Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 4469-4476 [2013-01010]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 78, Number 14 (Tuesday, January 22, 2013)]
[Notices]
[Pages 4469-4476]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-01010]


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NUCLEAR REGULATORY COMMISSION

[NRC-2013-0012]


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

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 or 
combined license, as applicable, 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 December 27, 2012 to January 9, 2013. The 
last biweekly notice was published on January 8, 2013 (78 FR 1267).

ADDRESSES: You may access information and comment submissions related 
to this document, which the NRC possesses and are publically available, 
by searching on https://www.regulations.gov under Docket ID NRC-2013-
0012. You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2013-0012. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: Carol.Gallagher@nrc.gov.
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
     Fax comments to: RADB at 301-492-3446.
    For additional direction on accessing information and submitting 
comments, see ``Accessing Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID NRC-2013-0012 when contacting the NRC 
about the availability of information regarding this document. You may 
access information related to this document by any of the following 
methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2013-0012.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly-available documents online in the NRC 
Library at https://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to pdr.resource@nrc.gov. Documents may be viewed in 
ADAMS by performing a search on the document date and docket number.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2013-0012 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information that that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.
Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses and Combined 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 section 50.92 of Title 10 of the Code 
of Federal Regulations (10 CFR), 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 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

[[Page 4470]]

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.
    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 or 
combined 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 NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. The NRC regulations are accessible electronically from the NRC 
Library on the NRC's Web site at https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or a presiding 
officer designated by the Commission or by the Chief Administrative 
Judge of the Atomic Safety and Licensing Board Panel, will rule on the 
request and/or petition; and the Secretary or the Chief Administrative 
Judge of the Atomic Safety and Licensing Board will issue a notice of a 
hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then 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 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital identification (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 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 the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in,

[[Page 4471]]

is available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with the NRC guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at https://www.nrc.gov/site-help/e-submittals.html, by email at 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866 672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, 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. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the following three 
factors in 10 CFR 2.309(c)(1): (i) The information upon which the 
filing is based was not previously available; (ii) the information upon 
which the filing is based is materially different from information 
previously available; and (iii) the filing has been submitted in a 
timely fashion based on the availability of the subsequent information.
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through ADAMS in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to 
ADAMS or who encounter problems in accessing the documents located in 
ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209, 
301-415-4737, or by email to pdr.resource@nrc.gov.
    Detroit Edision, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan
    Date of amendment request: November 13, 2012.
    Description of amendment request: The proposed amendment would 
modify Technical Specification requirements to operate ventilation 
systems with charcoal filters for 10 hours each in accordance with 
Technical Specifications Task Force (TSTF)-522, Revision 0, ``Revise 
Ventilation System Surveillance Requirements to Operate for 10 hours 
per Month.''
    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 replaces an existing Surveillance 
Requirement to operate the SGT System and CREF System equipped with 
electric heaters for a continuous 10 hour period every 31 days with 
a requirement to operate the systems for 15 continuous minutes with 
heaters operating.
    These systems are not accident initiators and therefore, these 
changes do not involve a significant increase in the probability of 
an accident. The proposed system and filter testing changes are 
consistent with current regulatory guidance for these systems and 
will continue to assure that these systems perform their design 
function which may include mitigating accidents. Thus the change 
does not involve a significant increase in the consequences of an 
accident.
    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 change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change replaces an existing Surveillance 
Requirement to operate the SGT System and CREF System equipped with 
electric heaters for a continuous 10 hour period every 31 days with 
a requirement to operate the systems for 15 continuous minutes with 
heaters operating.

[[Page 4472]]

    The change proposed for these ventilation systems does not 
change any system operations or maintenance activities. Testing 
requirements will be revised and will continue to demonstrate that 
the Limiting Conditions for Operation are met and the system 
components are capable of performing their intended safety 
functions. The change does not create new failure modes or 
mechanisms and no new accident precursors are generated.
    Therefore, it is concluded that this change does 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 replaces an existing Surveillance 
Requirement to operate the SGT System and CREF System equipped with 
electric heaters for a continuous 10 hour period every 31 days with 
a requirement to operate the systems for 15 continuous minutes with 
heaters operating.
    The design basis for the ventilation systems' heaters is to heat 
the incoming air which reduces the relative humidity. The heater 
testing change proposed will continue to demonstrate that the 
heaters are capable of heating the air and will perform their design 
function. The proposed change is consistent with regulatory 
guidance.

