Biweekly Notice, Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 55507-55516 [2014-21833]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 79, Number 179 (Tuesday, September 16, 2014)]
[Notices]
[Pages 55507-55516]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2014-21833]


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NUCLEAR REGULATORY COMMISSION

[NRC-2014-0917]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: 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 August 21, 2014 to September 3, 2014. The 
last

[[Page 55508]]

biweekly notice was published on September 2, 2014.

DATES: Comments must be filed by October 16, 2014. A request for a 
hearing must be filed by November 17, 2014.

ADDRESSES: You may submit comments 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-2014-0917. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: 3WFN-06-A44M, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Beverly A. Clayton, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-3475, email: Beverly.Clayton@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2014-0917 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
     Federal rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2014-0917.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection 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 (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section.
     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-2014-0917 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.

II. Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses and Combined Licenses and Proposed No Significant 
Hazards Consideration Determination

    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.

A. Opportunity To Request a Hearing and Petition for Leave To Intervene

    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's 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.

[[Page 55509]]

    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.

B. Electronic Submissions (E-Filing)

    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 
ten 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/getting-started.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 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'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 NRC'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 public 
Web site at https://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-

[[Page 55510]]

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 these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

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

    Date of amendment request: June 30, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14184B384.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TS) by reducing the allowed maximum rated 
thermal power (RTP) at which the unit can operate when select High 
Pressure Injection (HPI) System equipment is inoperable.
    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 TS changes do not modify the reactor coolant system 
pressure boundary, nor make any physical changes to the facility 
design, material, or construction standards. The probability of any 
design basis accident (DBA) is not affected by this change, nor are 
the consequences of any DBA affected by this change. The new small 
break loss-of-coolant accident (SBLOCA) partial-power analysis 
demonstrates that all 10 CFR 50.46 acceptance criteria are 
satisfied. Radiological consequences for loss-of-coolant accident 
(LOCA) events are evaluated in ONS Updated Final Safety Analysis 
Report Section 15.15 for the Maximum Hypothetical Accident. The 
proposed changes will not impact assumptions and conditions 
previously used in the radiological consequence evaluations for the 
Maximum Hypothetical Accident. The proposed changes do not involve 
changes to any structures, systems, or components (SSCs) that can 
alter the probability for initiating a LOCA event.
    Therefore, the proposed TS changes do not significantly increase 
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 TS changes reduce the allowed power level that the 
unit may be operated at with select HPI equipment out-of-service. 
The changes do not alter the plant configuration (no new or 
different type of equipment will be installed) or make changes in 
methods governing normal plant operation. No new failure modes are 
identified, nor are any SSCs required to be operated outside the 
design bases.
    Therefore, the possibility of a new or different kind of 
accident from any kind of accident previously evaluated is not 
created.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed TS changes are supported by SBLOCA analyses which 
demonstrate that the acceptance criteria of 10 CFR 50.46 are 
satisfied. These analyses were performed in accordance with the 
Evaluation Model described in AREVA Topical Report BAW-10192P-A. The 
new SBLOCA analysis assumes a lower initial core power level (50% of 
rated thermal power (RTP)) than what was previously analyzed in 
support of TS 3.5.2 (i.e., 75% of RTP). The resulting peak cladding 
temperature results for the new SBLOCA analysis are lower than the 
existing analysis. In addition, a supplemental evaluation 
demonstrated that failure to perform a desired operator action of 
maintaining secondary-side pressure at 300 psig by throttling the 
atmospheric dump valve during a SBLOCA did not result in adverse 
affects to the new SBLOCA analysis results. Therefore, it is 
concluded that the proposed amendment request will not result in a 
significant decrease 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: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Robert J. Pascarelli.

Duke Energy Progress Inc., Docket No. 50-261, H.B. Robinson Steam 
Electric Plant, Unit 2 (HBRSEP2), Darlington County, South Carolina

    Date of amendment request: June 20, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14188B015.
    Description of amendment request: The amendment would revise 
Technical

[[Page 55511]]

