[Federal Register Volume 75, Number 239 (Tuesday, December 14, 2010)]
[Notices]
[Pages 77906-77919]
From the Federal Register Online via the Government Publishing 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, http://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 
<a href="/cdn-cgi/l/email-protection#0e666b6f7c676069206a616d656b7a4e607c6d20696178"><span class="__cf_email__" data-cfemail="e78f8286958e8980c98388848c8293a7899584c9808891">[email protected]</span></a>, 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 http://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 
http://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 http://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 http://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 http://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
<a href="/cdn-cgi/l/email-protection#db9688939ff589bea8b4aea9b8be9bb5a9b8f5bcb4ad"><span class="__cf_email__" data-cfemail="2469776c600a7641574b51564741644a56470a434b52">[email protected]</span></a>, 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 
http://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, 
http://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 <a href="/cdn-cgi/l/email-protection#dfafbbadf1adbaacb0aaadbcba9fb1adbcf1b8b0a9"><span class="__cf_email__" data-cfemail="e0908492ce9285938f95928385a08e9283ce878f96">[email protected]</span></a>.

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 of an accident previously evaluated are 
not significantly affected by this change. The addition of a 
requirement to assess and manage the risk introduced by this change 
will further minimize possible concerns.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

Criterion 2: The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to an unavailable barrier, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns.
    Thus, this change does not create the possibility of a new or 
different kind of accident from an accident previously evaluated.

Criterion 3: The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The proposed change allows a delay time for entering a supported 
system 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. The risk impact of the proposed TS changes was assessed 
following the three-tiered approach recommended in [Regulatory 
Guide] RG 1.177. A bounding risk assessment was performed to justify 
the proposed TS changes. This application of LCO 3.0.9 is predicated 
upon the licensee's performance of a risk assessment and the 
management of plant risk. The net change to the margin of safety is 
insignificant as indicated by the anticipated low levels of 
associated risk (ICCDP and ICLERP) as shown in Table 1 of Section 
3.1.1 in the [model] Safety Evaluation [on page 71 FR 58450 of the 
Federal Register dated October 3, 2006].
    Therefore, this 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. M. S. Ross, Florida Power & Light 
Company, P. O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Robert J. Pascarelli.

South Carolina Electric and Gas Company (SCE and G), South Carolina 
Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear 
Station, Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: November 11, 2010.
    Description of Amendment Request: The licensee proposes to amend 
the operating license for Virgil C. Summer Nuclear Station (VCSNS), by 
revising

[[Page 77915]]

the Technical Specifications (TS) and SCE&G proposes to provide 
surveillance enhancements that will improve operation and testing of 
the Emergency Diesel Generators (EDG). The changes will provide a more 
restrictive voltage and frequency band for operation when not connected 
in parallel with the offsite sources.
    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?

    No.
    The changes proposed by this license amendment will revise the 
Surveillance Requirements of Technical Specification \3/4\.8.1, AC 
SOURCES--OPERATING, to expand the continuous rated load 
specification to a range of 90% to 100% of the continuous rated 
load, specify an overload range of 105% to 110% of the continuous 
rated load, add a power factor limit while testing, allow gradual 
loading and unloading of the EDG, specify a maximum frequency for 
the overspeed limit, specify a maximum allowable overspeed voltage, 
and add a more restrictive voltage and frequency band for testing 
during steady state operation.
    The majority of these changes are being proposed in order to 
implement recommendations contained in [Institute of Nuclear Power 
Operations] INPO Significant Operating Experience Report (SOER) 03-
01, Emergency Power Reliability, Recommendation Number 5, which 
recommends that the utility review testing practices for emergency 
power systems to verify that the practices are representative of 
actual demand conditions and appropriately exercise equipment that 
is expected to respond in an actual demand condition. These changes 
are based on the guidance provided by Regulatory Guide 1.9, Revision 
3, Selection, Design, Qualification, and Testing of Emergency Diesel 
Generator Units Used as Class 1E Onsite Electric Power Systems at 
Nuclear Power Plant.
    The more restrictive voltage and frequency band for testing 
during steady state operation is proposed to ease the impact of EDG 
voltage and frequency that are being incorporated into the Charging 
Pump performance requirements. The allowable voltage and frequency 
uncertainty limits for steady state operation are being reduced. 
This will ensure that the Charging Pumps continue to operate within 
their analyzed range.
    These changes do not affect the probability or consequences of 
an accident previously evaluated because the proposed changes do not 
make a change to any accident initiator, initiating condition, or 
assumption. The proposed changes do not involve a significant change 
to the plant design or operation. These changes do not invalidate 
assumptions used in evaluating the radiological consequences of an 
accident, do not alter the source term or containment isolation, and 
do not provide a new radiation release path or alter a potential 
radiological release. 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?