    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: Bruce R. Masters, DTE Energy, General 
Counsel--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.
    NRC Branch Chief: Robert D. Carlson.
    Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Power Station, Unit 2, New London County, Connecticut
    Date of amendment request: December 17, 2012.
    Description of amendment request: The proposed amendment would 
revise the Millstone Power Station, Unit 2 (MPS2) Technical 
Specification (TS) Surveillance Requirement 4.4.3.2 to remove the 
requirement to perform the quarterly surveillance for a pressurizer 
power-operated relief valve (PORV) block valve that is being maintained 
closed in accordance with TS 3.4.3 Action a. The proposed change is 
consistent with the requirements of the standard Technical 
Specification for Combustion Engineering plants (NUREG-1432).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), of Title 10 of the Code 
of Federal Regulations (10 CFR), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:

Criterion 1

    Will operation of the facility in accordance with the proposed 
change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The block valve for the pressurizer PORV is not a potential 
accident initiator. Therefore, not requiring a surveillance of the 
block valve while it is being used to isolate its associated PORV 
will not increase the probability of an accident previously 
evaluated. Not requiring the surveillance of the block valve may 
slightly reduce the probability of a loss of coolant accident from a 
stuck open PORV since it will eliminate the challenge to the PORV 
from the pressure transient that results from cycling the block 
valve.
    The PORVs or the PORV block valves are not credited in the MPS2 
Final Safety Analysis Report (FSAR), Chapter 14, ``Safety 
Analysis,'' for event mitigation. If pressurizer spray is not 
available or is not effective, either one or the two pressurizer 
PORVs may be manually actuated to depressurize the RCS in response 
to certain transients. Not performing the surveillance on the block 
valve is not relevant to the primary system for depressurizing the 
RCS (pressurizer spray). The block valves have been demonstrated by 
operating experience to be reliable and are also subject to the 
motor-operated valve testing program. Consequently, the proposed 
change does not significantly reduce the confidence that the block 
valve can be opened to permit manual actuation of the PORV to 
depressurize the RCS.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Criterion 2

    Will operation of the facility in accordance with the proposed 
change create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    Response: No.
    The proposed change only affects the performance of the 
surveillance test for the block valve and does not involve any 
physical alteration of plant equipment or introduce any operating 
configurations not previously evaluated. The pressurizer PORV block 
valves provide isolation for a postulated stuck-open or leaking 
PORV. Isolation is satisfied with the block valve closed in 
accordance with SR 4.4.3.2. PORV block valve closure is not credited 
in FSAR Chapter 14 for inadvertent opening of the PORV event 
mitigation.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.

Criterion 3

    Will operation of the facility in accordance with the proposed 
change involve a significant reduction in the margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident. These barriers include the fuel 
cladding, the reactor coolant system, and the containment system. 
These barriers are not significantly affected by the changes 
proposed herein. 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 establishment 
of setpoints for the actuation of equipment relied upon to respond 
to an event, and thereby protect the fission product barriers. The 
proposed change to the surveillance requirement for the presurrizer 
PORV block valve does not affect the assumptions in any accident 
analysis.
    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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: George A. Wilson.
    Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County, 
Texas
    Date of amendment request: October 2, 2012.
    Brief description of amendments: The amendments would revise 
Technical Specification (TS) 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' and TS 3.3.2, ``Engineered Safety Feature Actuation 
System (ESFAS) Instrumentation,'' to relocate the TS requirements for 
the following instruments to the Technical Requirements Manual (TRM), a 
licensee-controlled document, under 10 CFR 50.59:
     Pressurizer Water level--High (RTS Function No. 9)
     Trip of all Main Feedwater Pumps (ESFAS Function No. 6.g)
     ESFAS Interlock, Reactor Trip, P-4 (ESFAS Function No. 
8.a)
    The proposed changes would relocate the TS requirements in their 
entirety and not result in deletion or alteration of any RTS or ESFAS 
requirements. The proposed relocation of the TS requirements for these 
RTS and ESFAS instrument Functions is based on the

