[Federal Register Volume 78, Number 161 (Tuesday, August 20, 2013)]
[Notices]
[Pages 51219-51234]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-20154]


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NUCLEAR REGULATORY COMMISSION

[NRC-2013-0191]


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 (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 July 25, 2013 to August 7, 2013. The last 
biweekly notice was published on August 6, 2013 (78 FR 47785).

ADDRESSES: You may submit comment by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0191. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected].
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: 3WFN, 06A44M, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001.
    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-2013-0191 when contacting the NRC 
about the availability of information regarding this document. You may 
access

[[Page 51220]]

publicly-available information related to this action by the following 
methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0191.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly-available documents online in the NRC 
Library at http://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 [email protected]. 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-2013-0191 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 you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering 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 Sec.  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 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 ``Agency 
Rules of Practice and Procedure'' 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 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

[[Page 51221]]

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 the NRC's 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's 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 [email protected], 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 http://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 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 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, 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 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's Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by email to 
[email protected], 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 
http://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

[[Page 51222]]

information. However, a request to intervene will require including 
information on local residence in order to demonstrate a proximity 
assertion of interest in the proceeding. 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)-(iii).
    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's Library at 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's PDR Reference staff at 1-800-397-4209, 
301-415-4737, or by email to [email protected].

Carolina Power and Light Company, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit 2, (HBRSEP) Darlington County, South 
Carolina

    Date of amendment request: June 7, 2013.
    Description of amendment request: The proposed change would delete 
the current HBRSEP Surveillance Requirements (SRs) 3.1.7.1, 3.1.7.2, 
and 3.1.7.3 of Technical Specification 3.1.7, ``Rod Position 
Indication,'' and renumber current SR 3.1.7.4 as SR 3.1.7.1. This 
change deletes a redundant SR and eliminates a minimum of eight 
reactivity manipulations per year.
    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 initiating conditions and assumptions for dose consequences 
of accidents described in the Updated Final Safety Analyses Report 
remain as previously analyzed. The proposed change does not 
introduce a new accident initiator nor does it introduce changes to 
any existing accident initiators described in the Updated Final 
Safety Analyses Report. The proposed change eliminates requirements 
to periodically demonstrate agreement of individual rod position 
with average rod position and group demand step counter position 
during control rod movement while maintaining less frequent 
requirements for control rod movement associated with verification 
of control rod freedom of movement (SR 3.1.4.2) and confirmation 
that the two rod position indication systems are within alignment 
limits (SR 3.1.4.1). Control rod movement is a potential accident 
initiator and less frequent surveillances involving less control rod 
movement will not increase the probability or consequences of an 
accident.
    The proposed change also eliminates surveillance requirements 
which are redundant to the requirements of SR 3.1.4.1 and modifies 
SR 3.1.7.4 to renumber it as SR 3.1.7.1. The elimination of 
redundant surveillance requirements does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Current SR 3.1.7.4 and proposed SR 3.1.7.1 involve the 
maintenance and configuration of instrumentation used to indicate 
rod position. The proposed change renumbers SR 3.1.7.4 as SR 3.1.7.1 
and maintains the requirement to perform a Channel Calibration on an 
18 Month-Frequency which does not change the means and manner of 
control of control rod movement and therefore does not involve a 
significant increase in the probability of consequences of an 
accident previously evaluated.
    Based on the above, 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 previously evaluated?
    Response: No.
    The proposed change will not introduce any new failure modes to 
the required protection functions. The proposed change modifies 
surveillance requirements associated with operation and function of 
instrumentation indicating rod position that is part of the control 
rod control system (demand step counter position) and individual 
analog rod position indication instrumentation. The proposed change 
does not alter the manner in which the respective rod position 
indications function or the control system controls control rod 
movement such that the modified surveillance requirements of TS 
3.1.7 cannot create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed amendment does not involve revisions to any safety 
analysis limits or safety system settings that will adversely impact 
plant safety. The proposed amendment does not alter the functional 
capabilities assumed in a safety analysis for any system, structure, 
or component important to the mitigation and control of design bases 
accident conditions within the facility. Nor does this amendment 
revise any parameters or operating restrictions that are assumptions 
of a design basis accident. In addition, the proposed amendment does 
not affect the ability of safety systems to ensure that the facility 
can be placed and maintained in a shutdown condition for extended 
periods of time.
    The Technical Specifications continue to assure that the 
applicable operating parameters and systems are maintained within 
the design requirements and safety analysis assumptions. Therefore, 
the proposed changes which eliminate surveillance requirements that 
are either redundant or inconsistent with industry standards for the 
partial movement of control rods and rod position indication system 
surveillance and add a new requirement that the rod position 
indication systems agree within a prescribed value will not result 
in a significant reduction in the margin of safety as defined in the 
Updated Final Safety Analyses Report or Technical Specifications.
    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, 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: David T. Conley, Manager--Senior Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Acting Branch Chief: Douglas A. Broaddus.

Detroit Edison, Docket No. 50-341, Fermi 2, Monroe County, Michigan

    Date of amendment request: April 17, 2013.
    Description of amendment request: The proposed amendment would 
modify the Fermi 2 technical specification (TS) related to control room 
envelope habitability in accordance with NRC-approved Technical 
Specifications Task Force (TSTF) change traveler TSTF-448, ``Control 
Room Habitability,'' Revision 3. The proposed amendment is consistent 
with the Consolidated Line

[[Page 51223]]

Item Improvement Process that adopts changes to TS Section 3.7.3, 
``Control Room Emergency Filtration (CREF) System,'' and adds TS 
Section 5.5.14, ``Control Room Envelope Habitability Program.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration, is presented below. The 
licensee incorporated, by reference, the proposed no significant 
hazards consideration published in the Federal Register on January 9, 
2007 (72 FR 2032).