    No.
    These changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed changes do not introduce a new or different 
accident initiator or introduce a new or different equipment failure 
mode or mechanism.
    No changes are being made in equipment hardware or software, 
operational philosophy, testing frequency, or how the system 
actually operates. Therefore, the proposed amendment will 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?

    No.
    These changes do not involve a significant reduction in a margin 
of safety because the proposed changes do not reduce the margin of 
safety that exists in the present Technical Specifications or 
Updated Final Safety Analysis Report. The operability requirements 
of the Technical Specifications are consistent with the initial 
condition assumptions of the safety analyses. The proposed changes 
do not affect the Action statement requirements for the various 
levels of degradation in the EDG. Therefore, the proposed change 
does not involve a significant reduction in a margin of safety.
    Based on the above, SCE&G concludes that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    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: J. Hagood Hamilton, Jr., South Carolina 
Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 
29218.
    NRC Branch Chief: Gloria Kulesa.

Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, Houston 
County, Alabama

    Date of amendment request: October 29, 2010.
    Description of amendment request: The proposed amendments request 
the adoption of an approved change to the standard technical 
specifications for Westinghouse Plants (NUREG-1431), to allow 
relocation of specific Technical Specifications (TS) surveillance 
frequencies to a licensee-controlled program. The proposed change is 
described in Technical Specification Task Force (TSTF) Traveler, TSTF-
425, Revision 3, ``Relocate Surveillance Frequencies to Licensee 
Control--RITSTF Initiative 5b'' (Agencywide Documents Access and 
Management System (ADAMS) Accession No. ML080280275), and was described 
in the Notice of Availability published in the Federal Register (FR) on 
July 6, 2009 (74 FR 31996). The proposed changes are consistent with 
NRC-approved TSTF-425, Revision 3. The proposed change relocates 
surveillance frequencies to a licensee-controlled program, the 
surveillance frequency control program. This change is applicable to 
licensees using probabilistic risk guidelines contained in NRC-approved 
[Nuclear Energy Institute] NEI 04-10, ``Risk-Informed Technical 
Specifications Initiative 5b, Risk-Informed Method for Control of 
Surveillance Frequencies,'' (ADAMS Accession No. 071360456).
    The licensee affirmed the applicability to the FNP of the model no 
significant hazards consideration determination provided in the FR on 
July 6, 2009 (74 FR 31996), in its application dated October 29, 2010.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the analysis of the 
issue of no significant hazards consideration 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 specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program [SFCP]. Surveillance 
frequencies are not an initiator to any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not significantly increased. The systems and components 
required by the Technical Specifications for which the surveillance 
frequencies are relocated are still required to be operable, meet 
the acceptance criteria for the surveillance requirements, and be 
capable of performing any mitigation function assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly increased.

[[Page 77916]]

    Therefore, the proposed change does 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.
    No new or different accidents result from utilizing the proposed 
change. The 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. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    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 the 
margin of safety?

    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, the 
licensee will perform a probabilistic risk evaluation using the 
guidance contained in NRC approved NEI 04-10, Rev. 1, in accordance 
with the TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable 
acceptance guidelines and methods for evaluating the risk increase 
of proposed changes to surveillance frequencies consistent with 
Regulatory Guide 1.177.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based upon the reasoning presented above, licensee concludes 
that the requested change does not involve a significant hazards 
consideration as set forth in 10 CFR 50.92(c), Issuance of 
Amendment.

    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: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Branch Chief: Gloria J. Kulesa.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant (HNP), Units 1 and 2, Appling County, Georgia

