Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 67196-67207 [2014-26556]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 79, Number 218 (Wednesday, November 12, 2014)]
[Notices]
[Pages 67196-67207]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2014-26556]


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NUCLEAR REGULATORY COMMISSION

[NRC-2014-0243]


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 October 16, 2014 to October 29, 2014. The 
last biweekly notice was published on October 28, 2014.

DATES: Comments must be filed by December 12, 2014. A request for a 
hearing must be filed by January 12, 2015.

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-2014-0243. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     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: Angela Baxter, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-2976, email: Angela.Baxter@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2014-0243 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-2014-0243.
     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-2014-0243 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

[[Page 67197]]

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.
    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 unless the 
Commission finds an imminent danger to the health or safety of the 
public, in which case it will issue an appropriate order or rule under 
10 CFR Part 2.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave

[[Page 67198]]

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-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

[[Page 67199]]

Submitting Comments'' section of this document.

DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: September 16, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14259A564.
    Description of amendment request: The proposed amendment would 
modify the technical specifications (TS) by relocating specific 
surveillance frequencies to a licensee-controlled program with the 
adoption of Technical Specification Task Force (TSTF)--425, Revision 3, 
``Relocate Surveillance Frequencies to Licensee Control--RITSTF 
Initiative 5b'' (ADAMS Accession No. ML080280275). Additionally, the 
change would add a new program, the Surveillance Frequency Control 
Program (SFCP), to Section 5.5, ``Programs and Manuals'' of the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of any accident previously evaluated?
    Response: No.
    The proposed changes relocate 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 
TSs 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 changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    No new or different accidents result from utilizing 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. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, DTE 
Electric Company (DTE) will perform a probabilistic risk evaluation 
using the guidance contained in NRC approved NEI 04-10, Revision 1, 
in accordance with the TS SFCP. 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.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bruce R. Maters, DTE Energy, General 
Counsel--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.
    NRC Branch Chief: David L. Pelton.

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina; Docket 
Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2, 
Mecklenburg County, North Carolina; and Docket Nos. 50-269, 50-270, and 
50-287, Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina

    Date of amendment request: July 21, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14212A502.
    Description of amendment request: The amendment would revise the 
licensed operator training requirements to be consistent with the 
National Academy for Nuclear Training (NANT) program. Additionally, the 
amendment would make administrative changes to Technical Specification 
Sections 5.1, ``Responsibility''; 5.2, ``Organization''; 5.3, ``Unit 
Staff Qualifications''; 5.5, ``Programs and Manuals''; and for Catawba 
and McGuire, Section 5.7, ``High Radiation Area.''
    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, with NRC edits in square brackets, 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 Technical Specification (TS) changes regarding 
organization, unit staff responsibility and unit staff 
qualifications are administrative changes to clarify the current 
requirements for Duke Energy's licensed operator qualifications and 
training program. With this change, the TSs continue to meet the 
current requirements of 10 CFR 55. Although licensed operator 
qualifications and training may have an indirect impact on accidents 
previously evaluated, the [Nuclear Regulatory Commission (NRC)] 
considered this impact during the rulemaking process, and by 
promulgation of the revised 10 CFR 55 rule, concluded that this 
impact remains acceptable as long as the licensed operator training 
programs are certified to be accredited and are based on a systems 
approach to training. The proposed TS change takes credit for the 
National Academy for Nuclear Training (NANT) accreditation of the 
licensed operator training program.
    The proposed TS change regarding responsibility, organization 
and high radiation area is administrative in nature to reflect the 
current titles and responsibilities of station personnel and is 
consistent with Standard Technical Specifications (STS).
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously analyzed.
    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 are administrative changes to clarify 
the current requirements for Duke Energy's licensed operator 
qualifications and training program and to conform to the revised 10 
CFR 55. Similar to the discussion above, although licensed operator 
qualifications and training may have an indirect impact on the 
possibility of a new or different kind of accident from any accident 
previously evaluated, the [NRC] considered this impact during the 
rulemaking process, and by promulgation of the revised rule 
concluded that this impact remains acceptable as long as licensed 
operator training programs are certified to be accredited and based 
on a systems approach to training. As previously noted, the Duke 
Energy licensed operator training program is accredited by NANT and