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits.
    The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not 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.
    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice.
    Therefore, 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 
the margin of safety?
    Response: No.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition.
    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 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 Energy Kewaunee (DEK), Docket No. 50-305, Kewaunee Power 
Station (KPS), Kewaunee County, Wisconsin

    Date of amendment request: April 16, 2013.
    Description of amendment request: The proposed amendment would 
revise the Renewed Facility Operating License by deleting a license 
condition associated with license renewal.
    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 amendment would modify the KPS renewed facility 
operating license by deleting a license condition that pertains to 
plant operation during the period of extended operation. KPS is 
permanently ceasing operation and will permanently defuel the 
reactor vessel prior to the start of the period of extended 
operation. Therefore, the probability of occurrence of previously 
evaluated accidents is not affected, since the original license did 
not contain this license condition. The license condition being 
deleted pertains to operation beyond the term of the original 
license. Additionally, the occurrence of postulated accidents 
associated with reactor operation is no longer credible in a 
permanently defueled reactor.
    Since KPS is permanently ceasing operation, the generation of 
fission products will cease and the remaining source term will 
decay. This significantly reduces the consequences of the remaining 
applicable postulated accident. Therefore, the proposed amendment 
does not involve a significant increase in the consequences of a 
previously evaluated accident.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits.
    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 activities and programs that were the subject of this 
license condition were intended to ensure that systems, structures, 
and components (SSCs) continue to respond properly in the event of a 
previously analyzed accident during the period of extended operation 
of the renewed facility operating license. However, the reactor will 
not operate during the period of extended operation.
    The proposed amendment does not involve a physical alteration of 
the plant. No new or different types of equipment will be installed 
and there are no physical modifications to existing equipment 
associated with the proposed amendment. Similarly, the proposed 
amendment would not physically change any SSCs involved in the 
mitigation of any postulated accidents. Thus, no new initiators or 
precursors of a new or different kind of accident are created. 
Furthermore, the proposed amendment does not create the possibility 
of a new failure mode associated with any equipment or personnel 
failures.
    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.
    Because the 10 CFR part 50 license for KPS will no longer authorize 
operation

[[Page 51224]]

of the reactor or emplacement or retention of fuel into the reactor 
vessel, as specified in 10 CFR 50.82(a)(2), the occurrence of 
postulated accidents associated with reactor operation is no longer 
credible. The remaining credible accident (90 days after shutdown) is a 
fuel handling accident (FHA) in the auxiliary building. The proposed 
amendment does not affect the inputs or assumptions of any of the 
design basis analyses that impact a FHA in the auxiliary building and 
the current design limits continue to be met for the accident of 
concern.
    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 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., Counsel for Dominion Energy Kewaunee, Inc., 
120 Tredegar Street, Richmond, VA 23219.
    NRC Branch Chief: Robert D. Carlson.

Dominion Energy Kewaunee (DEK), Docket No. 50-305, Kewaunee Power 
Station (KPS), Kewaunee County, Wisconsin

    Date of amendment request: May 29, 2013.
    Description of amendment request: The proposed amendment would 
revise the operating license and revise the associated technical 
specifications (TSs) to the permanently defueled technical 
specifications (PDTSs) consistent with the permanent cessation of 
reactor operation.
    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.
    KPS has permanently ceased operation. The proposed amendment 
would modify the KPS renewed facility operating license and TS by 
deleting the portions of the license and TS that are no longer 
applicable to a permanently defueled facility, while modifying the 
remaining portions to correspond to the permanently shutdown 
condition. This change is consistent with the Standard TS and with 
the criteria set forth in 10 CFR 50.36 for the contents of TS.
    Section 14 of the KPS Updated Safety Analysis Report (USAR) 
described the design basis accident (DBA) and transient scenarios 
applicable to KPS during power operations. With the reactor in a 
permanently defueled condition, the spent fuel pool and its systems 
have been isolated and are dedicated only to spent fuel storage. In 
this condition the spectrum of credible accidents is much smaller 
than for an operational plant. As a result of the certifications 
submitted by DEK in accordance with 10 CFR 50.82(a)(1), and the 
consequent removal of authorization to operate the reactor or to 
place or retain fuel in the reactor in accordance with 10 CFR 
50.82(a)(2), most of the accident scenarios postulated in the USAR 
are no longer possible.
    The definition of safety-related structures, systems, and 
components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are 
those relied on to remain functional during and following design 
basis events to assure:
    1. The integrity of the reactor coolant boundary;
    2. The capability to shutdown the reactor and maintain it in a 
safe shutdown condition; or
    3. The capability to prevent or mitigate the consequences of 
accidents which could result in potential offsite exposures 
comparable to the applicable guideline exposures set forth in 10 CFR 
50.43(a)(1) or 100.11.
    The first two criteria (integrity of the reactor coolant 
pressure boundary and safe shutdown of the reactor) are not 
applicable to a plant in a permanently defueled condition. The third 
criterion is related to preventing or mitigating the consequences of 
accidents that could result in potential offsite exposures exceeding 
limits. However, after the termination of reactor operations at KPS 
and the permanent removal of the fuel from the reactor vessel 
(following 90 days of decay time after shutdown) and purging of the 
contents of the waste gas decay tanks and liquid waste tanks, none 
of the SSCs at KPS are required to be relied on for accident 
mitigation. Therefore, none of the SSCs at KPS meet the definition 
of a safety-related SSC stated in 10 CFR 50.2 (with the exception of 
the passive spent fuel pool structure).
    The deletion of TS definitions and rules of usage and 
application, that are currently not applicable in a defueled 
condition, has no impact on facility SSCs or the methods of 
operation of such SSCs. The deletion of design features and safety 
limits not applicable to the permanently shutdown and defueled 
status of KPS has no impact on the remaining DBA (the fuel handling 
accident in the auxiliary building). The removal of limiting 
conditions for operation (LCOs) or surveillance requirements (SRs) 
that are related only to the operation of the nuclear reactor or 
only to the prevention, diagnosis, or mitigation of reactor-related 
transients or accidents do not affect the applicable DBAs previously 
evaluated since these DBAs are no longer applicable in the defueled 
mode. The safety functions involving core reactivity control, 
reactor heat removal, reactor coolant system inventory control, and 
containment integrity are no longer applicable at KPS as a 
permanently defueled plant. The analyzed accidents involving damage 
to the reactor coolant system, main steam lines, reactor core, and 
the subsequent release of radioactive material are no longer 
possible at KPS.
    Since KPS has permanently ceased operation, the future 
generation of fission products has ceased and the remaining source 
term will decay. The radioactive decay of the irradiated fuel since 
shutdown of the reactor will have reduced the consequences of the 
fuel handling accident to levels well below those previously 
analyzed. The relevant parameter (water level) associated with the 
fuel pool provides an initial condition for the fuel handling 
accident analysis and is included in the permanently defueled TS.
    The spent fuel pool water level, spent fuel pool boron 
concentration, and spent fuel pool storage LCOs are retained to 
preserve the current requirements for safe storage of irradiated 
fuel.
    Fuel pool cooling and makeup related equipment and support 
equipment (e.g., electrical power systems) are not required to be 
continuously available since there is sufficient time to effect 
repairs, establish alternate sources of makeup flow, or establish 
alternate sources of cooling in the event of a loss of cooling and 
makeup flow to the spent fuel pool.
    The deletion and modification of provisions of the 
administrative controls do not directly affect the design of SSCs 
necessary for safe storage of irradiated fuel or the methods used 
for handling and storage of such fuel in the fuel pool. The changes 
to the administrative controls are administrative in nature and do 
not affect any accidents applicable to the safe management of 
irradiated fuel or the permanently shutdown and defueled condition 
of the reactor.
    The probability of occurrence of previously evaluated accidents 
is not increased, since extended operation in a defueled condition 
is the only operation currently allowed, and therefore bounded by 
the existing analyses. Additionally, the occurrence of postulated 
accidents associated with reactor operation is no longer credible in 
a permanently defueled reactor. This significantly reduces the scope 
of applicable accidents.
    Therefore, the proposed amendment 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 changes have no impact on facility SSCs affecting 
the safe storage of irradiated fuel, or on the methods of operation 
of such SSCs, or on the handling and storage of irradiated fuel 
itself. These changes are consistent with the standard TS. The 
removal of TS that are related only to the operation of the nuclear 
reactor or only to the prevention, diagnosis, or mitigation of 
reactor-related transients or accidents cannot result in different 
or more adverse failure modes or accidents than previously evaluated 
because the reactor is permanently shutdown and defueled and KPS is 
no longer authorized to operate the reactor.

