[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
[email protected].
Mail comments to: Secretary, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attn: Rulemakings and Adjudications Staff.
E-mail comments to: [email protected]. 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
[email protected].
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: [email protected].
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-\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-\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-\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 10-5 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 10-5 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 (10-5 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 10-5 per year and
10-6 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 10-5 and
10-6 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 10-5 and 10-6 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
(10-5 to 10-6 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
10-6 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 10-5 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 10-5 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, 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
10-5 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 10-5 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 10-5 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 10-5 or 10-6 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 10-5 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.
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\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
10-7 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 10-7 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 10-7 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-\5\ or
10-\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-\5\ or 10-\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-\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-\5\ to 10-\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-\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-\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-\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-\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:
PCT less than 2200 [deg]F;
Maximum local cladding oxidation (MLO) less than 17
percent;
Maximum hydrogen production--core wide cladding oxidation
less than one percent;
Maintenance of coolable geometry; and
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:
Maintenance of coolable geometry, and
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-\5\ per reactor year for CDF
and 10-\6\ per reactor year for LERF
[[Page 40038]]
are normally acceptable until the baseline risk increases to reference
values of approximately 10-\4\ per reactor year and
10-\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-\6\ per
reactor year and 10-\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-\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-\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-\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
[email protected]. 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
[email protected].
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
[email protected] 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 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
[email protected] 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