Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 70589-70596 [2013-28225]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 78, Number 228 (Tuesday, November 26, 2013)]
[Notices]
[Pages 70589-70596]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-28225]


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NUCLEAR REGULATORY COMMISSION

[NRC-2013-0257]


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 to 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 October 31, 2013 to November 13, 2013. The 
last biweekly notice was published on November 12, 2013 (78 FR 67402).

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 https://www.regulations.gov and search for Docket ID NRC-2013-0257. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individual(s) listed in the FOR FURTHER INFORMATION CONTACT section 
of this document.
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: 3WFN, 06-44M, 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-2013-0257 when contacting the NRC 
about the availability of information regarding this document. You may 
access publicly-available information related to this action by the 
following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2013-0257.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly-available documents online in the NRC 
Library at https://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to pdr.resource@nrc.gov. The ADAMS accession number 
for each document referenced in this notice (if that document is 
available in ADAMS) is provided the first time that a document is 
referenced.
     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-2013-0257 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 https://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 section 50.92 of Title 10 of the Code 
of Federal Regulations (10 CFR), this means that operation of the 
facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances 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

[[Page 70590]]

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 https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed by the above date, the Commission or a presiding officer 
designated by the Commission or by the Chief Administrative Judge of 
the Atomic Safety and Licensing Board Panel, will rule on the request 
and/or petition; and the Secretary or the Chief Administrative Judge of 
the Atomic Safety and Licensing Board will issue a notice of a hearing 
or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC'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 hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then

[[Page 70591]]

submit a request for hearing or petition for leave to intervene. 
Submissions should be in Portable Document Format (PDF) in accordance 
with the NRC guidance available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered 
complete at the time the documents are submitted through the NRC's E-
Filing system. To be timely, an electronic filing must be submitted to 
the E-Filing system no later than 11:59 p.m. Eastern Time on the due 
date. Upon receipt of a transmission, the E-Filing system time-stamps 
the document and sends the submitter an email notice confirming receipt 
of the document. The E-Filing system also distributes an email notice 
that provides access to the document to the NRC's Office of the General 
Counsel and any others who have advised the Office of the Secretary 
that they wish to participate in the proceeding, so that the filer need 
not serve the documents on those participants separately. Therefore, 
applicants and other participants (or their counsel or representative) 
must apply for and receive a digital ID certificate before a hearing 
request/petition to intervene is filed so that they can obtain access 
to the document via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's Web site 
at  https://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
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 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 Library at https://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to 
ADAMS or who encounter problems in accessing the documents located in 
ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209, 
301-415-4737, or by email to pdr.resource@nrc.gov.

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

    Date of amendment request: October 30, 2012, as supplemented on 
January 21, June 11, September 3, and October 21, 2013.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs) to allow operation of a 
reverse osmosis system during normal plant operation to purify the 
water in the borated water storage tanks and the spent fuel pools. 
Automatic isolation valves would be installed in the Spent Fuel Pool 
Cooling (SFPC) system upstream of the Reverse Osmosis (RO) system 
borated water storage tank suction connections.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), in its supplemental 
letter dated October 21, 2013, the licensee provided a revised analysis 
of the issue of no significant hazards consideration, which is 
presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change requests Nuclear Regulatory Commission (NRC) 
approval of design features and controls that will be used to ensure 
that unisolating the SFPC Purification System and the Reverse 
Osmosis (RO) System during Unit operations does not significantly 
impact the Borated Water Storage Tank (BWST) or other plant 
equipment and that periodic limited operation of the RO System when 
aligned to a SFP during Unit operation does not significantly impact 
the Spent Fuel Pool (SFP) function or other plant equipment. The 
proposed change also requests NRC to approve proposed Technical 
Specification (TS) requirements that will impose operating 
restrictions and isolation requirements for the SFPC Purification 
System and the RO System.
    The new high energy piping and non-seismic piping being 
installed for the RO System is non-QA1 and is postulated to fail. 
Adequate measures have been provided to isolate the flood source 
(BWST or SFP) prior to affecting SSCs important to safety.
    The BWST will be automatically isolated prior to going below the 
TS water volume requirement. For the SFP, the suction to the RO 
system is above the required TS water level, therefore, the design 
ensures the required TS water level is maintained.
    Procedural controls will ensure that the boron concentration 
does not go below the TS limit as a result of water returned from 
the RO System with lower boron concentration. Thus, no adverse 
effects from decreased boron concentration will occur.
    The RO System takes suction from the top of the SFP to protect 
SFP inventory. Plant

