, the PRA update periodicity was not changed in the 
revised proposed rule.
    Comment. The description of the risk-informed process should not be 
included in the application for a license amendment to implement Sec.  
50.46a. NEI provided complete alternative rule language in its 
comments. At the June 28, 2006, public meeting to clarify the comments, 
NEI emphasized that the proposed rule provided in their comments did 
not require that the RISP process be submitted for review because they 
felt that such a review was unnecessary. Although this comment was not 
formally submitted, several other participants at the June 2006 public 
meeting agreed with this comment.
    NRC response. The NRC disagrees with the comment that a description 
of a licensee's risk-informed assessment process need not be submitted 
for NRC review as part of the licensee's application to adopt Sec.  
50.46a. However, the NRC believes that the amount and complexity of the 
process description that must be submitted will vary appropriately 
depending on which, and how many, facility changes enabled by the rule 
a licensee chooses to make.
    As discussed, the NRC has revised the proposed rule by reducing the 
requirement that all future facility changes be evaluated using a risk-
informed evaluation to only requiring that facility changes enabled by 
the rule be evaluated. Licensees who make limited facility changes 
under the rule, may chose to not submit a request to make future 
facility changes enabled by the rule without prior NRC approval as 
would be permitted in paragraph (c)(1)(iv). Licensees who make one or 
more risk-informed submittals without requesting the authority 
permitted under Sec.  50.46a(c)(1)(iv) would only need to demonstrate 
that the process used to evaluate the specific change(s) described in 
each submittal provides confidence that the requirements of Sec.  
50.46a(f)(2) are satisfied. The content of these submittals is expected 
to be similar to, and consistent with, risk-informed license amendment 
requests currently accepted for review by the NRC.
    A licensee requesting authority to make future changes without NRC 
review as permitted by Sec.  50.46a(c)(1)(iv) must submit for NRC 
review and approval additional information, i.e., the licensee's 
process including its risk assessment models and methods that will be 
used for making future risk-informed changes. Section 50.46a(c)(3)(iii) 
provides that the NRC may approve an application if, in part, the 
licensee's risk-informed evaluation process is adequate for determining 
whether the acceptance criteria in Sec.  50.46a(f) have been met. As 
described in RG 1.174, the technical acceptability of a PRA should be 
commensurate with the application for which it is intended; the level 
of detail required of the PRA should be sufficient to model the impact 
of the proposed change; and the effects of the changes should be 
appropriately accounted for. A licensee's submittal to make future 
changes must provide sufficient information on both the risk assessment 
models and how future changes will be reflected in these models, to 
allow the NRC to conclude that the requirement in Sec.  
50.46a(c)(3)(iii) is met.
    Comment. Requirements on late containment failure should be 
removed. It is inappropriate to require licensees to retain a level of 
mitigation for late containment failure and late radiological releases, 
because these releases constitute a very small fraction of overall 
plant risk. Therefore, these references should be removed.
    NRC response. The NRC is proposing changes in the revised proposed 
rule that would make this topic moot. The commenter was remarking on 
the parenthetical ``(early and late)'' that was added to the 
containment related defense in depth element described in RG 1.174 when 
three of the elements were incorporated as acceptance criteria in the 
proposed rule. The NRC has removed the defense-in-depth acceptance 
criteria in the revised proposed rule, including the reference to early 
and late containment failures, but has retained the general criterion 
that defense-in-depth be maintained.
    The NRC will continue to follow the guidelines in RG 1.174 to 
address defense-in-depth when evaluating whether a licensee has 
satisfied the rule criterion that defense-in-depth has been maintained. 
The RG 1.174 guidelines for defense-in-depth in risk-informed 
applications have been used successfully by the NRC for more than a 
decade and do not need further clarification through rulemaking. 
Retaining the defense-in-depth guidelines in a regulatory guide instead 
of promulgating acceptance criteria in the rule would also allow the 
NRC to more effectively update its guidance as new information becomes 
available or if the Commission changes its policy.
    Comment. Section 50.46a(f)(4) contradicts Sec.  50.46a(f)(5). One 
commenter stated that Sec.  50.46a(f)(4) implies that only a PRA 
meeting the requirements of the following paragraphs may be used in the 
risk-informed assessment. This was seen as contradictory to Sec.  
50.46a(f)(5), which allows non-PRA risk assessment methods.
    NRC response. The NRC disagrees that the rule language is 
contradictory. The relevant phrase in Sec.  50.46a(f)(4) states that 
``* * * to the extent that a PRA is used in the risk-informed 
assessment, it must * * *,'' meet the following PRA requirements. If a 
PRA need not be used according to Sec.  50.46a(f)(1)(i) and (f)(2)(ii), 
and a PRA is not used, then non-PRA risk assessment methods that 
satisfy the requirements in Sec.  50.46a(f)(5) may be used. No changes 
were made in the revised proposed rule.
    Comment. Performance monitoring is already covered by Appendix B to 
Part 50. One commenter stated that the proposed requirement for a 
monitoring program designed to detect and prevent degradation of 
systems, structures, and components (SSCs) before plant safety is 
compromised is unnecessary. The commenter stated that 10 CFR Part 50, 
Appendix B, Criterion XVI for corrective action already contains this 
requirement.
    NRC response. The NRC does not agree. Appendix B to 10 CFR Part 50 
applies to safety-related SSCs and activities. The risk-informed 
decision process includes risk models that consider a much broader set 
of accidents and can credit a larger set of equipment and actions to 
mitigate these accidents than the set of safety-related equipment or 
actions. The NRC believes that performance measurement is an important 
part of risk-informed decision making that must be applied irrespective 
of the classification of an SSC or activity as ``safety-related.'' The 
performance monitoring requirement remains in the revised proposed 
rule.
    Comment. Power uprates and relaxation of the single failure 
criteria for breaks larger than a TBS LOCA could result in a situation 
when all emergency power supplies are needed to successfully mitigate a 
break larger than the TBS when accompanied by a loss-of-offsite power. 
The potential consequences of relying on the availability of offsite 
power supply in a deregulated environment or a requirement to have both 
divisions of onsite power available (without single failure capability) 
to mitigate the uprated reactor accident would not appear to be offset 
by any compensatory factors.
    NRC response. The NRC agrees that licensees who adopt Sec.  50.46a 
could potentially make changes to the facility such that all emergency 
onsite power

[[Page 40018]]

supplies were required to demonstrate successful mitigation of a break 
larger than the TBS when accompanied by a loss-of-offsite power. Such 
an operating configuration would not be permitted by the current 
regulations. Licensees who adopt Sec.  50.46a would have the 
flexibility to make facility changes that would not normally be 
permitted by current ECCS regulations but must comply with all the 
requirements of Sec.  50.46a. One requirement is to demonstrate that 
all changes made under the rule meet the risk acceptance criteria in 
Sec.  50.46a(f) before the facility change may be implemented. Another 
requirement is that the change in risk from all changes to the facility 
must be periodically assessed and steps must be taken if the result 
exceeds the acceptance criteria in Sec.  50.46a(f)(2). If changes to 
the plant-specific emergency power configuration and/or grid 
reliability over time result in risk increases exceeding the acceptance 
criteria, the plant changes that would permit this operating 
configuration may not be implemented, or other steps must be taken to 
reduce overall facility risk.
    However, in response to the ACRS recommendation in the November 16, 
2006, letter from Graham Wallis to Chairman Dale E. Klein, 
(ML063190465), to increase the level of defense-in-depth provided by 
the rule for mitigating LOCAs larger than the TBS, the NRC has modified 
the revised proposed rule with respect to the availability of onsite 
electrical power. The NRC has added the requirement that all equipment 
needed to mitigate pipe breaks larger that the TBS must be designed so 
that onsite power can be provided to the equipment. Onsite power may be 
provided automatically or as the result of manual actions taken by 
facility staff within a time frame that provides mitigation of damage 
and accident consequences. Although the ECCS analyses for pipe breaks 
larger than the TBS may still assume the availability of offsite power, 
the availability of onsite power to the necessary equipment provides 
additional defense-in-depth for postulated large break accidents.

E. Comments Related to the Applicability of the Backfit Rule

    Comment. Commenters stated that the proposed rule provision 
limiting the applicability of the backfit rule is unnecessary. These 
commenters stated that the rule requires maintaining a mitigation 
capability up to the largest LOCA, regardless of the size of the TBS. 
The NRC should either apply the backfit rule to future changes in the 
TBS, or define a set of criteria defining how and when the NRC would 
determine that the TBS is no longer acceptable. Licensees should be 
provided with a great deal of latitude on achieving compliance 
following any change in the TBS, with the goal being that risk 
requirements are achieved with a reasonable mix of prevention and 
mitigation.
    NRC response. The NRC disagrees, for the most part, with the 
comments on this question. Because the estimated low LOCA frequency and 
corresponding low risk of large LOCAs is necessary to maintain 
assurance of public health and safety with this risk-informed 
regulation, the NRC believes that the exclusion of TBS changes from the 
backfit rule must be maintained in case future changes in estimated 
LOCA frequency require changes to the TBS.
    With respect to a commenter's argument about the continuing 
regulatory requirement for LOCA mitigative capability beyond the TBS, 
the NRC notes that even though mitigative capability is retained, the 
proposed beyond-TBS mitigative capability is reduced, as compared to 
the capability required under the current ECCS rule. In developing the 
proposed rule, the NRC recognized the open-ended nature of the backfit 
exclusion. The NRC attempted to develop criteria for assessing whether 
new information mandates a change to the TBS. Unfortunately, the NRC 
was unable to develop relatively clear criteria and it was concluded 
that adoption of generalized criteria for constraining the NRC in 
future changes to the TBS would not prove useful or practical. Thus, 
the proposed rule did not set forth proposed criteria for assessing 
whether new information mandates a change to the TBS. The NRC notes 
that no commenter suggested any criteria for assessing the need for, or 
desirability of, changes to the TBS based upon new information.
    The NRC agrees that the proposed amendment should provide licensees 
with substantial flexibility to determine the manner in which they 
would come back into compliance with applicable regulatory requirements 
following any future change in the TBS. Licensees who must take actions 
to come back into compliance need not return the plant to the precise 
conditions and circumstances in effect immediately before 
implementation of the Sec.  50.46a regulation. Rather, licensees should 
be afforded the flexibility of deciding what actions to implement to 
comply with a revised TBS. Further, as one of the commenters suggests, 
the overall goal of any actions taken to restore compliance is to 
achieve a reasonable mix of prevention and mitigation. The NRC will 
consider making this clear in implementing guidance. For these reasons, 
the NRC has decided to adopt the exclusion of future TBS changes from 
the backfit rule by retaining the provisions of proposed Sec. Sec.  
50.46a(m) and 50.109(b)(2) in the revised proposed rule.
    Comment. Proposed Sec. Sec.  50.109(b)(2) and 50.46a(d)(5) should 
not be adopted, and any changes to the TBS should be accomplished by 
rulemaking, and evaluated under the backfit rule. Excluding future 
changes to the TBS from compliance with the backfit rule would defeat 
the goal of regulatory stability embodied in the backfit rule and may 
result in changes that are not cost-justified.
    NRC response. The NRC disagrees with the comment that the NRC's 
three reasons for excepting TBS changes and any consequent licensee 
reanalyses and changes from the backfit rule do not address how the 
objectives of the backfit rule are met. On the contrary, the NRC's 
first reason (consideration of costs and benefits in a regulatory 
analysis) and the third reason (flexibility may reduce impacts of 
changes in the TBS) directly address the underlying objectives of the 
backfit rule. In addition, the second reason (application of the 
backfit rule favors incremental increases in risk) is relevant to the 
backfit rule's ``substantial increase in protection'' criterion. A 
backfitting standard that limits increases in protection to public 
health and safety or common defense and security to those which are 
both substantial and cost-justified, but ignores (or allows) 
incremental decreases in protection without restriction does not seem 
to be a justifiable regulatory approach. Hence, the NRC believes that 
adoption of criteria to control these incremental decreases is 
justifiable and appropriate, even if inconsistent with the objective of 
regulatory stability, which is, arguably, the primary objective of the 
backfit rule.
    Finally, the NRC agrees that the goal of regulatory stability is 
not negated by the fact that a licensee's decision to comply with Sec.  
50.46a rule would be optional or voluntary. On the contrary, the NRC 
believes that regulatory stability should be an important factor in 
developing a rule. However, the NRC disagrees with the commenter's 
implicit assertion that, absent consideration under the backfit rule, 
regulatory stability would not be appropriately considered in any 
future revisions to the TBS. As the NRC stated in the statement of 
considerations in the proposed rule, a regulatory analysis would be 
required for any revision to the TBS. (See 70 FR 67617-67618.) This 
regulatory tool provides an appropriate means of

[[Page 40019]]

ensuring that regulatory stability is considered by the NRC when 
determining whether to revise the TBS.
    Comment. The NRC should not adopt the backfitting exclusion 
provision in Sec.  50.46a(d), which would require that any facility 
changes made necessary by the maintenance and upgrading of risk 
assessments, would not be deemed to be backfitting.
    NRC response. The NRC disagrees with this comment, which was part 
of a broader comment opposing the proposed rule's provision excluding 
from backfit consideration changes to a plant and its procedures that 
are necessitated by any future TBS changes mandated by the NRC (see the 
immediately-preceding comment analysis). The commenter did not provide 
a separate basis supporting its position that licensee changes 
necessitated by the periodic risk assessment maintenance and upgrading 
(as contrasted with NRC-mandated TBS changes) should be subject to 
backfitting consideration.
    The NRC believes that the policy and regulatory considerations with 
respect to backfitting of changes stemming from future TBS changes are 
irrelevant to the policy and regulatory considerations with respect to 
backfitting of changes required to maintain compliance with updated 
risk analyses. The NRC regards plant changes necessitated by periodic 
risk assessments under Sec.  50.46a to be analogous (from a backfitting 
standpoint) to the 120-month updating of inservice inspection (ISI) and 
inservice testing (IST) under Sec.  50.55a(f) and (g). Under those 
provisions, a licensee must update its ISI and IST program every 120 
months to the latest version of the ASME Code in effect 12 months 
before the beginning of the next inspection interval. The NRC has 
stated that the 120-month updating does not constitute backfitting, in 
part because the regulatory requirement for updating is known to the 
operating license applicant before it receives its license, which 
addresses the policy of regulatory stability and predictability 
embodied in the backfit rule. See 69 FR 58804, 58817 (third column) 
(October 1, 2004); 67 FR 60520, 60536-60537 (September 26, 2002). This 
logic also applies to the periodic risk assessment maintenance and 
upgrading under Sec.  50.46a(d)(4) and any necessary licensee actions 
necessary to maintain compliance with the relevant 50.46a acceptance 
criteria. The NRC also notes that Sec.  50.46a does not prescribe any 
specific manner or approach for achieving compliance following the 
periodic risk assessment maintenance and upgrading under Sec.  
50.46a(d)(4); this performance-based approach to regulation affords the 
licensee substantial flexibility and gives the licensee control over 
how best to achieve compliance. This further tends to reduce the impact 
of Sec.  50.46a(d)(4) on licensees, which is an implicit objective of 
the backfit rule. For these reasons, the NRC declines to adopt the 
commenter's recommendation.
    Comment. The fact that the proposed rule provides an alternative or 
voluntary approach for LOCA analysis does not negate either the backfit 
rule itself or the policy of regulatory stability.
    NRC response. The NRC disagrees with the comment. As discussed 
elsewhere in the backfitting discussion, the backfit rule's protections 
apply only when the NRC is imposing (directly or indirectly) a change 
to the activities authorized by a license; it does not apply when the 
NRC is providing a regulatory approach as an alternative to compliance 
with an existing regulatory requirement. As a general matter, the 
regulatory stability and predictability afforded to a licensee by the 
backfit rule applies to the scope of activities approved by the 
license. If a licensee seeks a change to its licensing basis--which is 
what a transition to a voluntary alternative is--the licensee is 
seeking to do something that is not within the scope of activities 
authorized by its license. It is the NRC's view that, in such a 
circumstance, the licensee has no reasonable expectation that the NRC's 
criteria for judging the acceptability of that proposed change remains 
the same as the criteria used by the NRC in judging the original 
license application. Thus, the protections of the backfit rule do not 
apply either when a licensee seeks a voluntary change to its licensing 
basis, or when the NRC develops a voluntary alternative.
    Comment. The NRC set forth three justifications for excepting TBS 
changes from backfitting protection: the consideration of alternatives 
will occur in the required regulatory analysis; application of the 
backfitting rule effectively favors increases in risk; and the 
flexibility provided by the rule will tend to reduce the burden of any 
changes in the TBS. However, even if these justifications are true, 
they do not address how the objective of the backfit rule will be met 
or that this objective does not apply.
    NRC response. The NRC disagrees in part with this comment. The NRC 
views the backfit rule as having three underlying objectives: 
regulatory stability and predictability for a licensee; reasoned agency 
decisionmaking (that NRC's decision to impose a backfit is assessed 
against rational criteria); and transparency of agency decisionmaking 
(that the reasons for the NRC's determination on the overall backfiting 
criteria are publicly available). The second and third objectives would 
be met if the NRC imposes future TBS changes by rulemaking (which is by 
far the most likely course), inasmuch as such a rulemaking must include 
preparation of a regulatory analysis. A regulatory analysis which is 
performed in accordance with the NRC's ``Regulatory Analysis 
Guidelines'', NUREG/BR-0058, Revision 4 (2004), provides for a 
disciplined agency decisionmaking process. The draft regulatory 
analysis is published and made available for public comment as part of 
the proposed rule. The final regulatory analysis, which addresses 
public comments, is also made available to the public as part of the 
final rulemaking. Hence, the NRC believes that the backfit rule's 
objectives of reasoned decisionmaking and transparency of agency 
decisionmaking will be satisfied by any rulemaking changes to the TBS. 
With respect to the first objective of the backfit rule, the NRC 
recognizes that exclusion of future changes to the TBS from the backfit 
rule could lead to reduced regulatory stability and predictability 
because neither the adequate protection, compliance, or substantial 
safety increase criteria would be binding as checks against unwarranted 
agency action. However, the NRC believes that this is offset to some 
extent by two factors. First, by explicitly excluding future TBS 
changes and necessary changes from the backfit rule, licensees who 
choose to adopt Sec.  50.46a are aware that the NRC may revise the TBS 
in the future (the argument here is similar to the Commission's 
determination that the backfit rule does not apply to rulemakings 
endorsing more recent editions and addenda of the ASME Code for 
mandatory use in the 120-month interval process for ISI and IST in 
Sec. Sec.  50.55a(f) and (g)). Second, the NRC acknowledges that plant-
specific orders imposing TBS changes would not necessarily meet all of 
the backfit rule objectives. However, the NRC's internal process 
governing the development and issuance of orders should, at minimum, 
result in reasoned decisionmaking. Moreover, as is the case with 
rulemaking changes to the TBS, regulatory predictability for changes to 
the TBS by order is addressed somewhat by explicitly stating in both 
Sec. Sec.  50.109 and 50.46a that the backfit rule does not apply if a 
revised TBS is imposed by order. These provisions provide notice to 
licensees considering adoption of Sec.  50.46a of the special 
backfitting

[[Page 40020]]

process under Sec.  50.46a. Licensees contemplating adoption of Sec.  
50.46a may then factor this limited exclusion from the backfit rule 
into their decision whether to adopt Sec.  50.46a.
    Comment. The Commission-proposed exclusion of TBS changes from 
backfitting protection would leave licensees who voluntarily adopt 
Sec.  50.46a without recourse to a backfit appeal process.
    NRC response. The NRC disagrees with the comment. Licensees who 
adopt Sec.  50.46a would continue to have access to the backfitting 
appeals process with respect to licensee-claims of backfit for all 
matters other than those attributable to TBS changes.
    Further, affected licensees would have an opportunity to raise 
concerns about the cost and expected benefits of proposed TBS changes, 
whether the TBS changes are imposed by rulemaking or by order. If the 
TBS were accomplished through rulemaking, all licensees would have an 
opportunity to comment on the proposed rule, including the associated 
regulatory analysis. By contrast, if the NRC imposes a TBS change by 
order, the affected licensee would have an opportunity to request a 
hearing on the order. During this hearing any issues could be raised on 
costs and benefits for the TBS change as applied to that licensee. 
Although these opportunities do not constitute, strictly speaking, a 
backfit appeal process, the NRC believes that they are the functional 
equivalent of a backfit appeal process.
    Finally, as noted earlier, it is the NRC's expectation that should 
it mandate a change in the TBS, that licensees would have substantial 
discretion and flexibility with respect to how they would address that 
TBS change. Accordingly, the NRC sees no additional benefit from 
providing a licensee with a plant-specific backfitting appeal process 
related to TBS changes in addition to the public comment and hearing 
opportunities already provided for by law.

F. Comments on Topics Requested by the Commission

    In the initial proposed rule, the NRC identified 16 significant 
topics associated with the proposal and invited the public to submit 
specific comments on those issues. (See 70 FR 6718--6719.)
    NRC Topic 1. In proposed Sec.  50.46a(b), the NRC specifically 
precludes the application of the Sec.  50.46a alternative requirements 
to future reactors. However, future light water reactors might benefit 
from Sec.  50.46a. The NRC requests specific public comments regarding 
whether Sec.  50.46a should be made available to future light water 
reactors.
    Comments. Framatome commented that Sec.  50.46a should be available 
to nuclear power plants licensed after the publication of the rule that 
are of similar design to the current generation of operating BWRs and 
PWRs. Framatome stated that the advanced LWR designs previously 
certified (ABWR, System 80+, AP 600, AP 1000), under design 
certification review (ESBWR) and in the pre-review process (US EPR), 
all fit into this category and can realize benefits from Sec.  50.46a. 
However, for Sec.  50.46a to apply to a new design, the NRC must first 
make a determination that the design is substantially similar to 
currently operating LWRs. The applicability to the new design of the 
frequency of pipe rupture versus break size curves used as a basis for 
establishing the TBS in Sec.  50.46a must be established. The WOG 
stated that future PWRs and BWRs operating with materials, pressures 
and temperatures similar to operating LWRs should be able to use Sec.  
50.46a because there is no technical reason that new plants should have 
to meet outdated requirements for which existing plants can opt out. 
The BWROG and three other commenters also stated that Sec.  50.46a 
should be made available to future light water reactors.
    NRC response. The NRC agrees with the commenters who stated that 
there are no technical reasons which prevent the new Sec.  50.46a 
regulations from being applied to new light water reactor designs that 
are similar in nature (with respect to design and expected LOCA pipe 
break frequency) to current operating reactors. However, it would be 
difficult to apply the new regulation to certified reactor designs 
which have already received NRC approval. These design approvals were 
completed as rulemaking activities for the particular standardized 
design as of the date of the application, as amended. Changes may not 
be made to these designs unless the designers choose to resubmit the 
designs for reevaluation and reopen the design approval/rulemaking 
process to address Sec.  50.46a. Moreover, it is not clear that these 
changes could be made under the special backfitting criteria in Sec.  
52.63, because it does not appear that there is an issue related to 
adequate protection, compliance with requirements in effect at the time 
of certification, reduction of unnecessary burden, providing detailed 
design information, correcting material errors in the certification 
information, increasing standardization, or providing a substantial 
increase in overall safety, reliability, or security.
    Three new standardized LWR designs and one resubmitted LWR design 
are now being considered by the NRC. Although the NRC has not performed 
a detailed analysis of these new designs in the manner done for 
establishing the technical basis of this rule for existing designs, the 
frequency of large LOCAs at these facilities could be as low as it is 
at current LWRs. Thus, it may be appropriate to apply the alternative 
Sec.  50.46a requirements to these future designs. Accordingly, the 
revised proposed rule has been modified to apply to new reactor 
designs, e.g. facilities other than those which are currently licensed 
to operate. Applicants for design certification or combined licenses, 
holders of combined licenses under Part 52, or future licensees of 
operating new light-water reactors who wish to apply Sec.  50.46a must 
submit an analysis for NRC approval, demonstrating why it would be 
appropriate to apply the alternative ECCS requirements and what the 
appropriate TBS would be for the new design to meet the intent of Sec.  
50.46a.
    In its analysis, the applicant, holder, or licensee must 
demonstrate that the proposed reactor facility is similar to reactors 
licensed before the effective date of the rule. In addressing 
similarity of the proposed reactor design to current reactor designs 
licensed before the effective date of the rule, the applicant, holder, 
or licensee would need to address design, construction and fabrication, 
and operational factors that include, but are not limited to:
    (1) The similarity of the piping materials of construction and 
construction techniques for new reactors to those in the currently 
operating fleet;
    (2) The similarity of service conditions and operational programs 
(e.g., in-service inspection and testing, leak detection, quality 
assurance etc.) for new reactors to those for operating plants;
    (3) The similarity of piping design, e.g. pipe sizes and pipe 
configuration, for new reactors to those found in operating plants;
    (4) Adherence to existing regulatory requirements, regulatory 
guidance, and industry programs related to mitigation and control of 
age-related degradation (e.g., aging management, fatigue monitoring, 
water chemistry, stress corrosion cracking mitigation etc.); and
    (5) Any plant-specific attributes that may increase LOCA 
frequencies compared to the generic results in NUREG-1829 and NUREG-
1903.
    The analysis must also include a recommendation for an appropriate 
TBS and a justification that the