    Date of amendment request: October 29, 2010.
    Description of amendment request: The proposed amendments request 
the adoption of an approved change to the standard technical 
specifications for General Electric Plants, BWR/4 (NUREG-1433), to 
allow relocation of specific Technical Specification (TS) surveillance 
frequencies to a licensee-controlled program. The proposed change is 
described in Technical Specification Task Force (TSTF) Traveler, TSTF-
425, Revision 3, ``Relocate Surveillance Frequencies to Licensee 
Control--RITSTF Initiative 5b.'' (Agencywide Documents Access and 
Management System (ADAMS) Accession No. ML080280275), and was described 
in the Notice of Availability published in the Federal Register (FR) on 
July 6, 2009 (74 FR 31996). The proposed changes are consistent with 
NRC-approved TSTF-425, Revision 3. The proposed change relocates 
surveillance frequencies to a licensee-controlled program, the 
surveillance frequency control program. This change is applicable to 
licensees using probabilistic risk guidelines contained in NRC-approved 
[Nuclear Energy Institute] NEI 04-10, ``Risk-Informed Technical 
Specifications Initiative 5b, Risk-Informed Method for Control of 
Surveillance Frequencies,'' (ADAMS Accession No. 071360456). The 
licensee affirmed the applicability to the HNP of the model no 
significant hazards consideration determination provided in the FR on 
July 6, 2009 (74 FR 31996) in its application dated October 29, 2010.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the analysis of the 
issue of no significant hazards consideration 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 specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program [SFCP]. Surveillance 
frequencies are not an initiator to any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not significantly increased. The systems and components 
required by the Technical Specifications for which the surveillance 
frequencies are relocated are still required to be operable, meet 
the acceptance criteria for the surveillance requirements, and be 
capable of performing any mitigation function assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly increased.
    Therefore, the proposed change does 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.
    No new or different accidents result from utilizing the proposed 
change. The 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. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    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 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, SNC will 
perform a probabilistic risk evaluation using the guidance contained 
in NRC approved NEI 04-10, Rev. 1, in accordance with the TS SFCP. 
NEI 04-10, Rev.1, methodology provides reasonable acceptance 
guidelines and methods for evaluating the risk increase of proposed 
changes to surveillance frequencies consistent with Regulatory Guide 
1.177.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based upon the reasoning presented above, licensee concludes 
that the requested change does not involve a significant hazards 
consideration as set forth in 10 CFR 50.92(c), Issuance of 
Amendment.

    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

[[Page 77917]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Branch Chief: Gloria J. Kulesa.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of amendment request: February 18, 2010, as supplemented on 
November 12, 2010 (TS-468).
    Description of amendment request: The proposed amendment would 
modify Technical Specification 3.8.1 to extend the completion time (CT) 
for the return of an inoperable emergency diesel generator (DGs) to 
operable status from 7 days to 14 days, based on the availability of 
two non-safety related temporary diesel generators (TDGs). Commensurate 
changes to the maximum completion times were also proposed, extending 
the times from 14 to 21 days in Required Actions A.3 and B.4. The 
change also eliminates a historical footnote for a previous CT for Unit 
3 only that is no longer needed.
    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.
    The proposed changes do not affect the design of the DGs, the 
operational characteristics or function of the DGs, the interfaces 
between the DGs and other plant systems, or the reliability of the 
DGs. Required Actions and their associated CTs are not considered 
initiating conditions for any UFSAR [updated final safety analysis 
report] accident previously evaluated, nor are the DGs considered 
initiators of any previously evaluated accidents. The DGs are 
provided to mitigate the consequences of previously evaluated 
accidents, including a loss of off-site power.
    The consequences of previously evaluated accidents will not be 
significantly affected by the extended DG CT, because a sufficient 
number of onsite Alternating Current [AC] power sources will 
continue to remain available to perform the accident mitigation 
functions associated with the DGs, as assumed in the accident 
analyses. In addition, as a risk mitigation and defense-in-depth 
action, an independent AC power source, via two available TDGs, will 
be available to support the ESF [engineered safety feature] bus with 
the inoperable DG during a SBO [station blackout].
    Therefore, the proposed changes do 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 does not involve a change in the permanent 
design, configuration, or method of operation of the plant. The 
proposed changes will not alter the manner in which equipment 
operation is initiated, nor will the functional demands on credited 
equipment be changed. The proposed changes allow operation of the 
unit to continue while a DG is repaired and retested with the TDGs 
in standby to mitigate a SBO event. The proposed extensions do not 
affect the interaction of a DG with any system whose failure or 
malfunction can initiate an accident. As such, no new failure modes 
are being introduced. 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 amendment involve a significant reduction in a 
margin of safety?