[[Page 4473]]

application of the TS criteria of 10 CFR 50.36(c)(2)(ii).
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the TS does not affect the initiators of 
any analyzed accident. In addition, operation in accordance with the 
proposed TS change will continue to ensure that the previously 
evaluated accidents will be mitigated as analyzed. Thus, the 
proposed change does not adversely affect the design function or 
operation of any structures, systems, and components important to 
safety.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes 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 physical alteration of 
the plant (no new or different type of equipment will be installed). 
The proposed change does not create any new failure modes for 
existing equipment or any new limiting single failures. Additionally 
the proposed change does not involve a change in the methods 
governing normal plant operation and all safety functions will 
continue to perform as previously assumed in accident analyses. 
Thus, the proposed change does not adversely affect the design 
function or operation of any structures, systems, and components 
important to safety.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will not adversely affect the operation of 
plant equipment or the function of equipment assumed in the accident 
analyses. The proposed changes to the RTS and ESFAS TS requirements 
do not change the RTS or ESFAS design and capability to perform the 
required safety functions consistent with the assumptions of the 
applicable safety analyses. In addition, operation in accordance 
with the proposed TS change will continue to ensure that the 
previously evaluated accidents will be mitigated as analyzed.
    Therefore, the proposed change does not involve a 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: Timothy P. Matthews, Esq., Morgan, Lewis and 
Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Branch Chief: Michael T. Markley.
    NextEra Energy Seabrook, LLC Docket No. 50-443, Seabrook Station, 
Unit 1, Rockingham County, New Hampshire
    Date of amendment request: December 20, 2012.
    Description of amendment request: The proposed amendment will 
revise the Seabrook Technical Specifications (TS) TS 6.7.6.m, ``Reactor 
Coolant Pump Flywheel Inspection Program.'' The proposed amendment will 
extend the reactor coolant pump (RCP) motor flywheel examination 
frequency from the currently approved 10-year inspection interval, to 
an interval not to exceed 20 years. The changes are consistent with 
Industry/Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-421, ``Revision to RCP Flywheel 
Inspection Program (WCAP-15666).'' The availability of this TS 
improvement was announced in the Federal Register on October 22, 2003, 
as part of the consolidated line item improvement process (CLIIP).
    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) through incorporation by reference of the NSHC 
published in the Federal Register Notice dated June 24, 2003 (68 FR 
37590), 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 to the RCP flywheel examination frequency 
does not change the response of the plant to any accidents. The RCP 
will remain highly reliable and the proposed change will not result 
in a significant increase in the risk of plant operation. Given the 
extremely low failure probabilities for the RCP motor flywheel 
during normal and accident conditions, the extremely low probability 
of a loss-of-coolant accident (LOCA) with loss of offsite power 
(LOOP), and assuming a conditional core damage probability (CCDP) of 
1.0 (complete failure of safety systems), the core damage frequency 
(CDF) and change in risk would still not exceed the NRC's acceptance 
guidelines contained in RG 1.174 (<1.0E-6 per year). Moreover, 
considering the uncertainties involved in this evaluation, the risk 
associated with the postulated failure of an RCP motor flywheel is 
significantly low. Even if all four RCP motor flywheels are 
considered in the bounding plant configuration case, the risk is 
still acceptably low.
    The proposed change does not adversely affect accident 
initiators or precursors, nor alter the design assumptions, 
conditions, or configuration of the facility, or the manner in which 
the plant is operated and maintained; alter or prevent the ability 
of structures, systems, components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits; or affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the type or amount of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposure. The proposed change is 
consistent with the safety analysis assumptions and resultant 
consequences.
    Therefore, the proposed 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 Accident Previously

    The proposed change in flywheel inspection frequency does not 
involve any change in the design or operation of the RCP. Nor does 
the change to examination frequency affect any existing accident 
scenarios, or create any new or different accident scenarios. 
Further, the change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or alter the methods governing normal plant operation. In 
addition, the change does not impose any new or different 
requirements or eliminate any existing requirements, and does not 
alter any assumptions made in the safety analysis. The proposed 
change is consistent with the safety analysis assumptions and 
current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    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 impacted by this change. The proposed change will not result 
in plant operation in a configuration outside of the design basis. 
The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in RG 1.174. There are no significant 
mechanisms for inservice degradation of the RCP flywheel. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the analysis and, based on this review, 
it appears that the three standards of

[[Page 4474]]