[[Page 67200]]

is based on a systems approach to training. The proposed TS change 
takes credit for the NANT accreditation of the licensed operator 
training program.
    The proposed TS change regarding responsibility, organization 
and high radiation area does not impact any plant systems that are 
accident initiators nor does the proposed change adversely impact 
any accident mitigating system. No physical changes are being made 
to the plant. This change is administrative in nature to reflect the 
current titles and responsibilities of station personnel and to be 
consistent with STS.
    The proposed amendment does not impact plant design, hardware, 
system operation or procedures, and therefore 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 the margin of safety?
    Response: No.
    The proposed [TS] change regarding unit staff qualifications is 
an administrative change to clarify the current requirements 
applicable to Duke Energy's licensed operator qualifications and 
training program. With this change, the TS continue to meet the 
current requirements of 10 CFR 55. Although licensed operator 
qualifications and training may have an indirect impact on accidents 
previously evaluated, the NRC considered this impact during the 
rulemaking process, and by promulgation of the revised 10 CFR 55 
rule, concluded that this impact remains acceptable as long as the 
licensed operator training programs are certified to be accredited 
and are based on a systems approach to training. As noted 
previously, the Duke Energy licensed operator training program is 
accredited by NANT and is based on a systems approach to training.
    The NRC has concluded per NUREG-1262, that the standards and 
guidelines provided by the Institute for Nuclear Power Operations' 
NANT are equivalent to those put forth or endorsed by the NRC. As a 
result, maintaining a NANT accredited, systems approach based 
licensed operator training program is equivalent to maintaining an 
NRC approved licensed operator training program. Furthermore, the 
NRC published Regulatory Issue Summary (RIS) 2001-001 to familiarize 
licensees with the NRC's current guidelines for the qualification 
and training of Reactor Operator and Senior Operator license 
applicants. This document again acknowledges that the NANT 
guidelines for education and experience outline acceptable methods 
for implementing the NRC's regulations in this area. The margin of 
safety is maintained by virtue of maintaining the NANT accredited 
licensed operator training program.
    The proposed TS change regarding responsibility, organization 
and high radiation area is administrative in nature to reflect the 
current titles and responsibilities of station personnel and is 
consistent with STS. Systems and components are not impacted and 
therefore are capable of performing as designed. The performance of 
fission product barriers will not be impacted by the proposed 
change.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    Based on the above discussion, Duke Energy concludes that the 
proposed amendment presents no significant hazards consideration 
under the standards set forth in 10 CFR 50.92(c) and, accordingly, a 
finding of ``no significant hazards consideration'' is justified.

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

Energy Northwest, Docket No. 50-397, Columbia Generating Station (CGS), 
Benton County, Washington

