Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 77906-77919 [2010-31086]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 75, Number 239 (Tuesday, December 14, 2010)]
[Notices]
[Pages 77906-77919]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2010-31086]


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NUCLEAR REGULATORY COMMISSION

[NRC-2010-0382]


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC) is publishing this regular biweekly notice. The Act 
requires the Commission publish notice of any amendments issued, or 
proposed to be issued and grants the Commission the authority to issue 
and make immediately effective any amendment to an operating license 
upon a determination by the Commission that such amendment involves no 
significant hazards consideration, notwithstanding the pendency before 
the Commission of a request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from November 18, 2010, to December 1, 2010. The 
last biweekly notice was published on November 30, 2010 (75 FR 74091).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in title 10 of the Code of Federal 
Regulations (10 CFR), section 50.92, this means that operation of the 
facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this

[[Page 77907]]

proposed determination for each amendment request is shown below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules, 
Announcements and Directives Branch (RADB), TWB-05-B01M, Division of 
Administrative Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be faxed to the RADB at 301-492-3446. 
Documents may be examined, and/or copied for a fee, at the NRC's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's ``Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR part 2. Interested person(s) 
should consult a current copy of 10 CFR 2.309, which is available at 
the Commission's PDR, located at One White Flint North, Room O1-F21, 
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly 
available records will be accessible from the Agencywide Documents 
Access and Management System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC Web site, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or a presiding 
officer designated by the Commission or by the Chief Administrative 
Judge of the Atomic Safety and Licensing Board Panel, will rule on the 
request and/or petition; and the Secretary or the Chief Administrative 
Judge of the Atomic Safety and Licensing Board will issue a notice of a 
hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, any hearing held 
would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the Internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten (10) days prior to the filing deadline, the participant should 
contact the Office of the Secretary by e-mail at 
hearing.docket@nrc.gov, or by telephone at 301-415-1677, to request (1) 
a digital ID certificate, which allows the participant (or its counsel 
or representative) to digitally sign documents and access the E-
Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for

[[Page 77908]]

hearing (even in instances in which the participant, or its counsel or 
representative, already holds an NRC-issued digital ID certificate). 
Based upon this information, the Secretary will establish an electronic 
docket for the hearing in this proceeding if the Secretary has not 
already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in NRC's ``Guidance for Electronic 
Submission,'' which is available on the agency's public Web site at 
https://www.nrc.gov/site-help/e-submittals.html. Participants may 
attempt to use other software not listed on the Web site, but should 
note that the NRC's E-Filing system does not support unlisted software, 
and the NRC Meta System Help Desk will not be able to offer assistance 
in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through Electronic Information Exchange System, users 
will be required to install a Web browser plug-in from the NRC Web 
site. Further information on the Web-based submission form, including 
the installation of the Web browser plug-in, is available on the NRC's 
public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
E-Filing system also distributes an e-mail notice that provides access 
to the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at https://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, or the presiding officer. Participants 
are requested not to include personal privacy information, such as 
social security numbers, home addresses, or home phone numbers in their 
filings, unless an NRC regulation or other law requires submission of 
such information. With respect to copyrighted works, except for limited 
excerpts that serve the purpose of the adjudicatory filings and would 
constitute a Fair Use application, participants are requested not to 
include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the Commission's PDR, located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the ADAMS 
Public Electronic Reading Room on the Internet at the NRC Web site, 
https://www.nrc.gov/reading-rm/adams.html. Persons who do not have 
access to ADAMS or who encounter problems in accessing the documents 
located in ADAMS, should contact the NRC PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to pdr.resource@nrc.gov.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina

    Date of amendment request: June 29, 2009, as supplemented June 24, 
2010.
    Description of amendment request: The proposed amendments would 
approve changes to the updated final safety analysis report to allow 
the use of fiber reinforce polymer on masonry walls for uniform 
pressure loads resulting from a tornado event.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

(1) Involve a Significant Increase in The Probability or Consequences 
of an Accident Previously Evaluated

    Response: Physical protection from a tornado event is a design 
basis criterion rather than a requirement of a previously analyzed 
[updated final safety analysis report] UFSAR accident analysis. The 
current

[[Page 77909]]

licensing basis (CLB) for Oconee states that systems, structures, 
and components (SSC's) required to shut down and maintain the units 
in a shutdown condition will not fail as a result of damage caused 
by natural phenomena.
    The in-fill masonry walls to be strengthened using an FRP system 
are passive, non-structural elements. The use of a fiber reinforced 
polymer [FRP] system on existing Auxiliary Building masonry walls 
will allow them to resist uniform pressure loads resulting from a 
tornado and will not adversely affect the structure's ability to 
withstand other design basis events such as earthquakes or fires. 
Therefore, the proposed use of FRP on existing masonry walls will 
not significantly increase the probability or consequences of an 
accident previously evaluated.