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves NSHC.
    Attorney for licensee: James Petro, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Meena Khanna.
    PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating 
Station, Unit 1, Salem County, New Jersey
    Date of amendment request: May 8, 2012.
    Description of amendment request: The proposed amendment would 
revise Salem Unit 1 Technical Specification (TS) 6.8.4.i, ``Steam 
Generator (SG) Program,'' to permanently exclude portions of the tube 
below the top of the steam generator tubesheet from periodic steam 
generator tube inspections. In addition, this amendment proposes to 
revise TS 6.9.1.10, ``Steam Generator Tube Inspection Report,'' to 
provide permanent reporting requirements that have been previously 
established on an interim basis.
    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, with the NRC staff edits in 
square brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed change 
that alters the steam generator inspection criteria does not have a 
detrimental impact on the integrity of any plant structure, system, 
or component that initiates an analyzed event. The proposed change 
will not alter the operation of, or otherwise increase the failure 
probability of any plant equipment that initiates an analyzed 
accident.
    Of the applicable accidents previously evaluated, the limiting 
transients with consideration to the proposed change to the steam 
generator tube inspection and repair criteria are the steam 
generator tube rupture (SGTR) event, the steam line break (SLB) and 
the feedline break (FLB) postulated accidents.
    Addressing the SGTR event, the required structural integrity 
margins of the steam generator tubes and the tube-to-tubesheet joint 
over the H* distance will be maintained. Tube rupture in tubes with 
cracks within the tubesheet is precluded by the presence of the 
tubesheet and constraint provided by the tube-to-tubesheet joint. 
Tube burst cannot occur within the thickness of the tubesheet. The 
tube-to-tubesheet joint constraint results from the hydraulic 
expansion process, thermal expansion mismatch between the tube and 
tubesheet, from the differential pressure between the primary and 
secondary side, and tubesheet deflection. The structural margins 
against burst, as discussed in Regulatory Guide (RG) 1.121, ``Bases 
for Plugging Degraded PWR [pressurized-water reactor] Steam 
Generator Tubes,'' and TS 6.8.4.i are maintained for both normal and 
postulated accident conditions.
    The proposed change has no impact on the structural or leakage 
integrity of the portion of the tube outside of the tubesheet. The 
proposed change maintains structural and leakage integrity of the 
steam generator tubes consistent with the performance criteria in TS 
6.8.4.i. Therefore, the proposed change results in no significant 
increase in the probability of the occurrence of a SGTR accident.
    At normal operating pressures, leakage from tube degradation 
below the proposed limited inspection depth is limited by the tube-
to-tubesheet joint. Consequently, negligible normal operating 
leakage is expected from degradation below the inspected depth 
within the tubesheet region. The consequences of an SGTR event are 
not affected by the primary to secondary leakage flow during the 
event as primary to secondary leakage flow through a postulated tube 
that has been pulled out of the tubesheet is essentially equivalent 
to a severed tube. Therefore, the proposed changes do not result in 
a significant increase in the consequences of a SGTR.
    The consequences of a SLB or FLB are also not significantly 
affected by the proposed changes. The leakage analysis shows that 
the primary-to-secondary leakage during a SLB/FLB event would be 
less than or equal to that assumed in the Updated Safety Analysis 
Report.
    Primary-to-secondary leakage from tube degradation in the 
tubesheet area during the limiting accidents (i.e., SLB/FLB) is 
limited by flow restrictions. These restrictions result from the 
crack and tube-to-tubesheet contact pressures that provide a 
restricted leakage path above the indications and also limit the 
degree of potential crack face opening as compared to free span 
indications.
    The leakage factor for Salem Unit 1, for a postulated SLB/FLB, 
has been calculated as 2.16. Specifically, for the condition 
monitoring (CM) assessment, the component of leakage from the prior 
cycle from below the H* distance will be multiplied by a factor of 
2.16 and added to the total leakage from any other source and 
compared to the allowable accident induced leakage limit. For the 
operational assessment (OA), the difference in the leakage between 
the allowable leakage and the accident induced leakage from sources 
other than the tubesheet expansion region will be divided by 2.16 
and compared to the observed operational leakage.
    The probability of an SLB/FLB is unaffected by the potential 
failure of a steam generator tube as the failure of the tube is not 
an initiator for an SLB/FLB event. SLB/FLB leakage is limited by 
leakage flow restrictions resulting from the leakage path above 
potential cracks through the tube-to-tubesheet crevice. The leak 
rate during all postulated accident conditions that model primary-
to-secondary leakage (including locked rotor and control rod 
ejection) has been shown to remain within the accident analysis 
assumptions for all axial and or circumferentially orientated cracks 
occurring 15.21 inches below the top of the tubesheet. The accident 
analysis calculations have an assumption of 0.6 gpm [gallons per 
minute] at room temperature (gpmRT) primary-to-secondary leakage in 
a single SG and 1 gpm at room temperature (gpmRT) total primary-to-
secondary leakage for all SGs. This apportioned primary-to-secondary 
leakage is used in the Main Steam Line Break and Locked Rotor 
accidents. Primary-to-secondary leakage of 1 gpm at room temperature 
(gpmRT) from all SGs, conservatively modeled to be released from a 
single location to maximize control room dose consequences, is used 
in the Control Rod Ejection (CRE) accident. The TS operational leak 
rate limit is 150 gallons per day (gpd) (0.104 gpmRT). The maximum 
accident leak rate ratio for Salem Unit 1 is 2.16 (Revised Table 9-
7, Reference 15, [of the licensee's amendment request dated May 8, 
2012]). Consequently, this results in significant margin between the 
conservatively estimated accident leakage and the allowable accident 
leakage.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change alters the steam generator inspection and 
reporting criteria. It does not introduce any new equipment, create 
new failure modes for existing equipment, or create any new limiting 
single failures. Plant operation will not be altered, and safety 
functions will continue to perform as previously assumed in accident 
analyses.
    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 change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed change alters the steam generator inspection and 
reporting criteria. It maintains the required structural margins of 
the steam generator tubes for both normal and accident conditions. 
NEI 97-06 and RG 1.121, are used as the bases in the development of 
the limited tubesheet inspection depth methodology for determining 
that steam generator tube integrity considerations are maintained 
within acceptable limits. RG 1.121 describes a method acceptable to 
the NRC for meeting GDC [General Design Criteria] 14, ``Reactor 
Coolant Pressure Boundary,'' GDC 15, ``Reactor Coolant System 
Design,'' GDC 31, ``Fracture Prevention of Reactor Coolant Pressure 
Boundary,'' and GDC 32, ``Inspection of Reactor Coolant Pressure 
Boundary,'' by reducing the probability and consequences of a SGTR. 
RG 1.121 concludes that by determining the limiting safe conditions 
for tube wall degradation, the probability and consequences of a 
SGTR are