    Date of amendment request: August 12, 2014, as supplemented by 
letter dated September 9, 2014. Publicly-available versions are in 
ADAMS under Accession Nos. ML14234A457, and ML14268A233, respectively.
    Description of amendment request: The amendment would revise the 
CGS Technical Specifications (TSs) to risk-inform requirements 
regarding selected Required Action end states by incorporating TS Task 
Force (TSTF) traveler TSTF-423, Revision 1, ``Technical Specifications 
End States, NEDC-32988-A.'' The Notice of Availability for TSTF-423, 
Revision 1, was published in the Federal Register on February 18, 2011 
(76 FR 9164).
    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 allows a change to certain required end 
states when the TS Completion Times for remaining in power operation 
will be exceeded. Most of the requested technical specification (TS) 
changes are to permit an end state of hot shutdown (Mode 3) rather 
than an end state of cold shutdown (Mode 4) contained in the current 
TS. The request was limited to: (1) Those end states where entry 
into the shutdown mode is for a short interval, (2) entry is 
initiated by inoperability of a single train of equipment or a 
restriction on a plant operational parameter, unless otherwise 
stated in the applicable TS, and (3) the primary purpose is to 
correct the initiating condition and return to power operation as 
soon as is practical. Risk insights from both the qualitative and 
quantitative risk assessments were used in specific TS assessments. 
Such assessments are documented in Section 6 of topical report NEDC-
32988-A, Revision 2, ``Technical Justification to Support Risk 
Informed Modification to Selected Required Action End States for BWR 
[Boiling-Water Reactor] Plants.'' They provide an integrated 
discussion of deterministic and probabilistic Issues, focusing on 
specific TSs, which are used to support the proposed TS end state 
and associated restrictions. The risk insights support the 
conclusions of the specific TS assessments. Therefore, the 
probability of an accident previously evaluated is not significantly 
increased, if at all. The consequences of an accident after adopting 
TSTF-423 are no different than the consequences of an accident prior 
to adopting TSTF-423. Therefore, the consequences of an accident 
previously evaluated are not significantly affected by this change. 
The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns.
    Therefore, the proposed change does not involve a significant 
Increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
If risk is assessed and managed, allowing a change to certain 
required end states when the TS Completion Times for remaining in 
power operation are exceeded (i.e., entry into hot shutdown rather 
than cold shutdown to repair equipment) will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change and the commitment by Energy Northwest to adhere to the 
guidance in TSTF-IG-05-02, ``Implementation Guidance for TSTF-423, 
Revision 1, `Technical Specifications End States, NEDC-32988-A,' '' 
will further minimize possible concerns.
    Thus, based on the above, this change does not create the 
possibility of a new or different-kind of accident from an accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change allows, for some systems, entry into hot 
shutdown rather than cold shutdown to repair equipment, if risk is 
assessed and managed. The BWROG's [BWR Owners Group's] risk 
assessment approach is comprehensive and follows NRC staff guidance 
as documented in Regulatory

[[Page 67201]]

Guides (RG) 1.174 and 1.177. In addition, the analyses show that the 
criteria of the three-tiered approach for allowing TS changes are 
met. The risk impact of the proposed TS changes was assessed 
following the three-tiered approach recommended in RG 1.177.
    A risk assessment was performed to justify the proposed TS 
changes. The net change to the margin of safety is insignificant.
    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: William A. Horin, Esq., Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: July 9, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14212A396.
    Description of amendment request: The amendment would modify the 
RBS Surveillance Requirements (SRs) related to Technical Specification 
3.8.1, ``AC [Alternating Current] Sources--Operating.'' Specifically, 
the proposed changes will lower the upper bound of the frequency SR 
Acceptance Criteria Tolerance Band (ACTB), lower the upper bound of the 