[[Page 51225]]

    The proposed deletion of requirements of the KPS TS do not 
affect systems credited in the accident analysis for the fuel 
handling accident in the auxiliary building at KPS. The proposed 
permanently defueled TS (PDTS) continue to require proper control 
and monitoring of safety significant parameters and activities.
    The proposed restriction on the fuel pool level is fulfilled by 
normal operating conditions and preserves initial conditions assumed 
in the analyses of the postulated DBA. The spent fuel pool water 
level, spent fuel pool boron concentration, and spent fuel pool 
storage LCOs are retained to preserve the current requirements for 
safe storage of irradiated fuel.
    The proposed amendment does not result in any new mechanisms 
that could initiate damage to the remaining relevant safety barriers 
for defueled plants (i.e., fuel cladding and spent fuel cooling). 
Since extended operation in a defueled condition is the only 
operation currently allowed, and therefore bounded by the existing 
analyses, such a condition does not create the possibility of a new 
or different kind of accident.
    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
    Because the 10 CFR Part 50 license for KPS no longer authorizes 
operation of the reactor or emplacement or retention of fuel into 
the reactor vessel, as specified in 10 CFR 50.82(a)(2), the 
occurrence of postulated accidents associated with reactor operation 
is no longer credible. The only remaining credible accident is a 
fuel handling accident (FHA). The proposed amendment does not 
adversely affect the inputs or assumptions of any of the design 
basis analyses that impact a FHA.
    The proposed changes are limited to those portions of TS and 
license that are not related to the safe storage of irradiated fuel. 
The requirements for SSCs that have been deleted from the KPS TS are 
not credited in the existing accident analysis for the remaining 
applicable postulated accident; and as such, do not contribute to 
the margin of safety associated with the accident analysis. 
Postulated DBAs involving the reactor are no longer possible because 
the reactor is permanently shutdown and defueled and KPS is no 
longer authorized to operate the reactor.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety because the current design limits 
continue to be met for the accident of concern.

    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., Counsel for Dominion Energy Kewaunee, Inc., 
120 Tredegar Street, Richmond, VA 23219.
    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: May 3, 2013.
    Description of amendment request: The proposed amendment would 
revise the Millstone Power Station, Unit 2 (MPS2) Technical 
Specification (TS) 3/4.7.11, ``Ultimate Heat Sink'', to increase the 
current ultimate heat sink water temperature limit from 
75[emsp14][deg]F to 80[emsp14][deg]F and change the TS Action to state, 
``With the ultimate heat sink water temperature greater than 
80[emsp14][deg]F, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN 
within the following 30 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, which is presented below:

Criterion 1

    Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    Previously evaluated accident consequences are not impacted 
because credited mitigating equipment continues to perform its 
design function. The proposed change does not significantly impact 
the probability of an accident previously evaluated because those 
SSCs that can initiate an accident are not significantly impacted.
    Based on the above, DNC concludes that the proposed increased 
temperature limits do not involve a significant increase in the 
probability or consequences of an accident or transient previously 
evaluated in the safety analysis report.

Criterion 2

    Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    A new or different accident from any accident previously 
evaluated is not created because previously credited SSCs, are not 
impacted, there is no new reliance upon equipment not previously 
credited, there is no new equipment installed (except for monitoring 
equipment), there is no impact upon the existing failure modes and 
effects analysis, and conformance to the single failure criterion is 
maintained. The increased limits do not introduce any new mode of 
plant operation and will not result in a change to the design 
function or the operation of any SSC that is used for mitigating 
accidents.
    Based on the above, DNC concludes that the proposed changes do 
not create the possibility of a new or different kind of accident or 
transient from any previously evaluated.