(2) Create the Possibility of a New or Different Kind of Accident From 
Any Accident Previously Evaluated

    Response: The final state of the FRP system is passive in nature 
and will not initiate or cause an accident. More generally, this 
understanding supports the conclusion that the potential for new or 
different kinds of accidents is not created.

(3) Involve a Significant Reduction in a Margin of Safety

    Response: The application of an FRP system to existing Auxiliary 
Building masonry walls will act to enhance the margin of safety, 
e.g., the West Penetration Room walls, by increasing the walls' 
ability to resist tornado-induced differential pressure. 
Consequently, this change does not involve a significant reduction 
in a margin of safety.
    In summary, based upon the above evaluation, Duke has concluded 
that the proposed amendment involves no significant hazards 
consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Gloria Kulesa.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina

    Date of amendment request: July 14, 2010.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to adopt NRC Approved 
Technical Specification Task Force (TSTF) Change to the Standard TS, 
TSTF-52 concerning performance-based containment leakage testing 
requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

(1) Does the proposed amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    No. Implementation of these changes will provide continued 
assurance that specified parameters associated with containment 
integrity will remain within acceptance limits as delineated in 
[Title 10 of the Code of Federal Regulations (10 CFR) Part 50] 10 
CFR Part 50, Appendix J, Option B. The changes are consistent with 
current safety analyses. Although some of the proposed changes 
represent minor relaxation to existing [Technical Specifications] TS 
requirements, they are consistent with the requirements specified by 
Option B of 10 CFR Part 50, Appendix J. The systems affecting 
containment integrity related to this proposed amendment request are 
not assumed in any safety analyses to initiate any accident 
sequence. Therefore, the probability of any accident previously 
evaluated is not increased by this proposed amendment. The proposed 
changes maintain an equivalent level of reliability and availability 
for all affected systems. In addition, maintaining leakage within 
analyzed limits assumed in accident analyses does not adversely 
affect either onsite or offsite dose consequences.
    Therefore, adopting Appendix J, Option B does not significantly 
increase the probability or consequences of any accident previously 
evaluated.

(2) Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    No. No changes are being proposed which will introduce any 
physical changes to the existing plant design. The proposed changes 
are consistent with the current safety analyses. Some of the changes 
may involve revision in the testing of components; however, these 
are in accordance with the current safety analyses and provide for 
appropriate testing or surveillance that is consistent with 10 CFR 
Part 50, Appendix J, Option B. The proposed changes will not 
introduce new failure mechanisms beyond those already considered in 
the current accident analyses. No new modes of operation are 
introduced by the proposed changes. The proposed changes maintain, 
at minimum, the present level of operability of any system that 
affects containment integrity.
    Therefore, adoption of Appendix J, Option B will not create the 
possibility of a new or different kind of accident from any kind of 
accident previously evaluated.

(3) Does the proposed amendment involve a significant reduction in a 
margin of safety?

    No. The provisions specified in Option B of 10 CFR Part 50, 
Appendix J allow changes to Type B and Type C test intervals based 
upon the performance of past leak rate tests. 10 CFR Part 50, 
Appendix J, Option B allows longer intervals between leakage tests 
based on performance trends, but does not relax the leakage 
acceptance criteria. Changing test intervals from those currently 
provided in the TS to those provided in 10 CFR Part 50, Appendix J, 
Option B does not increase any risks above and beyond those that the 
[U S. Nuclear Regulatory Commission] NRC has deemed acceptable for 
the performance based option. In addition, there are risk reduction 
benefits associated with reduction in component cycling, stress, and 
wear associated with increased test intervals. The proposed changes 
provide continued assurance of leakage integrity of containment 
without adversely affecting the public health and safety and will 
not significantly reduce existing safety margins.
    Therefore, adoption of Appendix J, option B does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Gloria Kulesa.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station (ONS), Units 1, 2, and 3, Oconee County, South 
Carolina; Docket Nos. 50-369 and 50-370, McGuire Nuclear Station (MNS), 
Units 1 and 2, Mecklenburg County, North Carolina; Docket Nos. 50-413 
and 50-414, Catawba Nuclear Station (CNS), Units 1 and 2, York County, 
South Carolina

    Date of amendment request: September 16, 2010.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to update the qualification 
requirements for the Station Manager and Radiation Protection Manager 
to meet or exceed the minimum qualifications in ANSI/ANS-3.1-1993, 
``Selection, Qualification, and Training of Personnel for Nuclear Power 
Plants.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. Does the proposed amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    Response: No.