[[Page 40021]]

recommended TBS is consistent with the technical basis for this 
proposed rule. For new reactor designs that employ design features that 
effectively increase the break size, via opening of specially designed 
valves, to rapidly depressurize the reactor coolant system during any 
size loss of coolant accident, justification of the relevance of a TBS 
would be necessary. The methodology used to determine the proposed TBS 
should be described in the justification. Based on information 
currently available, new reactor designs may have similar piping 
materials, similar service conditions and operational programs, similar 
piping designs, and similar mitigation and control of age-related 
degradation programs to those found in currently operating plants. 
Therefore, based on information currently available, the NRC envisions 
that the TBS defined in the revised proposed rule could be applicable 
to the new reactor designs.
    In addition, a holder of an operating or combined license for a 
plant with a currently approved standard design could adopt Sec.  
50.46a if the design is demonstrated, by satisfying the five criteria 
above, to be similar to the designs of plants licensed before the 
effective date of the rule and the TBS proposed by the licensee is 
found acceptable by the NRC.
    In the revised proposed rule language and elsewhere in this 
document, whenever the NRC refers to similarity of the designs of new 
reactors to the designs of current operating reactors, the NRC intends 
for ``design'' to be broadly interpreted to encompass design, 
construction and fabrication, and operational factors that should be 
addressed, at a minimum, by considering the five similarity factors 
indentified above.
    NRC Topic 2. The TBS specified by the NRC in the proposed rule does 
not include an adjustment to address the effects of seismically-induced 
LOCAs. NRC is currently performing work to obtain better estimates of 
the likelihood of seismically-induced LOCAs larger than the TBS. By 
limiting the extent of degradation of reactor coolant system piping, 
the likelihood of seismically-induced LOCAs may not affect the basis 
for selecting the proposed TBS. However, if the results of the ongoing 
work indicate that seismic events could have a significant effect on 
overall LOCA frequencies, the NRC may need to develop a new TBS. To 
facilitate public comment on this issue, a report from this evaluation 
will be posted on the NRC rulemaking Web site at http://ruleforum.llnl.gov before the end of the comment period. Stakeholders 
should periodically check the NRC rulemaking Web site for this 
information. [The NRC published the report on December 20, 2005 (70 FR 
75501; ML053470439).] The NRC requests specific public comments on the 
effects of pipe degradation on seismically-induced LOCA frequencies and 
the potential for affecting the selection of the TBS. The NRC also 
requests public comments on the results of the NRC evaluation that will 
be made available during the comment period.
    NRC response. Comments received on this topic were previously 
discussed in Section IV.B. of this document, ``Comments on Seismic 
Considerations Related to the TBS.'' Because this topic was identified 
for public comment in the initial proposed rule, the NRC completed and 
published the study on the risks associated with seismically induced 
LOCAs larger than the TBS (NUREG-1903, ``Seismic Considerations for the 
Transition Break Size'' February 2008; ML080880140). The NRC considered 
the public comments received on seismic considerations in the final 
version of NUREG-1903. As previously discussed in Section IV.B of this 
document, the NRC has concluded that no adjustment to the TBS is needed 
to account for seismically-induced LOCAs.
    NRC Topic 3. Depending on the outcome of an ongoing NRC study, the 
final rule could include requirements for licensees to perform plant-
specific assessments of seismically-induced pipe breaks. These 
assessments would need to consider piping degradation that would not be 
prejudiced by implementation of the licensee's inspection and repair 
programs. The assessments would have to demonstrate that reactor 
coolant system piping will withstand earthquakes such that the seismic 
contribution to the overall frequency of pipe breaks larger than the 
TBS is insignificant. The NRC requests specific public comments on this 
and any other potential options and approaches to address this issue.
    NRC response. After this topic was identified, the NRC completed 
and published the study on the risks associated with seismically-
induced LOCAs larger than the TBS (NUREG-1903, ``Seismic Considerations 
for the Transition Break Size'' February 2008; ML080880140). Comments 
received on this topic were previously addressed in Section IV.B of 
this document, ``Comments on Seismic Considerations Related to the 
TBS.'' The NRC has concluded that applicants wishing to implement the 
alternative ECCS requirements should conduct a plant-specific 
assessment of the risk associated with seismically-induced failures of 
flawed piping. The NRC is currently preparing guidance for conducting 
these plant-specific assessments (``Plant-Specific Applicability of 10 
CFR 50.46 Technical Basis'' February 2009; ML090350757).
    NRC Topic 4. The ACRS noted that ``a better quantitative 
understanding of the possible benefits of a smaller break size is 
needed before finalizing the selection of the transition break size.'' 
The TBS to be included in the final rule should be selected to maximize 
the potential safety improvements. Thus, the NRC is soliciting comments 
on the relationship between the size of the TBS and potential safety 
improvements that might be made possible by reducing the maximum 
design-basis accident break size.
    NRC response. No comments were received which specifically 
addressed the relationship between the size of the TBS and potential 
safety improvements that might be made possible by reducing the maximum 
design-basis accident break size. However, the WOG stated, ``It is not 
appropriate to set the TBS on the basis of where the most benefit is, 
as this may change tomorrow and there will be no easy recourse.'' This 
comment and other related issues were previously discussed in Section 
III.A of this document, ``Comments on Selection of the TBS''. The NRC 
made no changes to the size of the TBS in the revised proposed rule.
    NRC Topic 5. Proposed Sec.  50.46a includes an integrated, risk-
informed change process to allow for changes to the facility following 
reanalysis of beyond design basis LOCAs larger than the TBS. However, 
because the current regulations in 10 CFR part 50 already have 
requirements addressing changes to the facility (Sec. Sec.  50.59 and 
50.90), it might be more efficient to include the integrated, risk-
informed change (RISP) requirements for plants that use Sec.  50.46a 
under these existing change processes. The NRC solicits specific public 
comments on whether to revise existing Sec. Sec.  50.59 and 50.90 to 
accommodate the requirements for making facility changes under Sec.  
50.46a.
    Comments. Three commenters responded directly to this question. One 
stated that Sec. Sec.  50.59 and 50.90 should not be revised to 
accommodate the requirements for making plant changes under Sec.  
50.46a. Another stated that Sec.  50.59 requirements could be augmented 
to address the risk evaluations but that the augmentation was not 
necessary. The third commenter stated that Sec. Sec.  50.59 and 50.90 
should contain change requirements for Sec.  50.46a but that these 
requirements

[[Page 40022]]

should not be the RISP requirements included in the proposed rule.
    NRC response. The NRC is not changing Sec. Sec.  50.59 and 50.90 to 
include integrated, risk-informed change requirements. The NRC has 
modified the risk-informed change control process to apply only to 
facility changes made under the rule, i.e., facility changes enabled by 
the rule as well as other facility changes unrelated to the rule but 
bundled together by the licensee for estimating the change in risk. 
Other facility changes would be unrelated insofar as the basis of the 
changes and NRC approval, when necessary, will rely on regulations, 
guidelines, or facility priorities that do not depend on the new TBS. 
The NRC changed the process to more closely follow the process 
described in RG 1.174, which has been used successfully for a wide 
variety of risk-informed applications. The NRC has concluded that this 
risk-informed change control process can be used to successfully and 
safely implement facility changes enabled by the new TBS LOCA in the 
Sec.  50.46a final rule.
    NRC Topic 6. The proposed rule would rely on risk information. The 
NRC has included specifically applicable PRA quality and scope 
requirements in the proposed rule. However, there are other NRC 
regulations that also rely on risk information (e.g. the maintenance 
rule in Sec.  50.65 and Sec.  50.69 pertaining to alternative special 
treatment requirements). Consistent with the Commission policy on a 
phased approach to PRA quality, it might be more efficient and 
effective to describe PRA requirements (e.g., contents, scope, 
reporting, changes, etc.) in one location in the regulations so that 
the PRA requirements would be consistent among all regulations. The NRC 
is seeking specific public comments on whether it would be better to 
consolidate all PRA requirements into a single location in the 
regulations so that they were consistent for all applications or to 
locate them separately with the specific regulatory applications that 
they support.
    Comments. Five commenters recommended that it would be preferable 
to collect all PRA requirements in a single location in the 
regulations, but they all also stated that it would be premature to use 
the Sec.  50.46a rulemaking to combine PRA requirements at the present 
time. Some commenters argued that different applications have different 
requirements for the supporting PRA analyses and cautioned that PRA 
requirements should not be based on the most demanding application.
    NRC response. The NRC takes note of the recommendation that PRA 
requirements be eventually collected into a single location in the 
regulations. The NRC agrees that the Sec.  50.46a rulemaking is not the 
appropriate vehicle to achieve this regulatory change. The NRC will 
include PRA requirements adequate to support this rulemaking in the 
Sec.  50.46a rule. After the NRC develops broad-based PRA requirements 
suitable for use on a generic basis in different applications, the NRC 
will be able to codify these generic PRA requirements in a single 
regulatory location and could remove the Sec.  50.46a specific PRA 
requirements (or limit them to existing licensees approved under Sec.  
50.46a to avoid backfitting).
    NRC Topic 7. Proposed Sec.  50.46a would include the requirement 
that all allowable at-power operating configurations be included in the 
analysis of LOCAs larger than the TBS and demonstrated to meet the ECCS 
acceptance criteria. Historically, operational restrictions have not 
been contained in Sec.  50.46 but were controlled through other 
requirements (e.g., technical specifications and maintenance rule 
requirements). It might be more practical to control the availability 
of equipment credited in the beyond design-basis LOCA analyses in a 
manner more consistent with other operational restrictions. As a 
result, the NRC is soliciting public comments on the most effective 
means for implementing appropriate operational restrictions and 
controlling equipment availability to ensure that ECCS acceptance 
criteria are continually met for beyond design-basis LOCAs.
    Comment. As previously discussed, all commenters stated that the 
NRC should not include the operational restriction that all allowable 
at-power operating configurations be demonstrated to meet the ECCS 
acceptance criteria. Several commenters proposed alternatives ranging 
from placing limits that might be required in licensee-controlled 
documentation to eliminating all operational restrictions associated 
with breaks greater than the TBS. Most commenters stated that 
operational restrictions negated the relief from the requirement to 
assume the worst single failure during the evaluation of beyond TBS 
breaks.
    NRC response. As discussed in Section III.D of this document, the 
NRC has decided that operational restrictions must be retained if it 
cannot be demonstrated in the analysis of LOCAs larger that the TBS 
that the ECCS acceptance criteria are met, but the restrictions would 
be reduced. The proposed rule prohibited at-power operation in a 
configuration without the demonstrated ability to mitigate a LOCA 
larger than the TBS. The revised proposed rule would require that at-
power operation in such a configuration shall not exceed a total of 
fourteen days in any 12-month period. The NRC believes that this change 
will satisfy the Commission's intention that mitigative capability be 
maintained for all breaks up to the double-ended rupture of the largest 
reactor coolant pipe and still allow a reasonable amount of time for 
licensees to make corrective actions needed to restore the plant to a 
fully analyzed configuration.
    NRC Topic 8. Given the Commission's intent (see SRM for SECY-04-
0037) that facility changes made possible by this proposed rule should 
be constrained in areas where the current design requirements 
``contribute significantly to the `built-in capability' of the plant to 
resist security threats,'' the NRC seeks examples on either side of 
this threshold (facility changes allowed versus facility changes 
prohibited), and additionally any examples of facility changes made 
possible by Sec.  50.46a that could enhance plant security and defense 
against radiological sabotage or attack. The NRC also solicits comments 
on whether the proposed Sec.  50.46a rule should explicitly include a 
requirement to maintain plant security when making facility changes 
under Sec.  50.46a or otherwise rely on a separate rulemaking now being 
considered by the NRC to more globally address safety and security 
requirements when making facility changes under Sec. Sec.  50.59 and 
50.90. Any examples of facility changes that involve safeguards 
information should be marked and submitted using the appropriate 
procedures.
    Comments. On the first question regarding examples of facility 
changes that should or should not be constrained in areas where the 
current design requirements ``contribute significantly to the `built-in 
capability' of the plant to resist security threats,'' NEI said that 
the proposed rule would not enable facility changes that reduce plant 
safety margins as well as the capacity to deal with security threats. 
NEI stated that the opposite is true because the proposed rule would 
increase the safety focus on risk-significant events and mitigating 
equipment, and improve the reliability and availability of this 
equipment by removing excessive conservatism from the design basis.
    On the second question as to whether the Sec.  50.46a rule should 
contain a security requirement, NEI said that

[[Page 40023]]

existing change control requirements in the regulations preclude 
significant reductions in safety or security. The BWROG supported the 
NEI position on this issue. The WOG stated that the security-related 
aspects of facility changes that might be enabled by this rule change 
should be addressed in the evaluation of those specific facility 
changes. The WOG also stated that the changes to Sec.  50.46a should 
not be tied to security issues. Making a ``security connection'' to 
this proposed amendment would introduce needless complications and be 
counterproductive. Issues related to preserving ``built-in capability'' 
of the plant to resist threats should be addressed centrally in a 
single location within the regulations. Maintaining all requirements 
related to security in one place, either in the regulations or in 
Commission policy, is the most appropriate way to avoid conflicting 
information and enhance the ease of change. Progress Energy stated that 
consideration for security concerns should be included in the 
consideration of safety concerns to avoid possible negative effects 
caused by these sometimes competing objectives. However, to simplify 
the processes and maintain consistency, the safety and security 
interface should be addressed globally by a separate rulemaking.
    NRC response. The NRC agrees with commenters that security 
requirements should be addressed by regulations separate from those in 
Sec.  50.46a. The NRC is not adding security requirements to proposed 
Sec.  50.46a. Security requirements will continue to be addressed by 
overall security requirements located elsewhere in the regulations. 
Specifically, 10 CFR 73.58, ``Safety/security Interface Requirements 
for Nuclear Power Reactors'' of the new Power Reactor Security Rule (74 
FR 13926; March 27, 2009), requires licensees to communicate plans for 
proposed plant changes that could impact plant security to security 
personnel who are qualified to analyze and identify potentially adverse 
impacts that the changes may have on safety and/or security programs. 
After security personnel analyze the changes for potential impacts, the 
regulation requires the licensee to take appropriate actions to 
mitigate the security impacts.
    NRC Topic 9. Given the potential impact to the licensee (because 
the backfit rule would not apply) of the NRC's periodic re-evaluation 
of estimated LOCA frequencies which could cause the NRC to increase the 
TBS, should the proposed rule require licensees to maintain the 
capability to bring the plant into compliance with an increased 
transition break size (TBS), within a reasonable period of time?
    Comments. NEI, the BWROG, and the WOG commented that licensees 
should be provided with a great deal of latitude on achieving 
compliance following any change in the TBS, with the goal being that 
risk requirements are achieved with a reasonable mix of prevention and 
mitigation.
    NRC response. The NRC agrees with commenters that the Sec.  50.46a 
rule should provide licensees with substantial flexibility to determine 
how they will come back into compliance with applicable regulatory 
requirements following any future change in the TBS. Licensees who must 
take actions to come back into compliance need not return the plant to 
the precise conditions and circumstances in effect immediately before 
implementation of Sec.  50.46a. Rather, licensees would be afforded the 
flexibility of deciding what actions they will implement to bring about 
compliance under any revised TBS. Further, as one of the commenters 
suggests, the overall goal of any actions taken to restore compliance 
is to achieve a reasonable mix of prevention and mitigation.
    NRC Topic 10. Is the proposed rule sufficiently clear as to be 
``inspectable?'' That is, does the rule language lend itself to timely 
and objective NRC conclusions regarding whether or not a licensee is in 
compliance with the rule, given all the facts? In particular, are the 
proposed requirements for PRA quality sufficient in this regard?
    Comment. On the question of whether the proposed rule is clear 
enough to be inspectable, NEI was particularly concerned that the 
operational restrictions would conflict with the existing technical 
specifications. The BWROG supported the NEI position on this topic.
    NRC response. To reduce potential conflict between plant technical 
specifications and the operability requirements in Sec.  50.46a, the 
NRC has also modified operability requirements to allow limited 
operation (for no more than a total of fourteen days in any 12-month 
period) in configurations where mitigation of LOCAs larger that the TBS 
has not been demonstrated. A detailed discussion on the basis for this 
new provision is provided below in Section V.F of this document, 
Operational Requirements.
    Comment. NEI stated that the rule would be difficult to inspect 
because it overlaps so many existing regulatory requirements. The WOG 
stated that the risk-informed aspects of the proposed rule, including 
the PRA quality requirements, should rely on the guidance of RG 1.174 
and RG 1.200. The WOG stated that proposed Sec.  50.46a should require 
no more ``inspectability'' than any other performance-based risk-
informed application. Another commenter stated that the NRC should 
clarify certain aspects of the proposed rule and that the rule 
appropriately includes language like ``reasonable balance'' that 
requires a knowledgeable individual to exercise judgment which should 
be informed by appropriate regulatory guidance documents.
    NRC response. The NRC has modified the proposed rule to provide 
greater operational flexibility and reduce the potential for conflict 
with plant technical specification requirements that might cause 
``inspectability'' problems. Although the WOG stated that the proposed 
rule would not have inspectability problems if it relied on the 
guidance in RG 1.174 and RG 1.200, the NRC notes that inspectors may 
not inspect licensees for compliance with regulatory guides because 
these guides are not regulatory requirements. The NRC has incorporated 
the important aspects of RG 1.174 and PRA quality guidance into the 
revised proposed rule itself so that inspectors would have a clear 
indication of the Sec.  50.46a requirements. Specific inspection 
guidance will be developed as necessary after the final rule is 
published.
    NRC Topic 11. Proposed Sec.  50.46a would impose no limitations on 
``bundling'' of different facility changes together in a single 
application. Facility changes which would increase plant risk 
substantially or create risk outliers could be grouped with other 
facility changes which would reduce risk so that the net change would 
meet the risk acceptance criteria. Are the net change in risk 
acceptance criteria in the proposed rule adequate or should some 
additional limitations be imposed to avoid allowing facility changes 
which are known to increase plant risk?
    Comments. Several commenters said that ``bundling'' is essential 
for meeting the objectives of this proposed rule which concerns overall 
plant risk. Bundling provides licensee management with the necessary 
flexibility to reallocate resources for implementation of the 
alternative requirements. The RG 1.174 criteria related to bundling 
(combined change request in RG 1.174) are sufficient and no additional 
criteria or restrictions on bundling should be imposed by this proposed 
rule.
    NRC response. The NRC agrees that bundling of facility changes is 
desirable because it appropriately permits licensees to credit risk 
beneficial facility changes and encourages licensees to identify and 
implement facility changes

[[Page 40024]]

that decrease risk. The NRC also agrees that the guidelines on combined 
changes in RG 1.174 are sufficient to avoid facility changes which 
would unacceptably increase plant risk.
    NRC Topic 12. Is there an alternative to tracking the cumulative 
risk increases associated with facility changes made after implementing 
Sec.  50.46a that is sufficient to provide reasonable assurance of 
protection to public health and safety and common defense and security?
    Comments. Four of the commenters who responded to the question 
stated that tracking cumulative risk increases was reasonable but they 
appeared to define cumulative tracking differently than as specified in 
the requirements of the proposed rule. NEI, whose comments were 
generally endorsed by most of the 12 commenters, recommended rule text 
stating ``[t]he licensee shall periodically assess the cumulative 
effect of changes to the plant design configuration and update as 
necessary, the PRA and other risk analyses.'' After discussing this 
proposed text at the June 28, 2006, public meeting, the NRC determined 
that the recommendation equated tracking cumulative risk increases with 
periodically updating the PRA and estimating the latest core damage 
frequency (CDF) and large early release frequency (LERF) using the 
updated PRA. NEI intended for these latest risk estimates themselves to 
represent the assessment of the cumulative increase. However, the 
proposed rule required that some previous estimates of CDF and LERF be 
subtracted from the latest estimates to obtain the amount by which the 
CDF and LERF has increased. One of the four commenters added that 
tracking the cumulative risk increase (as intended by the NRC in the 
proposed rule) was not necessary because the threshold for risk 
increase is low enough so that the cumulative effect is not 
significant. A fifth commenter argued that tracking cumulative risk 
should not be required by the rule because compliance with the guidance 
in RG 1.174 should be sufficient to ensure that cumulative risk does 
not impact the health and safety of the public.
    NRC response. The NRC has retained the requirement to track the 
total risk increases in CDF and LERF made under the proposed rule and 
has retained the definition of risk ``increase'' as being the amount by 
which risk increases. RG 1.174 provides guidance on judging the 
acceptability of proposed facility changes based primarily on the 
amount by which the facility changes increase CDF and LERF. The NRC has 
clarified what it has concluded must be tracked in Sec.  
50.46a(f)(2)(iv) utilizing the requirement for tracking the cumulative 
effect on risk of changes made under the NFPA-805 standard which was 
incorporated by reference into Sec.  50.48(c) (see, 69 FR 33536; June 
16, 2004). By utilizing the same language in both rules, the NRC 
intends that the implementation of both rules would be consistent.
    The NRC has concluded that the alternative proposed by the 
commenters (i.e. to track cumulative risk by simply updating the PRA) 
is not acceptable because the latest estimates of CDF and LERF alone 
provide insufficient information to be used in the risk-informed 
framework contained in RG 1.174. Two other commenters argued that risk 
tracking is not needed because controls external to proposed Sec.  
50.46a (e.g., in RG 1.174) would ensure that the cumulative effect 
would not be significant. The commenters provided no basis for their 
assertions that controls external to the rule would keep increases in 
risk small enough to ensure protection of public health and safety. RG 
1.174 does discuss tracking changes in cumulative risk, but regulatory 
guides are not enforceable requirements. The NRC has determined that it 
is necessary to establish a regulatory requirement to track the 
cumulative risk increases from all changes made under this proposed 
rule. The NRC continues to believe that risk tracking as described in 
the proposed rule is needed to ensure that facility changes permitted 
by the revised ECCS analyses under Sec.  50.46a do not result in 
greater increases in risk than were intended by the Commission.
    NRC Topic 13. The NRC requested specific public comments on the 
acceptability of applying the change in risk acceptance guidelines in 
RG 1.174 to the total cumulative change in risk from all changes in the 
plant after adoption of Sec.  50.46a. Should other risk guidelines be 
used and, if so, what guidelines should be used?
    Comments. As discussed, four commenters proposed tracking 
cumulative risk increases by periodically updating the PRA, estimating 
the latest CDF and LERF using the updated PRA, and equating these 
latest estimates with tracking the cumulative risk increase. Applying 
this definition for tracking cumulative risk increase, these commenters 
concluded that the change in risk acceptance guidelines should not be 
applied to the total cumulative change in risk which would not, under 
their proposals, be estimated.
    In general, most commenters' either explicitly or implicitly 
recommended that the rule should not include the acceptance criteria 
that ``the total increases in CDF and LERF should be small and the 
overall risk should remain small.'' Proposals for alternatives varied. 
NEI's proposed rule text did not include acceptance criteria related to 
increases in CDF and LERF. Instead, NEI proposed requiring the licensee 
to report the results of the updated PRA and other risk analyses to the 
NRC. One commenter argued that for facility changes enabled by the new 
Sec.  50.46a, compliance with RG 1.174 should be sufficient. Two 
commenters stated that risk tracking accomplished by updating the PRA 
and estimating the latest CDF and LERF can be used to ensure that the 
total risk as well as the risk from specific initiators or classes of 
accidents is not increasing.
    NRC response. The NRC has retained the requirement in the revised 
proposed rule that the total change in risk from facility changes, 
measured as the amount by which CDF and LERF (or LRF for new reactors) 
increase, be tracked and compared to the RG 1.174 acceptance criteria. 
However, the NRC has reduced the scope of facility changes that must be 
tracked from all changes to only those changes made to the plant under 
Sec.  50.46a. Implementation of all RG 1.174 guidelines can only be 
achieved using a process that includes an estimate of the cumulative 
change in risk. Also, consistent with the Commission's direction in the 
SRM for SECY-07-0082, the NRC has reduced the size of an acceptable 
risk increase from ``small'' to ``very small''. The revised proposed 
rule would continue to use the quantitative guidelines in RG 1.174.
    NEI's proposal for reporting the latest estimates of CDF and LERF 
to the NRC after each periodic assessment would not be useful because 
the NRC has no criteria for determining which CDF and LERF values would 
be acceptable. It would be a lengthy process to establish such 
acceptance criteria. Lack of acceptance criteria against which the 
latest CDF and LERF can be compared will result in different 
stakeholders applying different criteria to judge the acceptability of 
the results most likely leading to different conclusions.
    The NRC believes that the two comments proposing that the total CDF 
as well as the CDF from specific initiators or class of accidents could 
be tracked to ensure that risk from these scenarios is not increasing 
would satisfy the requirement that the total increase in risk remains 
very small provided that the appropriate initiators or class of 
accident is identified (and including LERF or LRF). The commenters did 
not

[[Page 40025]]

appear to be proposing that such a constraint be included in the rule, 
instead they were only making observations on what would be possible. 
Nevertheless, in an SRM on August 10, 2007, the Commission concluded 
that only a very small increase in risk is acceptable when implemented 
according to the requirements in this rule. Requiring that there be no 
risk increase, as hypothesized by the commenters, is more restrictive 
than the criteria in the revised proposed rule.
    Although the revised proposed rule would permit licensees to make 
plant changes that result in very small risk increases, the NRC 
requests stakeholder comments on whether any increase in risk should be 
allowed. Instead of the risk acceptance criteria allowing very small 
risk increases, should the acceptance criteria in the final rule 
require that the net effect of plant changes made under Sec.  50.46a be 
risk neutral or risk beneficial? The NRC requests stakeholders to 
provide comments on the use of risk acceptance criteria that would not 
allow a cumulative increase in risk for plant changes made under Sec.  
50.46a.
    NRC Topic 14. After approval to implement Sec.  50.46a, the 
proposed rule would require tracking risk associated with all proposed 
facility changes but would not require a licensee to include risk 
increases caused by previous risk-informed facility changes that were 
implemented before Sec.  50.46a was adopted. Licensees who adopt Sec.  
50.46a before implementing other risk-informed applications would have 
a smaller risk increase ``available'' compared to licensees who have 
already incorporated some risk-informed facility changes into their 
overall plant risk before adopting Sec.  50.46a. The NRC requests 
specific public comments on whether this potential inconsistency should 
be addressed and, if so, how?
    Comments. Three commenters stated that these potential 
inconsistencies in acceptable risk increases should be addressed by 
deleting the requirement that the cumulative risk increase be tracked 
and compared to the RG 1.174 acceptance guidelines. The commenters 
argued that licensees and the NRC have effectively managed incremental 
risk without the need for this structure and that any facility changes 
that seek to apply the revised design bases should be evaluated using 
the same methods proven effective in the past. A fourth commenter 
agreed with the others but proposed that inconsistencies among 
licensees created by the order of implementing risk-informed 
applications could be resolved by allowing a licensee to reestablish 
the baseline and removing some facility changes from tracking.
    NRC response. The NRC is proposing additional changes in the 
revised proposed rule that would make this topic moot. The proposed 
rule would have required tracking total risk from all facility changes. 
This requirement reflected a difficulty uniquely associated with 
comparing the total risk increases from all facility changes to the 
acceptance criteria. The revised proposed rule would only require that 
facility changes made under the rule be tracked. Other risk-informed 
facility changes referred to in Topic 14 would no longer be included in 
this change in risk estimate and therefore, the acceptability of those 
facility changes will be independent of facility changes made under 
this rule (aside from the indirect affect these facility changes have 
on the plant's risk profile).
    NRC Topic 15. Proposed Sec.  50.46a would require licensees to 
report every 24 months all ``minimal'' risk facility changes made under 
Sec.  50.46a(f)(1) without NRC review. Are there less burdensome or 
more effective ways of ensuring that the cumulative impact of an 
unbounded number of ``minimal'' changes remains inconsequential?
    Comments. Several commenters stated that the Sec.  50.46a(g)(3) 
report summarizing minimal risk changes every 24 months is redundant to 
reports required under Sec.  50.59(d)(2) as well as Sec.  50.71(e). 
Thus, Sec.  50.46a(g)(3) should be deleted. The requirement needlessly 
focuses licensee and NRC resources directly on a large set of 
information that by its very definition has no safety or risk 
significance.
    NRC response. The NRC agrees with the commenters that the reporting 
requirements in proposed Sec.  50.46a(g)(3) could be redundant to other 
reporting requirements for some facility changes because some changes 
made under the new rule might be reportable under both Sec.  50.59 and 
Sec.  50.46a(g)(3). The NRC has determined that breaks larger than the 
TBS should be removed from the design basis event category. Therefore, 
the NRC believes that some facility changes that may be made under the 
new rule would no longer be reportable under Sec.  50.59 because the 
change would no longer affect design basis events. The NRC is proposing 
to reduce the scope of facility changes that need to be evaluated under 
the new provision, from all changes made to the facility after adoption 
of the rule to only facility changes that are made under the new rule. 
This change would reduce the number of potentially redundant reports.
    To avoid the possibility that potentially risk-significant changes 
are not reported, the NRC has concluded that all facility changes made 
under the new rule should be reported because the NRC will rely on the 
risk evaluation to prevent facility changes that might not be 
protective of public health and safety. Therefore, the NRC has retained 
the reporting requirements in Sec.  50.46a(g)(3) because these 
requirements would ensure the reporting of all potentially risk-
significant facility changes made under the proposed rule.
    NRC Topic 16. Should the Sec.  50.46a rule itself include high-
level criteria and requirements for the risk evaluation process and 
acceptance criteria described in RG 1.174? If these criteria were 
included in the regulatory guide only, and not in Sec.  50.46a, how 
could the NRC take enforcement action for licensees who failed to meet 
the acceptance criteria?
    Comments. Four commenters stated that proposed Sec.  50.46a rule 
should not contain the high-level criteria and requirements for the 
risk evaluation process and acceptance criteria described in RG 1.174. 
These commenters did not specifically propose how the NRC could take 
enforcement action to ensure compliance with the criteria, but instead 
asserted that regulatory guidance documents and inspection guidelines 
are the appropriate places for the risk acceptance criteria.
    NRC response. The NRC does not agree with the commenters. The 
proposed rule would have to contain high-level requirements for the 
risk evaluation and acceptance criteria to establish the legally 
enforceable alternative regulatory requirements needed to ensure 
adequate protection of public health and safety in a manner which 
maximizes regulatory predictability and stability. The NRC believes 
that proposed Sec.  50.46a should build upon NRC and industry 
experience with the key principles of risk-informed decision making set 
forth in RG 1.174, but notes that RG 1.174 only contains guidance, not 
requirements. To be enforceable, proposed Sec.  50.46a must contain and 
does contain high-level requirements relating to risk, defense-in-
depth, safety margins, risk, and performance measurement. Specific, 
detailed guidance on how to meet the high-level requirements will be 
set forth in regulatory guidance and inspection guidelines, as 
appropriate.