Criterion 3

    Do the proposed changes involve a significant reduction in the 
margin of safety?
    Response: No.
    This change does not involve a significant reduction in margin 
of safety because the containment analysis acceptance criteria 
continue to be met when operating with the proposed increased UHS 
temperature limit. Containment integrity will not be challenged and 
will continue to meet its design basis acceptance criteria following 
a large break LOCA or MSLB. The proposed change has no impact upon 
fuel cladding or RCS fission product barrier margin because credited 
SSCs continue to perform their design functions with an 
80[emsp14][deg]F UHS temperature.
    Based on the above, DNC concludes that the proposed changes do 
not involve a significant reduction in the 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 Acting Branch Chief: Robert H. Beall.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power 
Station, Unit 3, New London County, Connecticut

    Date of amendment request: May 3, 2013.
    Description of amendment request: The proposed amendment would 
revise the Millstone Power Station, Unit 3 (MPS3) Technical 
Specification (TS) 3/4.7.5, ``Ultimate Heat Sink'', to increase the 
current ultimate heat sink water temperature limit from 
75[emsp14][deg]F to 80[emsp14][deg]F and change the TS Action to state, 
``With the ultimate heat sink water temperature greater than 
80[emsp14][deg]F, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN 
within the following 30 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, which is presented below:


[[Page 51226]]


    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    Previously evaluated accident consequences are not impacted 
because credited mitigating equipment continues to perform its 
design function. The proposed change does not significantly impact 
the probability of an accident previously evaluated because those 
SSCs that can initiate an accident are not significantly impacted.
    Based on the above, DNC concludes that the proposed increased 
temperature limits do not involve a significant increase in the 
probability or consequences of an accident or transient previously 
evaluated in the safety analysis report.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    A new or different accident from any accident previously 
evaluated is not created because previously credited SSCs, are not 
impacted; there is no new reliance upon equipment not previously 
credited; there is no new equipment installed (except for monitoring 
equipment); there is no impact upon the existing failure modes and 
effects analysis; and conformance to the single failure criterion is 
maintained.
    The increased limits do not introduce any new mode of plant 
operation and will not result in a change to the design function or 
the operation of any SSC that is used for mitigating accidents.
    Based on the above, DNC concludes that the proposed changes do 
not create the possibility of a new or different kind of accident or 
transient from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    This change doesn't involve a significant reduction in margin of 
safety because containment structure fission product barrier design 
margin is unaffected because peak pressure/temperature occurs early 
in the accident before UHS temperature can influence the containment 
response. The proposed change has no impact upon fuel cladding or 
RCS fission product barrier margin because credited SSCs continue to 
perform their design functions with an 80[emsp14][deg]F UHS 
temperature.
    Based on the above, DNC concludes that the proposed changes do 
not involve a significant reduction in the 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 Acting Branch Chief: Robert H. Beall.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana

    Date of amendment request: June 13, 2013.
    Description of amendment request: The amendment will adopt 
Technical Specification Task Force (TSTF)-423, Revision 1, ``Technical 
Specifications End States.'' Specifically, the proposed amendment would 
modify Technical Specifications (TSs) to risk-informed requirements 
regarding selected Required Action end states. The proposed changes are 
consistent with NRC-approved TSTF-423, Revision 1, with some deviations 
noted.
    The NRC issued a ``Notice of Availability of the Proposed Models 
for Plant-Specific Adoption of Technical Specifications Task Force 
(TSTF) Traveler TSTF-423, Revision 1, `Technical Specifications End 
States, NEDC-32988-A,' for Boiling Water Reactor Plants Using the 
Consolidated Line Item Improvement Process,'' published in the Federal 
Register on February 18, 2011 (76 FR 9614), which included the model no 
significant hazards consideration and safety evaluation for TSTF-423, 
Revision 1.
    Basis for proposed no significant hazards consideration 
determination: An analysis of the no significant hazards consideration 
was presented in the TSTF-423. The licensee has affirmed the 
applicability of the model no significant hazards consideration 
determination, 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.
    Required Actions are not an initiator of any accident previously 
evaluated. Therefore, the proposed changes do not affect the 
probability of any accident previously evaluated. NEDC-32988-A 
demonstrated that the proposed changes in the required end state do 
not significantly increase the consequences of any accidents 
previously evaluated.
    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.
    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.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    NEDC-32988-A demonstrated that the changed end states represent 
a condition of equal or lower risk than the original end states.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    Based on the above, the TSTF-423 concludes that the proposed change 
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.
    Attorney for licensee: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana

    Date of amendment request: July 16, 2013.
    Description of amendment request: The amendment would adopt 
Technical Specifications Task Force (TSTF) change traveler TSTF-535, 
Revision 0, ``Revise Shutdown Margin Definition to Address Advanced 
Fuel Designs.'' The Shutdown Margin (SDM) (i.e., the amount of 
reactivity by which the reactor is subcritical) is calculated under the 
conservative conditions that the reactor is Xenon free, the most 
reactive control rod is outside the reactor core, and the moderator 
temperature produces the maximum reactivity. For standard fuel designs, 
maximum reactivity occurs at a moderator temperature of 68 degrees 
Fahrenheit ([deg]F), which is reflected in the temperature specified in 
the Technical Specifications (TSs). New, advanced Boiling Water Reactor 
(BWR) fuel designs can have a higher reactivity at moderator shutdown 
temperatures above 68 [deg]F. Therefore, the proposed amendment, 
consistent with TSTF-535, Revision 0, seeks to modify the TSs to 
require the SDM to be calculated at whatever temperature produces the 
maximum reactivity (i.e., temperatures at or above 68 [deg]F).