[[Page 70592]]

procedures will prohibit the use of the RO System for the Units 1 
and 2 SFP during the time period directly after an outage that 
requires the Units 1 and 2 SFP level to be maintained higher than 
the TS Limiting Condition for Operation (LCO) 3.7.11 level 
requirement. The higher level is required to support TS LCO 3.10.1 
requirements for Standby Shutdown Facility (SSF) Reactor Coolant 
(RC) Makeup System operability (due to the additional decay heat 
from the recently offloaded spent fuel). Plant procedures will also 
specify the siphon be broken during this time period so the SFP 
water above the RO suction point cannot be siphoned off if the RO 
piping breaks. The proposed change does not impact the fuel 
assemblies, the movement of fuel, or the movement of fuel shipping 
casks. The SFP boron concentration, level, and temperature limits 
will not be outside of required parameters due to restrictions/
requirements on the system's operation. In addition, the proposed 
new Technical Specification will require the siphon be broken during 
movement of irradiated fuel assemblies in the SFP or movement of a 
cask over the SFP. Therefore, RO System operation cannot occur 
during these activities, effectively eliminating a Fuel Handling 
Accident (FHA) from occurring while the RO System is in operation.
    The BWST is used for mitigation of Steam Generator Tube Rupture 
(SGTR), Main Steam Line Break (MSLB), and Loss of Coolant Accidents 
(LOCAs). The SGTR and MSLB are bounded by the small break (SBLOCA) 
analyses with respect to the performance requirements for the High 
Pressure Injection (HPI) System. In the normal mode of Unit 
operation, the BWST is not an accident initiator. The SFP is 
evaluated to maintain acceptable criticality margin for all abnormal 
and accident conditions including FHAs and cask drop accidents. Both 
the BWST and SFP are specified by TS requirements to have minimum 
levels/volumes and boron concentrations. The BWST also has TS 
requirements for temperature. Prior to RO System operation, 
procedures will require the minimum required initial boron 
concentration and initial level/volume to be adjusted. Additionally, 
they will require the RO System operation to be restricted to a 
specified maximum time period before readjusting volume and boron 
concentration prior to another RO session. This ensures that the TS 
specified boron concentration and level/volume limits for both the 
SFP and the BWST are not exceeded during RO System operation. Thus, 
the design functions of the BWST and the SFP will continue to be met 
during RO System operation.
    The proposed TS will require the RO system to be isolated (by 
breaking the siphon) from the SFPs during fuel handling activities 
and will require the automatic isolation valves between the BWST and 
the SFPC Purification System, upstream of the branch line to the RO 
System branch line, be OPERABLE in MODES 1, 2, 3, and 4. The TS will 
also require manual valves in branch lines upstream of the SFPC 
Purification System automatic isolation valves to be closed and meet 
Inservice Testing (IST) Program leakage requirements.
    The additional controls imposed by the proposed Technical 
Specifications will provide additional assurance that isolation 
valves and operating restrictions credited to eliminate the need to 
analyze new release pathways will be in place.
    Therefore, allowing the SFPC Purification System and the RO 
System to be unisolated during Unit operation do not significantly 
increase the probability or consequences of any 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 RO System adds non-seismic piping in the Auxiliary Building. 
However, the break of a single non-seismic pipe in the Auxiliary 
Building has already been postulated as an event in the licensing 
basis. The RO System also does not create the possibility of a 
seismic event concurrent with a LOCA since a seismic event is a 
natural phenomenon event. The RO System does not adversely affect 
the Reactor Coolant System pressure boundary.
    Duke Energy also evaluated potential releases of radioactive 
liquid to the environment. Design features, controls imposed by the 
proposed Technical Specification, and procedural controls will 
preclude release of radioactive materials outside the Auxiliary 
Building by ensuring the SFPC Purification System and the RO System 
will be isolated when required.
    The SFP suction line is designed such that the SFP water level 
will not go below TS required levels, thus the fuel assemblies will 
have the TS required water level over them. Procedural controls will 
restrict the use of the RO System and require breaking vacuum on the 
Units 1 and 2 SFP suction line when the SSF conditions require the 
SFP level be raised to support SSF RC Makeup System operability. 
Thus, the SFP water level will not be reduced below required water 
levels for these conditions. RO System operating restrictions will 
prevent reducing the SFP boron concentration below TS limits.
    Since the BWST and SFP already have TS boron concentration and 
level/volume requirements and the RO System will be automatically 
isolated, the mitigation of a LOCA or FHA does not result in an 
increase in dose consequence. The design basis LOCA analysis for 
Oconee assumes 5 gpm back-leakage from the Reactor Building sump to 
the BWST. The automatic isolation valves will isolate on a BWST 
level prior to swapover to the recirculation phase and prior to 
going below the actual TS water level. The proposed TS will require 
the RO system to be isolated (by breaking the siphon) from the SFPs 
prior to movement of irradiated fuel assemblies in the SFP or 
movement of cask over the SFP and will require the automatic 
isolation valves between the BWST and RO System to be OPERABLE in 
MODES 1, 2, 3, and 4.
    The additional controls imposed by the proposed Technical 
Specifications will provide additional assurance that isolation 
valves and operating restrictions credited to eliminate the need to 
analyze new release pathways (introduced by allowing the SFPC 
Purification System and the RO system to be unisolated during Unit 
operation) will be in place.
    Therefore, operation of these systems unisolated will not create 
the possibility of a new or different kind of accident from any kind 
of accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Duke Energy evaluated the impact of allowing the SFPC 
Purification System and the RO System to be unisolated during Unit 
operation on SSCs important to safety and determined that the 
proposed TS controls and procedural controls will ensure that TS 
limits for SFP and BWST volume, temperature, and boron concentration 
will continue to be met. For the BWST, these controls will ensure 
the TS minimum BWST boron concentration and level are available to 
mitigate the consequences of a small break LOCA or a large break 
LOCA. For the SFP, these controls ensure the assumptions of the fuel 
handling and cask drop accident analyses are preserved. The proposed 
change does not significantly impact the condition or performance of 
SSCs relied upon for accident mitigation. This change does not alter 
the existing TS allowable values or analytical limits. The existing 
operating margin between Unit conditions and actual Unit setpoints 
is not significantly reduced due to these changes. The assumptions 
and results in any safety analyses are not impacted. Therefore, 
operation of the RO System during Unit operation does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Robert J. Pascarelli.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: June 7, 2013.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) 1.1 ``Definitions,'' for 
Shutdown Margin (SDM), to require calculation of the SDM at a reactor 
moderator temperature of 68 [deg]F or a higher temperature that is 
determined to represent the most reactive state throughout the 
operating cycle of the reactor. This change is needed to address new 
Boiling Water Reactor (BWR) fuel designs which may be more reactive at 
shutdown temperatures above 68 [deg]F.