[[Page 40026]]

V. Revised Proposed Rule

A. Overview

    The NRC's revised proposed rule would establish an alternative set 
of risk-informed requirements with which licensees may choose to comply 
in lieu of meeting the current emergency core cooling system 
requirements in 10 CFR 50.46. Using the alternative ECCS requirements 
would provide some licensees with opportunities to change other aspects 
of facility design.
    As was the case in the initial proposed rule, the revised proposed 
rule divides the current spectrum of LOCA break sizes into two regions. 
The division between the two regions is delineated by the TBS. The 
first region includes small size breaks up to and including the TBS. 
The second region includes breaks larger than the TBS up to and 
including the DEGB of the largest RCS pipe. Break area for the TBS is 
not based on a double-ended offset break. Rather, it is based on the 
inside area of a single-sided circular pipe break. Pipe breaks in the 
smaller break size region are considered more likely than pipe breaks 
in the larger break size region. Consequently, each break size region 
will be subject to different ECCS requirements, commensurate with 
likelihood of the break. LOCAs in the smaller break size region must be 
analyzed by the same conservative methods, assumptions, and criteria 
currently used for LOCA analysis. Accidents in the larger break size 
region may be analyzed using more realistic methods and assumptions 
based on their lower likelihood. Although LOCAs for break sizes larger 
than the transition break would become ``beyond design-basis 
accidents,'' the revised proposed rule would require that licensees 
maintain the ability to mitigate all LOCAs up to and including the DEGB 
of the largest RCS pipe. However, mitigation analyses for LOCAs larger 
than the TBS need not assume the loss-of-offsite power or the 
occurrence of a single failure.
    Licensees who perform LOCA analyses using the risk-informed 
alternative requirements may find that their plant designs are no 
longer limited by certain parameters associated with previous DEGB 
analyses. Reducing the DEGB limitations could enable licensees to 
propose a wide scope of design or operational changes up to the point 
of being limited by some other parameter associated with any of the 
other required accident analyses. Potential design changes include 
modification of containment spray designs, modifying core peaking 
factors, modifying setpoints on accumulators or removing some from 
service, eliminating fast starting of one or more emergency diesel 
generators, and increasing power, etc. Some of these design and 
operational changes could increase plant safety because a licensee 
could modify its systems to better mitigate the more likely LOCAs. 
Other changes, such as increasing power, could increase overall risk to 
the public. The risk-informed Sec.  50.46a option would include risk 
acceptance criteria for evaluating future design changes to ensure that 
any risk increases are acceptably small. These acceptance criteria 
would be consistent with the guidelines for risk-informed license 
amendments in RG 1.174 and would ensure both the acceptability of the 
changes from a risk perspective and the maintenance of sufficient 
defense-in-depth, safety margins, and performance monitoring. The 
requirements for the risk-informed evaluation process are discussed in 
detail in Section V.E of this document.
    The NRC will periodically evaluate LOCA frequency information. 
Should estimated LOCA frequencies increase causing a significant 
increase in the risk associated with breaks larger than the TBS, the 
NRC would undertake rulemaking (or issue orders, if appropriate) to 
change the TBS. In such a case, the backfit rule (10 CFR 50.109) will 
not apply. If previous plant changes are invalidated because of a 
change to the TBS, licensees would have to modify or restore components 
or systems as necessary so that the facility would continue to comply 
with Sec.  50.46a acceptance criteria. The backfit rule (10 CFR 50.109) 
also would not apply in these cases.
    Changes consist of a new Sec.  50.46a and conforming changes to 
existing Sec. Sec.  50.34, 50.46, 50.46a (redesignated as Sec.  
50.46b), 50.109, 10 CFR Part 50, Appendix A, General Design Criteria 
17, 35, 38, 41, 44 and 50, and Sec. Sec.  52.47, 52.79, 52.137, and 
52.157.

B. Determination of the Transition Break Size

    To help establish the TBS, the NRC developed pipe break frequencies 
as a function of break size using an expert opinion elicitation process 
for degradation-related pipe breaks in typical BWR and PWR reactor 
coolant systems (NUREG-1829; ``Estimating Loss-of-Coolant Accident 
(LOCA) Frequencies through the Elicitation Process'' March 2008; 
ML082250436). The elicitation process is used for quantifying 
phenomenological knowledge when data or modeling approaches are 
insufficient. The elicitation focused solely on determining event 
frequencies that initiate unisolable primary system side failures 
related to material degradation.
    A baseline TBS was established from the expert elicitation results 
for each reactor type (i.e., PWR and BWR) that corresponded to a break 
frequency of once per 100,000 reactor years (1 x 10<SUP>-</SUP>\5\ or 
10<SUP>-</SUP>\5\ per reactor year). The NRC then considered 
uncertainty in the elicitation process, other potential mechanisms that 
could cause passive component failure that were not explicitly 
considered in the expert elicitation process, and the higher 
susceptibility to rupture/failure of specific locations in the reactor 
coolant system (RCS); adjusting the TBS upwards to account for these 
factors. Other mechanisms that contribute to the overall LOCA frequency 
include LOCAs resulting from failures of non-passive components and 
LOCAs resulting from low probability events (earthquakes of magnitude 
larger than the safe shutdown earthquake and dropped heavy loads). 
These LOCAs have a strong dependency on plant-specific factors.
    LOCAs caused by failure of non-passive components, such as stuck-
open valves and blown out seals or gaskets have a greater frequency of 
occurrence than LOCAs resulting from the failure of passive components. 
LOCAs resulting from the failure of non-passive components would be 
small-break LOCAs, when considering the size of the opening that could 
result should components fail open or blow out (e.g., safety valves, 
pump seals). LOCAs resulting from stuck-open valves are limited by the 
size of the auxiliary pipe. In some PWRs, there are large loop 
isolation valves in the hot and cold leg piping. However, a complete 
failure of the valve stem packing is not expected to result in a large 
flow area, because the valves are back-seated in the open 
configuration. Based on these considerations, non-passive LOCAs are 
relatively small in size and are bounded by the selected TBS.
    LOCAs could also be caused by dropping heavy loads that could cause 
a breach of the RCS piping. During power operation, personnel entry 
into the containment is typically infrequent and of short duration. The 
lifting of heavy loads that if dropped would have the potential to 
cause a LOCA or damage safety-related equipment is typically performed 
while the plant is shutdown. The majority of heavy loads are lifted 
during refueling evolutions when the primary system is depressurized, 
further reducing the risk of a LOCA and a loss of core cooling. If 
loads are lifted during power operation, they would not be loads 
similar to the heavy loads lifted during plant

[[Page 40027]]

shutdown, e.g., vessel heads and reactor internals. In addition, the 
RCS is inherently protected by surrounding concrete walls, floors, 
missile shields, and biological shielding. Thus, the contribution of 
heavy load drops to overall LOCA frequency is not considered to be 
significant and would not affect the TBS.
    Seismically-induced LOCA break frequencies can vary greatly from 
plant to plant because of factors such as site seismicity, seismic 
design considerations, and plant-specific layout and spatial 
configurations. Seismic break frequencies are also affected by the 
amount of pipe degradation occurring prior to postulated seismic 
events. Seismic PRA insights have been accumulated from the NRC Seismic 
Safety Margins Research Program and the Individual Plant Examination of 
External Events submittals. Based on these studies, piping and other 
passive RCS components generally exhibit high seismic capacities and, 
therefore, are not significant risk contributors. However, these 
studies did not explicitly consider the effect of degraded component 
performance on the risk contributions. Therefore, the NRC conducted a 
study to evaluate the seismic performance of undegraded and degraded 
passive system components (NUREG-1903, ``Seismic Considerations for the 
Transition Break Size,'' February 2008; ML080880140). This effort 
examined operating experience, seismic PRA insights, and models to 
evaluate the failure likelihood of undegraded and degraded piping. The 
operating experience review considered passive component failures that 
have occurred as a result of strong motion earthquakes in nuclear and 
fossil power plants as well as other industrial facilities. No 
catastrophic failures of large pipes resulting from earthquakes between 
0.2g and 0.5g peak ground acceleration have occurred in power plants. 
However, piping degradation could increase the LOCA frequency 
associated with seismically-induced piping failures. The NUREG-1903 
report evaluated seismic loadings on degraded piping and concluded that 
a very large, pre-existing crack on the order of 30 percent through-
wall and 145 degrees around the piping circumference would have to be 
present during a 10<SUP>-</SUP>\5\ or 10<SUP>-</SUP>\6\ per year 
earthquake in order for pipe failure to occur. The NRC concluded that 
the likelihood of flaws large enough to fail during a seismic event is 
sufficiently low that the TBS need not be modified to address 
seismically-induced direct piping failures. In reaching its conclusion, 
the NRC considered the comments received as well as historical 
information related to piping degradation and the potential for the 
presence of cracks sufficiently large that pipe failure would be 
expected under loads associated with rare (10<SUP>-</SUP>\5\ per year) 
earthquakes.
    Indirect failures are primary system ruptures that are a 
consequence of failures in nonprimary system components or structural 
support failures (such as a steam generator support). Structural 
support failures could then cause displacements in components that 
stress and in turn, fail the piping. The NRC performed studies on two 
plants to estimate the conditional pipe failure probability due to 
structural support failure given a low return frequency earthquake 
(10<SUP>-</SUP>\5\ to 10<SUP>-</SUP>\6\ per year). The results 
indicated that the conditional probability was on the order of 0.1. 
These studies used seismic hazard curves from NUREG-1488 (NUREG-1488, 
``Revised Livermore Seismic Hazard Estimates for Sixty-Nine Nuclear 
Power Plant Sites East of the Rocky Mountains, April 1994; 
ML052640591). More recent studies were completed by EPRI on three 
plants using updated seismic hazard estimates. The updated seismic 
hazard increases the peak ground acceleration at some sites. The 
highest pipe failure probability calculated for the three plants in the 
industry analyses was 6 x 10<SUP>-</SUP>\6\ per year. The NRC noted in 
its report that indirect failure analyses are highly plant-specific. 
Therefore, it is possible that example plants assessed in the NRC and 
EPRI analyses are not limiting for all plants.
    The NRC has considered the importance of indirect failures on the 
selection of the TBS. For the cases considered in both the EPRI and NRC 
studies, the likelihood of indirectly induced piping failures resulting 
from major component support failures is less than 10<SUP>-</SUP>\5\ 
per reactor year, the frequency criterion used to select the TBS. Also, 
as noted in the public comments, the median seismic capacities for both 
the primary piping system and primary system components are typically 
higher than other safety related components within the nuclear power 
plant. Because of these relative capacities, it is expected that a 
seismic event of sufficient magnitude to cause consequential failure 
within the primary system would also induce failure of components in 
multiple trains of mitigation systems, or even induce multiple RCS pipe 
breaks. Consequently, the risk contribution from seismically induced 
indirect failures is expected to depend more heavily on the relative 
fragilities of plant components and systems than the size of the TBS. 
Therefore, the NRC believes that adjustment to the TBS for seismically 
induced indirect LOCAs is also not warranted.
    The final consideration in selecting the TBS was actual piping 
system design (e.g., sizes) and operating experience. For example, due 
to configuration and operating environment, certain piping is 
considered to be more susceptible than other piping in the same size 
range. For PWRs, the range of pipe break sizes determined from the 
various aggregations of expert opinion was 6 to 10 inches in diameter 
(i.e., inside dimension) for the 95th percentile. This is only slightly 
smaller than the PWR surge lines, which are attached to the RCS main 
loop piping and are typically 12- to 14-inch diameter Schedule 160 
piping (i.e., 10.1 to 11.2 inch inside diameter piping). The RCS main 
loop piping is in the range of 30 inches in diameter and has 
substantially thicker walls than the surge lines. The expert 
elicitation panel concluded that this main loop piping is much less 
likely to break than other RCS piping. The shutdown cooling lines and 
safety injection lines may also be 12- to 14-inch diameter Schedule 160 
piping and are likewise connected to the RCS. The difference in 
diameter and thickness of the reactor coolant piping and the piping 
connected to it forms a reasonable line of demarcation to define the 
TBS. Therefore, to capture the surge, shutdown cooling, and safety 
injection lines in the range of piping considered to be equal to or 
less than the TBS, the NRC specified the TBS for PWRs as the cross-
sectional flow area of the largest piping attached to the RCS main 
loop.
    For BWRs, the arithmetic and geometric means of the break sizes 
having approximately a 95th percentile probability of 10<SUP>-</SUP>\5\ 
per year ranged from values of approximately 13 inches to 20 inches 
equivalent diameter. The information gathered from the elicitation for 
BWRs showed that the estimated frequency of pipe breaks dropped 
markedly for break sizes beyond the range of approximately 18 to 20 
inches. After evaluating BWR designs, it was determined that typical 
residual heat removal piping connected to the recirculation loop piping 
and feedwater piping is about 18 to 24 inches in diameter. These pipe 
sizes are consistent with break sizes beyond which the pipe break 
frequency is expected to decrease markedly below 10<SUP>-</SUP>\5\ per 
year. It was also recognized that the sizes of attached pipes vary 
somewhat among plants. Thus, for

[[Page 40028]]

BWRs, the TBS is specified as the cross-sectional flow area of the 
larger of either the feedwater or the RHR piping inside primary 
containment.
    Because the effects of TBS breaks on core cooling vary with the 
break location, the NRC evaluated whether the frequency of TBS breaks 
varies with location and whether TBS breaks should, therefore, vary in 
size with location. In PWRs, the pressurizer surge line is only 
connected to one hot leg and the pipes attached to the cold legs are 
generally smaller than the surge line. The cold legs (including the 
intermediate legs) operate at slightly cooler temperatures. Thermally-
activated degradation mechanisms would be expected to progress more 
slowly in the cold leg than in the hot leg. Therefore, the NRC 
evaluated whether it may be appropriate to specify a TBS for the cold 
leg that would be smaller than the size of the surge line. The 
frequency of occurrence of a break of a given size is composed of both 
the frequency of a completely severed pipe of that size (a complete 
circumferential break) plus the frequency of a partial break of that 
size in an equal or larger size pipe (a partial circumferential or 
longitudinal break). Therefore, the NRC evaluated an option where the 
TBS for the hot and cold legs would be distinctly different and would 
be composed of two components: (1) Complete breaks of the pipes 
attached to the hot or cold legs at the limiting locations within each 
attached pipe, and (2) partial breaks of a constant size, as 
appropriate for either the hot or cold leg, at the limiting locations 
within the hot or cold legs. The NRC attempted to estimate the 
appropriate size of the partial break component for the TBS by 
reviewing the expert elicitation results to determine the frequencies 
of occurrence of partial breaks within hot and cold legs that would be 
equivalent to the frequency of a complete surge line break. The NRC 
found that frequencies of occurrence of partial breaks of a given size 
are generally lower for the cold leg than for the hot leg. However, 
other than this general trend, the elicitation results do not contain 
sufficient information to adequately quantify differences among the hot 
leg, cold leg, and surge line pipe break frequencies. Because it was 
not possible to establish a smaller partial break TBS criterion in the 
hot or cold legs, the NRC concluded that the TBS associated with 
partial breaks in the hot and cold legs should remain equivalent in 
size to the internal cross sectional area of the surge line. Similarly, 
the elicitation results do not contain sufficient detail to quantify 
break frequency differences among the BWR recirculation, residual heat 
removal, and feedwater system piping. Thus, a smaller partial break TBS 
criterion also could not be established for BWR recirculation piping.
    The NRC also evaluated whether TBS breaks should be analyzed as 
single-ended or double-ended breaks. To address this issue, the NRC 
reviewed the expert elicitation process and the guidance given to the 
experts in developing their frequency estimates. The NRC concluded that 
the expert elicitation LOCA frequency estimates correspond to a break 
area having an equivalent circular diameter at each break size. This 
correspondence is representative of a single-ended break. Additionally, 
the experts based their estimates on knowledge of postulated failure 
mechanisms in pressure boundary components and not on the flow rates 
emanating from the breaks. The flow rates are governed by the break 
location and system configuration which determines whether reactor 
coolant will be discharged from both ends of the break.
    The current design basis analysis for light water reactors requires 
analysis of a DEGB of the largest pipe in the RCS. Under the proposed 
rule, all breaks up to and including the TBS would be analyzed under 
existing requirements. A possible reason for specifying the TBS for 
PWRs as double-ended could be that a complete break of the pressurizer 
surge line would result in reactor coolant exiting both ends of the 
break. Although this occurs initially during a LOCA, core cooling 
requirements are dominated by the flow rate of coolant exiting from the 
hot leg side of the break, with much less contribution from the flow 
rate of coolant exiting from the pressurizer side. Therefore, 
specifying the TBS break as an area equivalent to a double-ended break 
of the surge line would be overly conservative. For BWRs, the effect of 
a double-ended break area is also considered to be overly conservative. 
The selected TBS for BWRs is based on the larger of the residual heat 
removal or main feedwater lines attached to the main recirculation 
piping. A single-ended break in these lines would bound double-ended 
breaks of the smaller lines in the reactor recirculation and feedwater 
system. Therefore, the NRC concluded that treating the TBS as a single-
ended break reasonably characterizes the expert elicitation results and 
represents the flow rates associated with postulated pipe breaks within 
the RCS.
    For the TBS to remain valid at a particular facility, future plant 
modifications must not significantly increase the LOCA pipe break 
frequency estimates generated during the expert elicitation and used as 
the basis for the TBS. For example, the expert elicitation panel did 
not consider the effects of power uprates in deriving the break 
frequency estimates. The expert elicitation panel assumed that future 
plant operating characteristics would remain consistent with past 
operating practices. The NRC recognizes that significant plant changes 
may change plant performance and relevant operating characteristics to 
a degree that they might impact future LOCA frequencies. The NRC will 
expect applicants for plant changes under revised proposed Sec.  50.46a 
to demonstrate that those changes do not significantly increase break 
frequencies. As discussed in Section V.C. of this document, the NRC is 
currently preparing guidance for applicants to use to demonstrate that 
proposed plant changes do not undermine the Sec.  50.46a technical 
basis (``Plant-Specific Applicability of 10 CFR 50.46 Technical Basis'' 
February 2009; ML090350757).
    The baseline TBS was adjusted upward to account for uncertainties 
and failure mechanisms leading to pipe rupture that were not considered 
in the expert elicitation process. As the NRC obtains additional 
information that may tend to reduce those uncertainties or allow for 
more structured consideration of degradation mechanisms, the NRC will 
assess whether the TBS (as defined in Sec.  50.46a) should be adjusted, 
and may initiate rulemaking to revise the TBS definition to account for 
this new information. The NRC will also continue to assess the failure 
precursors that might be indicative of an increase in pipe break 
frequencies in BWR and PWR plants to establish whether the TBS would 
need to be adjusted.
    However, these TBS values are within the range supported by the 
expert elicitation estimates when considering the uncertainty inherent 
in processing the degradation-related frequency estimates. In addition, 
the NRC believes that the TBS definitions in the proposed rule would 
provide necessary conservatism to compensate for possible future 
increases in break frequencies. The NRC expects that the TBS values 
would result in regulatory stability because future LOCA frequency 
reevaluations are less likely to make it necessary for the NRC to 
change the TBS and cause licensees to undo plant modifications made 
after implementing Sec.  50.46a.

[[Page 40029]]

C. Evaluation of the Plant-Specific Applicability of the Transition 
Break Size

    As discussed in Section V.B. of this document, the NRC has 
published two reports, NUREG-1829 (ML082250436), and NUREG-1903 
(ML080880140) that form part of the technical basis used to select the 
TBS for BWR and PWR plants. NUREG-1829 used expert elicitation to 
develop generic LOCA frequency estimates of passive system failure as a 
function of break size for both BWR and PWR plants and considered 
normal operational loading and transients expected over a 60-year plant 
life. NUREG-1903 assessed the likelihood that rare seismic events would 
induce primary system failures larger than the postulated TBS. NUREG-
1903 evaluated both direct failures of flawed and unflawed primary 
system pressure boundary components and indirect failures of nonprimary 
system components and supports that could lead to primary system 
failures. Because these studies were not intended to develop bounding 
estimates, unique plant attributes may result in plant-specific LOCA 
frequencies due to normal operational and/or seismic loading that are 
greater than reported in either NUREG-1829 or NUREG-1903. Consequently, 
the NRC has included a requirement that applicants wishing to implement 
Sec.  50.46a conduct an evaluation to demonstrate that the results in 
NUREG-1829 and NUREG-1903 are applicable to their individual plants.
    The NRC is preparing guidance for conducting the plant specific 
review to demonstrate the applicability of both the NUREG-1829 and 
NUREG-1903 results. The scope of this applicability guidance would be 
limited to primary system piping and other primary pressure boundary 
components that are large enough to result in LOCA break sizes larger 
than the TBS. This guidance is applicable to aspects of the facility 
design affecting compliance with ECCS requirements and would not 
pertain to design-bases or operational procedures associated with other 
aspects of the facility licensing basis.
    The plant applicability evaluation would require that Sec.  50.46a 
applicants first demonstrate that the applicable systems in the plant 
adhere to the current licensing basis. Additionally, the evaluation 
would require that licensees consider the effects of unique, plant-
specific attributes on the generic LOCA frequencies developed in NUREG-
1829. The licensee would also evaluate the effect of proposed plant 
changes on both direct and indirect system failures to demonstrate that 
NUREG-1829 results remain applicable after the proposed changes have 
been implemented. After a licensee is approved to implement revised 
proposed Sec.  50.46a requirements, it would also be necessary to 
evaluate the effect of future proposed plant changes to demonstrate 
that NUREG-1829 results remain applicable after enacting the proposed 
changes.
    An evaluation framework is also provided for determining the 
applicability of the NUREG-1903 assessment of direct piping failures. 
This framework identifies the aspects that applicants would consider in 
a plant-specific analysis, provides several options for conducting the 
analysis, and describes a systematic approach associated with each 
option. One important step is to determine whether the NUREG-1903 
results can be used directly or if a plant-specific analysis is 
required to determine the limiting flaw sizes under rare seismic 
loading. NUREG-1903 also addressed indirect piping failures caused by 
rare seismic loading. However, the risk of indirect failure is highly 
plant-specific and NUREG-1903 only considered the risks associated with 
two different plants. Consequently, the limited analysis of indirect 
piping failures does not provide a sufficient technical basis for 
allowing generic changes to the seismic design, testing, analysis, 
qualification, and maintenance requirements associated with any 
component under Sec.  50.46a. Any proposed changes to these criteria 
would be justified using a plant-specific analysis to assess the change 
in risk associated with seismically induced failures of the relevant 
component and/or system that results from the proposed plant changes. 
After receiving approval to implement revised proposed Sec.  50.46a 
requirements, it would also be necessary for licensees to demonstrate 
that the NUREG-1903 results remain applicable after implementing 
proposed changes.
    More specific details on how to conduct these applicability reviews 
are available in a white paper entitled, ``Plant-Specific Applicability 
of the 10 CFR 50.46 Technical Basis'' February 2009 (ML090350757). 
Commenters on this revised proposed rule may review this white paper to 
get a better understanding of the scope of the evaluation being 
considered by the NRC.