voltage SR ACTB for diesel generator (DG) 1A and DG 1B (existing DG 1C 
voltage SR ACTB is retained), and raise the lower bound of the test 
load SR ACTB.
    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 EDGs [emergency diesel generators] are not initiators for 
accidents evaluated in the USAR [Updated Safety Analysis Report]. 
The proposed changes do not alter the capability of the EDGs or 
their supporting systems to start, load and perform their intended 
functions as described in the USAR. The proposed changes do not 
impact the initiators of analyzed events, nor do they impact the 
mitigation of accidents.
    The proposed changes enable SR testing to demonstrate sufficient 
margin to ensure that the EDGs and equipment being powered by the 
EDGs will function as required to mitigate an accident as described 
in the USAR. Thus, the EDGs will be capable of performing their 
accident mitigation function as described in the USAR, and there is 
no impact on the consequences of accident analyses.
    The proposed changes increase the minimum EDG test loads, but 
the upper limits of the test loads are not changed. Furthermore, the 
test program (number and type of SR starts, test loads and run 
length) is not changed. Therefore, the effect of the proposed 
changes on EDG wear and/or reliability is negligible, and the 
proposed changes will not reduce EDG reliability from the current 
value of 95%.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve any physical alteration of 
the plant (e.g., no new or different type of equipment will be 
installed), or a change in the methods governing EDG operation. The 
changes ensure margin between the EDG SR test loads and the EDG 
maximum calculated loads and that the EDGs operate as assumed in the 
accident analyses.
    The purposes of the EDG surveillance tests are to confirm the 
capability of each EDG to start and achieve the minimum conditions 
required to accept the loads in the accident analysis. No changes 
are being made in operating philosophy, testing frequency, how EDGs 
operate or how EDGs are physically tested. The proposed changes do 
not affect the EDGs' ability to supply minimum voltage and frequency 
within 10 seconds (DG 1A and DG 1B), 13 seconds (DG 1C) or the 
minimum steady state voltage and frequency. The EDGs will continue 
to perform their intended safety function in accordance with the 
safety analysis. Therefore, the proposed changes do not affect 
safety analysis assumptions.
    The proposed changes do not degrade the EDGs, the circuits 
connected to the EDGs or the equipment powered by the EDGs. 
Therefore, no new failure modes or effects are introduced that could 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    The proposed changes do not affect the initiators of analyzed 
events, nor do they affect the mitigation of accidents.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes enable SR testing to demonstrate sufficient 
margin between demonstrated EDG capability in the surveillance tests 
and maximum calculated EDG loads to ensure that the EDGs and 
equipment being powered by the EDGs will function as required to 
mitigate an accident as described in the USAR. Thus the proposed 
changes do not involve a significant reduction in the EDG electrical 
load margin.
    The proposed changes increase the minimum EDG test loads, but 
the upper limits of the test loads are not changed. Furthermore, the 
test program (number and type of SR starts, test loads and run 
length) is not changed. Therefore, the effect of the proposed 
changes on EDG wear and/or reliability is negligible and the 
proposed changes do not involve a significant reduction in the EDG 
physical margin.
    The margin of safety is related to the confidence in the ability 
of the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The proposed changes do not directly affect these barriers, 
nor do they involve any adverse impact on the EDGs that serve to 
support these barriers in the event of an accident concurrent with a 
loss of offsite power. The proposed changes do not affect the EDG's 
capabilities to provide emergency power to plant equipment that 
mitigates the consequences of the accident. In summary: the proposed 
changes have no affect the ability of the EDGs to start and load; no 
change is made to the accident analysis assumptions; no margin of 
safety is reduced as part of this change; and the margin between the 
calculated emergency loads and minimum test load is ensured.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Douglas A. Broaddus.