[[Page 70593]]

    The NRC staff announced the availability of Technical 
Specifications (TSs) Task Force (TSTF) Traveler TSTF-535, Revision 0, 
``Revise Shutdown Margin Definition to Address Advanced Fuel Designs.'' 
The TSTF-535, Revision 0 provides guidance for plant-specific adoption 
of changes needed to address BWR fuel designs which may be more 
reactive at shutdown temperatures greater than 68 [deg]F, using the 
agency's Consolidated Line Item Improvement Process'' (CLIIP). The 
availability and the model safety evaluation of TSTF-535, Revision 0, 
was provided under ADAMS Accession No. ML12355A772, and published in 
the Federal Register dated November 19, 2012 (77 FR 69507).
    The licensee has reviewed the information provided by the NRC staff 
in TSTF-535, and the model safety evaluation, as announced in the 
Federal Register (FR) Notice of availability. The licensee concluded 
that the justification presented in the FR Notice of availability of 
TSTF-535, Revision 0 and the model safety evaluation, prepared by the 
NRC staff, is applicable to the James A. FitzPatrick Nuclear Power 
Plant and justifies the current request for amendment to TS 1.1, 
``Definitions'' for SDM.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed [amendment] involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No.
    The proposed change 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 [amendment] 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 [amendment] 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: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Acting Branch Chief: R. Beall.