D. Alternative ECCS Analysis Requirements and Acceptance Criteria

    The revised proposed rule would require licensees to analyze ECCS 
cooling performance for breaks up to and including a double-ended 
rupture of the largest pipe in the RCS. These analyses would have to be 
performed by methods acceptable to the NRC and must demonstrate that 
ECCS cooling performance conforms to the acceptance criteria set forth 
in the rule. For breaks at or below the TBS, Sec.  50.46a(e)(1) would 
specify requirements identical to the existing ECCS analysis 
requirements set forth in Sec.  50.46. However, commensurate with the 
lower probability of breaks larger than the TBS, Sec.  50.46a(e)(2) of 
the revised proposed rule specifies less conservatism for the analyses 
and associated acceptance criteria for breaks larger than the TBS. LOCA 
analyses for break sizes equal to or smaller than the TBS would be 
applied to all locations in the RCS to find the limiting break 
location. LOCA analyses for break sizes larger than the TBS (but using 
the more realistic analysis requirements) would also be applied to all 
locations in the RCS to find the limiting break size and location. This 
analytical approach is consistent with current NRC regulatory positions 
and industry practice.
1. Acceptable Methodologies and Analysis Assumptions
    Under existing Sec.  50.46 requirements, prior NRC approval is 
required for ECCS evaluation models. Acceptable evaluation models are 
currently of two types; those that realistically describe the behavior 
of the RCS during a LOCA, and those that conform with the required and 
acceptable features specified in Appendix K to Part 50. Appendix K 
evaluation models incorporate conservatism as a means to justify that 
the acceptance criteria are satisfied by an ECCS design. In contrast, 
the realistic or best-estimate models attempt to accurately simulate 
the expected phenomena. As a result, comparisons to applicable 
experimental data must be made and uncertainty in the evaluation model 
and inputs must be identified and assessed. This is necessary so that 
the uncertainty in the results can be estimated so that when the 
calculated ECCS cooling performance is compared to the acceptance 
criteria, there is a high level of probability that the criteria would 
not be exceeded. Appendix K, Part II, contains the documentation 
requirements for evaluation models. All of these existing requirements 
are included in Sec.  50.46a(e)(1) of the revised proposed rule for 
breaks at or below the TBS.
    As currently required under Sec.  50.46, the ECCS analysis 
performed with a model other than one based on Appendix K must 
demonstrate with a high level of probability that the

[[Page 40030]]

acceptance criteria will not be exceeded. The position taken in RG 
1.157 has been that 95 percent probability constitutes an acceptably 
high probability. Section 50.46a(e)(1) of the revised proposed rule 
would retain the high level of probability as the statistical 
acceptance criterion.
    Revised proposed Sec. Sec.  50.46a(e)(1) and (e)(2) would require 
that the worst break size and location be calculated separately for 
breaks at or below the TBS and for breaks larger than the TBS up to and 
including a double-ended rupture of the largest pipe in the RCS. 
Different methodologies, analytical assumptions, and acceptance 
criteria may be used for each break size region. Consistent with 
current Sec.  50.46 requirements, licensees would be required to 
analyze breaks at or below the TBS by assuming the worst single failure 
concurrent with a loss-of-offsite power, limiting operating conditions, 
and only crediting safety systems. For breaks larger than the TBS, 
licensees may take credit for operation of any equipment supported by 
availability data provided that onsite power (either safety or non-
safety) can be reliably provided to that equipment through manual 
actions within a reasonable time after a loss of offsite power. All 
non-safety equipment that is credited for analyses of breaks larger 
than the TBS would have to be identified as such and listed in the 
plant technical specifications. Analyses of breaks larger than the TBS 
could assume nominal operating conditions rather than technical 
specification limits. This would also include combining actual fuel 
burnup in decay heat predictions with the corresponding operating 
peaking factors at the appropriate time in the fuel cycle. The 
assumptions of loss-of-offsite power and the worst single failure would 
not be required because breaks larger than the TBS are very unlikely; 
therefore, less margin would be needed in the analysis of breaks in 
this region. A capability to provide onsite power to non-safety 
equipment in a reasonable time following a loss of offsite power (e.g. 
approximately 30 minutes) is a defense-in-depth consideration for 
severe accident management.
2. Acceptance Criteria
    ECCS acceptance criteria in Sec.  50.46a(e)(3) for breaks at or 
below the TBS would be the same as those currently required in Sec.  
50.46. Therefore, licensees would be required to use an approved 
methodology to demonstrate that the following acceptance criteria are 
met for the limiting LOCA at or below the TBS:
    <bullet> PCT less than 2200 [deg]F;
    <bullet> Maximum local cladding oxidation (MLO) less than 17 
percent;
    <bullet> Maximum hydrogen production--core wide cladding oxidation 
less than one percent;
    <bullet> Maintenance of coolable geometry; and
    <bullet> Maintenance of long-term cooling.
    Commensurate with the lower probability of occurrence, the 
acceptance criteria in Sec.  50.46a(e)(4) for breaks larger than the 
TBS would be less prescriptive:
    <bullet> Maintenance of coolable geometry, and
    <bullet> Maintenance of long-term cooling.
    The revised proposed rule would allow licensees flexibility in 
establishing appropriate metrics and quantitative acceptance criteria 
for maintenance of coolable geometry. A licensee's metrics and 
acceptance criteria must realistically demonstrate that coolable core 
geometry and long-term cooling will be maintained. Unless data or other 
valid justification criteria are provided, licensees should use 2200 
[deg]F and 17 percent for the limits on PCT and MLO, respectively, as 
metrics and quantitative acceptance criteria for meeting the rule. 
Other less conservative criteria would be acceptable if properly 
justified by licensees.
    However, the NRC acknowledges that it would be expensive and time-
consuming for industry to develop the necessary experimental and 
analytical data to justify alternative acceptance criteria as a 
surrogate for demonstrating coolable geometry. Because of the 
difficulty in demonstrating alternative metrics, the NRC is requesting 
stakeholder comments on whether the final Sec.  50.46a rule should 
retain the coolable geometry criterion for beyond-TBS breaks. Retaining 
coolable geometry would give licensees the option to demonstrate 
alternative coolable geometry metrics or use the current metric (2200 
[deg]F PCT and 17 percent MLO). If the NRC removed the coolable 
geometry criterion, the beyond-TBS acceptance criteria would be the 
same as the acceptance criteria for TBS and smaller breaks (2200 [deg]F 
PCT and 17 percent MLO). The NRC will evaluate stakeholder comments on 
this question before deciding which beyond-TBS acceptance criteria to 
include in the final rule.
    As previously discussed in Section IV.C of this document, the NRC 
is working to revise the ECCS acceptance criteria in Sec.  50.46(b) to 
account for new experimental data on cladding ductility and to allow 
for the use of advanced cladding alloys. The NRC will soon issue an 
ANPR seeking public comments on a planned regulatory approach. The NRC 
expects that this rulemaking (Docket ID NRC-2008-0332) will establish 
new cladding embrittlement acceptance criteria in Sec.  50.46(b) for 
design basis LOCAs. As these new acceptance criteria are established, 
the NRC will also make conforming changes to Sec.  50.46a as necessary 
for both below and above TBS breaks.
3. Restriction of Reactor Operation
    Section 50.46a(e)(5) would allow the Director of the Office of 
Nuclear Reactor Regulation to impose restrictions on reactor operation 
if it is determined that the evaluations of ECCS cooling performance 
are not consistent with the requirements for evaluation models and 
analysis methods specified in revised proposed Sec.  50.46a(e)(1) 
through (e)(4). Non-compliance may be due to factors such as lack of a 
sufficient data base upon which to assess model uncertainty, use of a 
model outside the range of an appropriate data base, models 
inconsistent with the requirements of Appendix K of Part 50, or 
phenomena unknown at the time of approval of the methodology. Lack of 
compliance with methodological requirements would not necessarily 
result in failure to meet the acceptance criteria of revised proposed 
Sec. Sec.  50.46a(e)(3) and (e)(4), but, rather, would provide results 
that could not be relied upon to demonstrate compliance with the 
appropriate acceptance criteria. Thus, depending upon the specific 
circumstances, it might be necessary for the NRC to impose restrictions 
on operation until these issues are resolved. This requirement is 
included in the revised proposed rule for consistency with the current 
ECCS regulations, because it is comparable to existing Sec.  
50.46(a)(2).

E. Risk-Informed Changes to the Facility, Technical Specifications, or 
Procedures

    Licensees who adopt Sec.  50.46a would use a risk-informed 
evaluation process to demonstrate, before implementation, that facility 
changes will satisfy the risk-informed acceptance criteria in revised 
proposed Sec.  50.46a(f). Changes that must be evaluated are specified 
in revised proposed Sec.  50.46a(d)(3) and would include all 
``enabled'' changes that satisfy the alternative ECCS analysis 
requirements in Sec.  50.46a but do not satisfy the current ECCS 
analysis requirements in Sec.  50.46. Also, changes in risk from 
facility changes not enabled by the alternative ECCS requirements could 
be combined with changes in risk

[[Page 40031]]

from facility changes enabled by Sec.  50.46a if the licensee chooses 
to combine the changes in its application of the risk-informed change 
process defined in the rule. In this case, the changes made under Sec.  
50.46a would include those enabled by Sec.  50.46a and those not 
enabled by Sec.  50.46a but included in the risk-informed application.
    Licensees would be required to periodically maintain and upgrade 
the PRA used in the risk assessments and ensure that over time all 
changes made under Sec.  50.46a continue to meet the risk-informed 
acceptance criteria. If necessary, revised proposed Sec.  50.46a(g)(2) 
would require the licensee to propose steps and a schedule to bring the 
facility back into compliance with the acceptance criteria in Sec.  
50.46a(f)(2)(ii) or Sec.  50.46a(f)(2)(iii), as applicable.
    The risk-informed evaluation would be required to demonstrate that 
increases in plant risk (if any) meet appropriate risk acceptance 
criteria, defense-in-depth is maintained, adequate safety margins are 
maintained, and adequate performance-measurement programs are 
implemented. The NRC believes that all changes to a plant, its 
technical specifications, or its procedures which are based upon the 
analyses of ECCS performance permitted under Sec.  50.46a(e)(2)--with 
the exception of those changes permitted under Sec.  50.46a(f)(1)--must 
be reviewed and approved by the NRC for two reasons. First, a wide 
range of changes could be implemented under Sec.  50.46a, which, if 
improperly implemented by licensees, could result in significant 
adverse impacts on public health and safety or common defense and 
security. NRC review and approval would provide verification that a 
licensee has properly evaluated each proposed change against the 
acceptance criteria in Sec.  50.46a. Second, changes involving 
technical specifications must receive NRC review and approval in the 
form of a license amendment, as required by the Atomic Energy Act of 
1954, as amended. Accordingly, the NRC's revised proposed rule would 
require NRC review and approval of all changes initiated under Sec.  
50.46a(f)(2).
1. Requirements for the Risk-Informed Evaluation
    The revised proposed rule is based upon the regulatory premise that 
the acceptability of all licensee-initiated changes made under the rule 
should be judged in a risk-informed manner. The risk-informed 
assessment process must include methods for evaluating compliance with 
the risk criteria, defense-in-depth criteria, safety margin criteria, 
and performance measurement criteria in Sec.  50.46a(f). These 
attributes have been identified by the Commission as a necessary set of 
risk evaluation tools to ensure that changes to the facility do not 
endanger public health and safety.
    Compliance with the risk criteria plays a key role in the 
regulatory structure of the proposed rule. A risk-assessment must be 
used to determine the change in risk associated with facility changes. 
Inasmuch as PRA methodologies are generally recognized as the best 
current approach for conducting risk assessments suitable for making 
decisions in areas of potential safety significance, Sec.  50.46a(f)(4) 
of the revised proposed rule would require that a technically adequate 
PRA be used in demonstrating compliance with the requirements of Sec.  
50.46a that would affect the regulatory decision in a substantive 
manner. However, the NRC recognizes that non-quantitative PRA 
assessment methodologies and approaches could also be used to 
complement or supplement the quantitative aspects of a PRA, especially 
when performance of a quantitative PRA methodology of the level needed 
to support a particular decision is not justifiable because the safety 
significance of the decision does not warrant the level of technical 
sophistication inherent in a PRA. Accordingly, Sec.  50.46a(f)(5) is 
written to recognize that non quantitative risk assessment may also be 
utilized.
a. Probabilistic Risk Assessment Requirements
    Sections 50.46a(f)(4)(i) through (iv) set forth the four general 
attributes of an acceptable PRA for the purposes of this rule. Section 
50.46a(f)(4)(i) would require that the PRA address initiating events 
from internal and external sources, and for all modes of operation, 
including low power and shutdown, that would affect the regulatory 
decision in a substantial manner. Failure to consider sources of risk 
from internal and external events, or from anticipated operating modes, 
could result in an inaccurate characterization of the level of risk 
associated with a plant change. Therefore, initiating events from 
internal and external sources and during all modes of operation would 
have to be considered by the PRA when the change in risk would affect 
the regulatory decision, in order to ensure that the effect on risk 
from licensee-initiated changes is adequately characterized in a manner 
sufficient to support a technically defensible determination of the 
level of risk.
    Section 50.46a(f)(4)(ii) states that the PRA must reasonably 
represent the current configuration and operating practices at the 
plant. A plant's risk may vary as plant configuration and/or plant 
procedures change. Failure to update the PRA based upon these 
configuration or procedure changes may result in inaccurate or invalid 
PRA results. Accordingly, to ensure that estimates of risk adequately 
reflect the facility for which a decision must be made, the rule would 
require that the PRA address current plant configuration and operating 
practices.
    Section 50.46a(f)(4)(iii) would require that the PRA have 
``sufficient technical adequacy'' including consideration of 
uncertainty, as well as a sufficient level of detail to provide 
confidence that the calculated risk and the changes in risk adequately 
reflect the proposed facility change. The revised proposed rule would 
require the PRA to consider uncertainty because the decision maker must 
understand the limitations of the particular PRA that was performed to 
ensure that the decision is robust and accommodates relevant 
uncertainties. With respect to level of detail, failure to model the 
plant (or relevant portion of the plant) at the appropriate level of 
detail may result in calculated risk values that do not appropriately 
capture the risk significance of the proposed change.
    Finally, Sec.  50.46a(f)(4)(iv) would require that, to the extent 
that the PRA is used, the PRA must meet NRC-approved industry 
standards. The NRC has prepared a regulatory guide (RG 1.200) on 
determining the technical adequacy of PRA results for risk-informed 
activities. As one step in the assurance of technical quality, the PRA 
would be subjected to a peer review process assessed against an 
industry standard or set of acceptance criteria that is endorsed by the 
NRC. Industry standards for all initiators and operating modes are 
under development but not yet complete. The NRC will develop review 
guidelines that endorse criteria for considering the sufficiency of a 
PRA peer review process for this application in Sec.  50.46(c) if this 
guidance becomes necessary before industry standards have been 
completed and endorsed in RG 1.200.
b. Requirements for Risk Assessments Other Than PRA
    Risk assessment need not always be performed using PRA. The rule 
explicitly recognizes the possibility of using risk assessment methods 
other than PRA to demonstrate compliance with various acceptance 
criteria in the rule. However, as with PRA

[[Page 40032]]

methodologies, the NRC believes that minimum quality requirements for 
PRAs and risk assessments used by a licensee in implementing the rule 
must be established. Accordingly, Sec.  50.46a(f)(5) would establish 
the minimum requirement for risk assessment methodologies other than 
PRA. The NRC believes that this requirement provides flexibility to 
licensees to use the non-PRA risk methodology (or combination of 
different methodologies) when these methodologies produce results that 
are sufficient upon which to base decisions that the various acceptance 
criteria in the proposed rule have been met.
2. Aggregation of Plant Changes When Evaluating Changes in Risk
    Licensees often make changes to the facility, technical 
specifications, and procedures. Some changes that the licensees could 
make after adopting this rule would not have been permitted without the 
new Sec.  50.46a (related or enabled changes). Other changes would be 
unrelated insofar as the basis of the changes and NRC approval, when 
necessary, will rely on regulations, guidelines, or facility priorities 
that do not depend on the new ECCS requirements in Section 50.46a. 
Unrelated changes will indirectly influence the change in risk of the 
Sec.  50.46a related changes insofar as they change the risk profile of 
the facility. If unrelated changes are combined with related changes in 
determining the Sec.  50.46a change in risk estimates (bundling), the 
result will normally be different than if the unrelated changes are 
considered as part of the baseline risk associated with the current 
design and operation of the facility. If bundling is permitted, a 
licensee could implement facility changes that would decrease risk to 
offset increased risk from Sec.  50.46a enabled changes. These changes 
would increase the safety of the facility and are expected to result in 
a reallocation of resources to areas where safety can be improved. 
Current NRC practice, consistent with RG 1.174, is to compare the total 
or cumulative risk increase from all related changes, and only related 
changes, to the acceptance guidelines. RG 1.174 does, however, permit 
bundling changes (referred to as combined changes in RG 1.174) and 
provides additional acceptance guidelines that must be met when 
permitting unrelated plant changes that might decrease risk to be 
combined together with a group of related changes in a change in risk 
estimate that would be compared to the acceptance guidelines.
    The NRC believes that allowing bundling of unrelated changes into 
the Sec.  50.46a change in risk estimates will encourage licensees to 
use risk-informed methods to take advantage of opportunities to reduce 
risk, and not just eliminate requirements that a licensee deems as 
undesirable. However, in some situations, bundling could mask the 
creation of significant risk outliers. To ensure that outliers are not 
created, and that the additional guidelines in RG 1.174 are 
appropriately applied, the rule would not permit bundling of changes 
without previous review and approval. Therefore, the revised, proposed 
Sec.  50.46a(f)(2)(iv) would allow changes not enabled by Sec.  50.46a 
to be combined with changes enabled by Sec.  50.46a in the calculation 
of the change in risk when a licensee submits an application for a 
change under 50.90.
3. NRC Approval of a Licensee Process for Making Changes to a 
Licensee's Facility or Procedures Without NRC Review and Approval
    As a general matter, the licensee must obtain NRC review and 
approval (through a license amendment application) for any changes to 
the facility, technical specifications, or procedures that may be 
implemented under this section. However, the NRC believes that there is 
a subset of plant and procedure changes that would be made possible by 
Sec.  50.46a involving minimal changes in risk which also have no 
significant impact upon defense-in-depth capabilities. Prior NRC review 
and approval of these changes on an individual basis would be 
unnecessary if the NRC has previously concluded that the licensee has 
an adequate technical process for appropriately identifying this subset 
of changes. In the NRC's view, plant changes which involve minimal 
changes in risk and have no significant impact upon defense-in-depth 
(and do not involve a change to the license), by definition, do not 
result in significant issues involving public health and safety or 
common defense and security.
    Expending licensee resources to prepare an application for approval 
of plant changes involving minimal changes in risk and NRC resources to 
review and approve these applications is not an efficient use of 
resources. Rather, the NRC believes that if it reviews and approves in 
advance the licensee's processes (including the adequacy of the 
licensee's PRA and other risk assessment methods) and criteria for 
identifying changes which are both minimal from a risk standpoint and 
do not significantly affect defense-in-depth or plant physical 
security, then there is no need to review and approve each of the 
changes individually. Further, the NRC believes that these minimal 
changes are unlikely to impact the built-in capability of the facility 
to resist security threats. Accordingly, the NRC has proposed an 
approach in Sec.  50.46a(f)(1) allowing a licensee to obtain ``pre-
approval'' of a process for identifying minimal plant and procedure 
changes made possible under Sec.  50.46a.
    The revised proposed Sec.  50.46a(f)(1) states that a licensee may 
make changes based upon the provisions of this section without prior 
review and approval if the stated requirements in paragraphs (f)(1) and 
(f)(3) of this section are met. The revised proposed rule also states 
that the provisions of Sec.  50.59 would apply. Licensees with a pre-
approved change process would be allowed to make facility changes 
without NRC approval if they met Sec.  50.59 and Sec.  50.46a 
requirements. Compliance with the Sec.  50.59 requirements is necessary 
to ensure that facility changes made without NRC approval do not result 
in plant conditions that could impact public health and safety. 
Compliance with the Sec.  50.46a(f) requirements for risk assessments 
is required to ensure that facility changes result in acceptable 
changes in risk, adequate defense-in-depth, that safety margins will be 
maintained, and that adequate performance-measurement programs are 
implemented.
4. Risk Acceptance Criteria for Plant Changes
    Sections 50.46a(f)(2)(ii) and (f)(2)(iii) would require that the 
total increases in risk are very small and that the overall plant risk 
remains small. Two sets of metrics are used to measure risk depending 
on when the applicant's operating license was issued. For reactors 
licensed before the effective date of the rule, Sec.  50.46a(f)(2)(ii) 
would apply and CDF and LERF would be used. For new reactors licensed 
after the effective date of the rule, Sec.  50.46a(f)(2)(iii) would 
apply and CDF and large release frequency (LRF) are used. The NRC 
believes that this requirement is a necessary element for ensuring that 
changes which would be permitted by the revised Sec.  50.46a ECCS 
analyses do not result in a greater change in risk than intended by the 
Commission.
a. Risk Estimate
    To satisfy the Commission's requirements in Sec. Sec.  
50.46a(f)(2)(ii) and (f)(2)(iii) that the total increases in risk

[[Page 40033]]

are very small would require that the change in risk for each facility 
change be evaluated and shown to meet the acceptance guidelines. If a 
series of changes are made over time, Sec.  50.46a(f)(2)(iv) would 
require that cumulative effect of these changes be evaluated and shown 
to meet the acceptance criteria. Section 50.46a(f)(2)(iv) would also 
permit changes in risk from facility changes not enabled by Sec.  
50.46a to be combined by the licensee with facility changes that are 
enabled by this section for the purposes of meeting the acceptance 
guidelines. The total change in risk from all facility changes made 
under the rule after the adoption of Sec.  50.46a must be evaluated and 
compared to the ``very small'' acceptance criterion before each change 
requiring a risk-informed evaluation and after the periodic PRA 
maintenance and upgrading. Requiring that the total change in risk from 
all facility changes made under the rule after the adoption of Sec.  
50.46a be compared to the Sec.  50.46a acceptance criteria instead of 
allowing the changes in risk to be partitioned and individually 
compared to the acceptance criteria would ensure that the total risk 
increase of all changes, as they are implemented over time, would not 
constitute more than a very small increase in risk. If the total 
increase in the applicable risk metrics were not compared to the 
acceptance criteria, a number changes where every individual change's 
risk increase is kept below the proposed rule's risk acceptance 
criteria could, considered cumulatively, result in a significant 
increase in risk. A significant increase would not satisfy the 
Commission's criteria that the overall plant risk remains small. Also, 
comparing the risk increase from each change to the acceptance criteria 
independently of all previous changes would render the use of the 
``very small'' criterion inadequate to monitor and control increases in 
risk from a series of plant changes implemented over time.
    Comparing the total risk increase to the risk increase criterion, 
and allowing bundling of unrelated changes in the change in risk 
estimate, will support the NRC's philosophy that, consistent with the 
principles of risk-informed integrated decision making, licensees 
should have a risk management philosophy in which risk insights are not 
just used to systematically increase risk, but also to help reduce risk 
where appropriate and where it is shown to be cost effective.
b. Acceptance Criteria
    In Sec.  50.46a(f)(2)(ii), CDF and LERF are used as surrogates for 
early and latent health effects, which are used in the Commission's 
Policy Statement on Safety Goals (51 FR 30028; August 4, 1986). The NRC 
has used CDF and LERF in making regulatory decisions for over 20 years. 
The NRC endorsed the use of CDF and LERF as appropriate measures for 
evaluating risk and ensuring safety in nuclear power plants when it 
adopted RG 1.174 in 1997. After the adoption of RG 1.174, the NRC has 
had eleven years of experience in applying risk-informed regulation to 
support a variety of applications, including amending facility 
procedures and programs (e.g., IST and ISI programs), amending facility 
operating licenses (e.g., power up-rates, license renewals, and changes 
to the FSAR), and amending technical specifications. On the basis of 
this experience, for current operating reactors, the NRC has determined 
that CDF and LERF are acceptable measures for evaluating changes in 
risk as the result of changes to a facility, technical specifications, 
and procedures, with the exception of certain changes that affect 
containment performance but do not affect CDF or LERF. Changes that 
affect containment performance are considered as part of the defense-
in-depth evaluation.
    For new reactors, CDF and LRF (instead of LERF) would apply as 
indicated in Sec.  50.46a(f)(2)(iii). For new reactor licensing the 
Commission has established a goal based on LRF (see SRM on SECY-89-
102--Implementation of the Safety Goals, June 15, 1990; and SRM on 
SECY-90-016--Evolutionary Light Water Reactor (LWR) Certification 
Issues and Their Relationship to Current Regulatory Requirements, June 
26, 1990).
    The Commission has concluded that changes under this rule should be 
restricted to very small risk increases. As discussed in RG 1.174, a 
very small risk increase is independent of a plant's overall risk as 
measured by the current CDF and LERF. Increases in CDF of 10-6 per 
reactor year or less, and increases in LERF of 10-7 per reactor year or 
less are very small risk increases for existing reactor facilities.
    For new reactors, the same CDF metric is used and the same 
definition of very small increase (i.e., less than 10-6 per reactor 
year) would be used. The revised proposed rule uses LRF instead of LERF 
as a metric for new reactors. RG 1.174 provides no guidelines for LRF. 
The Commission has approved the overall mean frequency of a large 
release of radioactive material to the environment (LRF) to be less 
than 10-6 per reactor year. The revised proposed rule requires the 
total increase in LRF to be no more than very small. The NRC proposes 
that increases in LRF of 10-8 per reactor year or less are very small 
risk increases for new reactors. Because of the difference between the 
LERF acceptance criteria for existing reactors and the LRF acceptance 
criteria for new reactors, the NRC is seeking specific public comments 
on this topic. Additional background information on how the NRC is 
addressing this issue and how the NRC is soliciting public input on 
this topic in this revised proposed rule and in other regulatory areas 
is provided in Section J.2. of this document.
    After adopting RG 1.174 in 1997, the NRC has applied the 
quantitative change in risk guidelines to individual plant changes and 
to sequences of plant changes implemented over time. The NRC has found 
these guidelines and the CDF and LERF values (when used together with 
the defense in depth, safety monitoring, and performance measurement 
criteria) are capable of differentiating between changes, and sequences 
of changes, that are not expected to endanger public health and safety 
from those that might. The NRC believes that applying the LRF guideline 
for determining very small risk increases would also be protective of 
public health and safety.
    Section 50.46a(f)(1) would permit licensees to make changes under 
this provision without prior review and approval if the changes involve 
minimal increases in risk which also have no significant impact upon 
defense-in-depth capabilities. A minimal risk increase is one which, 
when considered qualitatively by itself or in combination with all 
other minimal increases, would never become significant. Logically, a 
minimal increase is less than the very small increase in CDF and in 
LERF, and was chosen as an increase of less than 10-7 per reactor year 
for CDF and an increase in LERF of less than 10-8 per reactor year. 
Similarly, for new reactor licensing, an increase in LRF less than 10-9 
per reactor year is a minimal increase. Although ten of these changes 
could cause the combination of minimal increases to exceed the very 
small criteria, the NRC believes that most of these changes will have a 
much smaller (and, in some cases, an unmeasurable) increase in risk. 
Regardless of whether a licensee makes changes under Sec.  50.46a(f)(1) 
instead of Sec.  50.46a(f)(2), the total cumulative risk including all 
the individually minimal risk increases as well as any increases 
approved by the NRC under Sec.  50.46a(f)(2), would have to

[[Page 40034]]

be considered in the periodic reporting required by Sec.  50.46a(g)(2). 
If a licensee implements an unexpectedly large number of minimal risk 
changes, the periodic reporting requirements in Sec.  50.46a(g)(2) 
would provide adequate notice to ensure that the NRC is aware of 
potentially significant changes (or any collective impact), so that the 
NRC may undertake additional oversight actions as deemed necessary and 
appropriate.
    Additionally, although the revised proposed rule would permit 
licensees to make plant changes that result in very small risk 
increases, the NRC is requesting stakeholder comments on whether the 
rule should allow plant changes that increase risk at all. Instead of 
the risk acceptance criteria allowing very small risk increases, should 
the risk acceptance criteria in final rule require that the net effect 
of plant changes made under Sec.  50.46a be risk neutral or risk 
beneficial? The NRC requests stakeholders to provide comments on the 
use of risk acceptance criteria that would not allow a cumulative 
increase in risk for plant changes made under Sec.  50.46a.
5. Defense-in-Depth
    Section 50.46a(f)(3)(i) would require that the risk-informed 
evaluation demonstrate that defense-in-depth is maintained. Defense-in-
depth is an element of the NRC's safety philosophy that employs 
successive measures to prevent accidents or mitigate damage if a 
malfunction, accident, or naturally caused event occurs at a nuclear 
facility. As conceived and implemented by the NRC, defense-in-depth 
provides redundancy in addition to a multiple barrier approach against 
fission product releases. Defense-in-depth continues to be an effective 
way to account for uncertainties in equipment and human performance. 
The NRC has determined that retention of adequate defense-in-depth must 
be ensured in all risk-informed regulatory activities.
6. Safety Margins
    Section 50.46a(f)(3)(ii) would require that adequate safety margins 
be retained to account for uncertainties. These uncertainties include 
phenomenology, modeling, and how the plant was constructed or is 
operated. The NRC's concern is that plant changes could inappropriately 
reduce safety margins, resulting in an unacceptable increase in risk or 
challenge to plant SSCs. This provision would ensure that an adequate 
safety margin exists to account for these uncertainties, such that 
there are no unacceptable results or consequences (e.g., structural 
failure) if an acceptance criterion or limit is exceeded.
7. Performance Measuring Programs
    Section 50.46a(f)(3)(iii) would require that adequate performance 
measurement programs and feedback strategies be implemented to ensure 
that the risk-informed evaluation continues to reflect actual plant 
design and operation. The risk-informed evaluation includes the risk 
assessment, maintenance of defense-in-depth, and adequacy of safety 
margins. Results from implementation of monitoring and feedback 
strategies can provide an early indication of unanticipated degradation 
of performance of plant elements that may invalidate the demonstration 
by the risk-informed evaluation that the change satisfied all the 
acceptance criteria. This section would require that the monitoring 
programs be designed to detect degradation of SSCs before plant safety 
is compromised. Permitting degradation to advance until plant safety 
could be compromised would be inconsistent with the NRC's regulatory 
responsibility of protecting public safety. The NRC expects that 
licensees will integrate existing programs for monitoring equipment 
performance and other operating experience on their site and throughout 
industry with the performance measuring programs required by this 
section.