Specification (TS) 5.5.9.b.2 for the Steam Generator (SG) Program 
accident-induced leakage performance criterion to correct an editorial 
error in the accident-induced leakage rate value for any design-basis 
accident other than a SG tube rupture.
    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 is a correction to an editorial error in the 
specified accident induced leakage performance criterion of TS 
5.5.9.b.2. The error in TS 5.5.9.b.2 being addressed by this 
proposed change was introduced at the time of the HBRSEP2 submittal 
of the NRC-approved Technical Specification Task Force (TSTF) 
traveler 449, Rev. 4, Steam Generator Tube Integrity. The accident-
induced leakage performance criterion will continue to be within the 
limit assumed in the accident analysis. As a result, neither the 
probability nor the consequences of any accident previously 
evaluated will be affected.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new or different accidents result from the proposed changes. 
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 or eliminate 
any existing requirements. The changes do not alter assumptions made 
in the safety analysis, it only corrects an editorial error in the 
accident-induced leakage performance criterion specified in the SG 
Program. The proposed changes are consistent with the safety 
analysis assumptions and current plant operating practice.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change will have no effect on the margin of safety. This 
proposed change corrects an editorial error in the accident-induced 
leakage performance criterion specified in the SG Program.

    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, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, 
Charlotte, NC 28202.
    Acting NRC Branch Chief: Lisa M. Regner.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: November 14, 2013. A publicly-available 
version is in ADAMS under Accession No. ML13323A516.
    Description of amendment request: The proposed amendment would 
eliminate operability requirements for secondary containment when 
handling sufficiently decayed irradiated fuel or a fuel cask following 
a minimum of 13 days after 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.
    The proposed changes do not modify the design or operation of 
equipment used to move spent fuel or to perform core alterations. 
The proposed changes cannot increase the probability of any 
previously analyzed accident because they are based on changes in 
Source Term, atmospheric dispersion and dose consequence analysis 
methodology, not in procedures or equipment used for fuel handling.
    The conservative re-analysis of the FHA [fuel-handling accident] 
concludes that the radiological consequences are within the 
regulatory limits established 10 CFR 50.67. This conclusion is based 
on the Alternate Source Term and guidance provided in Appendix B of 
Regulatory Guide 1.183 and analyses of fission product release and 
transport path that does not take credit for dose mitigation 
provided by engineered safeguards including secondary containment 
and the SGT system. The results of the core alteration events, other 
than the FHA, remain unchanged from the original design-basis that 
showed these events do not result in fuel cladding damage or 
radioactive release.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not introduce any new modes of plant 
operation and do not involve physical modifications to the plant.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Regulation in 10 CFR 50.67 permits licensees to voluntarily 
revise the accident source term used in design-basis radiological 
consequence analyses. This license amendment application evaluates 
the consequences of a design-basis fuel handling accident in 
accordance with this regulation and Regulatory Guide 1.183. The 
revised analysis concludes that the radiological consequences of the 
fuel handling accident are less than the regulatory allowable 
limits. Safety margins and analytical conservatisms are retained to 
ensure the analysis adequately bounds all postulated event 
scenarios. The selected assumptions and release models provide an 
appropriate and prudent safety margin against unpredicted events in 
the course of an accident and compensates for large uncertainties in 
facility parameters, accident progression, radioactive material 
transport and atmospheric dispersion. The proposed TS applicability 
statements continue to ensure that the total effective dose 
equivalent (TEDE) at the boundaries of the control room, the 
exclusion area, and low population zone boundaries are below the 
corresponding regulatory allowable limits in 10 CFR 50.67(b)(2).
    Therefore, the proposed change 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: Ms. Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Douglas A. Broaddus.

Exelon Generation Company, LLC, Docket Nos. STN 50-456, STN 50-457 and 
72-73, Braidwood Station, Units 1 and 2, Will County, Illinois
Exelon Generation Company, LLC, Docket Nos. STN 50-454, STN 50-455 and 
72-68, Byron Station, Units 1 and 2, Ogle County, Illinois
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit 1, DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-10, 50-237, 50-249 and 
72-37, Dresden Nuclear Power Station, Units 1, 2 and 3, Grundy County, 
Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373, 50-374 and 72-70, 
LaSalle County Station, Units 1 and 2, LaSalle County, Illinois

[[Page 55512]]

Exelon Generation Company, LLC, Docket Nos. 50-352, 50-353 and 72-65, 
Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania
Exelon Generation Company, LLC, et al., Docket No. 50-219 and 72-15, 
Oyster Creek Nuclear Generating Station, Ocean County, New Jersey
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
171, 50-277, 50-278 and 72-29, Peach Bottom Atomic Power Station, Units 
1, 2 and 3, York and Lancaster Counties, Pennsylvania
Exelon Generation Company, LLC, Docket Nos. 50-254, 50-265 and 70-53, 
Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Exelon Generation Company, LLC, Docket No. 50-320, Three Mile Island 
Nuclear Station, Unit 2, Dauphin County, Pennsylvania