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

    Date of amendment request: April 24, 2012, as supplemented by 
letters dated July 12 and August 23, 2012, and January 14, February 12, 
March 13, and June 13, 2013.
    Description of amendment request: The proposed amendment would 
adopt National Fire Protection Association (NFPA) 805, ``Performance-
Based Standard for Fire Protection for Light Water Reactor Generating 
Plants'' (2001 Edition). Implementation of the regulatory actions 
presented in the attachments to the license amendment request will 
enable Cooper Nuclear Station to adopt a new fire protection licensing 
basis which complies with the requirements in 10 CFR 50.48(a), 10 CFR 
50.48(c), and the guidance in Regulatory Guide (RG) 1.205, Revision 1.
    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.
    Operation of the Cooper Nuclear Station (CNS) in accordance with 
the proposed amendment does not result in a significant increase in 
the probability or consequences of accidents previously evaluated. 
The proposed amendment does not affect accident initiators or 
precursors as described in the CNS Updated Safety Analysis Report 
(USAR), nor does it adversely alter design assumptions, conditions, 
or configurations of the facility, and it does not adversely impact 
the ability of structures, systems, or components (SSCs) to perform 
their intended function to mitigate the consequences of an 
initiating event within the assumed acceptance limits. The proposed 
changes do not affect the way in which safety-related systems 
perform their functions as required by the accident analysis. The 
SSCs required to safely shut down the reactor and to maintain it in 
a safe shutdown condition will remain capable of performing their 
design functions.
    The purpose of this amendment is to permit CNS to adopt a new 
risk-informed, performance-based fire protection licensing basis 
that complies with the requirements in 10 CFR 50.48(a) and 10 CFR 
50.48(c), as well as the guidance contained in Regulatory Guide (RG) 
1.205. The NRC considers that NFPA 805 provides an acceptable 
methodology and performance criteria for licensees to identify fire 
protection requirements that are an acceptable alternative to the 10 
CFR part 50, appendix R, fire protection features (69 FR 33536; June 
16, 2004). Engineering analyses, which may include engineering 
evaluations, probabilistic risk assessments, and fire modeling 
calculations, have been performed to demonstrate that the 
performance-based requirements of NFPA 805 have been met.
    NFPA 805, taken as a whole, provides an acceptable alternative 
for satisfying General Design Criterion 3 (GDC 3) of appendix A to 
10 CFR part 50. It meets the underlying intent of the NRC's existing 
fire protection regulations and guidance, and achieves defense-in-
depth along with the goals, performance objectives, and performance 
criteria specified in NFPA 805, Chapter 1. In addition, if there are 
any increases in core damage frequency (CDF) or risk as a result of 
the transition to NFPA 805, the increase will be small, governed by 
the delta risk requirements of NFPA 805, and consistent with the 
intent of the Commission's Safety Goal Policy.
    Based on the above, the implementation of this amendment to 
transition the Fire Protection Plan (FPP) at CNS to one based on 
NFPA 805, in accordance with 10 CFR 50.48(c), does not result in a 
significant increase in the probability of any accident previously 
evaluated. In addition, all equipment required to mitigate an 
accident remains capable of performing the assumed function.