[[Page 51227]]

    The notice of availability of this TS improvement ``Models for 
Plant-Specific Adoption of Technical Specifications Task Force Traveler 
TSTF-535, Revision 0, `Revise Shutdown Margin Definition to Address 
Advanced Fuel Designs,' Using the Consolidated Line Item Improvement 
Process,'' was published in Federal Register on February 26, 2013 (78 
FR 13100), which included a model no significant hazards consideration 
(NSHC) determination and safety evaluation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
affirmed the applicability of the model no significant hazards 
consideration determination included in TSTF-535, Revision 0, and 
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 revises the definition of SDM. SDM is not an 
initiator to any accident previously evaluated. Accordingly, the 
proposed change to the definition of SDM has no effect on the 
probability of any accident previously evaluated. SDM is an 
assumption in the analysis of some previously evaluated accidents 
and inadequate SDM could lead to an increase in consequences for 
those accidents. However, the proposed change revises the SDM 
definition to ensure that the correct SDM is determined for all fuel 
types at all times during the fuel cycle. As a result, the proposed 
change does not adversely affect the consequences of any accident 
previously evaluated.
    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.
    The proposed change revises the definition of SDM. The change 
does 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 operations. The change does not alter 
assumptions made in the safety analysis regarding SDM.
    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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the definition of SDM. The proposed 
change does not alter the manner in which safety limits, limiting 
safety system settings or limiting conditions for operation are 
determined. The proposed change ensures that the SDM assumed in 
determining safety limits, limiting safety system settings or 
limiting conditions for operation is correct for all BWR fuel types 
at all times during the fuel cycle.
    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: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

Florida Power and Light Company, Docket Nos. 50-250, and 50-251, Turkey 
Point Nuclear Generating Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: March 22, 2013.
    Description of amendment request: The license amendment request 
proposes to revise the Technical Specifications (TS) to allow the use 
of Optimized ZIRLO\TM\ fuel rod cladding material. The proposed change 
would revise TS 5.3.1 to add Optimized ZIRLO\TM\ to the approved fuel 
rod cladding materials and TS 6.9.1.7 to add Westinghouse Electric 
Company LLC topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-
A, ``Optimized ZIRLO\TM\,'' to the analytical methods used to determine 
the core operating limits.
    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 would allow the use of Optimized ZIRLO\TM\ 
clad nuclear fuel in the reactors. The NRC approved topical report 
WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A ``Optimized 
ZIRLO\TM\,'' prepared by Westinghouse Electric Company LLC 
(Westinghouse), addresses Optimized ZIRLO\TM\ and demonstrates that 
Optimized ZIRLO\TM\ has essentially the same properties as currently 
licensed ZIRLO.[supreg] The fuel cladding itself is not an accident 
initiator and does not affect accident probability. Use of Optimized 
ZIRLO\TM\ fuel cladding will continue to meet all 10 CFR 50.46 
acceptance criteria and, therefore, will not increase the 
consequences of an accident.
    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.
    Use of Optimized ZIRLO\TM\ clad fuel will not result in changes 
in the operation or configuration of the facility. Topical Report 
WCAP-12610-PA and CENPD-404-P-A demonstrated that the material 
properties of Optimized ZIRLO\TM\ are similar to those of standard 
ZIRLO.[supreg] Therefore, Optimized ZIRLO\TM\ fuel rod cladding will 
perform similarly to those fabricated from standard ZIRLO,[supreg] 
thus precluding the possibility of the fuel becoming an accident 
initiator and causing a new or different type of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed change will not involve a significant reduction in 
the margin of safety because it has been demonstrated that the 
material properties of the Optimized ZIRLO\TM\ are not significantly 
different from those of standard ZIRLO.[supreg] Optimized ZIRLO\TM\ 
is expected to perform similarly to standard ZIRLO[supreg] for all 
normal operating and accident scenarios, including both loss of 
coolant accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, 
where the slight difference in Optimized ZIRLO\TM\ material 
properties relative to standard ZIRLO\TM\ could have some impact on 
the overall accident scenario, plant-specific LOCA analyses using 
Optimized ZIRLO properties demonstrates that the acceptance criteria 
of 10 CFR 50.46 has been satisfied.
    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: James Petro, Managing Attorney--Nuclear, 
Florida Power & Light Company, P.O. Box 14000, Juno Beach, Florida 
33408-0420.
    NRC Acting Branch Chief: Douglas A. Broaddus.

NextEra Energy Seabrook, LLC., Docket No. 50-443, Seabrook Station, 
Unit 1, Rockingham County, New Hampshire

    Date of amendment request: May 28, 2013.

[[Page 51228]]

    Description of amendment request: The proposed amendment will 
modify the Seabrook Technical Specifications (TSs). Specifically, the 
proposed amendment will modify the TS by relocating specific 
surveillance frequencies to a licensee-controlled program with 
implementation of Nuclear Energy Institute 04-10, ``Risk-Informed 
Technical Specification Initiative 5B, Risk-Informed Method for Control 
of Surveillance Frequencies.'' The changes are consistent with NRC-
approved Technical Specifications Task Force (TSTF) Standard Technical 
Specifications (STS) change TSTF-425, ``Relocate Surveillance 
Frequencies to Licensee Control--Risk Informed Technical Specifications 
Task Force (RITSTF) Initiative 5b,'' Revision 3, (ADAMS Accession No. 
ML090850642). The Federal Register notice published on July 6, 2009 (74 
FR 31996), announced the availability of this TSTF improvement, and 
included a model no significant hazards consideration and safety 
evaluation.
    Basis for proposed no significant hazards consideration 
determination: An analysis of the no significant hazards consideration 
was presented in the TSTF-425. The licensee has affirmed the 
applicability of the model 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 specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to an 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 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, NextEra 
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.

    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. James Petro, Managing Attorney, Florida 
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Acting Branch Chief: Veronica Rodriguez.