    Date of amendment request: May 30, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14164A054.
    Description of amendment request: The proposed changes revise the 
Emergency Plans for the affected facilities to adopt the Nuclear Energy 
Institute's (NEl's) revised Emergency Action Level (EAL) schemes 
described in NEI 99-01, Revision 6, ``Development of Emergency Action 
Levels for Non-Passive Reactors,'' which has been endorsed by the NRC 
in a letter dated March 28, 2013. A publicly-available version can be 
found in ADAMS under Accession No. ML12346A463.
    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:

    The proposed changes have been reviewed considering the 
applicable requirements of 10 CFR 50.47, 10 CFR 50, Appendix E, and 
other applicable NRC documents. Exelon has evaluated the proposed 
changes to the affected sites' Emergency Plans and determined that 
the changes do not involve a Significant Hazards Consideration. In 
support of this determination, an evaluation of each of the three 
(3) standards, set forth in 10 CFR 50.92, ``Issuance of amendment,'' 
is provided below.
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to Exelon's EAL schemes to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, ``Development of 
Emergency Action Levels for Non-Passive Reactors,'' do not reduce 
the capability to meet the emergency planning requirements 
established in 10 CFR 50.47 and 10 CFR Part 50, Appendix E. The 
proposed changes do not reduce the functionality, performance, or 
capability of Exelon's ERO [Emergency Response Organization] to 
respond in mitigating the consequences of any design basis accident.
    The probability of a reactor accident requiring implementation 
of Emergency Plan EALs has no relevance in determining whether the 
proposed changes to the EALs reduce the effectiveness of the 
Emergency Plans. As discussed in Section D, ``Planning Basis,'' of 
NUREG-0654, Revision 1, ``Criteria for Preparation and Evaluation of 
Radiological Emergency Response Plans and Preparedness in Support of 
Nuclear Power Plants'';
    ``. . . The overall objective of emergency response plans is to 
provide dose savings (and in some cases immediate life saving) for a 
spectrum of accidents that could produce offsite doses in excess of 
Protective Action Guides (PAGs). No single specific accident 
sequence should be isolated as the one for which to plan because 
each accident could have different consequences, both in nature and 
degree. Further, the range of possible selection for a planning 
basis is very large, starting with a zero point of requiring no 
planning at all because significant offsite radiological accident 
consequences are unlikely to occur, to planning for the worst 
possible accident, regardless of its extremely low likelihood . . . 
.''
    Therefore, Exelon did not consider the risk insights regarding 
any specific accident initiation or progression in evaluating the 
proposed changes.
    The proposed changes do not involve any physical changes to 
plant equipment or systems, nor do they alter the assumptions of any 
accident analyses. The proposed changes do not adversely affect 
accident initiators or precursors nor do they alter the design 
assumptions, conditions, and configuration or the manner in which 
the plants are operated and maintained. The proposed changes do not 
adversely affect the ability of Structures, Systems, or Components 
(SSCs) to perform their intended safety functions in mitigating the 
consequences of an initiating event within the assumed acceptance 
limits.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to Exelon's EAL schemes to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not involve any 
physical changes to plant systems or equipment. The proposed changes 
do not involve the addition of any new plant equipment. The proposed 
changes will not alter the design configuration, or method of 
operation of plant equipment beyond its normal functional 
capabilities. All Exelon ERO functions will continue to be performed 
as required. The proposed changes do not create any new credible 
failure mechanisms, malfunctions, or accident initiators.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from those that have been 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to Exelon's EAL schemes to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not alter or exceed a 
design basis or safety limit. There is no change being made to 
safety analysis assumptions, safety limits, or limiting safety 
system settings that would adversely affect plant safety as a result 
of the proposed changes. There are no changes to setpoints or 
environmental conditions of any SSC or the manner in which any SSC 
is operated. Margins of safety are unaffected by the proposed 
changes to adopt the NEI 99-01, Revision 6 EAL scheme guidance. The 
applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E 
will continue to be met.
    Therefore, the proposed changes do not involve any reduction in 
a margin of safety.
    In conclusion, and based on the considerations discussed above:
    (1) There is reasonable assurance that the health and safety of 
the public will not be endangered by the proposed changes to adopt 
the EAL schemes established in NEI 99-01, Revision 6, as endorsed by 
the U.S. Nuclear Regulatory Commission (NRC); (2) the changes will 
be in compliance with the NRC's regulations; and (3) the issuance of 
the amendments will not be inimical to the common defense and 
security or to the health and safety of the public.