[[Page 4475]]

reduced. This RG uses safety factors on loads for tube burst that 
are consistent with the requirements of Section III of the American 
Society of Mechanical Engineers (ASME) Code.
    For axially-oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially-oriented cracking, the H* Analysis documented in 
Section 3, [of the licensee's amendment request dated May 8, 2012,] 
defines a length of degradation-free expanded tubing that provides 
the necessary resistance to tube pullout due to the pressure induced 
forces, with applicable safety factors applied. Application of the 
limited hot and cold leg tubesheet inspection criteria will preclude 
unacceptable primary to secondary leakage during all plant 
conditions. The methodology for determining leakage provides for 
large margins between calculated and actual leakage values in the 
proposed limited tubesheet inspection depth criteria.
    Therefore, the proposed change does not involve a significant 
reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, and with the changes noted above in square brackets, 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: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Meena K. Khanna.
Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined 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.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, 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 NRC's Public Document Room (PDR), located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through the Agencywide Documents Access and 
Management System (ADAMS) in the NRC Library at https://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's Reference staff at 1-800-397-4209, 301-415-4737 or by email to 
pdr.resource@nrc.gov.
    Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Power Station, Unit 2, New London County, Connecticut
    Date of amendment request: July 31, 2012.
    Description of amendment request: The proposed amendment would 
revise the Millstone Power Station, Unit 2 Technical Specification 
requirements regarding steam generator tube inspections and reporting 
as described in TSTF-510, Revision 2, ``Revision to Steam Generator 
Program Inspection Frequencies and Tube Sample Selection;'' however, 
Dominion Nuclear Connecticut, Inc. is proposing minor variations and 
deviations from TSTF-510.
    Date of issuance: January 4, 2013.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 312.
    Renewed Facility Operating License No. DPR-65: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: September 4, 2012 (77 
FR 53926).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 4, 2013.
    No significant hazards consideration comments received: No.
    Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1 (RBS), West Feliciana 
Parish, Louisiana
    Entergy Nuclear Operations, Inc., Docket Nos. 50-155 and 72-043 
(ISFSI), Big Rock Point Plant (Big Rock), Charlevoix County, Michigan
    Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247 and 
50-286, Indian Point Nuclear Generating Units 1, 2 and 3 (IP1, IP2, and 
IP3), Westchester County, New York
    Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant (FitzPatrick), Oswego County, New York
    Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades 
Nuclear Plant (Palisades), Van Buren County, Michigan
    Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim 
Nuclear Power Station (Pilgrim), Plymouth County, Massachusetts
    Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station (VY), 
Vernon, Vermont
    Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Units 1 and 2 (ANO1 and ANO2), Pope County, Arkansas
    Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne 
County, Mississippi
    Entergy Operations, Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3 (Waterford), St. Charles Parish, Louisiana
    Date of application for amendment: December 13, 2011, as 
supplemented by letters dated May 21, and November 20, 2012.
    Brief description of amendment: The amendments approved changes to 
the Quality Assurance Program Manual (QAPM) and Technical 
Specifications (TSs) for the above specified plants. The proposed 
changes standardize unit staff qualification requirements for the 
Entergy fleet. Certain changes to the QAPM are a reduction in 
commitment and, in accordance with 10 CFR 50.54(a)(4), NRC approval is 
required prior to implementation.
    Date of issuance: December 28, 2012.
    Effective date: As of the date of issuance and shall be implemented 
120 days from the date of issuance.
    Amendment Nos.: ANO1--248; ANO2--296; FitzPatrick--304; GGNS--193; 
IP2--271; IP3--248; Palisades--