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 3, 2014. A publicly-
available version is in ADAMS under Accession No. ML14247A522.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to eliminate the Main Steam Line 
Radiation Monitor (MSLRM) from initiating: (1) A Reactor Protection

[[Page 67202]]

System automatic reactor scram; and (2) a Primary Containment Isolation 
System isolation including automatic closure of the Main Steam Line 
Isolation Valves (MSIVs), Main Steam Line (MSL) drain valves, MSL 
sample line valves, Residual Heat Removal (RHR) system sample line 
valves, and Reactor Recirculation loop sample line valves. Existing 
requirements for the Mechanical Vacuum Pump (MVP) would be retained in 
the Technical Requirements Manual (TRM).
    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 eliminate the MSLRM trip and isolation 
function from initiating an automatic reactor scram and automatic 
closure of the MSIVs. The justification for eliminating the MSLRM 
trip and isolation functions is based on the NRC-approved evaluation 
provided in General Electric's (GE's) Licensing Topical Report (LTR) 
NEDO-31400A, ``Safety Evaluation for Eliminating the Boiling Water 
Reactor Main Steam Line Isolation Valve Closure Function and Scram 
Function of the Main Steam Line Radiation Monitor,'' dated October 
1992. The proposed changes also include the elimination of the MSLRM 
isolation function from closing the MSL drain valves, MSL sample 
line valves, RHR system sample line valves, and Reactor 
Recirculation loop sample line valves. The identified sample lines 
are small in comparison to the size of MSLs, and therefore, the 
effects of not isolating these lines for at least one hour is 
considered small and is supported by the dose analyses. The MSLRM 
system is not an initiator of any accident previously evaluated. 
Retaining requirements for the MVP in the TRM will ensure that 
appropriate measures and requirements are in place such that any 
release of radioactive material released from a gross fuel failure 
will be contained in the Main Condenser and processed through the 
Offgas System.
    The proposed changes do not introduce new equipment or new 
equipment operating modes. The proposed changes do not increase 
system or component pressures, temperatures, or flowrates for 
systems designed to prevent accidents or mitigate the consequences 
of an accident. There are no changes or modifications to the MVP. 
The MVP will continue to function as designed in all required modes 
of operation. Since these conditions do not change, the likelihood 
of a failure or malfunction of a Structure, System, or Component 
(SSC) is not increased. As a result, the probability of any accident 
previously evaluated is not significantly increased. The 
consequences of an accident previously evaluated (i.e., the Control 
Rod Drop Accident (CRDA)), have been evaluated consistent with the 
PBAPS licensing basis, which is based on Alternative Source Term (10 
CFR 50.67). As demonstrated by the supporting dose analyses, the 
consequences of the accident are within the regulatory acceptance 
criterion. As a result, the consequences of any accident previously 
evaluated are not significantly increased.
    Based on the above, Exelon concludes that 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.
    No new or different accidents result from the proposed changes. 
The proposed changes do not involve a change in the method of 
operation of plant SSC. The proposed changes do not increase system 
or component pressures, temperatures, or flowrates. There is no new 
system component being installed, no construction of a new facility, 
and no performance of a new test or maintenance function. The MVP 
will continue to function as designed in all required modes of 
operation. Since these conditions do not change, the proposed 
changes will not create the possibility of a new or different kind 
of accident. Retaining requirements for the MVP in the TRM will 
ensure that appropriate measures and requirements are in place such 
that any release of radioactive material released from a gross fuel 
failure will be contained in the Main Condenser and processed 
through the Offgas System. The elimination of the MSLRM trip and 
isolation functions as described is only credited in the CRDA 
analysis and no other event in the safety analysis. The proposed 
changes are consistent with the revised safety analysis assumptions 
for a CRDA as described in this license amendment request.
    Based on the above discussion, Exelon concludes that the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes eliminate the MSLRM trip and isolation 
functions from initiating an automatic reactor scram and automatic 
closure of the MSIVs along with closing of the MSL drain valves, MSL 
sample line valves, RHR system sample line valves, and Reactor 
Recirculation loop sample line valves and are justified based on the 
NRC-approved LTR NEDO-31400A and supporting dose analysis. Retaining 
requirements for the MVP in the TRM will ensure that appropriate 
measures and requirements are in place such that any release of 
radioactive material from a gross fuel failure will be contained in 
the Main Condenser and processed through the Offgas System.
    The proposed changes do not increase system or component 
pressures, temperatures, or flowrates for systems designed to 
prevent accidents or mitigate the consequences of an accident. 
Analyses performed consistent with the PBAPS licensing basis, 
demonstrate that the removal of the trip and isolation functions as 
described will not cause a significant reduction in the margin of 
safety, as the resulting offsite dose consequences are being 
maintained within regulatory limits. The proposed changes do not 
exceed or alter a design basis or a safety limit for a parameter to 
be described or established in the Updated Final Safety Analysis 
Report (UFSAR) or the Renewed Facility Operating License (FOL).
    As a result, Exelon concludes that the proposed changes do not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for Licensee: J. Bradley Fewell, Esquire, Vice President 
and Deputy General Counsel, Exelon Generation Company, LLC, 200 Exelon 
Way, Kennett Square, PA 19348.
    NRC Branch Chief: Meena K. Khanna.

NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowac 
County, Wisconsin

    Date of amendment request: July 2, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14183A944.
    Description of amendment request: The proposed amendment would 
modify the technical specifications (TSs) to address U.S. Nuclear 
Regulatory Commission (NRC) Generic Letter 2008-01, ``Managing Gas 
Accumulation in Emergency Core Cooling, Decay Heat Removal, and 
Containment Spray Systems,'' by adoption of Technical Specifications 
Task Force (TSTF) Traveler TSTF-523, ``Generic Letter 2008-01, Managing 
Gas Accumulation,'' Revision 2. The proposed change revises and adds TS 
surveillance requirements (SRs) to verify that the system locations 
susceptible to gas accumulation are sufficiently filled with water and 
to provide allowances which permit performance of the verification.
    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 provided below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?