[[Page 77910]]

    The proposed change to [Technical Specifications] TS 5.3.1 is an 
administrative change to update the minimum qualification 
requirements for Station Manager and Radiation Protection Manager to 
meet or exceed ANSI/ANS 3.1-1993 as endorsed by Regulatory Guide 
1.8, Revision 3, dated May 2000. This update for Station Manager and 
Radiation Protection Manager qualifications will also provide 
Oconee, McGuire, and Catawba the needed flexibility to appoint 
Station Managers and Radiation Protection Managers from a larger 
candidate pool. The current qualification requirements restrict the 
pool of personnel capable of performing the Station Manager and 
Radiation Protection Manager functions. This change will also revise 
the current Oconee, McGuire, and Catawba TS 5.3.1 qualification 
requirements for Station Manager and Radiation Protection Manager to 
be consistent among all three stations. The proposed change does not 
impact the physical configuration or function of plant structures, 
systems, or components or the manner in which structures, systems, 
or components are operated, maintained, modified, tested, or 
inspected. Updating the minimum qualification requirements for 
Station Manager and Radiation Protection Manager is not an initiator 
of any accident previously evaluated. Updating the minimum 
qualification requirements for Station Manager and Radiation 
Protection Manager is not an assumption in the consequence 
mitigation of any accident previously evaluated. Therefore, it is 
concluded that this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

2. Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    Response: No.
    The proposed change to TS 5.3.1 is an administrative change to 
update the minimum qualification requirements for Station Manager 
and Radiation Protection Manager to meet or exceed ANSI/ANS 3.1-1993 
as endorsed by RG 1.8, Revision 3, dated May 2000. This represents 
an update to current guidance. This update for Station Manager and 
Radiation Protection Manager qualifications will also provide 
Oconee, McGuire, and Catawba the needed flexibility to appoint 
Station Manager and Radiation Protection Manager from a larger 
candidate pool. The current qualification requirements restrict the 
pool of personnel capable of performing the Station Manager and 
Radiation Protection Manager functions. This change will also revise 
the current Oconee, McGuire and Catawba TS 5.3.1 qualification 
requirements for Station Manager and Radiation Protection Manager to 
be consistent among all three stations.
    The proposed change does not impact the physical configuration 
or function of plant structures, systems, or components or the 
manner in which structures, systems, or components are operated, 
maintained, modified, tested, or inspected. In addition, there is no 
change in the types or increases in the amounts of effluents that 
may be released offsite, and there is no increase in individual or 
cumulative occupational radiation exposure.
    As the proposed change is administrative in nature, operation of 
the facility in accordance with the proposed amendment does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a 
margin of safety?

    Response: No.
    The proposed change to TS 5.3.1 is an administrative change to 
update the minimum qualification requirements for Station Manager 
and Radiation Protection Manager to meet or exceed ANSI/ANS 3.1-1993 
as endorsed by RG 1.8, Revision 3, dated May 2000. This update for 
Station Manager and Radiation Protection Manager qualifications will 
also provide Oconee, McGuire, and Catawba the needed flexibility to 
appoint Station Manager and Radiation Protection Manager from a 
larger candidate pool. The current qualification requirements 
restrict the pool of personnel capable of performing the Station 
Manager and Radiation Protection Manager functions. This change will 
also revise the current ONS, MNS, and CNS TS 5.3.1 qualification 
requirements for Station Manager and Radiation Protection Manager to 
be consistent among all three stations. The proposed change does not 
impact the physical configuration or function of plant structures, 
systems, or components or the manner in which structures, systems, 
or components are operated, maintained, modified, tested, or 
inspected. The proposed change does not alter the manner in which 
safety limits, limiting safety system settings or limiting 
conditions for operation are determined. The safety analysis 
acceptance criteria are not affected by this change. The proposed 
change will not result in plant operation in a configuration outside 
the design basis. The proposed change does not adversely affect 
systems that respond to safely shutdown the plant and to maintain 
the plant in a safe shutdown condition. The proposed change is 
administrative in nature; thus operation of the facility in 
accordance with the proposed amendment does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Gloria Kulesa.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina

    Date of amendment request: November 8, 2010.
    Description of amendment request: The proposed amendments would 
approve revisions to the updated final safety analysis report to 
incorporate the licensee's reactor vessel internals inspection plan.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