F. Operational Requirements

    The revised proposed rule includes five specific operational 
requirements that apply to licensees who are approved to implement 
Sec.  50.46a. These requirements are set forth in Sec.  50.46a(d) and 
would remain in effect as long as the facility is subject to the Sec.  
50.46a alternative ECCS requirements until such time as the licensee 
permanently ceases operations by submitting the decommissioning 
certifications required under Sec.  50.82(a). They are:
    1. Maintain ECCS models and/or analysis methods that demonstrate 
compliance with the ECCS acceptance criteria.
    2. Maintain reactor coolant leak detection equipment available at 
the facility and identify, monitor, and quantify leakage to ensure that 
adverse safety consequences do not result from leakage from piping and 
components larger than the transition break size.
    3. Perform a risk-informed evaluation for each potentially risk-
significant change (or group of changes) to the facility enabled by 
Sec.  50.46a.
    4. Periodically assess the cumulative effect of changes to the 
plant, operational practices, equipment performance, and plant 
operational experience.
    5. Do not operate the plant for more than fourteen days in any 12 
month period in an at-power operating configuration that has not been 
demonstrated to meet the ECCS acceptance criteria for breaks larger 
than the TBS.
    Each of the five operational requirements is discussed in detail 
below.
    1. Maintain ECCS models and/or analysis methods that demonstrate 
compliance with the ECCS acceptance criteria.
    Calculated results of licensee ECCS models and/or analysis methods 
must demonstrate compliance with the ECCS acceptance criteria 
throughout the operating lifetime of the plant. Licensees must also 
update ECCS models and/or analysis methods by modifying them as needed 
to address any plant design changes affecting ECCS performance during 
this time period.
    2. Maintain reactor coolant leak detection equipment available at 
the facility and identify, monitor, and quantify leakage to ensure that 
adverse safety consequences do not result from leakage from piping and 
components larger than the transition break size.
    In a Staff Requirements Memorandum dated August 10, 2007, 
responding to SECY-07-0082--``Rulemaking To Make Risk Informed Changes 
to Loss-of-Coolant Accident Technical Requirements; 10 CFR 50.46a, 
`Alternative Acceptance Criteria for Emergency Core Cooling Systems for 
Light-Water Nuclear Power Reactors' '', the Commission directed the NRC 
staff to evaluate various approaches for enhancing the 10 CFR 50.46a 
rule with requirements for improved leak detection methods. This SRM 
also directed the NRC staff to ``strengthen the assurance of defense-
in-depth [provided by the Sec.  50.46a rule] for breaks beyond the 
transition break size (TBS).''
    In response to a recommendation made by the Davis-Besse Lessons 
Learned Task Force (DBLLTF), (see memorandum from Arthur T. Howell to 
William F. Kane, ``Degradation of the Davis-Besse Nuclear Power Station 
Reactor Pressure Vessel Head Lessons-Learned Report; September 30, 
2002; ADAMS Accession No. ML022740211) the NRC evaluated whether it 
should impose new requirements on licensees in the areas of tighter 
reactor coolant leakage limits and new leakage monitoring requirements. 
Specifically, the DBLLTF Recommendation 3.1.5(1) said that the NRC 
should determine whether PWR plants should install on-line enhanced 
leakage detection systems

[[Page 40035]]

on critical plant components which would be capable of detecting 
leakage rates of significantly less than 1 gallon per minute.
    The evaluation identified techniques that could improve localized 
leak detection and on-line monitoring and several areas of possible 
improvements to leakage detection requirements that could provide 
increased confidence that plants are not operated at power with reactor 
coolant pressure boundary leakage. Although the NRC concluded that 
there was not a sufficient basis to require reduced technical 
specification leakage for existing licensees, the NRC recommended 
updating Regulatory Guide 1.45 on leak detection. This RG was revised 
in 2008.
    RG 1.45, Revision 1 incorporates progress in reactor coolant 
pressure boundary leakage detection technology; addresses the effect on 
radiation monitoring, and, subsequently, on leak detection from reduced 
activity levels of coolant resulting from improved fuel integrity; and 
incorporates lessons learned from operating experience. The title of 
the Regulatory Guide 1.45, Revision 1, has been changed from ``Reactor 
Coolant Pressure Boundary Leakage Detection Systems'' to ``Guidance on 
Monitoring and Responding to Reactor Coolant System Leakage,'' to 
reflect its broader scope. Revision 1 provides detailed guidance for 
timely detection and location of leaks, continuous monitoring, 
quantifying and trending of leak rates, assessing safety significance, 
and specifying plant actions following confirmation of an adverse trend 
in unidentified leak rate. Revision 1 describes acceptable leakage 
detection systems and methods, using risk-informed and performance-
based criteria to the extent practical. It retains the recommendations 
for monitoring of sump level or flow, airborne particulate activity, 
and condensate flow rate from air coolers. Other supplementary 
detection methods are recommended for use where and when appropriate.
    Paragraph 50.46a(d)(2) in the revised proposed rule contains new 
enhanced leak detection requirements. Enhanced leak detection is 
expected to provide increased defense-in-depth against large pipe 
breaks for licensees who implement the alternative ECCS rule. The NRC 
has concluded that implementing the guidance in Regulatory Guide 1.45, 
Revision 1, by licensees choosing to comply with 10 CFR 50.46a will 
result in improved monitoring and response to leaks in the reactor 
coolant system and will provide an acceptable method to satisfy the 
requirements of Section 50.46a(d)(2).
    3. Perform a risk-informed evaluation for each change (or group of 
changes) to the facility enabled by Sec.  50.46a.
    In addition to meeting all other applicable requirements, a risk-
informed evaluation required by Sec.  50.46a(d)(3) would have to be 
performed for changes enabled by Sec.  50.46a. If a licensee has a 
change methodology that was submitted under Sec.  50.46a(f)(1) and 
approved by the NRC, that licensee could make some changes without NRC 
approval, if the acceptance criteria in Sec.  50.46a(f)(1) are met. 
Otherwise, the licensee would be required to submit the results of its 
risk-informed evaluation for prior NRC review and approval in a license 
amendment request subject to the requirements of Sec.  50.90. The 
licensee would have to retain the results of all risk-informed 
evaluations made under Sec.  50.46a(f)(1) and periodically submit a 
summary of the results to the NRC as required under Sec.  50.46a(g)(3).
    4. Periodically assess the cumulative effect of changes to the 
facility.
    Key components of risk-informed regulation are the monitoring of 
changes in plant risk and feedback to the risk assessment and/or plant 
design activities and processes which are the subject of the risk 
assessment. Section 50.46a(d)(4) would require that after adopting 
Sec.  50.46a, a licensee would be required to periodically maintain and 
upgrade the risk assessments (both PRA and non-PRA) required under 
Sec. Sec.  50.46a(f)(4) and (f)(5). In particular, it is necessary that 
the PRA be maintained to reflect all plant changes; such as 
modifications, procedure changes, or changes in plant performance data. 
This maintenance enables the licensee to demonstrate that the total 
increases in CDF and LERF (or LRF for new reactors) after adopting 
Sec.  50.46a continue to meet the acceptance criteria in Sec.  
50.46a(f)(2). The risk assessments would have to continue to meet the 
minimum quality requirements in Sec. Sec.  50.46a(f)(4) and (f)(5) to 
support reasoned decision making under the rule.
    The revised proposed rule would specify that the maintenance and 
upgrading be conducted periodically ``but no less often than once every 
two refueling outages.'' The NRC believes that this is an appropriate 
period because the uncertainty of risk changes occurring during the two 
refueling outage period is tolerable and unlikely to result in high 
risk situations developing as a result of the implementation of plant 
changes. The NRC's determination is based upon the stringent acceptance 
criteria governing changes made under Sec.  50.46a, as well as the 
existing deterministic criteria in the substantive technical 
requirements in Part 50 and the criteria utilized in determining the 
acceptability of plant changes. The updating period specified in the 
rule is also comparable to other NRC requirements governing updating 
and reporting of safety information, e.g., Sec. Sec.  50.59, 50.71(e).
    If the assessment of the cumulative effect of changes made under 
the rule demonstrates that the acceptance criteria in Sec.  
50.46a(f)(2) are not met, Sec.  50.46a(g)(2) would require the licensee 
to develop steps and a schedule to bring the facility design and 
operation back into compliance with the acceptance criteria. These 
actions may include (but are not limited to) corrections to the risk 
analyses to demonstrate compliance, implementation of facility changes 
to offset adverse changes in risk, or reversal of changes previously 
made under the provisions of Sec.  50.46a(f). The NRC believes that 
this requirement provides appropriate flexibility for the licensee to 
determine the actions necessary to ensure continued compliance with the 
Sec.  50.46a(f) acceptance criteria, and is consistent with the concept 
of performance-based regulation.
    5. Do not operate the plant for more than a total of fourteen days 
in any 12 month period in an operating configuration that has not been 
demonstrated to meet the ECCS acceptance criteria for breaks larger 
than the TBS.
    As previously discussed in the supplementary information of this 
document, the NRC has included restrictions in the revised proposed 
rule on plant operation in configurations where licensees have not 
demonstrated that LOCAs larger that the TBS will be mitigated. The 
initial proposed rule (November 2005) would have completely prohibited 
at-power operation in any configuration without the demonstrated 
ability to mitigate a beyond-TBS LOCA. The revised proposed rule would 
restrict operation in such a configuration to not exceed fourteen days 
in any twelve month period. The NRC believes it is unlikely that 
licensees will experience circumstances where they would consider 
operating in such a condition for more than fourteen days, but has 
concluded that the establishing a limit on the allowable time is 
necessary to support the defense-in-depth philosophy. Even though the 
LOCA frequencies on which the TBS is founded indicate that the expected 
frequency of breaks larger than the TBS is low, the restriction is 
needed because there are large uncertainties associated with these 
frequency estimates. The

[[Page 40036]]

Commission concluded that the consequences of a challenge to the 
facility from an unmitigated break larger than the TBS are severe 
enough to warrant some confidence that the break could be mitigated. 
Thus the revised proposed rule will limit the allowed time period for 
operation in an unanalyzed condition to fourteen days in any twelve 
month period to ensure that mitigation capability is maintained except 
for occasional brief periods long enough to perform online maintenance 
of mitigation structures, systems and components.

G. Reporting Requirements

1. ECCS Analysis Reporting Requirements
    Section 50.46a(g)(1) sets forth reporting requirements with respect 
to changes or errors in LOCA evaluation models. For each change to or 
error discovered in an ECCS evaluation model or analysis method or in 
the application of such a model that affects the calculated results, 
the licensee shall report the nature of the change or error and its 
estimated effect on the limiting ECCS analysis to the NRC at least 
annually as specified in Sec.  50.4. If the change or error is 
significant, the licensee shall provide this report within 30 days and 
include with the report a proposed schedule for providing a reanalysis 
or taking other action as may be needed to show compliance with Sec.  
50.46a requirements. The 30 day period ensures sufficient time for the 
licensee to complete its evaluation and explanation of the changes and 
determine the course of action necessary to address compliance issues. 
For breaks smaller than the TBS a significant change is one which 
results in a calculated peak fuel cladding temperature different by 
more than 50 degrees Fahrenheit from the temperature calculated for the 
limiting transient using the last acceptable 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 degrees 
Fahrenheit. This requirement is the same as in Sec.  50.46. The NRC 
will also apply these reporting criteria to LOCAs involving pipe breaks 
larger than the TBS unless a specific alternative is proposed by a 
licensee and is approved by the NRC.
2. Risk Assessment Reporting Requirements
    Section 50.46a(g)(2) would set forth reporting requirements with 
respect to the PRA maintenance and upgrading that would be required by 
Sec.  50.46a(d)(4). When updating and upgrading the PRA, Sec.  
50.46a(g)(2) would require the licensee to report changes to the NRC 
within 60 days if the acceptance criteria in Sec. Sec.  
50.46a(f)(2)(ii) or (f)(2)(iii) (for new reactors) are exceeded. This 
provision would also require the report to include a schedule for 
implementation of any corrective actions necessary to bring plant 
operation or design back into compliance with the acceptance criteria. 
The 60-day period would ensure sufficient time for the licensee to 
complete its evaluation and explanation of the changes and determine 
the course of action necessary to address adverse changes in risk, 
while not unduly delaying the report to the NRC and thereby delaying 
NRC oversight. The NRC believes it should be informed of the licensee's 
implementation schedule so the NRC can ensure that the licensee takes 
corrective action on a timely basis, consistent with the safety 
significance of the change.
    Section 50.46a(g)(3) would require periodic reports of changes that 
required a risk-informed evaluation under Sec.  50.46a(d)(3) and were 
implemented without prior NRC approval under paragraph (f)(1) of this 
section. This process is comparable in many respects to the Sec.  50.59 
process which requires similar reports.

H. Documentation Requirements

    Section 50.46a(h) of the revised proposed rule would require that 
licensees maintain records sufficient to demonstrate compliance with 
Sec.  50.46a requirements. When making plant changes under Sec.  
50.46a(f) and when updating its PRA and/or other risk assessments, 
licensees would be required to document the bases for concluding that 
the acceptance criteria in Sec. Sec.  50.46a(f)(1) and (f)(2) are 
satisfied and that they continue to be satisfied throughout the 
operating lifetime of the facility. Licensees are also required under 
Part II of Appendix K to Part 50 to document the bases of evaluation 
models used to perform ECCS calculations. Licensees would also be 
required to document the time spent in an operating configuration not 
demonstrated to meet the ECCS acceptance criteria in Sec.  50.46a(c)(3) 
to demonstrate compliance with the fourteen days in any twelve month 
period limit in paragraph (d)(5) of this section. This documentation 
could be reviewed during NRC inspections and/or audits to ensure that 
the risk criteria in Sec.  50.46a(f) would be satisfied.

I. Submittal and Review of Applications

1. Initial Application for Implementing Alternative Sec.  50.46a 
Requirements
    When a licensee first applies to adopt the alternative Sec.  50.46a 
requirements, that licensee must submit an application under Sec.  
50.90 for NRC review and approval of a license amendment request. The 
initial application must contain the information as specified in 
Sec. Sec.  50.46a(c)(1)(i) through (v). This includes information 
related to the applicability to the facility of the NUREG-1829 and 
NUREG-1903 results; information identifying the ECCS analysis methods 
to be used; information describing the licensee's risk-informed 
evaluation process; information describing the licensee's proposed 
process for making risk-informed changes without prior NRC approval (if 
the licensee is seeking approval of such a process); and information 
describing non safety equipment to be credited for compliance with the 
ECCS acceptance criteria in Sec.  50.46a(e). A licensee's initial 
change from its existing ECCS analysis need not be reviewed by the 
licensee under the provisions of Sec.  50.59. Because the rule requires 
NRC review and approval of the initial license amendment application 
for compliance with the alternative Sec.  50.46a requirements, there is 
no purpose served by also requiring licensees to perform a Sec.  50.59 
evaluation, because Sec.  50.59 is a process to determine the need for 
prior NRC approval of a change to a facility or its procedures as 
described in the FSAR. After the Sec.  50.46a evaluation models and 
initial ECCS LOCA analyses are established by approval of the license 
amendment implementing Sec.  50.46a, subsequent changes to ECCS 
analyses would be controlled by the existing process in Sec.  50.59 
(which provides criteria for determining which changes are within the 
licensee's authority) and the requirements in Sec.  50.46a(g) for 
reporting when changes to evaluation models and analysis methods 
(whether from correction of errors or changes) is significant.
    The initial application may request one or more facility changes. 
The initial application may also include a request for NRC approval of 
a process for evaluating the acceptability of future changes enabled by 
Sec.  50.46a using the provisions in paragraph (f)(1) of this section. 
If approval of a process for evaluating future changes is requested, 
the application must include the information described in Sec.  
50.46a(c)(1)(iv). Otherwise, this information would not need to be 
submitted in the initial application.

[[Page 40037]]

2. Subsequent Applications for Plant Changes Under Sec.  50.46a
    After NRC approval of a licensee's initial license amendment 
application addressing ECCS analyses and the risk-informed evaluation 
processes, licensees may submit individual license amendment 
applications for plant changes under Sec.  50.90. These individual 
license amendment applications must contain:
    a. The information required by Sec.  50.90;
    b. Information from the risk-informed evaluation demonstrating that 
the risk criteria, defense-in-depth criteria, safety margins, and 
performance monitoring criteria in Sec. Sec.  50.46a(f)(2) and (f)(3) 
are met;
    c. Information demonstrating that the ECCS acceptance criteria in 
Sec. Sec.  50.46a(e)(3) and (e)(4) are met; and
    d. Information demonstrating that the proposed change will not 
increase the LOCA frequency of the facility by an amount that would 
invalidate the applicability to the facility of the generic NUREG-1829 
and NUREG-1903 reports.
    After reviewing the individual plant change license amendment 
application, the NRC may approve the change if it complies with the 
above criteria and all other applicable NRC regulations, including 
requirements for plant physical security. The NRC would evaluate 
potential impacts of the proposed change on facility security to ensure 
that the change does not significantly reduce the ``built-in 
capability'' of the plant to resist security threats, thus ensuring 
that the change is not inimical to the common defense and security and 
provides adequate protection to public health and safety.
    Licensees who have not submitted a request for NRC approval of a 
process for evaluating the acceptability of future changes enabled by 
Sec.  50.46a using the provisions in paragraph (f)(1) of that section 
may do so at any time by submitting the information described in 
paragraph (c)(1)(iv).

J. Applicability to New Reactor Designs

    As previously discussed under NRC Topic 1, the NRC has evaluated 
public comments and agrees with commenters who stated that there are no 
technical reasons which prevent the revised proposed Sec.  50.46a 
regulations from being applied to new light water reactor designs that 
are similar in nature (with respect to design and expected LOCA pipe 
break frequency) to current operating reactors.
1. Similarity of New Reactor Designs to Existing Reactor Designs
    There are several new LWR designs for which the NRC expects that 
the frequency of large LOCAs could be as low as it is at current LWRs. 
Thus, it could be appropriate to allow applicants to apply the Sec.  
50.46a requirements to these future designs. Accordingly, the revised 
proposed rule has been modified to apply to new LWR reactor designs; 
i.e. facilities other than those which are currently licensed to 
operate. Applicants for design certification or combined licenses, 
holders of combined licenses under 10 CFR part 52, or future licensees 
of operating light-water reactors who wish to apply Sec.  50.46a must 
submit an analysis for NRC approval demonstrating why it would be 
appropriate to apply the alternative ECCS requirements and what the 
appropriate transition break size (TBS) would be in order for the new 
design to meet the intent of the Sec.  50.46a rule.
    In its analysis, the applicant, holder, or licensee must 
demonstrate that the proposed reactor facility is similar to reactors 
licensed before the effective date of the rule. In addressing 
similarity of the proposed design to reactors licensed before the 
effective date of rule, the applicant, holder, or licensee would need 
to address design, construction and fabrication, and operational 
factors that include, but are not limited to:
    (1) The similarity of the piping materials of construction and 
construction techniques for new reactors to those in the currently 
operating fleet;
    (2) The similarity of service conditions and operational programs 
(e.g., in-service inspection and testing, leak detection, quality 
assurance etc.) for new reactors to those for operating plants;
    (3) The similarity of piping design, e.g. pipe sizes and pipe 
configuration, for new reactors to those found in operating plants;
    (4) Adherence to existing regulatory requirements, regulatory 
guidance, and industry programs related to mitigation and control of 
age-related degradation (e.g., aging management, fatigue monitoring, 
water chemistry, stress corrosion cracking mitigation etc.); and
    (5) Any plant-specific attributes that may increase LOCA 
frequencies compared to the generic results in NUREG-1829 and NUREG-
1903.
    The analysis must also include a recommendation for an appropriate 
TBS and a justification that the recommended TBS is consistent with the 
technical basis for this proposed rule. For those new reactor designs 
that employ design features that effectively increase the break size 
via opening of specially designed valves to rapidly depressurize the 
reactor coolant system during any size loss of coolant accident, 
justification of the relevance of a TBS would also be necessary. The 
methodology used to determine the proposed TBS should be described in 
the justification.
    Based on information currently available, new reactor designs may 
have similar piping materials, similar service conditions and 
operational programs, similar piping designs, and similar mitigation 
and control of age-related degradation programs to those found in 
currently operating plants. Therefore, the TBS defined in the proposed 
rule for currently operating reactors could potentially be applicable 
to some new reactor designs.
    In addition, after obtaining an operating or combined license for a 
plant with a currently-approved standard design, a licensee could adopt 
Sec.  50.46a if the design is demonstrated to be similar to the designs 
of plants licensed before the effective date of the rule (by evaluating 
the criteria above) and the TBS proposed by the licensee is found 
acceptable by the NRC.
2. NRC Request for Public Comments on the Use of Large Release 
Frequency (LRF) as the Risk Acceptance Criteria Metric for New Reactors
    Regulatory Guide 1.174, ``An Approach for Using Probabilistic Risk 
Assessment in Risk Informed Decisions on Plant Specific Changes to the 
Licensing Basis,'' was originally issued in July 1998. This RG provides 
guidance for a multitude of risk-informed applications and improves 
consistency in regulatory decisions in areas where the results of risk 
analyses are used to help justify regulatory action. The guide is the 
foundation for many other risk-informed programs (e.g., inservice 
testing, inservice inspection of piping) at the agency.
    Regulatory Guide 1.174 describes five key principles of the risk-
informed, integrated decision making process. In Principle 4--When 
proposed changes result in an increase in core damage frequency or 
risk, the increases should be small and consistent with the intent of 
the Commission's Safety Goal Policy Statement--the regulatory guide 
presents quantitative guidelines for acceptably small increases in CDF 
and LERF, as depicted in Figures 3 and 4 of the guide. The magnitude of 
acceptably small increases varies stepwise with the baseline CDF and 
LERF. A small increase up to 10<SUP>-</SUP>\5\ per reactor year for CDF 
and 10<SUP>-</SUP>\6\ per reactor year for LERF

[[Page 40038]]

are normally acceptable until the baseline risk increases to reference 
values of approximately 10<SUP>-</SUP>\4\ per reactor year and 
10<SUP>-</SUP>\5\ per reactor year for CDF and LERF respectively. 
Plants with baseline CDF and LERF which exceed the reference values, or 
with baseline risks that are not known with precision, would normally 
be limited to very small risk increases of up to 10<SUP>-</SUP>\6\ per 
reactor year and 10<SUP>-</SUP>\7\ per reactor year for CDF and LERF, 
respectively. Before RG 1.174 was issued, the Commission's SRM dated 
June 26, 1990, prepared in response to SECY-90-016, ``Evolutionary 
Light Water Reactor Certification Issues and their Relationships to 
Current Regulatory Requirements,'' established a goal for large release 
frequency (LRF) of less than 10<SUP>-</SUP>\6\ per reactor year for new 
reactor design certification and licensing. These goals are discussed 
further in Standard Review Plan (NUREG-0800) Chapter 19, and RG 1.206 
``Combined License Applications for Nuclear Power Plants'' Section 
C.I.19.
    In light of this difference in the risk metrics used for currently 
operating reactors (LERF) and new reactors (LRF), the NRC is seeking 
public comments on whether LRF should be the metric of concern in lieu 
of LERF for new reactor applicants (or licensees) implementing the 
Sec.  50.46a alternative ECCS requirements. Because the LRF goal for 
new reactors is a decade lower than the 10<SUP>-</SUP>\5\ per reactor 
year LERF reference value above which a facility would be limited to 
very small increases, should the definition of what constitutes ``very 
small increase'' and ``minimal increase'' for LRF (for new reactors) be 
a full decade lower than those defined for LERF (for existing reactors) 
or should the definition be based on relative change in LRF?
    The NRC has previously sought stakeholder input on the issue of 
risk metrics for new light-water reactors. A memorandum dated February 
12, 2009, from R. W. Borchardt, Executive Director for Operations, to 
the Commissioners, ``Alternative Risk Metrics for New Light-Water 
Reactor Risk-Informed Applications'' (Adams Accession No. ML090160008), 
provides a discussion of the issues. The white paper attached to that 
memorandum presents a full discussion of the issues and options for 
applying or modifying the current set of reactor risk metrics to new 
reactors. The paper discusses the issues posed by the lower risk 
estimates of new reactors in risk-informed applications, including 
changes to the licensing basis and the reactor oversight process, and 
describes the advantages and disadvantages of each option.
    On February 18, 2009, the NRC held a public meeting with 
stakeholders on the topic of risk metrics for new light-water reactors 
(see meeting summary; Adams Accession No. ML090570356). Additionally, 
both the NRC and industry representatives provided a briefing on the 
topic at the April 3, 2009, meeting of the ACRS.
    As discussed in these documents, the NRC is considering several 
options regarding risk metrics for new reactor risk-informed 
applications. The options include applying the existing operating 
reactor acceptance guidelines to new reactors, using new guidelines and 
thresholds for new reactors, or postponing any significant change to 
the process and evaluating new reactors on a case-by-case basis for an 
indeterminate period. As described in the NEI paper, ``Risk Metrics for 
Operating New Reactors'' (ML090900674; March 27, 2009), NEI has 
expressed its preference for applying the existing operating reactor 
acceptance guidelines to new reactors (which is referred to as Option 1 
in the NRC white paper).
    As part of the public comment process for this revised proposed 
rule, public stakeholders are invited to comment on the use of any of 
the alternative risk metric approaches for determining compliance with 
the risk acceptance criteria in Sec.  50.46a.