    Response: No.
    The proposed changes do not alter the permanent plant design, 
including instrument set points, nor does it change the assumptions 
contained in the safety analyses. The standby TDG alternate AC 
system is designed with sufficient redundancy such that a DG may be 
removed from service for maintenance or testing. The remaining seven 
DGs are capable of carrying sufficient electrical loads to satisfy 
the UFSAR requirements for accident Mitigation or unit safe 
shutdown. The proposed changes do not impact the redundancy or 
availability requirements of offsite power supplies or change the 
ability of the plant to cope with station blackout events. 
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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, Tennessee 37902.
    NRC Branch Chief: Douglas A. Broaddus.
mitigation or unit safe shutdown. The proposed changes do not impact 
the redundancy or availability requirements of offsite power supplies 
or change the ability of the plant to cope with station blackout 
events. 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, Tennessee 37902.
    NRC Branch Chief: Douglas A. Broaddus.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: October 21, 2010.
    Description of amendment request: The proposed amendment would 
correct a typographical error in Section 5, Administrative Controls, of 
the Technical Specifications (TSs). The current TSs, on page 5.0-31, 
has two paragraphs numbered as 5.7.2d.3. The amendment proposes to 
renumber the second paragraph as 5.7.2d.4. The typographical error was 
introduced in Amendment No. 123 issued on March 31, 1999.
    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.
    The proposed change is administrative in nature. The change 
involves correcting a typographical error. This change does not 
affect possible initiating events for accidents previously evaluated 
or alter the configuration or operation of the facility. The 
Limiting Safety System Settings and Safety Limits specified in the 
TS remain unchanged.
    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 previously evaluated?

    Response: No.
    The proposed change is administrative in nature. The safety 
analysis of the facility remains complete and accurate. There are no 
physical changes to the facility and the plant conditions for which 
the design basis accidents have been evaluated are still valid. The 
operating procedures and emergency procedures are unaffected. 
Consequently no new failure modes are introduced as a result of the 
proposed change.

[[Page 77918]]

    Therefore, the proposed change 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 is administrative in nature. Since there [are] 
no changes to the operation of the facility or the physical design, the 
Updated Safety Analysis Report (USAR) design basis, accident 
assumptions, or TS Bases are not affected.
    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: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Indiana Michigan Power Company (IandM), Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendment: September 8, 2010.
    Brief description of amendment: The licensee proposed to delete the 
Technical Specification requirements related to the containment 
hydrogen recombiners and the hydrogen monitors, in accordance with 
Nuclear Energy Institute Technical Specification Task Force (TSTF) 
initiative designated as TSTF-447.
    Date of publication of individual notice in Federal Register: 
October 14, 2010 (75 FR 63209).
    Expiration date of individual notice: December 13, 2010.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (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 (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#98e8fceab6eafdebf7edeafbfdd8f6eafbb6fff7ee"><span class="__cf_email__" data-cfemail="88f8ecfaa6faedfbe7fdfaebedc8e6faeba6efe7fe">[email protected]</span></a>.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendment: November 30, 2009, as 
supplemented by letter dated July 22, 2010.
    Brief description of amendment: The amendments revised Table 3.3.5-
1 of Technical Specification (TS) 3.3.5, ``Engineered Safety Features 
Actuation System (ESFAS) Instrumentation,'' to raise the refueling 
water tank (RWT) low level allowable values for the recirculation 
actuation signal; raised the minimum required RWT volume shown in TS 
Figure 3.5.5-1 of TS 3.5.5, ``Refueling Water Tank (RWT)''; and 
implemented a time-critical operator action to close the RWT isolation 
valves, including consideration of a potentially more limiting single 
failure of a low-pressure safety injection pump to automatically stop, 
as designed, on a recirculation actuation signal.
    Date of issuance: November 24, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: Unit 1--182; Unit 2--182; Unit 3--182.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendment revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: April 20, 2010 (75 FR 
20629). The supplemental letter dated July 22, 2010, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 24, 2010.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, LLC, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of application for amendments: April 5, 2010.
    Brief description of amendments: The amendment made title changes 
and corrections within Technical Specification (TS) 5.0, 
``Administrative

[[Page 77919]]

Controls.'' Specifically, the changes included:
    (1) Replacement of the use of plant specific titles to generic 
titles consistent with TS Task Force (TSTF) Traveler TSTF-65, Revision 
1, ``Use of Generic Titles for Utility Positions,''
    (2) Changes made to more closely align selected TSs with the 
Improved Standard TSs, and
    (3) Administrative changes to specified TSs.
    Date of issuance: November 22, 2010.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 296 for Unit 1 and 272 for Unit 2.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the License and Technical Specifications.
    Date of initial notice in Federal Register: June 1, 2010 (75 FR 
30443).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated November 22, 2010.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 2nd day of December 2010.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2010-31086 Filed 12-13-10; 8:45 am]
BILLING CODE 7590-01-P


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