(1) Involve a significant increase in the probability or consequences 
of an accident previously evaluated

    No. The proposed license amendment request provides the Reactor 
Vessel Internals Inspection Plan report. The report also provides a 
description of the inspection plan as it relates to the management 
of aging effects consistent with previous commitments. The 
inspection plan is based on MRP-227, Revision 0, ``Pressurized Water 
Reactors Internals Inspection and Evaluation Guidelines'' and 
describes using the ten Aging Management Program (AMP) elements in 
the current revision of NUREG-1801 ``Generic Aging Lessons Learned'' 
(GALL, Revision 1) report.
    The inspection plan contains a discussion of the background of 
the Babcock and Wilcox designed plant Reactor Vessel Internals 
programs, first sponsored by the utilities through the Babcock and 
Wilcox Owner's Group and later by the Pressurized Water Reactor 
Owner's Group, culminating in a submittal to the Nuclear Regulatory 
Commission through the Electric Power Research Institute Materials 
Reliability Program. The inspection plan also contains a discussion 
of operational experience, time-limited aging analyses, and relevant 
existing programs.
    The Reactor Vessel Internals Aging Management Program includes 
the inspection plan and demonstrates that the program adequately 
manages the effects of aging for Reactor Vessel Internals components 
and establishes the basis for providing reasonable assurance the 
Reactor Vessel Internals components will remain functional through 
the license renewal period of extended operation.
    This license amendment request provides an inspection plan based 
on industry work and experiences as agreed to in Duke Energy's 
license renewal commitments for Reactor Vessel Internals Inspection. 
It is not an accident initiator; therefore, it will not increase the 
probability or consequences of an accident previously evaluated.

(2) Create the possibility of a new or different kind of accident from 
any accident previously evaluated

    No. The proposed Reactor Vessel Internals Inspection Plan does 
not change the methods governing normal plant operation, nor are the 
methods utilized to respond to plant transients altered. The revised 
inspection plan is not an accident/event initiator. No new 
initiating events or transients result from the use of the Reactor 
Vessel Internals Inspection plan.

[[Page 77911]]

(3) Involve a significant reduction in a margin of safety

    No. The proposed safety limits have been preserved. The License 
Amendment Request requests review and approval for the Reactor 
Vessel Internals Inspection plan that Duke Energy committed to 
provide prior to commencing inspections.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Gloria Kulesa.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina

    Date of amendment request: November 15, 2010.
    Description of amendment request: The proposed amendments would 
approve changes to the updated final safety analysis report to allow 
operation of a reverse osmosis system during normal plant operation to 
remove silica from borated water storage tank and the spent fuel pool.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