[[Page 70594]]

    Therefore, the consequences of any accident previously evaluated 
are not significantly increased with the implementation of this 
License Amendment Request.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Operation of CNS in accordance with the proposed license 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. Any scenario or 
previously analyzed accident with offsite dose consequences was 
included in the evaluation of design basis accidents (DBA) 
documented in the USAR as a part of the transition to NFPA 805. The 
proposed amendment does not impact these accident analyses. The 
proposed change does not alter the requirements or functions for 
systems required during accident conditions, nor does it alter the 
required mitigation capability of the fire protection program, or 
its functioning during accident conditions as assumed in the 
licensing basis analyses and/or DBA radiological consequences 
evaluations.
    The proposed amendment does not adversely affect accident 
initiators nor alter design assumptions, or conditions of the 
facility. The proposed amendment does not adversely affect the 
ability of SSCs to perform their design function. SSCs required to 
maintain the unit in a safe and stable condition remain capable of 
performing their design functions.
    The purpose of the proposed amendment is to permit CNS to adopt 
a new fire protection licensing basis which complies with the 
requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c) and the guidance 
in Revision 1 of RG 1.205. As indicated in the Statements of 
Consideration, the NRC considers that NFPA 805 provides an 
acceptable methodology and performance criteria for licensees to 
identify fire protection systems and features that are an acceptable 
alternative to the 10 CFR part 50, appendix R fire protection 
features.
    The requirements in NFPA 805 address only fire protection and 
the impacts of fire effects on the plant have been evaluated. The 
proposed fire protection program changes do not involve new failure 
mechanisms or malfunctions that could initiate a new or different 
kind of accident beyond those already analyzed in the USAR. Based on 
this, as well as the discussion above, the implementation of this 
amendment to transition the FPP at CNS to one based on NFPA 805, in 
accordance with 10 CFR 50.48(c), 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.
    Operation of CNS in accordance with the proposed license 
amendment does not involve a significant reduction in a margin of 
safety. The transition to a new risk-informed, performance-based 
fire protection licensing basis that complies with the requirements 
in 10 CFR 50.48(a) and 10 CFR 50.48(c) does not alter the manner in 
which safety limits, limiting safety system settings, or limiting 
conditions for operation are determined. The safety analysis 
acceptance criteria are not affected by this change. The proposed 
license amendment does not adversely affect existing plant safety 
margins or the reliability of equipment assumed in the USAR to 
mitigate accidents. The proposed change does not adversely impact 
systems that respond to safely shut down the plant and maintain the 
plant in a safe shutdown condition. In addition, the proposed 
license amendment will not result in plant operation in a 
configuration outside the design basis for an unacceptable period of 
time without implementation of appropriate compensatory measures.
    The purpose of the proposed license amendment is to permit CNS 
to adopt a new fire protection licensing basis which complies with 
the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c) and the 
guidance in Regulatory Guide 1.205. The NRC considers that NFPA 805 
provides an acceptable methodology and performance criteria for 
licensees to identify fire protection systems and features that are 
an acceptable alternative to the 10 CFR part 50, appendix R required 
fire protection features (69 FR 33536; June 16, 2004).
    The risk evaluations for plant changes, in part as they relate 
to the potential for reducing a safety margin, were measured 
quantitatively for acceptability using the delta risk guidance 
contained in RG 1.205. Engineering analyses, which may include 
engineering evaluations, probabilistic safety assessments, and fire 
modeling calculations, have been performed to demonstrate that the 
performance-based methods of NFPA 805 do not result in a significant 
reduction in the margin of safety.
    As such, the proposed changes are evaluated to ensure that risk 
and safety margins are kept within acceptable limits. Based on the 
above, the implementation of this amendment to transition the FPP at 
CNS to one based on NFPA 805, in accordance with 10 CFR 50.48(c), 
will not significantly reduce 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. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: Michael T. Markley.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: February 28, 2013.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3.6.5, ``Containment Air 
Temperature,'' to increase the allowable containment average 
temperature from 120 [deg]F to 125 [deg]F. The revised TS Section 3.6.5 
would read as follows: ``Containment average air temperature shall be 
<=125 [deg]F.''
    The licensee supports the proposed change by revising the analyses 
for Loss of Coolant Accident (LOCA) and a Main Steam Line break, and 
evaluating the containment response by either increase in initial 
containment air temperature or increase in the temperature of safety 
injection accumulators, which are located in the Ginna containment, and 
are expected to be at the same temperature as containment air.
    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 to increase the containment average air 
temperature limit to 125 [deg]F, from 120 [deg]F, does not alter the 
assumed initiators to any analyzed event. The probability of an 
accident previously evaluated will not be increased by this proposed 
change. This proposed change will not affect radiological dose 
consequence analyses. The radiological dose consequence analyses 
assume a certain containment atmosphere leak rate based on the 
maximum allowable containment leakage rate, which is not affected by 
the change in allowable average containment air temperature 
resulting in a higher calculated peak containment pressure. The 10 
CFR part 50, appendix J containment leak rate testing program will 
continue to ensure that containment leakage remains within the 
leakage assumed in the offsite dose consequence analyses. The 
acceptable leakage corresponds to the peak allowable containment 
pressure of 60 psig. The radiological dose consequence analyses 
assume a certain source term, which is not affected by the change in 
allowable average containment air temperature. All core limitations 
set forth in 10 CFR 50.46 continue to be met. The consequences of an 
accident previously evaluated will not be increased by this proposed 
change.
    Therefore, operation of the facility in accordance with the 
proposed change to the containment average air temperature limit 
will 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 provides for a higher allowable containment 
average air