    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: Bradley Fewell, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: Travis L. Tate.

FirstEnergy Nuclear Operating Company (FENOC), Docket No. 50-440, Perry 
Nuclear Power Plant, Unit 1, Perry, OH

    Date of amendment request: March 25, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14084A165.
    Description of amendment request: The proposed changes are 
consistent with the NRC-approved Industry/Technical Specifications Task 
Force (TSTF) Traveler, TSTF-425, Revision 3, ``Relocate Surveillance 
Frequencies to

[[Page 55513]]

Licensee Control--RITSTF Initiative 5b.'' The proposed change relocates 
surveillance frequencies to a licensee controlled program, the 
Surveillance Frequency Control Program. This change is applicable to 
licensees using probabilistic risk guidelines contained in NRC-approved 
Nuclear Energy Institute (NEI) 04-10, ``Risk-Informed Technical 
Specifications Initiative 5b, Risk-Informed Method for Control of 
Surveillance Frequencies.''
    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 any 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 any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
significantly increased. The systems and components required by the 
technical specifications for which the surveillance frequencies are 
relocated are still required to be operable, meet the acceptance 
criteria for the surveillance requirements, and be capable of 
performing any mitigation function assumed in the accident analysis. 
As a result, the consequences of any accident previously evaluated 
are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the 
plant (that is, 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 change 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 
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 the 
TS [technical specification]), 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, FENOC will perform a probabilistic risk 
evaluation using the guidance contained in NRC approved Nuclear 
Energy Institute (NEI) 04-10, Revision 1, in accordance with the TS 
Surveillance Frequency Control Program. NEI 04-10, Revision 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, ``An Approach 
for Plant-Specific, Risk-Informed Decision-making: Technical 
Specifications.''
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based upon the reasoning presented above, FENOC concludes that 
the requested change does not involve a significant hazards 
consideration as set forth in 10 CFR 50.92(c), Issuance of 
Amendment.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Travis L. Tate.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, San 
Diego County, California

    Date of amendment request: March 21, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14085A141.
    Description of amendment request: The proposed amendment would 
revise the Operating License and associated Technical Specifications 
(TS) to reflect the permanent cessation of power operation. Because the 
licenses for SONGS, Units 2 and 3 no longer authorize emplacement or 
retention of fuel in the reactor vessel, the limiting conditions for 
operation and associated surveillance requirements that do not apply in 
the defueled condition are being proposed for deletion. The remaining 
portions of the TS are being proposed for revision and incorporation as 
the permanently defueled TS to provide a continuing acceptable level of 
safety, which addresses the reduced scope of postulated design basis 
accidents associated with a defueled plant, as described in the SONGS, 
Units 2 and 3 safety analyses.
    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.
    SONGS Units 2 and 3 have permanently ceased operation. The 
proposed amendment would modify the SONGS Units 2 and 3 facility 
operating licenses and TS by deleting the portions of the licenses 
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 criteria set forth in 10 CFR 50.36 for the contents of TS.
    Section 15 of the SONGS Updated Final Safety Analysis Report 
(UFSAR) described the design basis accident (DBA) and transient 
scenarios applicable to SONGS Units 2 and 3 during power operations. 
With the reactors in a permanently defueled condition, the fuel 
storage pools and their 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 SCE [Southern California 
Edison] in accordance with 10 CFR 50.82(a)(1), and the consequent 
removal of authorization to operate the reactors or to place or 
retain fuel in the reactors in accordance with 10 CFR 50.82(a)(2), 
most of the accident scenarios postulated in the UFSAR 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 shut down 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 shut down 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

[[Page 55514]]