NextEra Energy Seabrook, LLC., Docket No. 50-443, Seabrook Station, 
Unit 1, Rockingham County, New Hampshire

    Date of amendment request: June 25, 2013.
    Description of amendment request: The proposed amendment will 
revise the Seabrook Technical Specifications. Specifically, the 
proposed amendment will allow the use of Optimized ZIRLO\TM\ as fuel 
rod cladding.
    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, along with the NRC's 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 proposed change would allow the use of Optimized ZIRLO\TM\ 
clad nuclear fuel in the reactors. The NRC approved topical report 
WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A ``Optimized 
ZIRLO,\TM\'' prepared by Westinghouse Electric Company LLC 
(Westinghouse), addresses Optimized ZIRLO\TM\ and demonstrates that 
Optimized ZIRLO\TM\ has essentially the same properties as currently 
licensed ZIRLO.[supreg] The fuel cladding itself is not an accident 
initiator and does not affect accident probability. Use of Optimized 
ZIRLO\TM\ fuel cladding will continue to meet all [Title 10 of the 
Code of Federal Regulations] 10 CFR 50.46 acceptance criteria and, 
therefore, will not increase the consequences of an accident.
    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.
    Use of Optimized ZIRLO\TM\ clad fuel will not result in changes 
in the operation or configuration of the facility. Topical Report 
WCAP-12610-P-A and CENPD-404-P-A demonstrated that the material 
properties of Optimized ZIRLO\TM\ are similar to those of standard 
ZIRLO.[supreg] Therefore, Optimized ZIRLO\TM\ fuel rod cladding will 
perform similarly to those fabricated from standard ZIRLO,[supreg] 
thus precluding the possibility of the fuel cladding becoming an 
accident initiator and causing a new or different type of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed changes involve a significant reduction in 
a margin of safety?
    Response: No.
    The proposed change will not involve a significant reduction in 
the margin of safety because it has been demonstrated that the 
material properties of the Optimized ZIRLO\TM\ are not significantly 
different from those of standard ZIRLO.[supreg] Optimized ZIRLO\TM\ 
is expected to perform similarly to standard ZIRLO[supreg] for all 
normal operating and accident scenarios, including both loss of 
coolant accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, 
where the slight difference in Optimized ZIRLO\TM\ material 
properties relative to standard [ZIRLO[supreg]], ZIRLO\TM\ could 
have some impact on the overall accident scenario, plant-specific 
LOCA analyses using Optimized ZIRLO\TM\ properties will demonstrate 
that the acceptance criteria of 10 CFR 50.46 have been satisfied.

[[Page 51229]]

    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. James Petro, Managing Attorney, Florida 
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Acting Branch Chief: Veronica Rodriguez.

Northern States Power Company--Minnesota, Docket Nos. 50-282, and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota

    Date of amendment request: February 2, 2013, as supplemented by 
letter dated June 25, 2013.
    Description of amendment request: The proposed amendments would 
remove Technical Specification (TS) 3.5.3 ``[Emergency Core Cooling 
Systems (ECCS)]--Shutdown'' Limiting Condition for Operation (LCO) Note 
1 to eliminate information to the plant operators that could cause non-
conservative operation.
    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.
    This license amendment request proposes to revise the Technical 
Specification for ECCS operability requirements in Mode 4 by 
removing the LCO Note which allows the RHR subsystem to be 
considered operable for ECCS when aligned for shutdown cooling. 
These changes will require one train of RHR to be aligned for ECCS 
operation throughout the mode and other specified conditions of 
applicability.
    The proposed changes do not affect the ECCS and RHR subsystem 
design, the interfaces between the RHR subsystem and other plant 
systems' operating functions, or the reliability of the RHR 
subsystem. The proposed changes do not change or impact the 
initiators and assumptions of the analyzed accidents. Therefore, the 
ECCS and RHR subsystems will be capable of performing their accident 
mitigation functions, and the proposed removal of the LCO Note does 
not involve an increase in the probability of an accident.
    The proposed removal of the LCO Note will require that one train 
of RHR is aligned for ECCS operation during the mode and other 
specified conditions of applicability which assures that one train 
of ECCS is operable to mitigate the consequences of a loss of 
coolant accident. Thus the proposed removal of the LCO Note does not 
involve a significant increase in the consequences of an accident.
    Therefore, the proposed Technical Specification 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.
    This license amendment request proposes to revise the Technical 
Specification for ECCS operability requirements in Mode 4 by 
removing the LCO Note which allows the RHR subsystem to be 
considered operable for ECCS when aligned for shutdown cooling. 
These changes will require one train of RHR to be aligned for ECCS 
operation throughout the mode and other specified conditions of 
applicability.
    The proposed Technical Specification changes to remove the LCO 
Note involve changes to when system trains are operated, but they do 
not change any system functions or maintenance activities. The 
changes do not involve physical alteration of the plant, that is, no 
new or different type of equipment will be installed. The changes do 
not alter assumptions made in the safety analyses but ensure that 
one train of ECCS is operable to mitigate the consequences of a loss 
of coolant accident. These changes do not create new failure modes 
or mechanisms which are not identifiable during testing and no new 
accident precursors are generated.
    Therefore, the proposed Technical Specification changes do 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.
    This license amendment request proposes to revise the Technical 
Specification for ECCS operability requirements in Mode 4 by 
removing the LCO Note which allows the RHR subsystem to be 
considered operable for ECCS when aligned for shutdown cooling. 
These changes will require one train of RHR to be aligned for ECCS 
operation throughout the mode and other specified conditions of 
applicability.
    This license amendment proposes Technical Specification changes 
which assure that the ECCS--Shutdown TS LCO requirements are met if 
a Mode 4 LOCA were to occur. With these changes, other TS 
requirements for shutdown cooling in Mode 4 will continue to be met. 
Based on review of plant operating experience, there is no 
[discernible] change in cooldown rates when utilizing a single train 
of RHR for shutdown cooling. Thus, no margin of safety is reduced as 
part of this change.
    Therefore, the proposed Technical Specification 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 requests involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401
    NRC Branch Chief: Robert D. Carlson.

Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota

    Date of amendment request: May 23, 2013.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs) for Prairie Island Nuclear 
Generating Plant, Units 1 and 2, to add a methodology to TS 5.6.5 
``Core Operating Limits Report (COLR).''
    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.
    This license amendment request proposes to revise the Technical 
Specifications to reference and allow use of WCAP-16045-P-A, 
``Qualification of the Two-Dimensional Transport Code PARAGON'', and 
WCAP-16045-P-A, Addendum 1-A, ``Qualification of the NEXUS Nuclear 
Data Methodology'', for determining core operating limits.
    The methodologies which this license amendment proposes for 
determination of core operating limits are improvements over the 
current methodologies in use at the Prairie Island Nuclear 
Generating Plant.
    The NRC staff reviewed and approved these methodologies and 
concluded that these analysis codes are acceptable as a replacement 
for the current analysis code. Thus core operating limits determined 
using the proposed codes continue to assure that the reactor 
operates safely and, thus, the proposed changes do not involve an 
increase in the probability of an accident.
    Operation of the reactor with core operating limits determined 
by use of the proposed analysis codes does not increase the reactor 
power level, does not increase the core fission product inventory, 
and does not change any transport assumptions. Therefore the 
proposed methodology and Technical Specification changes do not 
involve a significant increase in the consequences of an accident.

[[Page 51230]]

    Therefore, the proposed methodology change and associated 
Technical Specification 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.
    This license amendment request proposes to revise the Technical 
Specifications to reference and allow use of WCAP-16045-P-A, 
``Qualification of the Two-Dimensional Transport Code PARAGON'', and 
WCAP-16045-P-A, Addendum 1-A, ``Qualification of the NEXUS Nuclear 
Data Methodology,'' for determining core operating limits.
    The proposed changes provide revised methodology for determining 
core operating limits, but they do not change any system functions 
or maintenance activities. The changes do not involve physical 
alteration of the plant, that is, no new or different type of 
equipment will be installed. The changes do not alter assumptions 
made in the safety analyses but ensure that the core will operate 
within safe limits. These changes do not create new failure modes or 
mechanisms which are not identifiable during testing, and no new 
accident precursors are generated.
    Therefore, the proposed methodology change and associated 
Technical Specification changes do 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.
    This license amendment request proposes to revise the Technical 
Specifications to reference and allow use of WCAP-16045-P-A, 
``Qualification of the Two-Dimensional Transport Code PARAGON'', and 
WCAP-16045-P-A, Addendum 1-A, ``Qualification of the NEXUS Nuclear 
Data Methodology,'' for determining core operating limits.
    This license amendment proposes revised methodology for 
determining core operating limits. The proposed methodology is an 
improvement that allows more accurate modeling of core performance. 
The NRC has reviewed and approved this methodology for use in lieu 
of the current methodology, thus, the margin of safety is not 
reduced due to this change.
    Therefore, the proposed methodology change and associated 
Technical Specification 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 requests involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401
    NRC Branch Chief: Robert D. Carlson.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: April 12, 2013.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 5.9.2. ``Annual Radiological 
Environmental Operating Report,'' to delete the reference to collocated 
dosimeters in relation to the NRC thermo luminescent dosimeters 
program. This change is consistent with NRC-approved Technical 
Specification Task Force (TSTF) change TSTF-348. In addition, it would 
correct a cross-reference error in TS 5.9.8, ``Post Accident Monitoring 
System (PAMS) Report.''
    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 consequence of an accident previously evaluated?
    Response: No.
    The proposed changes do not require physical changes to plant 
systems, structures, or components. The proposed changes are 
administrative in nature and therefore, do not change the 
fundamental requirements of the Technical Specifications. Removal of 
the discussion of the NRC environmental monitoring program with the 
State reflects the cancellation of that program with the State. It 
does not alter any other environmental monitoring requirements. 
Therefore, the changes do not affect accident or transient 
initiation or consequences. As described above, the proposed changes 
are administrative in nature and do not impact the operation of any 
equipment needed for the mitigation of an accident or any known 
accident initiators.
    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 changes are administrative in nature and therefore, 
do not change the fundamental requirements of the Technical 
Specifications. The proposed changes would not require any new or 
different accidents to be postulated, since no changes are being 
made to the plant that would introduce any new accident causal 
mechanisms. This license amendment request does not impact any plant 
systems that are potential accident initiators; nor does it have any 
significantly adverse impact on any accident mitigating systems.
    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.
    Since the proposed changes are administrative in nature, they do 
not change the fundamental requirements of the Technical 
Specifications. 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. Removal of the 
discussion of the NRC environmental monitoring program with the 
State reflects the cancellation of that program with the State. It 
does not alter any other environmental monitoring requirements.
    Therefore, the proposed amendment does not involve a significant 
reduction in the 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, ET 11A, Knoxville, Tennessee 37902.
    NRC Acting Branch Chief: Douglas A. Broaddus.

Previously Published Notices 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 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.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: February 12, 2013.
    Brief description of amendment request: The proposed amendment 
would modify Cooper Nuclear Station

[[Page 51231]]

license condition 2.E to require incorporation of the commitments 
listed in Appendix A of NUREG-1944, ``Safety Evaluation Report Related 
to the License Renewal of Cooper Nuclear Station,'' in the updated 
safety analysis report (USAR) to be managed in accordance with 10 CFR 
50.59.
    Date of publication of individual notice in Federal Register: July 
5, 2013 (78 FR 40519).
    Expiration date of individual notice: August 5, 2013 (public 
comments); September 3, 2013 (hearing requests).

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 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's Reference staff at 1-800-397-4209, 301-415-4737 or by email to 
[email protected].

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

    Date of application for amendments: July 2, 2012, as supplemented 
by letters dated March 6 and May 28, 2013.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 5.5.16 ``Containment Leakage Rate Testing Program'' 
by increasing the peak calculated containment internal pressure 
(Pa) from 49.4 pounds per square inch gauge (psig) to 49.7 
psig for the design basis loss-of-coolant accident. In support of the 
revised Pa, the amendments also revise TS 3.6.4 
``Containment Pressure'' by decreasing the upper bound internal 
containment pressure limit from 1.8 psig to 1.0 psig.
    Date of issuance: July 31, 2013.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 303 and 281.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the License and TSs.
    Date of initial notice in Federal Register: September 4, 2012 (77 
FR 53926). The supplements dated March 6 and May 28, 2013, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the Nuclear Regulatory Commission staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated July 31, 2013.
    No significant hazards consideration comments received: No.

Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling Water 
Reactor, Vernon County, Wisconsin

    Date of application for amendment: December 10, 2012, and 
supplemented February 25, 2013.
    Brief description of amendment: The amendment revises the La Crosse 
Boiling Water Reactor License and Technical Specifications, as a result 
of the completion of the transfer of the spent fuel to dry cask 
storage.
    Date of issuance: July 31, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 72.
    Facility Operating License No. DPR-7: This amendment revises the 
License and Technical Specifications.
    Date of initial notice in Federal Register: March 19, 2013 (78 FR 
16879).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 31, 2013.
    No significant hazards consideration comments received: No.

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

    Date of application of amendments: October 5, 2012.
    Brief description of amendments: The amendments revised the 
Technical Specifications related to the integrated leak rate test of 
the reactor containment buildings.
    Date of Issuance: August 5, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 383 and 382.
    Renewed Facility Operating License Nos. DPR-47 and DPR-55: 
Amendments revised the license and the technical specifications.
    Date of initial notice in Federal Register: December 11, 2012, 77 
FR 73688.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 5, 2013.
    No significant hazards consideration comments received: No.

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, Claiborne 
County, Mississippi

    Date of application for amendment: November 9, 2012, as 
supplemented by letter dated on January 30, 2013.
    Description of amendment: The amendment revised the Technical 
Specifications (TSs) to support the correction of a non-conservative TS 
allowable value in TS Table 3.3.6.1-1, ``Allowable Value for Primary 
Containment and Drywell Isolation Instrumentation,'' Function 3.c, 
``Reactor Core Isolation Cooling (RCIC) Steam Supply Line Pressure--
Low.'' This TS allowable value is changed

[[Page 51232]]

from greater than or equal to 53 pounds per square inch (psig) to 
greater than or equal to 57 psig.
    Date of issuance: August 5, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No: 194.
    Facility Operating License No. NPF-29: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: February 5, 2013 (78 FR 
8200). The supplemental letter dated January 30, 2013, 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 August 5, 2013.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 21, 2012.
    Brief description of amendment: The amendment adopts NRC-approved 
Technical Specification Task Force (TSTF)--522, ``Revise Ventilation 
System Surveillance Requirements to Operate For 10 Hours Per Month.'' 
The amendment revises the Surveillance Requirement (SR) which currently 
requires operating the Standby Gas Treatment (SGT) System, with the 
electrical heaters operating, for a continuous 10 hour period at a 
frequency specified in the Surveillance Frequency Control Program. This 
Surveillance Requirement (SR 3.6.4.3.1) is revised to require operation 
of the system for 15 continuous minutes without the heaters operating.
    In addition, the requirements for testing the SGT System specified 
in the Ventilation Filter Testing Program (VFTP) in Section 5.5.7, are 
revised accordingly to remove the electric heater output test 
(Specification 5.5.7.e) and to increase the specified relative humidity 
(RH) for the charcoal testing from the current 70% to 95% RH in 
Specification 5.5.7.c.
    Date of issuance: July 25, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 285.
    Renewed Facility Operating License No. DPR-49: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 16, 2013 (78 FR 
22571).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 25, 2013.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined Licenses and Final Determination of No Significant Hazards 
Consideration and Opportunity for a Hearing (Exigent Public 
Announcement or Emergency Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual notice of 
consideration of issuance of amendment, proposed no significant hazards 
consideration determination, and opportunity for a hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 51.12(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 
application for amendment, (2) the amendment to Facility Operating 
License or Combined License, as applicable, 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's Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access

[[Page 51233]]

to ADAMS or if there are problems in accessing the documents located in 
ADAMS, contact the NRC's PDR Reference staff at 1-800-397-4209, 301-
415-4737 or by email to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. 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 ``Agency Rules of Practice and Procedure'' 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, and electronically on the Internet at the NRC's Web 
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR's Reference staff 
at 1-800-397-4209, 301-415-4737, or by email to [email protected]. 
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 petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the 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 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. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    All documents filed in the NRC's 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 [email protected], 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 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 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, 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 the NRC guidance 
available on the NRC's 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

[[Page 51234]]

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's Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by email to 
[email protected], 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 
http://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. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. 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.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit 1, Washington County, Nebraska

    Date of amendment request: July 21, 2013 (Agencywide Documents 
Access and Management System (ADAMS) Accession No.ML13203A136), as 
supplemented by letter dated July 24, 2013 ADAMS Accession No. 
ML13206A042).
    Description of amendment request: The amendment revised the Updated 
Safety Analysis Report (USAR) for the design basis tornado and tornado 
missiles to include Regulatory Guide 1.76, Revision 1, ``Design-Basis 
Tornado and Tornado Missiles for Nuclear Power Plants,'' and Bechtel 
Power Corporation, Topical Report BC-TOP-9A, Revision 2, September 
1974, ``Design of Structures for Missile Impact.'' The changes revise 
the current licensing basis pertaining to protection from tornadoes and 
tornado-generated missiles. RG 1.76, Revision 1 provides guidance for 
licensees to use in selecting the DBT and DBT-generated missiles that a 
nuclear power plant should be designed to withstand to prevent undue 
risk to public health and safety. BC-TOP-9A, Revision 2 provides a 
methodology for evaluating the impact of tornado missiles. The changes 
provide a means to analyze and document that the plant will be able to 
withstand, without loss of the capability to protect the public, the 
additional forces that might be imposed by a tornado.
    Date of issuance: July 26, 2013.
    Effective date: As of its issuance date and shall be implemented 
upon approval.
    Amendment No.: 272.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the facility operating license.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (Omaha-World Herald, located in Omaha, Nebraska, on 
July 24 and 25, 2013). The notice provided an opportunity to submit 
comments on the Commission's proposed NSHC determination. One comment 
was received and evaluated.
    The supplemental letter dated July 24, 2013, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination as published in the Omaha-World Herald on July 24 and 25, 
2013.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
(including the comment received on the NSHC) are contained in a safety 
evaluation dated July 26, 2013 (ADAMS Accession No. ML13203A070).
    Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 
K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.

    Dated at Rockville, Maryland, this 12th day of August 2013.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2013-20154 Filed 8-19-13; 8:45 am]
BILLING CODE 7590-01-P