[[Page 67203]]

    Response: No.
    The proposed change revises and adds SRs that require 
verification that the Emergency Core Cooling System (ECCS), Residual 
Heat Removal (RHR) System, and the Containment Spray System are not 
rendered inoperable due to accumulated gas and to provide allowances 
which permit performance of the revised verification. Gas 
accumulation in the subject systems is not an initiator of any 
accident previously evaluated. As a Result, the probability of any 
accident previously evaluated is not significantly increased. The 
proposed SRs ensure that the subject systems continue to be capable 
of performing their assumed safety function and are not rendered 
inoperable due to gas accumulation. Thus, the consequences of an 
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.
    The proposed change revises and adds SRs that require 
verification that the ECCS, RHR System, and Containment Spray System 
are not rendered inoperable due to accumulated gas and to provide 
allowances which permit performance of the revised verification. The 
proposed change does not involve a physical alternation 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 proposed change does not impose any new or different 
requirements that could initiate an accident. The proposed change 
does not alter assumptions made in the safety analysis and is 
consistent with the safety analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, RHR System, and Containment Spray System 
are not rendered inoperable due to accumulated gas and to provide 
allowances which permit performance of the revised verification. The 
proposed change adds new requirements to manage gas accumulation in 
order to ensure that the subject systems are capable of performing 
their assumed safety functions. The proposed SRs are more 
comprehensive that the current SRs and will ensure that the 
assumptions of the safety analysis are protected. The proposed 
change does not adversely affect any current plant safety margins or 
the reliability of the equipment assumed in the safety analysis. 
Therefore, there are no changes being made to any safety analysis 
assumptions, safety limits, or limiting safety system settings that 
would adversely affect plant safety as a result of the proposed 
change.
    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: William Blair, Managing Attorney--Nuclear, 
Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard, 
Juno Beach, FL 33408-0420.
    NRC Branch Chief: David L. Pelton.

NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowac 
County, Wisconsin

    Date of amendment request: July 3, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14190A267.
    Description of amendment request: The proposed amendment would 
modify the technical specifications (TSs) by relocating specific 
surveillances to a licensee-controlled program by adoption of Technical 
Specifications Task Force (TSTF) Traveler TSTF-425, Revision 3, 
``Relocate Surveillance Frequencies to Licensee Control--Risk Informed 
Technical Specification Task Force (RITSTF) Initiative 5B.'' The 
proposed change would also add a new program, the Surveillance 
Frequency Control Program, to TS Section 5.0, ``Administrative 
Controls,'' Subsection 5.5, ``Programs and Manuals.''
    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 provided 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 any accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis assumptions and current plant operating 
practice.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, NextEra 
will perform a probabilistic risk evaluation using the guidance 
contained in NRC-approved NEI 04-10, 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 (RG) 1.177.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William Blair, Managing Attorney--Nuclear, 
Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard, 
Juno Beach, FL 33408-0420.
    NRC Branch Chief: David L. Pelton.

South Carolina Electric and Gas Company 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: September 25, 2014. A publicly-

[[Page 67204]]

available version is in ADAMS under Accession No. ML14268A388.
    Description of amendment request: The proposed changes would revise 
the Combined Licenses (COLs) by increasing the tolerances listed for 
four concrete thicknesses in COL Appendix C and plant-specific Tier 1 
Table 3.3-1, ``Definition of Wall Thicknesses for Nuclear Island 
Buildings, Turbine Building, and Annex Building,'' from 1'' 
to 1\1/4\'' for one wall and from 1'' to 1\5/8\'' for the remaining three walls.
    Because, this proposed change requires a departure from Tier 1 
information in the Westinghouse Advanced Passive 1000 Design Control 
Document (DCD), the licensee also requested an exemption from the 
requirements of the Generic DCD Tier 1 in accordance with 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.
    As indicated in the Updated Final Safety Analysis Report 
Subsection 3.8.3.1, the containment internal structures and 
associated modules support the reactor coolant system components and 
related piping systems and equipment. The increase in tolerance 
associated with the concrete thickness of four of these containment 
internal structure walls do not involve any accident initiating 
components or events, thus leaving the probabilities of an accident 
unaltered. The increased tolerance does not adversely affect any 
safety-related structures or equipment nor does the increased 
tolerance reduce the effectiveness of a radioactive material 
barrier. Thus, the proposed changes would not affect any safety-
related accident mitigating function served the containment internal 
structures.
    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 tolerance increases do not change the performance 
of the affected containment internal structures. As demonstrated by 
the continued conformance to the applicable codes and standards 
governing the design of the structures, the walls with an increased 
concrete thickness tolerance continue to withstand the same effects 
as previously evaluated. There is no change to the design function 
of the affected modules and walls, and no new failure mechanisms are 
identified as the same types of accidents are presented to the walls 
before and after the change.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to increase the concrete thickness tolerance 
does not alter any design code compliance, design function, design 
analysis, or safety analysis input or result. As such, because the 
system continues to respond to design basis accidents in the same 
manner as before without any changes to the expected response of the 
structure, no safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the proposed changes. 
Accordingly, no safety margin is reduced by the increase of the wall 
concrete thickness tolerance.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence J. Burkhart.