exposures exceeding limits. However, after the termination of 
reactor operations at SONGS Units 2 and 3 and the permanent removal 
of the fuel from the reactor vessels (following 17 months of decay 
time after shut down) and purging of the contents of the waste gas 
decay tanks, none of the SSCs at SONGS Units 2 and 3 are required to 
be relied on for accident mitigation. Therefore, none of the SSCs at 
SONGS Units 2 and 3 meet the definition of a safety-related SSC 
stated in 10 CFR 50.2 (with the exception of the passive fuel 
storage 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 shut down and defueled 
status of SONGS Units 2 and 3 has no impact on the remaining DBA. 
The removal of limiting conditions for operation (LCOs) or 
surveillance requirements (SRs) that are related only to the 
operation of the nuclear reactors 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 SONGS Units 2 and 3 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 
SONGS Units 2 and 3.
    Since SONGS Units 2 and 3 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 shut down of the reactor will have reduced the consequences of 
the FHA [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 FHA 
analysis and is included in the permanently defueled TS.
    The fuel storage pool water level, fuel storage pool boron 
concentration, and spent fuel assembly storage TS 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 fuel storage pool.
    The deletion and modification of provisions of the 
administrative controls does 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 shut down and defueled condition 
of the reactors.
    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. 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 shut down 
and defueled and SCE is no longer authorized to operate the 
reactors.
    The proposed deletion of requirements of the SONGS Unit 2 and 
Unit 3 TS do not affect systems credited in the accident analysis. 
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 fuel storage pool water 
level, fuel storage pool boron concentration, and spent fuel 
assembly storage TS 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 licenses for SONGS Units 2 and 3 no 
longer authorize operation of the reactors or emplacement or 
retention of fuel into the reactor vessels, 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 
accidents do not credit SSCs for mitigation. The proposed amendment 
does not adversely affect the inputs or assumptions of any of the 
design basis analyses that impact an accident.
    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 SONGS TS 
Units 2 and 3 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 reactors are no longer 
possible because the reactors are permanently shut down and defueled 
and SCE is no longer authorized to operate the reactors.
    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 accidents 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 requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: Douglas A. Broaddus.

Southern Nuclear Operating Company Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County, 
Georgia

    Date of amendment request: June 4, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14156A477.
    Description of amendment request: The purpose of the proposed 
license amendment request is to address proposed changes related to 
departure from the plant-specific Design Control Document (DCD) Tier 1 
(and corresponding Combined License Appendix C information) and Tier 2 
material to reconcile differences in the various valve table entries.
    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

[[Page 55515]]

consideration, which is presented below:

    1. Does the requested amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No.
    The proposed changes do not result in any physical changes to 
the plant, and therefore do not change any safety-related design 
requirement, qualification requirement or function. The proposed 
changes do not involve any accident initiating event or component 
failure, thus, the probabilities of the accidents previously 
evaluated are not affected. The proposed changes do not affect the 
radioactive material releases used in the accident analyses, thus, 
the radiological releases in the accident analyses are not affected. 
The proposed changes do not affect any postulated non-radioactive 
accident scenario as evaluated in UFSAR [Updated Final Safety 
Analysis Report] Chapter 15.
    Therefore, the requested amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the requested amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not result in any physical changes to 
the plant, and therefore do not adversely affect any structure, 
system or component. No safety-related equipment qualification or 
design function is affected. The proposed changes do not introduce a 
new failure mode or create a new fault or sequence of events that 
could result in a radioactive material release.
    Therefore, the requested amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not result in any physical changes to 
the plant, and therefore do not change valve performance, including 
containment isolation. No safety acceptance criterion would be 
exceeded or challenged. No safety related function would be 
affected. Valve qualification would not be affected.
    The proposed changes do not affect compliance with existing 
design codes and regulatory criteria and do not affect any safety 
analysis.
    Therefore, the requested 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: Mr. M. Stanford Blanton, Blach & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence Burkhart.

III. 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 can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: September 12, 2013, as 
supplemented by letters dated May 20 and July 22, 2014.
    Brief description of amendments: The amendments modify Technical 
Specification (TS) 3.3.2. Specifically, the change modifies setpoints 
associated with the auxiliary feedwater pump suction transfer on low 
suction pressure.
    Date of issuance: August 27, 2014.
    Effective date: This license amendment is effective as of its date 
of issuance and shall be implemented within 60 days of issuance.
    Amendment Nos.: 273 and 253. A publicly-available version is in 
ADAMS under Accession No. ML14211A403; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the licenses and technical specifications.
    Date of initial notice in Federal Register: December 10, 2013 (78 
FR 74179). The supplemental letters dated May 20 and July 22, 2014, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 27, 2014.
    No significant hazards consideration comments received: No

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

    Date of application for amendments: September 28, 2012, as 
supplemented by letters dated February 15, 2013, May 7, 2013, May 24, 
2013, June 4, 2013, June 27, 2013, July 30, 2013, July 31, 2013, August 
5, 2013, August 22, 2013, August 29, 2013, September 13, 2013, October 
11, 2013, October 15, 2013, October 31, 2013, December 6, 2013, 
December 20, 2013, January 17, 2014, January 31, 2014 (2 letters), 
February 20, 2014, February 28, 2014, March 10, 2014, March 17, 2014, 
April 11, 2014, April 18, 2014, May 6, 2014, June 5, 2014, and June 20, 
2014.
    Brief description of amendments: The amendments authorize an 
increase in the maximum licensed thermal power level for PBAPS, Units 2 
and 3, from 3514 megawatts thermal (MWt) to 3951 MWt, which is an 
increase of approximately 12.4 percent.
    Date of issuance: August 25, 2014.
    Effective date: For PBAPS, Unit 2, the amendment is effective as of 
its date of issuance and shall be implemented prior to startup from 
refueling outage P2R20. For PBAPS, Unit 3, the