[[Page 4476]]

249; Pilgrim--239; RBS--178; VY--253; and Waterford--240.
    Facility Operating License Nos. DPR-51, NPF-6, NPF-29, NPF-47, NPF-
38, DPR-59, DPR-35, DPR-26, DPR-64, DPR-20, and DPR-28: The amendments 
revise the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: March 20, 2012 (77 FR 
16274). The supplemental letters dated May 21 and November 20, 2012, 
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 December 28, 2012.
    No significant hazards consideration comments received: No.
    Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station, Unit 1, Washington County, Nebraska
    Date of amendment request: December 23, 2011, as supplement by 
letter dated June 18, 2012.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to incorporate a new Radial Peaking Factor 
definition and to clarify Limiting Condition for Operation 2.10.2(6), 
``Shutdown CEA [Control Element Assembly] Insertion Limit During Power 
Operation.'' Specifically, the amendment removed requirements for, and 
references to, the ``Unrodded Integrated Radial Peaking Factor.'' The 
amendment also added a definition of, and references to, the ``Maximum 
Radial Peaking Factor (FR\T\).'' Additional clarifications 
and editorial changes were made to TS 2.10, ``Reactor Core.''
    Date of issuance: December 31, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 269.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 17, 2012 (74 FR 
22816). The supplemental letter dated June 18, 2012, 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 amendment is contained 
in a safety evaluation dated December 31, 2012.
    No significant hazards consideration comments received: No.
    Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo 
County, California
    Date of application for amendment: January 5, 2012.
    Brief description of amendment: The amendments revised the Diablo 
Canyon Power Plant, Units 1 and 2, Final Safety Analysis Report Update 
Section 4.3.2.2, ``Power Distribution,'' to allow the use of the 
Westinghouse Electric Company LLC's Best Estimate Analyzer for the Core 
Operations-Nuclear (BEACON) Power Distribution Monitoring System 
methodology as described in WCAP-12472-P-A, Addendum 1-A, ``BEACON Core 
Monitoring and Operation Support System,'' January 2000.
    Date of issuance: January 9, 2013.
    Effective date: As of its date of issuance and shall be implemented 
within 120 days from the date of issuance. Implementation of the 
amendments shall also include revision of the Final Safety Analysis 
Report Update as described in the licensee's letter dated January 5, 
2012.
    Amendment Nos.: Unit 1--214; Unit 2--216.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: May 15, 2012 (77 FR 
28633).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 9, 2013.
    No significant hazards consideration comments received: No.
    Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama
    Date of amendment request: January 18, 2012.
    Brief description of amendment request: The amendment revises 
Technical Specification (TS) Surveillance Requirements 3.4.11.1 and 
3.4.11.4 by removing requirements no longer applicable to Joseph M. 
Farley Nuclear Plant, Unit 2.
    Date of issuance: December 27, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 186.
    Facility Operating License No. NPF-8:
    Date of initial notice in Federal Register: October 2, 2012 (77 FR 
60152).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 27, 2012.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 11th day of January 2013.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2013-01010 Filed 1-18-13; 8:45 am]
BILLING CODE 7590-01-P