[[Page 70595]]

temperature to that currently in the TS Section 3.6.5. The 
calculated peak containment temperature and pressure remain below 
the containment design temperature and pressure of 286[emsp14][deg]F 
and 60 psig. This change does not involve any alteration in the 
plant configuration (no new or different type of equipment will be 
installed) or make changes in the methods governing normal plant 
operation. The change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Therefore, operation of the facility in accordance with the 
proposed change to TS Section 3.6.5 would 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.
    The calculated peak containment pressure and temperature remain 
below the containment design pressure and temperature of 60 psig and 
286[emsp14][deg]F, respectively. The penalties applied to the BE 
LBLOCA [best estimate loss of coolant accident] analysis result in 
the limitations set forth in 10 CFR 50.46 continuing to be met. 
Since the radiological consequence analyses are based on the maximum 
allowable containment leakage rate, which is not being revised, the 
change in the calculated peak containment pressure and temperature 
and changes in core response do not represent a significant change 
in the margin of safety. The longterm impact of the peak containment 
temperature following a design basis accident exceeding the EQ 
profile by 2[emsp14][deg]F with respect to the current licensing 
basis is negligible.
    Therefore, operation of the facility in accordance with the 
proposed change to increase the allowable containment average air 
temperature from 120[emsp14][deg]F to 125[emsp14][deg]F 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: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Group, LLC, 750 East Pratt Street, 17 Floor, 
Baltimore, MD 21202.
    NRC Acting Branch Chief: Robert Beall.

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

    Date of amendment request: October 3, 2013.
    Description of amendment request: The proposed amendment would 
revise the scheduled completion date of the Cyber Security Plan 
Milestone 8.
    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 revises the Cyber Security Plan 
Implementation Schedule. This change does not alter accident 
analysis assumptions, add any initiators, or affect the function of 
plant systems or the manner in which systems are operated, 
maintained, modified, tested, or inspected.
    The proposed change is a change to the completion date of 
implementation milestone 8 that in itself does not require any plant 
modifications which affect the performance capability of the 
structures, systems, and components relied upon to mitigate the 
consequences of postulated accidents and have no impact on the 
probability or consequences of an 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 Cyber Security Plan 
Implementation Schedule. This proposed change to modify the 
completion date of implementation milestone 8 does not alter 
accident analysis assumptions, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected. The proposed 
change does not require any plant modifications which affect the 
performance capability of the structures, systems and components 
relied upon to mitigate the consequences of postulated accidents. 
This change also does not 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 accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Plant safety margins are established through limiting conditions 
for operation, limiting safety system settings, and safety limits 
specified in the technical specifications. The proposed change 
revises the Cyber Security Plan Implementation Schedule. Because 
there is no change to these established safety margins as result of 
this change, the proposed change does not involve a significant 
reduction in a margin of safety. 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: J. Hagood Hamilton, Jr., South Carolina 
Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 
29218.
    NRC Branch Chief: Robert J. Pascarelli.