[[Page 55516]]

amendment is effective as of its date of issuance and shall be 
implemented prior to startup from refueling outage P3R20.
    Amendments Nos.: 293 and 296. A publicly-available version is in 
ADAMS under Accession No. ML14133A046; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the Facility Operating Licenses and the Technical 
Specifications.
    Date of initial notice in Federal Register: April 9, 2013 (78 FR 
21168). The letters dated February 15, 2013, May 7, 2013, May 24, 2013, 
June 4, 2013, June 27, 2013, July 30, 2013, July 31, 2013, August 5, 
2013, August 22, 2013, August 29, 2013, September 13, 2013, October 11, 
2013, October 15, 2013, October 31, 2013, December 6, 2013, December 
20, 2013, January 17, 2014, January 31, 2014 (2 letters), February 20, 
2014, February 28, 2014, March 10, 2014, March 17, 2014, April 11, 
2014, April 18, 2014, May 6, 2014, June 5, 2014, and June 20, 2014, 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination or expand 
the application beyond the scope of the original Federal Register 
notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 25, 2014.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 28, 2013, as supplemented by 
letters dated June 19, and November 11, 2013 and January 22, March 14, 
March 26, and June 6, 2014.
    Brief description of amendment: The amendment revised the Ginna 
Nuclear Power Plant Technical Specifications (TSs) to revise the 
allowable containment average air temperature from ``<= 
120[emsp14][deg]F'' to ``<= 125[emsp14][deg]F'' for TS 3.6.5 
``Containment Air Temperature.''
    Date of issuance: August 12, 2014.
    Effective date: As of the date of issuance to be implemented within 
45 days.
    Amendment No.: 116. A publicly-available version is in ADAMS under 
Accession No. ML14232A125; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: November 26, 2013 (78 
FR 70594). The supplemental letters dated June 19, and November 11, 
2013, and January 22, March 14, March 26, and June 6, 2014, 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 12, 2014.
    No significant hazards consideration comments received: No.
    Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia
    Date of application for amendments: July 23, 2013, as supplemented 
August 5, 2014.
    Brief description of amendments: The amendments revise the 
Technical Specification (TS) requirements and add license conditions 
related to control room envelope habitability in accordance with the 
Nuclear Regulatory Commission approved Revision 3 of Technical 
Specification Task Force (TSTF) Standard Technical Specifications 
Change Traveler TSTF-448, ``Control Room Habitability.''
    Date of issuance: August 29, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: Unit 1-268 and Unit 2-212. A publicly-available 
version is in ADAMS under Accession No. ML14147A410; documents related 
to this amendment are listed in the Safety Evaluation enclosed with the 
amendment.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Renewed Facility Operating licenses and the 
Technical Specifications.
    Date of initial notice in Federal Register: September 3, 2013 (78 
FR 54290). The supplement dated August 5, 2014, 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.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 29, 2014.
    No significant hazards consideration comments received: No.
    Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County
    Date of application for amendment: February 22, 2013.
    Brief description of amendment: The amendment revised Technical 
Specification 3.1.6, ``Control Bank Insertion Limits,'' to include 
text, in Condition A, stating, ``for reasons other than Condition C.'' 
This text addition modifies Condition A, for control bank sequence or 
overlap limits, to include language currently in Condition B, for 
control bank insertion limits, this change would point to Condition C, 
which, if applicable, would allow the specified completion time to 
restore the control bank to within the insertion limit to be increased 
from 2 hours to 72 hours. This would align the description of the 
sequence and overlap limit of Condition A with the description of 
control bank insertion limit Condition B.
    Date of issuance: August 27, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 272 and 254. A publicly-available version is in 
ADAMS under Accession No. ML14188C453; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
revised the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 30, 2013 (78 FR 
25317).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 27, 2014.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 5th day of September 2014.

    For the Nuclear Regulatory Commission.
A. Louise Lund,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2014-21833 Filed 9-15-14; 8:45 am]
BILLING CODE 7590-01-P