VI. Specific Topics Indentified for Public Comment

    The NRC seeks specific public comments on three topics. These 
issues were discussed previously in this document, but are summarized 
again here to assist commenters.
    1. Although the revised proposed rule would permit licensees to 
make plant changes that result in very small risk increases, the NRC is 
requesting stakeholder comments on whether the rule should allow plant 
changes that increase risk at all. Instead of the risk acceptance 
criteria allowing very small risk increases, should the risk acceptance 
criteria in final rule require that the net effect of plant changes 
made under Sec.  50.46a be risk neutral or risk beneficial? The NRC 
requests stakeholders to provide comments on the use of risk acceptance 
criteria that would not allow a cumulative increase in risk for plant 
changes made under Sec.  50.46a. (See Section V.E.4.b of this 
document.)
    2. Because of the difference in the risk acceptance criteria 
metrics used for currently operating reactors (LERF) and new reactors 
(LRF), the NRC is seeking public comments on whether LRF should be the 
metric of concern in lieu of LERF for new reactor applicants (or 
licensees) implementing the Sec.  50.46a alternative ECCS requirements. 
Because the LRF goal for new reactors is a decade lower than the 
10<SUP>-</SUP>\5\ per reactor year LERF reference value above which a 
facility would be limited to very small increases, should the 
definition of what constitutes ``very small increase'' and ``minimal 
increase'' for LRF (for new reactors) be a full decade lower than those 
defined for LERF (for existing reactors) or should the definition be 
based on relative change in LRF? (See Section V.J of this document.)
    3. In Sec.  50.46a(e)(4)(i) of the revised proposed rule the NRC 
proposes coolable core geometry as a high level performance-based ECCS 
analysis acceptance criterion for beyond-TBS LOCAs. Applicants would be 
allowed to justify appropriate metrics to demonstrate coolable geometry 
or use the current metrics (2200 [deg]F PCT and 17 percent MLO). 
However, the NRC acknowledges that it would be expensive and time-
consuming for industry to develop the necessary experimental and 
analytical data to justify alternative acceptance criteria as a 
surrogate for demonstrating coolable geometry. Because of the 
difficulty in demonstrating alternative metrics, the NRC is requesting 
stakeholder comments on whether the final Sec.  50.46a rule should 
retain the coolable geometry criterion for beyond-TBS breaks. Retaining 
coolable geometry would give licensees the option to demonstrate 
alternative coolable geometry metrics or use the current metric (2200 
[deg]F PCT and 17 percent MLO). If the NRC removed the coolable 
geometry criterion, the beyond-TBS acceptance criteria would be the 
same as the acceptance criteria for TBS and smaller breaks (2200 [deg]F 
PCT and 17 percent MLO). The NRC will evaluate stakeholder comments on 
this question before deciding which beyond-TBS acceptance criteria to 
include in the final rule. (See Section V.D.2 of this document.)

VII. Petition for Rulemaking, PRM-50-75

    In February 2002, the Nuclear Energy Institute submitted a petition 
for rulemaking (PRM-50-75) requesting the NRC to revise ECCS 
requirements by redefining the large break LOCA (ML020630082). Notice 
of that petition was published in the Federal Register for public 
comment on April 8, 2002 (67 FR 16654). The petition requested the NRC 
to amend Sec.  50.46 and Appendices A and K of Part 50 to allow 
licensees to use as an alternative to the double-ended rupture of the 
largest pipe in the

[[Page 40039]]

RCS, a maximum LOCA break size of ``up to and including an alternate 
maximum break size that is approved by the Director of the Office of 
Nuclear Reactor Regulation.'' Seventeen sets of comments were received, 
mostly from the power reactor industry in favor of granting the 
petition. A few stakeholders were concerned about potential impacts on 
defense-in-depth or safety margins if significant changes were made to 
reactor designs based upon use of a smaller break size. The NRC 
considered the public comments, evaluated the petition, and published a 
notice in the Federal Register resolving the petition and closing the 
PRM-50-75 docket. (See 73 FR 66000; November 6, 2008.) The NRC 
concluded that the issue raised by the petitioner should be considered 
in the rulemaking process. Documents related to the resolution of PRM-
50-75 are available at http://www.regulations.gov under docket ID: NRC-
2002-0018. The NRC is addressing the issues raised by the petitioner 
and stakeholders in this rulemaking.

VIII. Section-by-Section Analysis of Changes

A. Section 50.34--Contents of Application; Technical Information

    Paragraph (a)(4)(i) of this section would specify that Sec.  50.46a 
contains alternative ECCS requirements that licensees could choose to 
apply to reactors whose construction permits were issued before the 
effective date of the rule. This section also states that applicants 
for construction permits for facilities which may be issued after the 
effective date of the rule could also choose to apply the Sec.  50.46a 
alternative ECCS requirements to preliminary analysis and evaluation of 
the design if the applicant demonstrates that the facility is similar 
to the designs of facilities licensed before the effective date of the 
rule.
    Paragraph (a)(4)(ii) would specify that applicants for construction 
permits for facilities which may be issued after the effective date of 
the rule who have not demonstrated that the facility is similar to the 
designs of facilities licensed before the effective date of the rule 
may not apply the Sec.  50.46a alternative ECCS requirements in the 
preliminary analysis and evaluation of the design.
    Paragraph (b)(4)(i) of this section would specify that applicants 
for operating licenses for facilities which may be issued before the 
effective date of the rule could choose to apply the Sec.  50.46a 
alternative ECCS requirements in the final analysis and evaluation of 
the design. This section also states that applicants for operating 
licenses for facilities which may be issued after the effective date of 
the rule could also choose to apply the Sec.  50.46a alternative ECCS 
requirements to final analysis and evaluation of the design if the 
applicant demonstrates that the facility is similar to the designs of 
facilities licensed before the effective date of the rule.
    Paragraph (b)(4)(ii) would specify that applicants for operating 
licenses for facilities which may be issued after the effective date of 
the rule who have not demonstrated that the design is similar to the 
designs of facilities licensed before the effective date of the rule 
may not apply the Sec.  50.46a alternative ECCS requirements in the 
final analysis and evaluation of the design.

B. Section 50.46--Acceptance Criteria for Emergency Core Cooling 
Systems for Light-Water Nuclear Power Plants

    Paragraph (a) of this section would specify that emergency core 
cooling systems of BWRs and PWRs licensed before the effective date of 
the rule must be designed under Sec.  50.46 or Sec.  50.46a. Paragraph 
(a) would also specify that emergency core cooling systems of BWRs and 
PWRs licensed after the effective date of the rule could also choose to 
comply with the Sec.  50.46a alternative ECCS requirements if the 
applicant or licensee demonstrates that the design is similar to the 
designs of LWR facilities licensed before the effective date of the 
rule.

C. Existing Section 50.46a--Acceptance Criteria for Reactor Coolant 
System Venting Systems, Is Administratively Redesignated as Section 
50.46b

D. Section 50.46a--Alternative Acceptance Criteria for Emergency Core 
Cooling Systems for Light-Water Reactors

    Paragraph (a) of this section would provide definitions for terms 
used in other parts of this section. The definition of evaluation model 
in Sec.  50.46a(a)(2) is the same as in Sec.  50.46. The definition of 
loss-of-coolant accidents in Sec.  50.46a(a)(3) is based on the 
existing definition in Sec.  50.46 but has been modified to indicate 
that pipe breaks larger than the TBS are beyond design-basis accidents.
    The new definitions are:
    (1) Changes enabled by this section, which means changes to the 
facility, technical specifications, or procedures that comply with 
Sec.  50.46a but do not comply with Sec.  50.46;
    (4) Operating configuration, which is used in Sec.  50.46a(d)(5) to 
specify plant equipment availability conditions that must be analyzed 
for conformance with acceptance criteria; and
    (5) Transition break size (TBS), which is used to distinguish 
between requirements applicable to pipe breaks at or below this size 
from those applicable to pipe breaks above this size.
    Paragraph (b) would provide the applicability and scope of the 
requirements of this section. Proposed Sec.  50.46a would apply to 
currently licensed light-water nuclear power reactors (licensed before 
the effective date of the rule). Proposed Sec.  50.46a would also apply 
to LWRs licensed after the effective date of the rule which have been 
demonstrated to be similar to the designs of LWR facilities licensed 
before the effective date of the rule. Its requirements would be in 
addition to any other requirements applicable to ECCS set forth in 10 
CFR 50, with the exception of Sec.  50.46.
    Paragraph (c)(1) would specify the contents of initial licensee 
applications for implementing the alternative ECCS requirements in 
Sec.  50.46a. Paragraph (c)(1)(i) would require that an application 
contain a written evaluation demonstrating applicability of the results 
in NUREG-1829 and NUREG-1903 to the licensee's facility. However, if 
the facility differs significantly from the facilities analyzed in 
NUREG-1903, the application must contain a plant specific analysis 
demonstrating that the risk of seismically-induced LOCAs larger than 
the TBS is comparable to or less than the seismically-induced LOCA risk 
associated with the NUREG-1903 results. Paragraph (c)(1)(ii) would 
require identification of the NRC-approved analysis methods to be used 
to comply with the ECCS analysis requirements and acceptance criteria 
in paragraph (e). Paragraph (c)(1)(iii) would require a description of 
the risk-informed evaluation process used to determine whether proposed 
changes to the facility meet the requirements for risk-informed 
evaluations in paragraph (f). Paragraph (c)(1)(iv) would require 
licensees who wish to make changes enabled by Sec.  50.46a without 
prior NRC approval to submit a description of the risk-informed 
evaluation process and the PRA or non-PRA risk-assessment methods to be 
used to determine the acceptability of such changes. The licensee's 
process must be capable of demonstrating that all of the acceptance 
criteria in paragraph (f) will be met for each change. Paragraph 
(c)(1)(v) would require licensees who wish to adopt the alternative 
ECCS requirements in Sec.  50.46a to submit a description of all non 
safety equipment to be relied on to mitigate the consequences of a LOCA 
larger than the TBS.

[[Page 40040]]

    Paragraph (c)(2) states that applicants for a construction permit, 
operating license, design approval, design certification, manufacturing 
license, or combined license seeking to implement the requirements of 
this section shall, in addition to the information that would be 
required by paragraph (c)(1) of this section, submit an analysis 
demonstrating why the proposed reactor design is similar to the designs 
of currently operating reactors.
    Paragraph (c)(3) specifies the acceptance criteria for approval of 
applications to comply with Sec.  50.46a. Paragraph (c)(3)(i) would 
require the evaluation submitted under paragraph (c)(1)(i) to 
demonstrate that the NUREG-1829 results are applicable to the facility, 
and the risk of seismically-induced LOCAs larger than the TBS is 
comparable to or less than the seismically-induced LOCA risk associated 
with the NUREG-1903 results. Paragraph (c)(3)(ii) would require that 
the method(s) for demonstrating compliance with the ECCS acceptance 
criteria in paragraphs (e)(3) and (e)(4) of this section meet the 
requirements in paragraphs (e)(1) and (e)(2). Paragraph (c)(3)(iii) 
would require that the risk-informed evaluation process the licensee 
proposes to use for making changes enabled by this section be adequate 
for determining whether the acceptance criteria in paragraph (f) of 
this section have been met. Paragraph (c)(3)(iv) would require that all 
non safety equipment credited for demonstrating compliance with the 
ECCS acceptance criteria is identified and listed as such in plant 
Technical Specifications. Paragraph (c)(3)(v) would require that the 
reactor design for all applicants other than those holding operating 
licenses issued before the effective date of the rule be similar to the 
designs of current operating reactors and the applicant's proposed TBS 
is consistent with the technical basis for Section 50.46a.
    Paragraph (d) specifies the requirements with which licensees would 
be required to comply during facility operation after implementing 
Sec.  50.46a.
    Paragraph (d)(1) would require that the ECCS models be maintained 
to comply with the ECCS acceptance criteria in paragraphs (e)(1) and 
(e)(2) of this section.
    Paragraph (d)(2) would require that the licensee maintain leak 
detection equipment available at the facility and identify, monitor, 
and quantify leakage to reduce the likelihood of a LOCA larger than the 
TBS.
    Paragraph (d)(3) would require that changes to the facility, 
technical specifications, or procedures enabled by Sec.  50.46a be 
evaluated by a risk-informed evaluation process which demonstrates that 
acceptance criteria in Sec.  50.46a(f) are met.
    Paragraph (d)(4), would require licensees to maintain and upgrade 
its PRA analyses no less often than once every 2 refueling outages. 
Maintaining a PRA involves the update of PRA models to reflect facility 
changes such as plant modifications, procedure changes, or changes in 
plant performance data. Upgrading a PRA involves incorporating into the 
PRA models a new methodology or significant changes in scope or 
capability that impact the significant accident sequences. Risk 
assessments would be required to continue to meet the quality 
requirements in Sec. Sec.  50.46a(f)(4) and (f)(5). Licensees would be 
required to take action to ensure that facility design and operation 
continue to be consistent with the risk assessment assumptions used to 
meet the acceptance criteria in Sec. Sec.  50.46a(f)(2) or (f)(3). Any 
necessary changes to the facility caused by maintaining or upgrading 
risk assessments would not be deemed backfitting.
    Paragraph (d)(5) would require licensees to control plant operation 
to ensure that for LOCAs larger than the TBS, operation in a plant 
operating configuration not demonstrated to meet the acceptance 
criteria in paragraph (e)(4) would not exceed a total of fourteen days 
in any 12 month period.
    Paragraph (d)(6) would require licensees to perform an evaluation 
to determine the effect of all planned facility changes and would 
prohibit licensees from implementing any facility change that would 
invalidate the evaluation performed pursuant to Sec.  50.46a(c)(1)(i) 
demonstrating the applicability to the licensee's facility of the 
generic results in NUREG-1829 and NUREG-1903.
    Paragraph (e) would provide the ECCS evaluation model requirements, 
analysis requirements, and acceptance criteria for the two LOCA break 
size regions.
    Paragraph (e)(1) would specify model and analysis requirements for 
breaks smaller than or equal to the TBS. These requirements are the 
same as the current requirements for LOCA analysis models in existing 
Sec.  50.46.
    Paragraph (e)(2) would specify model and analysis requirements for 
breaks larger than the TBS. Methods for evaluating ECCS cooling 
performance for breaks larger than the TBS must be approved by the NRC. 
However the analysis for breaks larger than the TBS may be performed 
using more realistic analysis inputs and assumptions than those 
required for breaks smaller than or equal to the TBS. Analysis of 
breaks larger than the TBS need not assume a coincident single failure 
of mitigation equipment or loss of offsite power. Non-safety grade 
equipment may also be credited in analyses of breaks larger than the 
TBS provided that onsite power can supplied to that equipment in a 
reasonable time in the event offsite power is lost.
    Paragraph (e)(3) would provide ECCS acceptance criteria for LOCAs 
smaller than or equal to the TBS. The criteria specified would be the 
same as the current requirements in Sec.  50.46(b).
    Paragraph (e)(4) would provide ECCS acceptance criteria for LOCAs 
larger than the TBS. These acceptance criteria would be based on 
maintaining a coolable geometry in the core and demonstrating long term 
cooling capability and are less prescriptive than the criteria 
presently used for LOCA analysis.
    Paragraph (e)(5) would provide that the Director of the Office of 
Nuclear Reactor Regulation may impose restrictions on reactor operation 
if ECCS requirements are not met. This paragraph would be added to be 
consistent with existing Sec.  50.46 which also contains this 
requirement.
    Paragraph (f) would provide requirements for implementing changes 
to the facility, technical specifications, and procedures under Sec.  
50.46a.
    Paragraph (f)(1) would specify that licensees may make changes 
without NRC approval if:
    (i) The changes are permitted under Sec.  50.59;
    (ii) A risk-informed evaluation process has been submitted by the 
licensee and reviewed and approved by the NRC under Sec.  
50.46a(c)(1)(iv); and
    (iii) The change does not invalidate the evaluation performed under 
Sec.  50.46a(c)(1)(i) of the applicability of the results in NUREG-1829 
and NUREG-1903 to the licensee's facility.
    Paragraph (f)(2) would state that for plant changes not permitted 
under paragraph (f)(1), licensees must submit an application for a 
license amendment under Sec.  50.90. The application must contain:
    (i) The information required under Sec.  50.90;
    (ii) For reactors licensed before the effective date of the rule, 
information from the risk-informed evaluation demonstrating that the 
total increases in core damage frequency and large early release 
frequency are very small and the overall risk remains small, and that 
the risk-informed change criteria in paragraph (f)(3) are met;

[[Page 40041]]

    (iii) For all applicants other than those holding operating 
licenses issued before the effective date of the rule, information from 
the risk-informed evaluation demonstrating that the total increases in 
core damage frequency and large release frequency are very small, the 
overall risk remains small, and the criteria in paragraph (f)(3) of 
this section are met;
    (iv) An evaluation of the cumulative effect of previous changes 
that have increased risk but have met the acceptance criteria. If more 
than one plant change is combined, including plant changes not enabled 
by Sec.  50.46a, into a group for the purposes of evaluating acceptable 
risk increases, the evaluation of each individual change shall be 
performed along with the evaluation of combined changes;
    (v) Information demonstrating that the ECCS analysis acceptance 
criteria in paragraphs (e)(3) and (e)(4) are met; and
    (vi) Information demonstrating that the proposed change will not 
increase the LOCA frequency of the facility (including the frequency of 
seismically-induced LOCAs) by an amount that would invalidate the 
applicability to the facility of the generic seismic studies (NUREG-
1829, ``Estimating Loss-of-Coolant Accident (LOCA) Frequencies through 
the Elicitation Process'', March 2008 and NUREG-1903, ``Seismic 
Considerations for the Transition Break Size'', February 2008) that 
support the technical basis for Sec.  50.46a.
    Paragraph (f)(3) would specify requirements for all plant changes. 
Paragraph (f)(3)(i) would require that defense-in-depth is maintained. 
Paragraph (f)(3)(ii) would require that adequate safety margins are 
maintained. Paragraph (f)(3)(iii) would require that adequate 
performance-measurement programs will be implemented. Paragraph 
(f)(3)(iii) provides criteria on the specific attributes required to 
meet the performance measurement requirements.
    Paragraph (f)(2) does not require use of PRA in assessing risks 
associated with the proposed changes. To the extent that PRA is used, 
paragraph (f)(4) of the revised proposed rule would identify specific 
technical requirements for the risk-informed assessment.
    (i) Address initiating events from sources both internal and 
external to the plant and for all modes of operation, including low 
power and shutdown modes, that would affect the regulatory decision in 
a substantial manner;
    (ii) Reasonably represent the current configuration and operating 
practices at the plant;
    (iii) Have sufficient technical adequacy (including consideration 
of uncertainty) and level of detail to provide confidence that the 
total risk estimate and the change in total risk estimate adequately 
reflect the plant and the effect of the proposed change on risk; and
    (iv) Be determined, through peer review, to meet industry standards 
for PRA quality that have been endorsed by NRC.
    Paragraph (f)(5) would require that to the extent that risk 
assessment methods other than PRA are used to develop quantitative or 
qualitative estimates of changes to risk in the risk-informed 
evaluation, an integrated, systematic process must be used. All aspects 
of the analyses must reasonably reflect the current plant configuration 
and operating practices, and applicable plant and industry operating 
experience.
    Paragraph (g) would provide the requirements for making reports to 
the NRC.
    Paragraph (g)(1) would require reporting of all errors or changes 
to ECCS analyses at least annually as specified in Sec.  50.4. For 
significant changes or errors, licensees would be required to report 
within 30 days including a schedule for reanalysis or other action as 
needed to show compliance with ECCS requirements. Under paragraph 
(g)(1)(i), for LOCAs involving pipe breaks equal to or smaller than the 
TBS, significant changes would be defined as a change in peak cladding 
temperature of greater than 50 [deg]F. Under paragraph (g)(1)(ii), for 
LOCAs involving pipe breaks larger than the TBS, a significant change 
would be defined as one resulting in a significant reduction in the 
capability to meet the ECCS acceptance criteria in Sec.  50.46a(e)(4).
    Paragraph (g)(2) would set forth reporting requirements with 
respect to the PRA maintenance and upgrading that would be required by 
Sec.  50.46a(d)(4). When maintaining and upgrading the PRA, Sec.  
50.46a(g)(2) would require the licensee to report changes to the NRC 
within 60 days if the acceptance criteria in Sec. Sec.  
50.46a(f)(2)(ii) or (f)(2)(iii) (for new reactors) are exceeded. This 
provision would also require the report to include a schedule for 
implementation of any corrective actions necessary to bring plant 
operation or design back into compliance with the acceptance criteria.
    Paragraph (g)(3) would contain reporting requirements for plant 
changes made under Sec.  50.46a(f)(1) involving minimal risk. A short 
description of these changes would be reported every 24 months.
    Paragraph (h) would provide documentation requirements for plant 
changes. Following implementation of Sec.  50.46a, licensees would be 
required to maintain records sufficient to demonstrate compliance with 
the requirements in Sec.  50.46a and Sec.  50.71.
    Paragraphs (i) through (l) would be reserved for future use.
    Paragraph (m) would provide that changes made by the NRC to the TBS 
and all changes required to return a facility to compliance with the 
acceptance criteria after a change in the TBS are not deemed to be 
backfitting under 10 CFR 50.109.

E. Section 50.109--Backfitting

    This section would be modified to provide that changes made by the 
NRC to the TBS and changes made by licensees to continue to comply with 
Sec.  50.46a are not deemed to be backfitting under 10 CFR 50.109.

F. Appendix A to Part 50--General Design Criteria for Nuclear Power 
Plants

    Five of the general design criteria contained in Appendix A would 
be modified to remove the requirement to assume a single failure and a 
loss-of-offsite power in the systems subject to these criteria for pipe 
breaks larger than the TBS up to and including the DEGB of the largest 
RCS pipe for those plants implementing Sec.  50.46a. The specific 
criteria are: GDC 17, Electrical power systems, GDC 35, Emergency core 
cooling, GDC 38, Containment heat removal, GDC 41, Containment 
atmosphere cleanup, and GDC 44, Cooling water systems. General Design 
Criterion 50, Containment design basis, would also be modified to 
specify that for plants under Sec.  50.46a, leak tight containment 
capability should be maintained for ``realistically'' calculated 
temperatures and pressures for LOCAs larger than the TBS.

G. Section 52.47--Contents of Applications; Technical Information

    Paragraph (a)(4) of this section would be amended to specify the 
technical information to be submitted in an application for a standard 
design certification for a nuclear power facility filed separately from 
the filing of an application for a construction permit or combined 
license for such a facility.
    New paragraph (a)(4)(i) would to specify that analyses of emergency 
core cooling systems and the need for high point vents for standard 
designs certified after the effective date of the Sec.  50.46a rule 
must be performed under the requirements of either Sec.  50.46 or Sec.  
50.46a (for ECCS performance) and Sec.  50.46b (for reactor coolant 
system high point vents) if the standard design is demonstrated to be 
similar to the

[[Page 40042]]

designs of reactors licensed before the effective date of Sec.  50.46a.
    New paragraph (a)(4)(ii) would specify that analyses of emergency 
core cooling systems and the need for high point vents for standard 
designs certified after the effective date of the Sec.  50.46a rule 
must be performed under the requirements of Sec.  50.46 (for ECCS 
performance) and Sec.  50.46b (for reactor coolant system high point 
vents) if the standard design is not demonstrated to be similar to the 
designs of reactors licensed before the effective date of Sec.  50.46a.

H. Section 52.79--Contents of Applications; Technical Information in 
Final Safety Analysis Report

    In this section paragraph (a)(5) would be amended to specify the 
technical information to be submitted in the final safety analysis 
report for an application for a combined license for a nuclear power 
facility.
    New paragraph (a)(5)(i) would specify that analyses of emergency 
core cooling systems and the need for high point vents for plants 
licensed after the effective date of the Sec.  50.46a rule must be 
performed under the requirements of either Sec.  50.46 or Sec.  50.46a 
(for ECCS performance) and Sec.  50.46b (for reactor coolant system 
high point vents) if the design is demonstrated to be similar to the 
designs of reactors licensed before the effective date of Sec.  50.46a.
    New paragraph (a)(5)(ii) would specify that analyses of emergency 
core cooling systems and the need for high point vents for plants 
licensed after the effective date of the Sec.  50.46a rule must be 
performed under the requirements of Sec.  50.46 (for ECCS performance) 
and Sec.  50.46b (for reactor coolant system high point vents) if the 
design is not demonstrated to be similar to the designs of reactors 
licensed before the effective date of Sec.  50.46a.

I. Section 52.137--Contents of Applications; Technical Information

    Paragraph (a)(4) of this section would be amended to specify the 
technical information to be submitted in an application for approval of 
a standard design for a nuclear power facility.
    New paragraph (a)(4)(i) would specify that analyses of emergency 
core cooling systems and the need for high point vents for designs 
approved after the effective date of the Sec.  50.46a rule must be 
performed under the requirements of either Sec.  50.46 or Sec.  50.46a 
(for ECCS performance) and Sec.  50.46b (for reactor coolant system 
high point vents) if the design is demonstrated to be similar to the 
designs of reactors licensed before the effective date of Sec.  50.46a.
    New paragraph (a)(4)(ii) would specify that analyses of emergency 
core cooling systems and the need for high point vents for designs 
approved after the effective date of the Sec.  50.46a rule must be 
performed under the requirements of Sec.  50.46 (for ECCS performance) 
and Sec.  50.46b (for reactor coolant system high point vents) if the 
design is not demonstrated to be similar to the designs of reactors 
licensed before the effective date of Sec.  50.46a.

J. Section 52.157--Contents of Applications; Technical Information in 
Final Safety Analysis Report

    Paragraph (f)(1) of this section would be amended to specify the 
technical information to be submitted in the final safety analysis 
report for an application for issuance of a license authorizing 
manufacture of nuclear power reactors to be installed at sites not 
identified in the manufacturing license application.
    New paragraph (f)(1)(i) would specify that analyses of emergency 
core cooling systems and the need for high point vents for a license 
authorizing manufacture of nuclear power reactors issued after the 
effective date of the Sec.  50.46a rule must be performed under the 
requirements of either Sec.  50.46 or Sec.  50.46a (for ECCS 
performance) and Sec.  50.46b (for reactor coolant system high point 
vents) if the design is demonstrated to be similar to the designs of 
reactors licensed before the effective date of Sec.  50.46a.
    New paragraph (f)(1)(ii) would specify that analyses of emergency 
core cooling systems and the need for high point vents for a license 
authorizing manufacture of nuclear power reactors issued after the 
effective date of the Sec.  50.46a rule must be performed under the 
requirements of Sec.  50.46 (for ECCS performance) and Sec.  50.46b 
(for reactor coolant system high point vents) if the design is not 
demonstrated to be similar to the designs of reactors licensed before 
the effective date of Sec.  50.46a.

IX. Criminal Penalties

    For the purposes of Section 223 of the Atomic Energy Act (AEA), as 
amended, the NRC is issuing the proposed rule to amend Sec.  50.46, add 
Sec.  50.46a, redesignate existing Sec.  50.46a as Sec.  50.46b and 
amend Sec. Sec.  52.47, 52.79, 52.137, and 52.157 under one or more of 
sections 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 Part 50, are discussed in Sec.  50.111 and as 
they apply to the regulations in Part 52, are discussed in Sec.  
52.303.

X. Compatibility of Agreement State Regulations

    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 ``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 Code of Federal Regulations, and although an Agreement State may 
not adopt program elements reserved to 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.