(1) Does the proposed amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    No. The proposed change requests Nuclear Regulatory Commission 
(NRC) approval of design features and controls that will be used to 
ensure that periodic limited operation of a Reverse Osmosis (RO) 
System during Unit operation does not significantly impact the 
Borated Water Storage Tank (BWST) or Spent Fuel Pool (SFP) function 
or other plant equipment. Duke Energy evaluated the effect of 
potential failures, identified precautionary measures that must be 
taken before and during RO System operation, and required operator 
actions to protect affected structures, systems, and components 
(SSCs) important to safety. The new high energy piping and non 
seismic piping being installed for the RO System is non QA-1 and is 
postulated to fail and cause an Auxiliary Building flood. Duke 
Energy determined that adequate time is available to isolate the 
flood source (BWST or SFP) prior to affecting SSCs important to 
safety.
    The existing Auxiliary Building Flood evaluation postulates a 
single break in the nonseismic piping occurring in a seismic event. 
The addition of the RO System will not increase the frequency of a 
seismic event. This event does not consider the amount of non-
seismic piping that is currently in the Auxiliary Building. The new 
piping is not more likely to fail as compared to the existing non-
seismic piping. The existing postulated source of the pipe break in 
the Auxiliary Building is due to the piping not being seismically 
designed. The new RO System piping is considered a potential source 
of a single pipe break for the same reason. Since the accident 
itself is defined as the failure of non-seismic pipe, the new non-
seismic piping does not increase the frequency of occurrence of an 
Auxiliary Building flood. The mitigation of an Auxiliary Building 
flood due to non seismic piping failure is by manual operator 
action. The same mitigation technique is used for the high energy 
line break.
    The RO System takes suction from the top of the SFP to protect 
SFP inventory. Plant procedures will prohibit the use of the RO 
System during the time period directly after an outage that requires 
the Unit 1 & 2 SFP level to be maintained higher than the Technical 
Specification (TS) Limiting Condition for Operation (LCO) 3.7.11 
level requirement. The higher level is required to support TS LCO 
3.10.1 requirements for Standby Shutdown Facility (SSF) Reactor 
Coolant (RC) Makeup System operability (due to the additional decay 
heat from the recently offloaded spent fuel). Plant procedures will 
also specify the siphon be broken during this time period so the SFP 
water above the RO suction point cannot be siphoned off if the RO 
piping breaks. The proposed change does not impact the fuel 
assemblies, the movement of fuel, or the movement of fuel shipping 
casks. The SFP boron concentration, level, and temperature limits 
will not be outside of required parameters due to restrictions/
requirements on the system's operation.
    The BWST is used for mitigation of Steam Generator Tube Rupture 
(SGTR), Main Steam Line Break (MSLB) and Loss of Coolant Accidents 
(LOCAs). The SGTR and MSLB are bounded by the [small-break] SBLOCA 
analyses with respect to the performance requirements for the [high 
pressure injection] HPI System. In the normal mode of Unit 
operation, the BWST is not an accident initiator. The SFP is assumed 
to maintain acceptable criticality margin for all abnormal and 
accident conditions including Fuel Handling Accidents (FHAs) and 
cask drop accidents. Both the BWST and SFP are specified by TS 
requirements to have minimum levels/volumes and boron 
concentrations. The BWST also has TS requirements for temperature. 
Prior to RO operation, procedures will require that minimum required 
initial boron concentration, and initial level/volume be adjusted 
and that the RO System be operated for a specified maximum time 
period before readjusting volume and boron concentration prior to 
another RO session to ensure that the TS specified boron 
concentration and level/volume limits for both the SFP and the BWST 
are not exceeded during RO System operation. Thus, the design 
functions of the BWST and the SFP will continue to be met during RO 
System operation.
    An Auxiliary Building flood due to a non-seismic RO System pipe 
break does not increase the consequences of the flood since the new 
non-seismic pipe break is bounded by the Auxiliary Building flood 
caused by existing non-seismic pipe breaks. Although the RO System 
will return water with lower boron concentration, procedural 
controls will ensure that the TS boron concentration level does not 
go below the limit. Thus, no adverse effects from decreased boron 
concentration levels will occur.
    Since the BWST and SFP will still have TS required boron 
concentration and level/volume, the mitigation of a LOCA or FHA does 
not result in an increase in dose consequence.
    Therefore, installation and operation of the RO System during 
Unit operation does not significantly increase the probability or 
consequences of any accident previously evaluated.

(2) Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    No. The RO System adds non-seismic piping in the Auxiliary 
Building. However, the break of a single non-seismic pipe in the 
Auxiliary Building has already been postulated as an event in the 
licensing basis. The RO System also does not create the possibility 
of a seismic event concurrent with a LOCA since a seismic event is a 
natural phenomena event. The RO System does not adversely affect the 
Reactor Coolant System pressure boundary. The suction to the RO 
System, when using the system for BWST purification, contains a 
normally closed manual seismic boundary valve so the seismic design 
criteria is met for separation of seismic/non-seismic piping 
boundaries.
    Duke Energy also evaluated potential releases of radioactive 
liquid to the environment due to RO System piping failures. Design 
features and administrative controls preclude release of radioactive 
materials outside the Auxiliary Building. Releases inside the 
Auxiliary Building are bounded by existing analyses.
    The SFP suction line is designed such that the SFP water level 
will not go below TS required levels, thus the fuel assemblies will 
have the TS required water level over them. Procedural controls will 
restrict the use of the RO System and require breaking vacuum on the 
SFP suction line when the SSF conditions require the SFP level be 
raised to support SSF RC Makeup System operability. Thus, the SFP 
water level will not be reduced below required water levels for 
these conditions. RO System operating restrictions will prevent 
reducing the SFP boron concentration below TS limits.
    Therefore, operation of the RO System during Unit operation will 
not create the possibility of a new or different kind of accident 
from any kind of accident previously evaluated.

(3) Does the proposed amendment involve a significant reduction in a 
margin of safety?