South Carolina Electric and Gas Docket Nos. 52-027 and 52-028, Virgil 
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County, 
South Carolina

    Date of amendment request: October 2, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer 
Nuclear Station (VCSNS) Units 2 and 3 by departing from the plant-
specific Design Control Document (DCD) Tier 1 (and corresponding 
Combined License Appendix C information) and Tier 2 material by making 
changes to the Non-Class 1E dc and Uninterruptible Power Supply System 
(EDS) and Uninterruptible Power Supply System (IDS) and making changes 
to the corresponding Tier 1 information in Appendix C to the Combined 
License. The proposed changes would:
    (1) Increase EDS total equipment capacity, component ratings, and 
protective device sizing to support increased load demand,
    (2) Relocate equipment and moving Turbine Building (TB) first bay 
EDS Battery Room and Charger Room. The floor elevation increases from 
elevation 148'-0'' to elevation 148'-10'' to accommodate associated 
equipment cabling with this activity, and
    (3) Remove the Class 1E IDS Battery Back-up tie to the Non-Class 1E 
EDS Battery. Because, this proposed change requires a departure from 
Tier 1 information in the Westinghouse Advanced Passive 1000 DCD, the 
licensee also requested an exemption from the requirements of the 
Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.

[[Page 70596]]

    The design function of the Turbine Building (TB) is to provide 
weather protection for the laydown and maintenance of major turbine/
generator components. The TB first bay is a seismic Category II 
structure designed to prevent the collapse under a safe shutdown 
earthquake (SSE) to protect the adjacent auxiliary building. The 
electrical system and air-handling units are designed to provide 
electrical power to plant loads and maintain acceptable temperatures 
for electrical equipment rooms and work areas. The electrical 
equipment continues to be in accordance with the same codes and 
standards stated in the Updated Final Safety Analysis Report 
(UFSAR). The proposed relocation of equipment, including the 
increase in floor elevation by 10 inches to accommodate overhead 
equipment cabling, does not impact the TB design function. The TB 
first bay continues to meet seismic Category II requirements. Based 
on this, the proposed changes would not increase the probability of 
an accident previously evaluated.
    The proposed changes do not involve any accident initiating 
event, thus the probabilities of the accidents previously evaluated 
are not affected. The relocation of equipment does not involve any 
safety-related structures, systems, or components; the affected 
rooms do not represent a radioactive material barrier; and this 
activity does not affect the containment of radioactive material. 
The radioactive material source terms and release paths used in the 
safety analyses are unchanged, thus the radiological releases in the 
accident analyses are not affected. Therefore, the consequences of 
an accident previously evaluated are not affected.
    Therefore, there is no 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 would use the same type of electrical 
equipment with higher ratings and capacity, change the source of a 
battery back-up, and relocate equipment. The electrical equipment 
will continue to perform its design functions because the same 
electrical codes and standards as stated in the UFSAR continue to be 
met. Therefore the proposed changes do not affect equipment failure 
probabilities or alter any accident initiator or initiating sequence 
of events. The proposed changes in location of equipment and 
elevation of the TB first bay floor do not affect the design 
function of the TB first bay to protect the adjacent auxiliary 
building by meeting seismic Category II structure requirements, or 
affect the operation of the relocated equipment, or the ability of 
the relocated equipment to meet its design functions. Because the 
SSCs and equipment affected by the proposed changes continue to meet 
their design functions, the structural codes and standards as stated 
in the UFSAR, the proposed changes do not introduce a different type 
of accident than those previously considered.
    Therefore, this activity does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The current seismic requirements applicable to the seismic 
Category II TB first bay structure, including the seismic modeling 
and analysis methods, will continue to apply to the TB first bay 
floor elevation increase. The proposed changes to relocate equipment 
and the increase in the floor elevation will continue to meet the 
fire rating requirements and will be in accordance with the same 
codes and standards currently identified in the UFSAR. The proposed 
changes to the electrical equipment will continue to meet existing 
electrical equipment industry standard recommendations identified in 
the UFSAR. Because no safety analysis or design basis acceptance 
limit/criterion is challenged or exceeded by these proposed changes, 
no margin of safety is reduced.
    Therefore, the 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: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence Burkhart.

    Dated at Rockville, Maryland, this 15th day of November 2013.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2013-28225 Filed 11-25-13; 8:45 am]
BILLING CODE 7590-01-P