XI. Availability of Documents

    Comments and other publicly available documents related to this 
rulemaking may be viewed electronically on the public computers located 
at the NRC's Public Document Room (PDR), O1 F21, One White Flint North, 
11555 Rockville Pike, Rockville, Maryland. The PDR reproduction 
contractor will copy documents for a fee.
    Publicly available documents are available electronically at the 
NRC's Electronic Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this site, the public can gain entry into the NRC's 
Agencywide Document Access and Management System (ADAMS), which 
provides text and image files of NRC's public documents. If you do not 
have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC Public Document Room (PDR) 
Reference staff at 1-800-397-4209, (301) 415-4737 or by e-mail to 
<a href="/cdn-cgi/l/email-protection#cebeaabc8ea0bcade0a9a1b8"><span class="__cf_email__" data-cfemail="38485c4a78564a5b165f574e">[email protected]</span></a>. The NRC is making the documents identified below available 
to interested persons through one or more of the following methods as 
indicated.
    Public Document Room (PDR). The NRC Public Document Room is located 
at Public File Area O-F21, One White Flint North, 11555 Rockville Pike, 
Rockville, Maryland.
    Federal eRulemaking Portal. Go to http://www.regulations.gov and 
search for documents filed under Docket ID NRC-2004-0006. Address 
questions about NRC dockets to Carol Gallagher (301) 415-5905; e-mail 
<a href="/cdn-cgi/l/email-protection#5a193b283536741d3b36363b3d3f281a342839743d352c"><span class="__cf_email__" data-cfemail="0a496b786566244d6b66666b6d6f784a647869246d657c">[email protected]</span></a>.
    NRC's Electronic Reading Room (ERR). The NRC's public electronic

[[Page 40043]]

reading room is located at http://www.nrc.gov/reading-rm.html.

----------------------------------------------------------------------------------------------------------------
               Document                    PDR                     Web                        Err (Adams)
----------------------------------------------------------------------------------------------------------------
Initial Proposed Rule (70 FR 67598)..          X   NRC-2004-0006.....................  ML091060434
NRC Report--Seismic Considerations             X   NRC-2004-0006.....................  ML053470439
 for the Transition Break Size
 (December 2006).
Letter from Graham B. Wallis (ACRS)            X   X.................................  ML063190465
 to Dale E. Klein, ``Draft Final Rule
 To Risk-Inform 10 CFR 50.46,
 `Acceptance Criteria For Emergency
 Core Cooling Systems For Light-Water
 Nuclear Power Reactors' '' (November
 16, 2006).
SECY-07-0082--Rulemaking to Make Risk-         X   X.................................  ML070180692
 Informed Changes to
 Loss[dash]of[dash]Coolant Accident
 Technical Requirements; 10 CFR
 50.46a ``Alternative Acceptance
 Criteria for Emergency Core Cooling
 Systems for Light[dash]Water Nuclear
 Power Reactors,'' (May 16, 2007).
Commission SRM on SECY-07-0082                 X   X.................................  ML072220595
 (August 10, 2007).
Memorandum from Luis A. Reyes to NRC           X   X.................................  ML080370355
 Commissioners, ``Plans And Schedule
 For The Rulemaking On Risk-Informed
 Changes To Loss-of-Coolant Accident
 Technical Requirements (April 1,
 2008).
NUREG-1488--Revised Livermore Seismic          X   X.................................  ML052640591
 Hazard Estimates for Sixty-Nine
 Nuclear Power Plant Sites East of
 the Rocky Mountains (April 1994).
NUREG-1829--Estimating Loss-of-                X   X.................................  ML051520574
 Coolant Accident (LOCA) Frequencies
 Through the Elicitation Process
 (Draft Report; June 2005).
NUREG-1829--Estimating Loss-of-                X   X.................................  ML082250436
 Coolant Accident (LOCA) Frequencies
 Through the Elicitation Process
 (Final Report; March 2008).
NUREG-1903--Seismic Considerations             X   X.................................  ML080880140
 for the Transition Break Size
 (February 2008).
NRC White Paper--Plant-Specific                X   X.................................  ML090350757
 Applicability of 10 CFR 50.46a
 Technical Basis (February 2009).
Memorandum from Arthur T. Howell to            X   X.................................  ML022740211
 William F. Kane, ``Degradation of
 the Davis-Besse Nuclear Power
 Station Reactor Pressure Vessel Head
 Lessons-Learned Report''; (September
 30, 2002).
Regulatory Analysis..................          X   X.................................  ML091050748
----------------------------------------------------------------------------------------------------------------

XII. Plain Language

    The Presidential memorandum dated June 1, 1998, entitled ``Plain 
Language in Government Writing'' directed that the Government's writing 
be in plain language. This memorandum was published on June 10, 1998 
(63 FR 31883). The NRC requests comments on the proposed rule 
specifically with respect to the clarity and reflectiveness of the 
language used. Comments should be sent to the address listed under the 
ADDRESSES caption of the preamble.

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. In this proposed rule, the 
NRC proposes to use the following Government-unique standard: 10 CFR 
50.46a. The NRC notes the ongoing development of voluntary consensus 
standards on PRAs, such as the ASME/ANS RA-Sa-2009 consensus standard 
on Probabilistic Risk Assessment for Nuclear Power Plant Applications. 
The Government standards would allow the use of voluntary consensus 
standards, but would not require their use. The NRC does not believe 
that these other standards are sufficient to specify the necessary 
requirements for licensees who wish to modify plant ECCS analysis 
methods and nuclear power reactor designs based on the results of 
probabilistic risk analysis. The NRC is not aware of any voluntary 
consensus standard addressing risk-informed ECCS design and consequent 
changes in a light-water power reactor facility, technical 
specifications, or procedures that could be used instead of the 
proposed Government-unique standard. The NRC will consider using a 
voluntary consensus standard if an appropriate standard is identified. 
If a voluntary consensus standard is identified for consideration, the 
submittal should explain how the voluntary consensus standard is 
comparable and why it should be used instead of the proposed 
Government-unique standard.

XIV. Finding of No Significant Environmental Impact: Environmental 
Assessment

    The NRC 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. The basis for this determination is as follows:
    This action stems from the NRC's ongoing efforts to risk-inform its 
regulations. If adopted, the proposed rule would establish a voluntary 
alternative set of risk-informed requirements for emergency core 
cooling systems. The alternative requirements are less stringent in the 
area of large break loss-of-coolant accidents (LOCAs). Using the 
alternative ECCS requirements will provide some licensees with 
opportunities to change various aspects of plant design to increase 
operational flexibility, increase power, or decrease costs. Licensee 
actions taken under the proposed rule could either decrease the 
probability of an accident or increase the probability of an accident 
by a very small amount. Mitigation of LOCAs of all sizes would still be 
required but with less redundancy and margin for the larger, low 
probability breaks. Increases in risk, if any, would be required to be 
very small so that adequate assurance of public health and safety is 
maintained. When considered together, the net effect of the licensee 
actions is expected to have an insignificant effect on accident 
probability.
    Thus, the proposed action would not significantly increase the 
probability or consequences of an accident, when considered in a risk-
informed manner. No changes would be made in the types or quantities of 
radiological effluents that may be released offsite, and there is no 
significant increase in public radiation exposure because there is no 
change to facility operations that could create a new or significantly 
affect a

[[Page 40044]]

previously analyzed accident or release path.
    With regard to non-radiological impacts, no changes would be made 
to non-radiological plant effluents and there would be no changes in 
activities that would adversely affect the environment. Therefore, 
there are no significant non-radiological impacts associated with the 
proposed action.
    The primary alternative would be the no action alternative. The no 
action alternative, at worst, would result in no changes to current 
levels of safety, risk, or environmental impact. The no action 
alternative would also prevent licensees from making certain plant 
modifications that could be implemented under the proposed rule that 
could increase plant safety, increase operational flexibility, or 
decrease costs. The no action alternative would also maintain existing 
regulatory burdens for which there could be little or no safety, risk, 
or environmental benefits.
    The determination of this environmental assessment is that there 
will be no significant offsite impact to the public from this action. 
However, public stakeholders should note that the NRC is seeking public 
participation on this assessment. Comments on any aspect of the 
environmental assessment may be submitted to the NRC as indicated under 
the ADDRESSES heading of this document.
    The NRC has sent a copy of the environmental assessment and this 
proposed rule to every State Liaison Officer and requested their 
comments on the environmental assessment.

XV. Paperwork Reduction Act Statement

    This proposed rule amends information collection requirements 
contained in 10 CFR part 50 that are subject to the Paperwork Reduction 
Act of 1995 (44 U.S.C. 3501 et seq). These information collection 
requirements have been submitted to the Office of Management and Budget 
(OMB) for approval. Existing requirements were approved by the Office 
of Management and Budget, control number 3150-0011.
    Type of submission: Revision.
    The title of the information collection: 10 CFR part 50--Domestic 
Licensing of Production and Utilization Facilities.
    The form number if applicable: Not applicable.
    How often the collection is required: Annually.
    Who will be required or asked to report: Licensees authorized to 
operate a nuclear power reactor or applicants for standard design 
certifications, combined licenses, standard design approvals or 
manufacturing licenses who have been approved to implement the risk-
informed alternative requirements in 10 CFR 50.46a for analyzing the 
performance of emergency core cooling systems during loss-of-coolant 
accidents.
    An estimate of the number of annual responses: 12.
    The estimated number of annual respondents: 6.
    An estimate of the total number of hours needed annually to 
complete the requirement or request: 53,388 hours total, including 
48,000 hours for reporting (an average of 8,000 hours per respondent) + 
5,388 hours recordkeeping (an average of 898 hours per recordkeeper).
    Abstract: The Nuclear Regulatory Commission (NRC) proposes to amend 
its regulations to permit applicants for and/or holders of power 
reactor operating licenses, standard design certifications, combined 
licenses, standard design approvals or manufacturing licenses to choose 
to implement a risk-informed alternative to the current requirements 
for analyzing the performance of emergency core cooling systems (ECCS) 
during loss-of-coolant accidents (LOCAs). In addition, the proposed 
rule would establish procedures and criteria for making changes in 
plant design and procedures based upon the results of the new analyses 
of ECCS performance during LOCAs. A licensee or applicant choosing to 
use the provisions of Section 50.46a would be required to submit a 
license amendment request with the required information, using the 
existing processes in Section 50.34 and Section 50.90.
    The U.S. Nuclear Regulatory Commission 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?
    A copy of the OMB clearance package may be viewed free of charge at 
the NRC Public Document Room, One White Flint North, 11555 Rockville 
Pike, Room O-1 F21, Rockville, MD 20852. The OMB clearance package and 
rule are available at the NRC worldwide Web site: http://www.nrc.gov/public-involve/doc-comment/omb/index.html for 60 days after the 
signature date of this notice.
    Send comments on any aspect of these proposed information 
collections, including suggestions for reducing the burden and on the 
above issues, by September 9, 2009 to the Records and FOIA/Privacy 
Services Branch (T-5 F53), U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, or by Internet electronic mail to 
<a href="/cdn-cgi/l/email-protection#3f767179707c7073737a7c6b6c116d5a4c504a4d5c5a7f716d7c11585049"><span class="__cf_email__" data-cfemail="7831363e373b3734343d3b2c2b562a1d0b170d0a1b1d38362a3b561f170e">[email protected]</span></a> and to the Desk Officer, Christine Kymn, 
Office of Information and Regulatory Affairs, NEOB-10202, (3150-0011), 
Office of Management and Budget, Washington, DC 20503. Comments on the 
proposed information collection may also be submitted via the Federal 
eRulemaking Portal http://www.regulations.gov, docket <greek-i> NRC-
2004-0006. 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. You may also e-mail comments to 
<a href="/cdn-cgi/l/email-protection#1a59726873696e73747f45503445516377745a757778347f756a347d756c"><span class="__cf_email__" data-cfemail="e7a48f958e94938e8982b8adc9b8ac9e8a89a7888a85c9828897c9808891">[email protected]</span></a> or comment by telephone at (202) 395-
4638.
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

    The NRC has prepared a draft regulatory analysis on this proposed 
regulation. The analysis examines the costs and benefits of the 
alternatives considered by the NRC. The NRC requests public comment on 
the draft regulatory analysis. Availability of the regulatory analysis 
is provided in Section X of this document. Comments on the draft 
analysis may be submitted to the NRC as indicated under the ADDRESSES 
heading of this document.

XVII. Regulatory Flexibility Certification

    Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the NRC 
certifies that this rule will not, if promulgated, have a significant 
economic impact on a substantial number of small entities. This 
proposed rule affects only the licensing and operation of nuclear power 
plants. The companies that own these plants do not fall within the 
scope of the definition of ``small entities'' set forth in the 
Regulatory Flexibility Act or

[[Page 40045]]

the size standards established by the NRC (10 CFR 2.810).

XVIII. Backfit Analysis

    The NRC has determined that the proposed rule generally does not 
constitute backfitting as defined in the backfit rule, 10 CFR 
50.109(a)(1), and that three provisions of the proposed rule 
effectively excluding certain actions from the purview of the backfit 
rule, viz., Sec.  50.109(b)(2); Sec.  50.46a(d)(4), and Sec.  
50.46a(m), are appropriate. The basis for each of these determinations 
follows.
    The NRC has determined that the proposed rule does not constitute 
backfitting because it provides a voluntary alternative to the existing 
requirements in 10 CFR 50.46 for evaluating the performance of an ECCS 
for light-water nuclear power plants. A licensee may decide to either 
comply with the requirements of Sec.  50.46a, or to continue to comply 
with the existing licensing basis of their plant with respect to ECCS 
analyses. Therefore, the backfit rule does not require the preparation 
of a backfit analysis for the proposed rule.
    As discussed in Section V.B of this document, the NRC may undertake 
future rulemaking to revise the TBS based upon re-evaluations of LOCA 
frequencies occurring after the effective date of a final rule. A 
proposed amendment to the backfit rule, Sec.  50.109(b)(2), would 
provide that future changes to the TBS would not be subject to the 
backfit rule. The NRC has determined that there is no statutory bar to 
the adoption of such a provision. The NRC also believes that the 
proposed exclusion of such rulemakings from the backfit rule is 
appropriate. The NRC intends to revise the TBS in Sec.  50.46a rarely 
and only if necessary based upon public health and safety and/or common 
defense and security considerations. The NRC also does not regard the 
proposed exclusion as allowing the NRC to adopt cost-unjustified 
changes to the TBS. The NRC prepares a regulatory analysis for each 
substantive regulatory action which identifies the regulatory 
objectives of the proposed action, and evaluates the costs and benefits 
of proposed alternatives for achieving those regulatory objectives. The 
NRC has also adopted guidelines governing treatment of individual 
requirements in a regulatory analysis (69 FR 29187; May 21, 2004). The 
NRC believes that a regulatory analysis performed in accordance with 
these guidelines will be effective in identifying unjustified 
regulatory proposals. In addition, this revised proposed rulemaking as 
applied to licensees who have not yet transferred to Sec.  50.46a would 
not constitute backfitting for those licensees, inasmuch as the backfit 
rule does not protect a future applicant who has no reasonable 
expectation that requirements will remain static. The policies 
underlying the backfit rule apply only to licensees who have already 
received regulatory approval. Accordingly, the NRC concludes that the 
proposed exclusion in Sec.  50.109(b)(2) of future changes to the TBS 
from the requirements of the backfit rule is appropriate.
    As discussed in Section V.E of this document, Sec.  50.46a(d)(4) 
would require that a PRA used to demonstrate compliance with the risk 
acceptance criteria in Sec.  50.46a(f)(1) or (f)(2) be periodically re-
evaluated and updated, and that the licensee implement changes to the 
facility and procedures as necessary to ensure that the acceptance 
criteria continue to be met. To ensure that such a re-evaluation and 
updating of the PRA and any necessary changes to a facility and its 
procedures under Sec.  50.46a(d)(4) are not considered backfitting, 
Sec.  50.46a(d)(4) would provide that such a re-evaluation, updating, 
and changes are not deemed to be backfitting. The NRC believes that 
this exclusion from the backfit rule is appropriate, inasmuch as 
application of the backfit rule in this context would effectively favor 
increases in risk. This is because most facility and procedure changes 
involve an up-front cost to implement a change which must be recovered 
over the remaining operating life of the facility in order to be 
considered cost-effective. For example, assume that after a change is 
implemented, subsequent PRA analyses suggest that the change should be 
``rescinded'' (either the hardware is restored to the original 
configuration or the new configuration is not credited in design bases 
analyses) in order to maintain the assumed risk level. The cost/benefit 
determination of the second, ``restoring'' change must address the 
unrecovered cost of the first change and the cost of the second, 
``restoring'' change. In most cases, application of cost/benefit 
analyses in evaluating the second, ``restoring'' change would skew the 
decision-making in favor of accepting the existing plant with the 
higher risk. Accumulation of these incremental increases in risk does 
not appear to be an appropriate regulatory approach. Accordingly, the 
NRC concludes that the backfitting exclusion in Sec.  50.46a(d)(4) is 
appropriate.
    Section 50.46a(m) would provide that if the NRC changes the TBS 
specified in Sec.  50.46a, licensees who have evaluated their ECCS 
under Sec.  50.46a shall undertake additional actions to ensure that 
the relevant acceptance criteria for ECCS performance are met with the 
new TBSs, and that these licensee actions are not to be considered 
backfitting. Consequently, the NRC may require licensees to take action 
under Sec.  50.46a(m) without consideration of the backfit rule. The 
NRC has determined that there is no statutory bar to the adoption of 
this provision, and that the proposed provision represents a justified 
departure from the principles underlying the backfit rule. First, the 
NRC's decision on this matter recognizes that any future rulemaking to 
alter the TBS will require preparation of a regulatory analysis. As 
discussed, the regulatory analysis will ordinarily include a cost/
benefit analysis addressing whether the costs of the TBS redefinition 
are justified in view of the benefits attributable to the redefinition. 
Second, the licensee has substantial flexibility under the proposed 
rule to determine the actions (reanalysis, procedure and operational 
changes, design-related changes, or a combination thereof) necessary to 
demonstrate compliance with the relevant ECCS acceptance criteria. The 
performance-based approach of the revised proposed rule lends 
substantial flexibility to the licensee and may tend to reduce the 
burden associated with changes in the TBS. Accordingly, the NRC 
concludes that the backfitting exclusion in Sec.  50.46a(m) is 
appropriate.

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

[[Page 40046]]

to adopt the following amendments to 10 CFR parts 50 and 52.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

    1. The authority citation for part 50 continues to read as follows:

    Authority:  Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 
68 Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 
234, 83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 
2201, 2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 
206, 88 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 
5846); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note); Energy 
policy Act of 2005, Pub. L. No. 109-58, 119 Stat. 194 (2005). 
Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat. 
2951 as amended by Pub. L. 102-486, sec. 2902, 106 Stat. 3123 (42 
U.S.C. 5841). Section 50.10 also issued under secs. 101, 185, 68 
Stat. 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L. 91-
190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd), and 
50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 
U.S.C. 2138).
    Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 
185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and 
Appendix Q also issued under sec. 102, Pub. L. 91-190, 83 Stat. 853 
(42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 
204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91, and 
50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42 U.S.C. 
2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 
U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184, 68 
Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued under 
sec. 187, 68 Stat. 955 (42 U.S.C. 2237)

    2. In Sec.  50.34, paragraphs (a)(4) and (b)(4) are revised to read 
as follows:


Sec.  50.34   Contents of application; technical information.

    (a) * * *
    (4) A preliminary analysis and evaluation of the design and 
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.
    (i) Analysis and evaluation of ECCS cooling performance and the 
need for high point vents following postulated loss-of-coolant 
accidents must be performed under the requirements of either Sec.  
50.46 or Sec.  50.46a, and Sec.  50.46b for facilities whose operating 
licenses were issued after December 28, 1974, but before [EFFECTIVE 
DATE OF RULE], and for facilities for which construction permits may be 
issued after [EFFECTIVE DATE OF RULE] and are demonstrated under Sec.  
50.46a(c)(2) to have designs that are similar to the designs of 
reactors licensed before [EFFECTIVE DATE OF RULE].
    (ii) Analysis and evaluation of ECCS cooling performance and the 
need for high point vents following postulated loss-of-coolant 
accidents must be performed under the requirements of Sec.  50.46 and 
Sec.  50.46b for facilities for which construction permits may be 
issued after [EFFECTIVE DATE OF RULE] and are not demonstrated under 
Sec.  50.46a(c)(2) to have designs that are similar to the designs of 
reactors licensed before [EFFECTIVE DATE OF RULE].
* * * * *
    (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.
    (i) Analysis and evaluation of ECCS cooling performance following 
postulated LOCAs must be performed under the requirements of either 
Sec.  50.46 or Sec.  50.46a, and Sec.  50.46b for facilities whose 
operating licenses were issued after December 28, 1974, but before 
[EFFECTIVE DATE OF RULE], and for facilities whose operating licenses 
are issued after [EFFECTIVE DATE OF RULE] and are demonstrated under 
Sec.  50.46a(c)(2) to have designs that are similar to the designs of 
reactors licensed before [EFFECTIVE DATE OF RULE].
    (ii) Analysis and evaluation of ECCS cooling performance following 
postulated LOCAs must be performed under the requirements of Sec. Sec.  
50.46 and 50.46b for facilities whose operating licenses are issued 
after [EFFECTIVE DATE OF RULE] and are not demonstrated under Sec.  
50.46a(c)(2) to have designs that are similar to the designs of 
reactors licensed before [EFFECTIVE DATE OF RULE].
* * * * *
    3. In Sec.  50.46, paragraph (a) is amended by adding an 
introductory paragraph and revising paragraph (a)(1)(i) to read as 
follows:


Sec.  50.46   Acceptance criteria for emergency core cooling systems 
for light-water nuclear power plants.

    (a) Each boiling or pressurized light-water nuclear power reactor 
fueled with uranium oxide pellets within cylindrical zircalloy or ZIRLO 
cladding must be provided with an emergency core cooling system (ECCS). 
The ECCS system must be designed under the requirements of this section 
or Sec.  50.46a for facilities whose operating licenses were issued 
before [EFFECTIVE DATE OF RULE]; for facilities whose operating 
licenses, combined licenses under part 52 of this chapter, or 
manufacturing licenses under part 52 of this chapter are issued after 
[EFFECTIVE DATE OF RULE] and are demonstrated under Sec.  50.46a(c)(2) 
to have designs that are similar to the designs of reactors licensed 
before [EFFECTIVE DATE OF RULE]; and for design approvals and design 
certifications under part 52 of this chapter issued after [EFFECTIVE 
DATE OF RULE] that are demonstrated under Sec.  50.46a(c)(2) to have 
designs that are similar to the designs of reactors licensed before 
[EFFECTIVE DATE OF RULE]. The ECCS system must be designed under the 
requirements of this section for facilities whose operating licenses, 
combined licenses under part 52 of this chapter, or manufacturing 
licenses under part 52 of this chapter are issued after [EFFECTIVE DATE 
OF RULE] and are not demonstrated under Sec.  50.46a(c)(2) to have 
designs that are similar to the designs of reactors licensed before 
[EFFECTIVE DATE OF RULE]; and for design approvals and design 
certifications under part 52 of this chapter that are not demonstrated 
under Sec.  50.46a(c)(2) to have designs that are similar to the 
designs of reactors licensed before [EFFECTIVE DATE OF RULE].
    (1)(i) The ECCS system must be designed so that its calculated 
cooling performance following postulated LOCAs conforms to the criteria 
set forth in paragraph (b) of this section. ECCS cooling performance 
must be calculated in accordance with an acceptable evaluation model 
and must be calculated for a number of postulated LOCAs of different 
sizes, locations, and other properties sufficient to provide assurance 
that the most severe postulated LOCAs are calculated. Except as 
provided in paragraph (a)(1)(ii) of this section, the evaluation model 
must include sufficient supporting justification to show that the 
analytical technique realistically describes the behavior of the 
reactor system during a LOCA. Comparisons to applicable experimental 
data must be made and 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,

[[Page 40047]]

so that, when the calculated ECCS cooling performance is compared to 
the criteria set forth in paragraph (b) of this section, there is a 
high level of probability that the criteria would not be exceeded. 
Appendix K, Part II Required Documentation, sets forth the 
documentation requirements for each evaluation model. This section does 
not apply to a nuclear power reactor facility for which the 
certifications required under Sec.  50.82(a)(1) have been submitted.
* * * * *
    4. Section 50.46a is redesignated as Sec.  50.46b, and a new Sec.  
50.46a is added to read as follows:


Sec.  50.46a   Alternative acceptance criteria for emergency core 
cooling systems for light-water nuclear power reactors.

    (a) Definitions. For the purposes of this section:
    (1) Changes enabled by this section means changes to the facility, 
technical specifications, and procedures that satisfy the alternative 
ECCS analysis requirements under this section but do not satisfy the 
ECCS requirements under 10 CFR 50.46.
    (2) Evaluation model means the calculational framework for 
evaluating the behavior of the reactor system during a postulated 
design-basis loss-of-coolant accident (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.
    (3) Loss-of-coolant accidents (LOCAs) means the hypothetical 
accidents 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. LOCAs involving breaks at 
or below the transition break size (TBS) are design-basis accidents. 
LOCAs involving breaks larger than the TBS are beyond design-basis 
accidents.
    (4) Operating configuration means those plant characteristics, such 
as power level, equipment unavailability (including unavailability 
caused by corrective and preventive maintenance), and equipment 
capability that affect plant response to a LOCA.
    (5) Transition break size (TBS) for reactors licensed before 
[EFFECTIVE DATE OF RULE] is a break area equal to the cross-sectional 
flow area of the inside diameter of the largest piping attached to the 
reactor coolant system for a pressurized water reactor, or the larger 
of the feedwater line inside containment or the residual heat removal 
line inside containment for a boiling water reactor. For reactors 
licensed after [EFFECTIVE DATE OF RULE], the TBS will be determined on 
a plant-specific basis.
    (b) Applicability and scope.
    (1) The requirements of this section may be applied to each boiling 
or pressurized light-water nuclear power reactor fueled with uranium 
oxide pellets within cylindrical zircalloy or ZIRLO cladding whose 
operating license was issued prior to [EFFECTIVE DATE OF RULE]; to each 
boiling or pressurized light-water nuclear power reactor fueled with 
uranium oxide pellets within cylindrical zircalloy or ZIRLO cladding 
whose operating license, combined license under part 52 of this chapter 
or manufacturing license under part 52 of this chapter is issued after 
[EFFECTIVE DATE OF RULE] and whose design is demonstrated under Sec.  
50.46a(c)(2) to be similar to the designs of reactors licensed before 
[EFFECTIVE DATE OF RULE]; and to each boiling or pressurized light-
water nuclear power reactor fueled with uranium oxide pellets within 
cylindrical zircalloy or ZIRLO cladding whose design approval or design 
certification under part 52 of this chapter is demonstrated under Sec.  
50.46a(c)(2) to be similar to the designs of reactors licensed before 
[EFFECTIVE DATE OF RULE]. The requirements of this section do not apply 
to a reactor for which the certification required under Sec.  
50.82(a)(1) has been submitted.
    (2) The requirements of this section are in addition to any other 
requirements applicable to ECCS set forth in this part, with the 
exception of Sec.  50.46. The criteria set forth in paragraphs (e)(3) 
and (e)(4) of this section, with cooling performance calculated in 
accordance with an acceptable evaluation model or analysis method under 
paragraphs (e)(1) and (e)(2) of this section, 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.
    (c) Application. (1) A licensee of a facility seeking to implement 
this section shall submit an application for a license amendment under 
Sec.  50.90 that contains the following information:
    (i) A written evaluation demonstrating applicability of the results 
in NUREG-1829, ``Estimating Loss-of-Coolant Accident (LOCA) Frequencies 
through the Elicitation Process''; March 2008 and NUREG-1903, ``Seismic 
Considerations for the Transition Break Size''; February 2008'' to the 
licensee's facility. As part of this evaluation, the application must 
contain a plant specific analysis demonstrating that the risk of 
seismically-induced LOCAs larger than the TBS is comparable to or less 
than the seismically-induced LOCA risk associated with the NUREG-1903 
results.
    (ii) Identification of the approved analysis method(s) for 
demonstrating compliance with the ECCS criteria in paragraph (e) of 
this section.
    (iii) A description of the risk-informed evaluation process used in 
evaluating whether proposed changes to the facility meet the 
requirements in paragraph (f) of this section.
    (iv) A licensee who wishes to make changes enabled by this section 
without prior NRC review and approval must submit for NRC approval a 
process to be used for evaluating the acceptability of these changes; 
including:
    (A) A description of the approach, methods, and decisionmaking 
process to be used for evaluating compliance with the acceptance 
criteria in paragraphs (f)(1), (f)(2), and (f)(3) of this section, and
    (B) A description of the licensee's PRA model and non-PRA risk 
assessment methods to be used for demonstrating compliance with 
paragraphs (f)(4) and (f)(5) of this section.
    (v) A description of non safety equipment that is credited for 
demonstrating compliance with the ECCS acceptance criteria in paragraph 
(e) of this section.
    (2) An applicant for a construction permit, operating license, 
design approval, design certification, manufacturing license, or 
combined license seeking to implement the requirements of this section 
shall, in addition to the information required by paragraph (c)(1) of 
this section, submit an analysis demonstrating why the proposed reactor 
design is similar to the designs of reactors licensed before [EFFECTIVE 
DATE OF RULE] such that the provisions of this section may properly 
apply. The analysis must also include a recommendation for an 
appropriate TBS and a justification that the recommended TBS is 
consistent with the technical basis for this section.