    No. The RO System adds non-seismic piping in the Auxiliary 
Building. Duke

[[Page 77912]]

Energy evaluated the impact of RO System operation on SSCs important 
to safety and determined that procedural controls will ensure that 
TS limits for SFP and BWST volume, temperature and boron 
concentration will continue to be met during RO operation. For the 
BWST, these controls will ensure the TS minimum BWST boron 
concentration and level are available to mitigate the consequences 
of a small break LOCA or a large break LOCA. For the SFP, these 
controls ensure the assumptions of the fuel handling and cask drop 
accident analyses are preserved. Additionally, the failure of non 
seismic RO System piping will not significantly impact SSCs 
important to safety. The BWST level may drop below the TS required 
level due to a rupture of the non seismic piping during a seismic 
event. However, due to the low probability of a seismic event 
coupled with the relatively short period of time the RO System will 
be aligned to the BWST, the possibility of dropping below the TS 
required level does not involve a significant reduction in the 
margin of safety. In addition, Oconee's licensing basis does not 
assume a design basis event occurs simultaneously with a seismic 
event. The proposed change does not significantly impact the 
condition or performance of SSCs relied upon for accident 
mitigation. This change does not alter the existing TS allowable 
values or analytical limits. The existing operating margin between 
Unit conditions and actual Unit setpoints is not significantly 
reduced due to these changes. The assumptions and results in any 
safety analyses are not impacted. Therefore, operation of the RO 
System during Unit operation does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Gloria Kulesa.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of amendment request: September 30, 2010.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 3.1.7, ``Standby Liquid Control 
(SLC) System,'' to add Surveillance Requirement (SR) 3.1.7.9 to verify 
sodium pentaborate enrichment prior to the addition to the SLC tank. 
The increase in boron-10 enrichment is needed to support future reloads 
of GE14 fuel by providing additional margin for preserving the shutdown 
objective of the SLC system. Reload analysis indicates that a core that 
is made up of a majority of GE14 fuel has a higher reactivity than 
previous Columbia Generating Station core designs warranting a 
corresponding increase in the shutdown capability of the SLC system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    Response: No.
    The SLC system is designed to provide sufficient negative 
reactivity to bring the reactor from full power to a subcritical 
condition at any time in a fuel cycle, without taking credit for 
control rod movement. The proposed changes to the SLC sodium 
pentaborate solution requirements maintain the capability of the SLC 
to perform this reactivity control function, and assure continued 
compliance with the requirements of 10 CFR 50.62 for ATWS [automatic 
transient without scram]. The proposed changes do not impact the 
LOCA [loss-of-coolant accident] suppression pool pH control function 
of SLC because single-pump minimum flow and sodium pentaborate 
solution concentration (weight percent) are not changed from the 
level credited in the LOCA analysis. The SLC is provided to mitigate 
ATWS events and LOCA and, as such, is not considered to be an 
initiator of the ATWS event, LOCA, or any other analyzed accident. 
The use of sodium pentaborate solution enriched with the boron-10 
isotope, which is chemically and physically similar to the current 
solution, does not alter the design or operation of the SLC or 
increase the likelihood of a system malfunction that could increase 
the consequences of an accident.
    Based on the above discussion, it is concluded that the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    Response: No.
    Injection of sodium pentaborate solution into the reactor vessel 
has been considered in the plant design. The proposed changes revise 
the SLC boron solution requirements such that the capability of the 
SLC system to bring the reactor to a subcritical condition without 
taking credit for control rod movement is maintained, considering 
operation with an equilibrium core of GE14 fuel. The use of sodium 
pentaborate solution enriched with the boron-10 isotope, which is 
chemically and physically similar to the current solution, does not 
alter the design, function, or operation of the SLC system. The 
correct boron-10 enrichment is assured by the proposed addition of 
an SR to the TS. The solution concentration and volume are not 
changed; thus, the existing minimum volume and solution and piping 
temperature specified in the TS will ensure that the boron remains 
in solution and does not precipitate out in the SLC storage tank or 
in the SLC pump suction piping. The minimum volume and concentration 
specified in the TS ensure that the LOCA suppression pool pH control 
function is not impacted.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.

3. Does the proposed change involve a significant reduction in a margin 
of safety?

    Response: No.
    The proposed changes revise the SLC boron solution requirements 
to maintain the capability of the SLC system to bring the reactor to 
a subcritical condition without taking credit for control rod 
movement. These changes support operation with an equilibrium core 
of GE14 fuel and assure continued compliance with the requirements 
of 10 CFR 50.62. The minimum required average boron-10 concentration 
in the reactor core, resulting from the injection of sodium 
pentaborate solution by the SLC system, has been determined using 
approved analytical methods. The analysis demonstrates that 
sufficient shutdown margin is maintained in the reactor such that 
the reactivity control function of the SLC system is assured. The 
additional quantity of boron included to account for imperfect 
mixing and leakage is maintained at 25 percent. No change in the 
solution pH or volume is made. Thus, the safety margin is maintained 
to bring the reactor subcritical in the event of an ATWS and to 
control suppression pool pH in the event of a LOCA.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan

    Date of amendment request: July 20, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.8.3, ``Diesel Fuel, Lube Oil, and 
Starting Air,'' by relocating the current stored

[[Page 77913]]

diesel fuel oil and lube oil numerical volume requirements from the TS 
to the TS Bases so that they may be modified under licensee control. 
The TS are modified so that the stored diesel fuel oil and lube oil 
inventory will require that a 7-day supply be available for either 
diesel generator. Condition A and Condition B in the Action table are 
revised and Surveillance Requirements (SR) 3.8.3.1 and 3.8.3.2 are 
revised to reflect the above change.
    The proposed changes also revise TS 3.8.3 by reducing the 
Completion Time for Condition C. Condition C currently requires that an 
inoperable fuel transfer system associated with fuel oil transfer pump 
P-18A be restored to operable status within 15 hours. The proposed TS 
change reduces the Completion Time for this Required Action from 15 to 
12 hours. The Completion Time is reduced to reflect the amount of time 
that an emergency diesel generator fuel oil day tank can support 
emergency diesel generator operation under design conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    Response: No.
    The proposed change relocates the volume of diesel fuel oil and 
lube oil required to support 7-day operation of the onsite emergency 
diesel generators, and the volume equivalent to a 6-day supply, to 
licensee control. The specific volume of fuel oil equivalent to a 7-
day and 6-day supply is calculated using the NRC approved 
methodology described in Regulatory Guide 1.137, Revision 1, ``Fuel 
Oil Systems for Standby Emergency diesel generators'' and ANSI N195-
1976, ``Fuel Oil Systems for Standby Diesel Generators.'' The 
specific volume of lube oil equivalent to a 7-day and 6-day supply 
is based on the emergency diesel generator manufacturer's 
consumption values for the run time of the diesel generator. Because 
the requirement to maintain a 7-day supply of diesel fuel oil and 
lube oil is not changed and is consistent with the assumptions in 
the accident analyses, and the actions taken when the volume of fuel 
oil and lube oil are less than a 6-day supply have not changed, 
neither the probability or the consequences of any accident 
previously evaluated will be affected.
    The proposed change also reduces the Completion Time for TS 
3.8.3 Condition C for an inoperable P-18A fuel transfer system from 
15 hours to 12 hours. Reducing the Completion Time to 12 hours 
bounds the 13.5-hour time duration that the emergency diesel 
generator day tank will support emergency diesel generator operation 
under accident loading conditions. The change in Completion Time 
does not affect required TS actions if the Completion Time is 
exceeded. The Completion Time change does not affect the probability 
or consequences of an accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    Response: No.
    The proposed fuel oil and lube oil changes do not involve a 
physical alteration of the plant (i.e., no new or different type of 
equipment will be installed) or a change in the methods governing 
normal plant operation. The change does not alter assumptions made 
in the safety analysis but ensures that the emergency diesel 
generator operates as assumed in the accident analysis. The proposed 
change is consistent with the safety analysis assumptions.
    The proposed change also reduces the Completion Time for TS 
3.8.3 Condition C for an inoperable P-18A fuel transfer system from 
15 hours to 12 hours. This change does not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed). This change does not create a condition in which 
a new or different kind of accident can occur. It does not alter 
assumptions made in the safety analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.

3. Does the proposed change involve a significant reduction in a margin 
of safety?

    Response: No.
    The proposed change relocates the volume of fuel oil and lube 
oil required to support 7-day operation of either emergency diesel 
generator, and the volume equivalent to a 6-day supply, to licensee 
control. As the bases for the existing limits on diesel fuel oil and 
lube oil are not changed, no change is made to the accident analysis 
assumptions and no margin of safety is reduced as part of this 
change.
    The proposed change also reduces the Completion Time for TS 
3.8.3 Condition C for an inoperable P-18A fuel transfer system from 
15 hours to 12 hours. There are no adverse affects on margins of 
safety since a more stringent operability requirement will be 
applied to the P-18A fuel transfer system.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White 
Plains, NY 10601.
    NRC Branch Chief: Robert J. Pascarelli.

Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and 
3, York and Lancaster Counties, Pennsylvania

    Date of amendment request: March 24, 2010, as supplemented by 
letter dated July 23, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3.1.7, ``Standby Liquid 
Control (SLC) System,'' to extend the completion time for Condition C 
(i.e., two SLC subsystems inoperable for reasons other than Condition 
A) from 8 hours to 72 hours.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (NSHC), which is presented below:

(1) Does the proposed amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    Response: No.
    The proposed amendment revises Technical Specification (TS) 
3.1.7, ``Standby Liquid Control (SLC) System,'' to extend the 
completion time (CT) for Condition C (i.e., ``Two SLC subsystems 
inoperable for reasons other than Condition A.'') from eight hours 
to 72 hours.
    The proposed change is based on a risk-informed evaluation 
performed in accordance with Regulatory Guides (RG) 1.174, ``An 
Approach for Using Probabilistic Risk Assessment in Risk-Informed 
Decisions On Plant-Specific Changes to the Licensing Basis,'' and RG 
1.177, ``An Approach for Plant-Specific, Risk-Informed Decision-
making: Technical Specifications.''
    The proposed amendment modifies an existing CT for a dual-train 
SLC System inoperability. The condition evaluated, the action 
requirements, and the associated CT do not impact any initiating 
conditions for any accident previously evaluated.
    The proposed amendment does not increase postulated frequencies 
or the analyzed consequences of an Anticipated Transient Without 
Scram (ATWS). Requirements associated with 10 CFR 50.62 will 
continue to be met. In addition, the proposed amendment does not 
increase postulated frequencies or the analyzed consequences of a 
large-break loss-of-coolant accident for which the SLC System is 
used for pH control. The new action requirement provides appropriate 
remedial actions to be taken in response to a dual-train SLC System

[[Page 77914]]

inoperability while minimizing the risk associated with continued 
operation. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

(2) Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    Response: No.
    The proposed amendment revises TS 3.1.7 to extend the CT for 
Condition C from eight hours to 72 hours. The proposed amendment 
does not involve any change to plant equipment or system design 
functions. This proposed TS amendment does not change the design 
function of the SLC System and does not affect the system's ability 
to perform its design function. The SLC System provides a method to 
bring the reactor, at any time in a fuel cycle, from full power and 
minimum control rod inventory to a subcritical condition with the 
reactor in the most reactive xenon free state without taking credit 
for control rod movement. Required actions and surveillance 
requirements are sufficient to ensure that the SLC System functions 
are maintained. No new accident initiators are introduced by this 
amendment. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.

(3) Does the proposed amendment involve a significant reduction in a 
margin of safety?

    Response: No.
    The proposed amendment revises TS 3.1.7 to extend the CT for 
Condition C from eight hours to 72 hours. The proposed amendment 
does not involve any change to plant equipment or system design 
functions. The margin of safety is established through the design of 
the plant structures, systems, and components, the parameters within 
which the plant is operated and the setpoints for the actuation of 
equipment relied upon to respond to an event.
    Safety margins applicable to the SLC System include pump 
capacity, boron concentration, boron enrichment, and system response 
timing. The proposed amendment does not modify these safety margins 
or the setpoints at which SLC is initiated, nor does it affect the 
system's ability to perform its design function. In addition, the 
proposed change complies with the intent of the defense-in-depth 
philosophy and the principle that sufficient safety margins are 
maintained consistent with RG 1.177 requirements (i.e., Section C, 
``Regulatory Position,'' paragraph 2.2,``Traditional Engineering 
Considerations''). Therefore, the proposed amendment does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves NSHC.
    Attorney for licensee: Mr. J. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center, Linn County, Iowa

    Date of amendment request: August 12, 2010.
    Description of amendment request: A change is proposed to the 
technical specifications to allow a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an unavailable barrier, if risk is assessed and managed 
consistent with the program in place for complying with the 
requirements of 10 CFR 50.65(a)(4). LCO 3.0.9 will be added to 
individual TS providing this allowance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration by affirming the applicability of the model analysis 
presented in the Federal Register notice dated October 3, 2006, 
starting on page 71 FR 58452, which is presented below:

Criterion 1: The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an unavailable barrier if risk is assessed and managed. 
The postulated initiating events which may require a functional 
barrier are limited to those with low frequencies of occurrence, and 
the overall TS system safety function would still be available for 
the majority of anticipated challenges. Therefore, the probability 
of an accident previously evaluated is not significantly increased, 
if at all. The consequences of an accident while relying on the 
allowance provided by proposed LCO 3.0.9 are no different than the 
consequences of an accident while relying on the TS required actions 
in effect without the allowance provided by proposed LCO 3.0.9. 
Therefore, the consequences