[[Page 40048]]

    (3) Acceptance criteria. The NRC may approve an application to use 
this section if:
    (i) The evaluation submitted under paragraph (c)(1)(i) of this 
section demonstrates that the NUREG-1829 results are applicable to the 
facility, and the risk of seismically-induced LOCAs larger than the TBS 
is comparable to or less than the seismically-induced LOCA risk 
associated with the NUREG-1903 results;
    (ii) The method(s) for demonstrating compliance with the ECCS 
acceptance criteria in paragraphs (e)(3) and (e)(4) of this section 
meet the requirements in paragraphs (e)(1) and (e)(2) of this section;
    (iii) The risk-informed evaluation process the licensee proposes to 
use for making changes enabled by this section is adequate for 
determining whether the acceptance criteria in paragraph (f) of this 
section have been met; and
    (iv) Non safety equipment that is credited for demonstrating 
compliance with the ECCS acceptance criteria in paragraph (e) of this 
section is identified in plant Technical Specifications.
    (v) For all applicants other than those holding operating licenses 
issued before [EFFECTIVE DATE OF RULE], the proposed reactor design is 
similar to the designs of reactors licensed before [EFFECTIVE DATE OF 
RULE] and the applicant's proposed TBS is consistent with the technical 
basis of this section.
    (d) Requirements during operation. A licensee whose application 
under paragraph (c) of this section is approved by the NRC shall comply 
with the following requirements as long as the facility is subject to 
the requirements in this section until the licensee submits the 
certifications required by Sec.  50.82(a):
    (1) The licensee shall maintain ECCS model(s) and/or analysis 
method(s) meeting the requirements in paragraphs (e)(1) and (e)(2) of 
this section;
    (2) The licensee shall have leak detection systems available at the 
facility and shall implement actions as necessary to identify, monitor 
and quantify leakage to ensure that adverse safety consequences do not 
result from primary pressure boundary leakage from piping and 
components that are larger than the transition break size.
    (3) A change enabled by this section must, in addition to meeting 
other applicable NRC requirements, be evaluated by a risk-informed 
evaluation demonstrating that the acceptance criteria in paragraph (f) 
of this section are met.
    (4) The licensee shall periodically maintain and upgrade, as 
necessary, its risk assessments to meet the requirements in paragraph 
(f)(4) and (f)(5) of this section. The maintenance and upgrading shall 
be consistent with NRC-endorsed consensus standards on PRA and must be 
completed in a timely manner, but no less often than once every two 
refueling outages. Based upon a re-evaluation of the risk assessments 
after the periodic maintenance and upgrading are completed, the 
licensee shall take appropriate action to ensure that the acceptance 
criteria in paragraphs (f)(2) or (f)(3) of this section, as applicable, 
are met. The PRA maintenance and upgrading required by this section, 
and any necessary changes to the facility, technical specifications and 
procedures as a result of this re-evaluation, shall not be deemed to be 
backfitting under any provision of this chapter.
    (5) For LOCAs larger than the TBS, operation in a plant operating 
configuration not demonstrated to meet the acceptance criteria in 
paragraph (e)(4) of this section may not exceed a total of fourteen 
days in any 12 month period.
    (6) The licensee shall perform an evaluation to determine the 
effect of all planned facility changes and shall not implement any 
facility change that would invalidate the evaluation performed pursuant 
to Sec.  50.46a(c)(1)(i) demonstrating the applicability to the 
licensee's facility of the generic results in NUREG-1829 and NUREG-
1903.
    (e) ECCS Performance. Each nuclear power reactor subject to this 
section must be provided with an ECCS that must be designed so that its 
calculated cooling performance following postulated LOCAs conforms to 
the criteria set forth in this section. The evaluation models for LOCAs 
must meet the criteria in this paragraph, and must be approved for use 
by the NRC. Appendix K, Part II, to 10 CFR Part 50, sets forth the 
documentation requirements for evaluation models.
    (1) ECCS evaluation for LOCAs involving breaks at or below the TBS. 
ECCS cooling performance at or below the TBS must be calculated in 
accordance with an evaluation model that meets the requirements of 
either section I to Appendix K of this part, or the following 
requirements, and must demonstrate that the acceptance criteria in 
paragraph (e)(3) of this section are satisfied. The evaluation model 
must be used for a number of postulated LOCAs of different sizes, 
locations, and other properties sufficient to provide assurance that 
the most severe postulated LOCAs involving breaks at or below the TBS 
are analyzed. The evaluation model must include sufficient supporting 
justification to show that the analytical technique realistically 
describes the behavior of the reactor system during a LOCA. Comparisons 
to applicable experimental data must be made and 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 criteria set forth in paragraph 
(e)(3) of this section, there is a high level of probability that the 
criteria would not be exceeded.
    (2) ECCS analyses for LOCAs involving breaks larger than the TBS. 
ECCS cooling performance for LOCAs involving breaks larger than the TBS 
must be calculated in accordance with an evaluation model that meets 
the requirements of either section I to Appendix K of this part, or the 
following requirements, and must demonstrate that the acceptance 
criteria in paragraph (e)(4) of this section are satisfied. The 
evaluation model must include sufficient supporting justification to 
show that the analytical technique realistically describes the behavior 
of the reactor system during a LOCA. Comparisons to applicable 
experimental data must be made and 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 criteria set forth in paragraph (e)(4) of this section, 
there is a high level of probability that the criteria would not be 
exceeded. The evaluation model must be used for a number of postulated 
LOCAs of different sizes, locations, and other properties sufficient to 
provide assurance that the most severe postulated LOCAs larger than the 
TBS up to the double-ended rupture of the largest pipe in the reactor 
coolant system are analyzed. These calculations may take credit for the 
availability of offsite power and do not require the assumption of a 
single failure. Realistic initial conditions and availability of 
safety-related or non safety-related equipment may be assumed if 
supported by plant-specific data or analysis, and provided that onsite 
power can be readily provided through simple manual actions to 
equipment that is credited in the analysis.
    (3) Acceptance criteria for LOCAs involving breaks at or below the 
TBS. The following acceptance criteria must be used in determining the 
acceptability of ECCS cooling performance:
    (i) Peak cladding temperature. The calculated maximum fuel element

[[Page 40049]]

cladding temperature must not exceed 2200 [deg]F.
    (ii) Maximum cladding oxidation. The calculated total oxidation of 
the cladding must not at any location exceed 0.17 times the total 
cladding thickness before oxidation. As used in this paragraph, total 
oxidation means the total thickness of cladding metal that would be 
locally converted to oxide if all the oxygen absorbed by and reacted 
with the cladding locally were converted to stoichiometric zirconium 
dioxide. If cladding rupture is calculated to occur, the inside 
surfaces of the cladding must be included in the oxidation, beginning 
at the calculated time of rupture. Cladding thickness before oxidation 
means the radial distance from inside to outside the cladding, after 
any calculated rupture or swelling has occurred but before significant 
oxidation. Where the calculated conditions of transient pressure and 
temperature lead to a prediction of cladding swelling, with or without 
cladding rupture, the unoxidized cladding thickness must be defined as 
the cladding cross-sectional area, taken at a horizontal plane at the 
elevation of the rupture, if it occurs, or at the elevation of the 
highest cladding temperature if no rupture is calculated to occur, 
divided by the average circumference at that elevation. For ruptured 
cladding the circumference does not include the rupture opening.
    (iii) Maximum hydrogen generation. The calculated total amount of 
hydrogen generated from the chemical reaction of the cladding with 
water or steam must 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.
    (iv) Coolable geometry. Calculated changes in core geometry must be 
such that the core remains amenable to cooling.
    (v) Long term cooling. After any calculated successful initial 
operation of the ECCS, the calculated core temperature must be 
maintained at an acceptably low value and decay heat must be removed 
for the extended period of time required by the long-lived 
radioactivity remaining in the core.
    (4) Acceptance criteria for LOCAs involving breaks larger than the 
TBS. The following acceptance criteria must be used in determining the 
acceptability of ECCS cooling performance:
    (i) Coolable geometry. Calculated changes in core geometry must be 
such that the core remains amenable to cooling.
    (ii) Long term cooling. After any calculated successful initial 
operation of the ECCS, the calculated core temperature must be 
maintained at an acceptably low value and decay heat must be removed 
for the extended period of time required by the long-lived 
radioactivity remaining in the core.
    (5) Imposition of restrictions. The Director of the Office of 
Nuclear Reactor Regulation may impose restrictions on reactor operation 
if it is found that the evaluations of ECCS cooling performance 
submitted are not consistent with paragraph (e) of this section.
    (f) Changes to facility, technical specifications, or procedures. A 
licensee who wishes to make changes to the facility or procedures or to 
the technical specifications enabled by this rule shall perform a risk-
informed evaluation.
    (1) The licensee may make such changes without prior NRC approval 
if:
    (i) The change is permitted under Sec.  50.59,
    (ii) The risk informed evaluation process described in paragraph 
(c)(1)(iii) of this section demonstrates that any increases in the 
estimated risk are minimal compared to the overall plant risk profile, 
and the criteria in paragraph (f)(3) of this section are met, and
    (iii) The change does not invalidate the evaluation performed 
pursuant to paragraph (c)(1)(i) of the applicability of the results in 
NUREG-1829 and NUREG-1903 to the licensee's facility.
    (2) For implementing changes which are not permitted under 
paragraph (f)(1) of this section, the licensee must submit an 
application for license amendment under Sec.  50.90. The application 
must contain:
    (i) The information required under Sec.  50.90;
    (ii) For applicants whose operating licenses were issued before 
[EFFECTIVE DATE OF RULE], information from the risk-informed evaluation 
demonstrating that the total increases in core damage frequency and 
large early release frequency are very small and the overall risk 
remains small, and the criteria in paragraph (f)(3) of this section are 
met;
    (iii) For applicants whose operating licenses were not issued 
before [EFFECTIVE DATE OF RULE], information from the risk-informed 
evaluation demonstrating that the total increases in core damage 
frequency and large release frequency are very small and the overall 
risk remains small, and the criteria in paragraph (f)(3) of this 
section are met;
    (iv) If previous changes have been made under Sec.  50.46a, 
information from the risk-informed evaluation on the cumulative effect 
on risk of the proposed change and all previous changes made under this 
section. If more than one plant change is combined; including plant 
changes not enabled by this section, into a group for the purposes of 
evaluating acceptable risk increases; the evaluation of each individual 
change shall be performed along with the evaluation of combined 
changes; and
    (v) Information demonstrating that the criteria in paragraphs 
(e)(3) and (e)(4) of this section are met.
    (vi) Information demonstrating that the proposed change will not 
increase the LOCA frequency of the facility (including the frequency of 
seismically-induced LOCAs) by an amount that would invalidate the 
applicability to the facility of the generic studies (NUREG-1829, 
``Estimating Loss-of-Coolant Accident (LOCA) Frequencies through the 
Elicitation Process'', March 2008 and NUREG-1903, ``Seismic 
Considerations for the Transition Break Size'', February 2008'') that 
support the technical basis for this section.
    (3) All changes enabled by this rule must meet the following 
criteria:
    (i) Adequate defense in depth is maintained;
    (ii) Adequate safety margins are retained to account for 
uncertainties; and
    (iii) Adequate performance-measurement programs are implemented to 
ensure the risk-informed evaluation continues to reflect actual plant 
design and operation. These programs shall be designed to detect 
degradation of the system, structure or component before plant safety 
is compromised, provide feedback of information and timely corrective 
actions, and monitor systems, structures or components at a level 
commensurate with their safety significance.
    (4) Requirements for risk assessment--PRA. Whenever a PRA is used 
in the risk-informed evaluation, the PRA must, with respect to the area 
of evaluation which is the subject of the PRA:
    (i) Address initiating events from sources both internal and 
external to the plant and for all modes of operation, including low 
power and shutdown modes, that would affect the regulatory decision in 
a substantial manner;
    (ii) Reasonably represent the current configuration and operating 
practices at the plant;
    (iii) Have sufficient technical adequacy (including consideration 
of uncertainty) and level of detail to provide confidence that the 
total risk estimate and the change in total risk

[[Page 40050]]

estimate adequately reflect the plant and the effect of the proposed 
change on risk; and
    (iv) Be determined, through peer review, to meet industry standards 
for PRA quality that have been endorsed by the NRC.
    (5) Requirements for risk assessment other than PRA. Whenever risk 
assessment methods other than PRAs are used to develop quantitative or 
qualitative estimates of changes to risk in the risk-informed 
evaluation, an integrated, systematic process must be used. All aspects 
of the analyses must reasonably reflect the current plant configuration 
and operating practices, and applicable plant and industry operating 
experience.
     (g) Reporting. (1) Each licensee shall estimate the effect of any 
change to or error in evaluation models or analysis methods or in the 
application of such models or methods to determine if the change or 
error is significant. For each change to or error discovered in an ECCS 
evaluation model or analysis method or in the application of such a 
model that affects the calculated results, the licensee shall report 
the nature of the change or error and its estimated effect on the 
limiting ECCS analysis to the Commission at least annually as specified 
in Sec.  50.4. If the change or error is significant, the licensee 
shall provide this report within 30 days and include with the report a 
proposed schedule for providing a reanalysis or taking other action as 
may be needed to show compliance with Sec.  50.46a requirements. This 
schedule may be developed using an integrated scheduling system 
previously approved for the facility by the NRC. For those facilities 
not using an NRC-approved integrated scheduling system, a schedule will 
be established by the NRC staff within 60 days of receipt of the 
proposed schedule. Any change or error correction that results in a 
calculated ECCS performance that does not conform to the criteria set 
forth in paragraphs (e)(3) or (e)(4) of this section is a reportable 
event as described in Sec. Sec.  50.55(e), 50.72 and 50.73. The 
licensee shall propose immediate steps to demonstrate compliance or 
bring plant design or operation into compliance with Sec.  50.46a 
requirements. For the purpose of this paragraph, a significant change 
or error is:
    (i) For LOCAs involving pipe breaks at or below the TBS, one which 
results either in a calculated peak fuel cladding temperature different 
by more than 50 [deg]F from the temperature calculated for the limiting 
transient using the last acceptable 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 [deg]F; or
    (ii) For LOCAs involving pipe breaks larger than the TBS, one which 
results in a significant reduction in the capability to meet the 
requirements of paragraph (e)(4) of this section.
    (2) As part of the PRA maintenance and upgrading under paragraph 
(d)(4) of this section, the licensee shall report to the NRC if the re-
evaluation results in exceeding the acceptance criteria in paragraphs 
(f)(1) or (f)(2) of this section, as applicable. The report must be 
filed with the NRC no more than 60 days after completing the PRA re-
evaluation. The report must describe and explain the changes in the PRA 
modeling, plant design, or plant operation that led to the increase(s) 
in risk, and must include a description of and implementation schedule 
for any corrective actions required under paragraph (d)(4) of this 
section.
    (3) Every 24 months, the licensee shall submit, as specified in 
Sec.  50.4, a short description of each change involving minimal 
changes in risk made under paragraph (f)(1) of this section after the 
last report and a brief summary of the basis for the licensee's 
determination pursuant to Sec.  50.46a(f)(2)(vi) that the change does 
not invalidate the applicability evaluation made under Sec.  
50.46a(c)(1)(i).
    (h) Documentation. Following implementation of the Sec.  50.46a 
requirements, the licensee shall maintain records sufficient to 
demonstrate compliance with the requirements in this section in 
accordance with Sec.  50.71.
    (i) through (l)--[RESERVED]
    (m) Changes to TBS. If the NRC increases the TBS specified in this 
section applicable to a licensee's nuclear power plant, each licensee 
subject to this section shall perform the evaluations required by 
paragraphs (e)(1) and (e)(2) of this section and reconfirm compliance 
with the acceptance criteria in paragraphs (e)(3) and (e)(4) of this 
section. If the licensee cannot demonstrate compliance with the 
acceptance criteria, then the licensee shall change its facility, 
technical specifications or procedures so that the acceptance criteria 
are met. The evaluation required by this paragraph, and any necessary 
changes to the facility, technical specifications or procedures as the 
result of this evaluation, must not be deemed to be backfitting under 
any provision of this chapter.
    5. In Sec.  50.109, paragraph (b) is revised to read as follows:


Sec.  50.109  Backfitting.

* * * * *
    (b) Paragraph (a)(3) of this section shall not apply to:
    (1) Backfits imposed prior to October 21, 1985; and
    (2) Any changes made to the TBS specified in Sec.  50.46a or as 
otherwise applied to a licensee.
* * * * *
    6. In Appendix A to 10 CFR Part 50, under the heading, 
``CRITERIA,'' Criterion 17, 35, 38, 41, 44, and 50 are revised to read 
as follows:

APPENDIX A TO PART 50--GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS

* * * * *

CRITERIA

* * * * *
    Criterion 17--Electrical power systems. An on-site electric 
power system and an offsite electric power system shall be provided 
to permit functioning of structures, systems, and components 
important to safety. The safety function for each system (assuming 
the other system is not functioning) shall be to provide sufficient 
capacity and capability to assure that (1) specified acceptable fuel 
design limits and design conditions of the reactor coolant pressure 
boundary are not exceeded as a result of anticipated operational 
occurrences and (2) the core is cooled and containment integrity and 
other vital functions are maintained in the event of postulated 
accidents.
    The onsite electric power supplies, including the batteries, and 
the onsite electrical distribution system, shall have sufficient 
independence, redundancy, and testability to perform their safety 
functions assuming a single failure, except for loss of coolant 
accidents involving pipe breaks larger than the transition break 
size under Sec.  50.46a, where a single failure of the onsite power 
supplies and electrical distribution system need not be assumed for 
plants under Sec.  50.46a. For those pipe breaks only, neither a 
single failure nor the unavailability of offsite power need be 
assumed.
    Electric power from the transmission network to the onsite 
electric distribution system shall be supplied by two physically 
independent circuits (not necessarily on separate rights of way) 
designed and located so as to minimize to the extent practical the 
likelihood of their simultaneous failure under operating and 
postulated accident conditions. A switchyard common to both circuits 
is acceptable. Each of these circuits shall be designed to be 
available in sufficient time following a loss of all onsite 
alternating current power supplies and the other offsite electric 
power circuit, to assure that specified acceptable fuel design 
limits and design conditions of the reactor coolant pressure 
boundary are not exceeded. One of these circuits shall be designed 
to be available within a few seconds following a LOCA to assure that 
core cooling, containment integrity, and other vital safety 
functions are maintained.

[[Page 40051]]

    Provisions shall be included to minimize the probability of 
losing electric power from any of the remaining supplies as a result 
of, or coincident with, the loss of power generated by the nuclear 
power unit, the loss of power from the transmission network, or the 
loss of power from the onsite electric power supplies.
* * * * *
    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 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, except for loss of coolant accidents 
involving pipe breaks larger than the transition break size under 
Sec.  50.46a. For those pipe breaks only, neither a single failure 
nor the unavailability of offsite power need be assumed.
* * * * *
    Criterion 38--Containment heat removal. 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 LOCA 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, except for analysis of loss of coolant 
accidents involving pipe breaks larger than the transition break 
size under Sec.  50.46a. For those pipe breaks only, neither a 
single failure nor the unavailability of offsite power need be 
assumed.
* * * * *
    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, except for analysis of loss of coolant accidents 
involving pipe breaks larger than the transition break size under 
Sec.  50.46a. For those pipe breaks only, neither a single failure 
nor the unavailability of offsite power need be assumed.
* * * * *
    Criterion 44--Cooling water. A system to transfer heat from 
structures, systems, and components important to safety, to an 
ultimate heat sink shall be provided. The system safety function 
shall be to transfer the combined heat load of these structures, 
systems, and components under normal operating and accident 
conditions.
    Suitable redundancy in components and features, and suitable 
interconnections, leak detection, and isolation 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, except for analysis of loss of coolant accidents 
involving pipe breaks larger than the transition break size under 
Sec.  50.46a. For those pipe breaks only, neither a single failure 
nor the unavailability of offsite power need be assumed.
* * * * *
    Criterion 50--Containment design basis. The reactor containment 
structure, including access openings, penetrations, and the 
containment heat removal system shall be designed so that the 
containment structure and its internal compartments can accommodate, 
without exceeding the design leakage rate and with sufficient 
margin, the calculated pressure and temperature conditions resulting 
from any loss-of-coolant accident. This margin shall reflect 
consideration of (1) the effects of potential energy sources which 
have not been included in the determination of the peak conditions, 
such as energy in steam generators and as required by Sec.  50.44 
energy from metal-water and other chemical reactions that may result 
from degradation but not total failure of emergency core cooling 
functioning, (2) the limited experience and experimental data 
available for defining accident phenomena and containment responses, 
and (3) the conservatism of the calculational model and input 
parameters.
    For licensees voluntarily choosing to comply with Sec.  50.46a, 
the structural and leak tight integrity of the reactor containment 
structure, including access openings, penetrations, and its internal 
compartments, shall be maintained for realistically calculated 
pressure and temperature conditions resulting from any loss of 
coolant accident larger than the transition break size.
* * * * *

PART 52--LICENSES, CERTIFICATIONS AND APPROVALS FOR NUCLEAR POWER 
PLANTS

    7. The authority citation for part 52 continues to read as follows:

    Authority: Secs. 103, 104, 161, 182, 183, 185, 186, 189, 68 
Stat. 936, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 
444, as amended (42 U.S.C. 2133, 2201, 2232, 2233, 2235, 2236, 2239, 
2282); secs. 201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended 
(42 U.S.C. 5841, 5842, 5846); sec. 1704, 112 Stat. 2750 (44 U.S.C. 
3504 note); Energy Policy Act of 2005, Pub. L. No. 109-58, 119 Stat. 
594 (2005), secs. 147 and 149 of the Atomic Energy Act.

    8. 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.
    (i) Analysis and evaluation of emergency core cooling system (ECCS) 
cooling performance and the need for high-point vents following 
postulated loss-of-coolant accidents may be performed under the 
requirements of either Sec.  50.46 or Sec.  50.46a and Sec.  50.46b of 
this chapter for designs certified after [EFFECTIVE DATE OF RULE] and 
demonstrated under Sec.  50.46a(c)(2) of this chapter to be similar to 
reactor designs licensed before [EFFECTIVE DATE OF RULE], or
    (ii) Analysis and evaluation of ECCS cooling performance and the 
need for high-point vents following postulated loss-of-coolant 
accidents must be performed under the requirements of Sec. Sec.  50.46 
and 50.46b of this chapter for designs that are not demonstrated under 
Sec.  50.46a(c)(2) of this chapter to be similar to reactor designs 
licensed before [EFFECTIVE DATE OF RULE].
* * * * *
    9. 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

[[Page 40052]]

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.
    (i) Analysis and evaluation of ECCS cooling performance and the 
need for high-point vents following postulated loss-of-coolant 
accidents must be performed under the requirements of either Sec.  
50.46 or Sec.  50.46a and Sec.  50.46b of this chapter for facilities 
licensed after [EFFECTIVE DATE OF RULE] and demonstrated under Sec.  
50.46a(c)(2) of this chapter to be similar to reactor designs licensed 
before [EFFECTIVE DATE OF RULE], or
    (ii) Analysis and evaluation of ECCS cooling performance and the 
need for high-point vents following postulated loss-of-coolant 
accidents must be performed under the requirements of Sec. Sec.  50.46 
and 50.46b of this chapter for facilities licensed after [EFFECTIVE 
DATE OF RULE] and not demonstrated under Sec.  50.46a(c)(2) of this 
chapter to be similar to reactor designs licensed before [EFFECTIVE 
DATE OF RULE].
* * * * *
    10. 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.
    (i) Analysis and evaluation of ECCS cooling performance and the 
need for high-point vents following postulated loss-of-coolant 
accidents must be performed under the requirements of either Sec.  
50.46 or Sec.  50.46a and Sec.  50.46b of this chapter for designs 
approved after [EFFECTIVE DATE OF RULE] and demonstrated under Sec.  
50.46a(c)(2) of this chapter to be similar to reactor designs licensed 
before [EFFECTIVE DATE OF RULE], or
    (ii) Analysis and evaluation of ECCS cooling performance and the 
need for high-point vents following postulated loss-of-coolant 
accidents must be performed under the requirements of Sec. Sec.  50.46 
and 50.46b of this chapter for designs that are not demonstrated under 
Sec.  50.46a(c)(2) of this chapter to be similar to reactor designs 
licensed before [EFFECTIVE DATE OF RULE].
* * * * *
    11. In Sec.  52.157, paragraph (f)(1) is revised to read as 
follows:


Sec.  52.157  Contents of applications; technical information in 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.
    (i) Analysis and evaluation of ECCS cooling performance and the 
need for high-point vents following postulated loss-of-coolant 
accidents must be performed under the requirements of either Sec.  
50.46 or Sec.  50.46a and Sec.  50.46b of this chapter for facilities 
licensed after [EFFECTIVE DATE OF RULE] and demonstrated under Sec.  
50.46a(c)(2) to be similar to reactor designs licensed before 
[EFFECTIVE DATE OF RULE], or
    (ii) Analysis and evaluation of ECCS cooling performance and the 
need for high-point vents following postulated loss-of-coolant 
accidents must be performed under the requirements of Sec. Sec.  50.46 
and 50.46b of this chapter for facilities licensed after [EFFECTIVE 
DATE OF RULE] and not demonstrated under Sec.  50.46a(c)(2) of this 
chapter to be similar to reactor designs licensed before [EFFECTIVE 
DATE OF RULE].
* * * * *

    Dated at Rockville, Maryland, this 6th day of July 2009.

    For the Nuclear Regulatory Commission.
Bruce S. Mallett,
Acting Executive Director for Operations.
[FR Doc. E9-18547 Filed 8-7-09; 8:45 am]
BILLING CODE 7590-01-P


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