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

    Date of amendment request: August 14, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14227A707.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-91 and NPF-92 for the Vogtle Electric 
Generating Plant (VEGP) Units 3 and 4. The requested amendment proposes 
changes to revise the VEGP Updated Final Safety Analysis Report 
(UFSAR), involving Tier 1 and associated Tier 2 departures that address 
the removal of an unneeded supply line from the Compressed and 
Instrument Air System (CAS) to the generator breaker package.
    Because this proposed change requires a departure from Tier 1 
information in the Westinghouse Advanced Passive 1000 design control 
document (DCD), the licensee also requested an exemption from the 
requirements of the Generic DCD Tier 1 in accordance with 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.
    The proposed change deletes a nonsafety-related air supply line 
to the (main) generator circuit breaker (GCB) from the CAS. The 
proposed changes do not involve any accident initiating component/
system failure or event, thus the probabilities of the accidents 
previously evaluated are not affected. The affected equipment does 
not affect or interact with safety-related equipment or a 
radioactive material barrier, and this activity does not involve the 
containment of radioactive material. Thus, the proposed changes 
would not affect any safety-related accident mitigating function. 
The radioactive material source terms and release paths used in the 
safety analyses are unchanged, thus the radiological releases in the 
UFSAR accident analyses are not affected.
    Therefore, the proposed amendment does not involve an 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 deletes a nonsafety-related air supply line 
to the GCB from CAS. No structure, system or component (SSC) or 
design function is affected, thus no equipment whose failure could 
initiate an accident is involved. No new interface with components 
that contain radioactive material is created. The proposed change 
does not create a new fault or sequence of events that could result 
in a radioactive material release.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change deletes a nonsafety-related air supply line 
to the GCB from CAS. The proposed changes do not affect any safety-
related equipment or function. The UFSAR Chapters 6 and 15 analyses 
are not affected. No safety analysis or design basis acceptance 
limit/criterion is challenged or exceeded by the proposed changes, 
thus a margin of safety is not directly nor indirectly affected.

    Therefore, the proposed amendment does not reduce 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.

[[Page 67205]]

    Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence J. Burkhart.

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

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

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

    Date of amendment request: August 14, 2013, as supplemented by 
letter dated May 14, 2014. Publicly-available versions are available in 
ADAMS under Accession Nos. ML13228A265, and ML14139A342, respectively.
    Brief description of amendment request: The amendment would modify 
the R.E. Ginna Nuclear Power Plant (Ginna) facility operating license, 
in accordance with Sec.  50.90 and as required under Order EA-13-092. 
The amendment would also modify the license to reflect a grant of 
Section 161A of the Atomic Energy Act, to permit the licensee's 
security personnel to possess and use weapons, devices, ammunition, or 
other firearms, notwithstanding state, local, and certain federal 
firearms laws that may prohibit such use. The NRC refers to this 
authority as ``stand-alone preemption authority.'' The licensee is 
seeking stand-alone preemption authority for standard weapons presently 
in use at the Ginna facility in accordance with the Ginna security 
plans.
    Date of publication of individual notice in Federal Register: 
October 27, 2014 (79 FR 63951).
    Expiration date of individual notice: November 26, 2014, for public 
comments; December 26, 2014, for hearing requests.

Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine 
Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York

    Date of amendment request: August 14, 2013, as supplemented by 
letter dated May 14, 2014. Publicly-available versions are available in 
ADAMS under Accession Nos. ML13228A265, and ML14139A342, respectively.
    Brief description of amendment request: The amendment would modify 
the Nine Mile Point Nuclear Station, Units 1 and 2 (Nine Mile Point) 
facility operating licenses, in accordance with Sec.  50.90 and as 
required under Order EA-13-092. The amendment would also modify the 
license to reflect a grant of Section 161A of the Atomic Energy Act, to 
permit the licensee's security personnel to possess and use weapons, 
devices, ammunition, or other firearms, notwithstanding state, local, 
and certain federal firearms laws that may prohibit such use. The NRC 
refers to this authority as ``stand-alone preemption authority.'' The 
licensee is seeking stand-alone preemption authority for standard 
weapons presently in use at the Nine Mile Point facility in accordance 
with the Nine Mile Point security plans.
    Date of publication of individual notice in Federal Register: 
October 27, 2014, (79 FR 63951).
    Expiration date of individual notice: November 26, 2014, for public 
comments; December 26, 2014, for hearing requests.

IV. 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. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina

    Date of application of amendments: May 20, 2014.
    Brief description of amendments: The amendments are administrative 
in nature to revise obsolete information that no longer pertains to the 
Technical Specifications related to the Reactor Protective System, the 
Engineered Safeguards Protective System, the Low Pressure Service Water 
Reactor Building Waterhammer Prevention Circuitry, and the Emergency 
Condenser Circulating Water System.
    Date of Issuance: October 21, 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, 388; Unit 2, 390; Unit 3, 389. A publicly-
available version is in ADAMS under Accession No. ML14195A355; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: August 5, 2014 (79 FR 
45473).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 21, 2014.
    No significant hazards consideration comments received: No.

[[Page 67206]]

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan

    Date of amendment request: December 11, 2013.
    Brief description of amendment: The amendment revises technical 
specification (TS) requirements to add a new Limiting Condition for 
Operation (LCO) Applicability requirement, LCO 3.0.9. The LCO 
establishes conditions under which TS systems would remain operable 
when required physical barriers are not capable of providing their 
related support function. The amendment is consistent with NRC-approved 
Technical Specification Task Force (TSTF) Standard Technical 
Specifications (STS) change TSTF-427, ``Allowance for Non-Technical 
Specification Barrier Degradation on Supported System OPERABILITY,'' 
Revision 2, using the consolidated line item improvement process.
    Date of issuance: October 22, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 252. A publicly-available version is in ADAMS under 
Accession No. ML13345B160; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-20: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 18, 2014 (79 FR 
15148).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 22, 2014.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois; Docket Nos. 
STN 50-454 and STN 50-455, Byron Station, Units 1 and 2, Ogle County, 
Illinois

    Date of application for amendment: March 18, 2014.
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) 3.4.15, ``RCS [Reactor Coolant System] Leakage 
Detection Instrumentation,'' to define a new time limit for restoring 
inoperable RCS leakage detection instrumentation to operable status and 
establish alternate methods of monitoring RCS leakage when one or more 
required monitors are inoperable. The changes are consistent with NRC-
approved Revision 3 to Technical Specification Task Force (TSTF) 
Improved Standard Technical Specification (STS) Change Traveler TSTF-
513, ``Revise PWR [pressurized-water reactor] Operability Requirements 
and Actions for RCS Leakage Instrumentation.'' The availability of this 
TS improvement was announced in the Federal Register on January 3, 2011 
(76 FR 189), as part of the consolidated line item improvement process.
    Date of issuance: October 20, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 179/185. A publicly-available version is in ADAMS 
under Accession No. ML14253A508; documents related to these amendments 
are listed in the Safety Evaluation enclosed with the amendments.
    Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: 
The amendments revised the Technical Specifications and License.
    Date of initial notice in Federal Register: June 24, 2014 (79 FR 
35804).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 20, 2014.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 31, 2013.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 5.5.9, ``Steam Generator (SG) Program,'' and TS 
5.6.9, ``Steam Generator Tube Inspection Report,'' to address 
implementation issues associated with the inspection periods. The 
amendments also revised TS 3.4.18, ``Steam Generator (SG) Tube 
Integrity,'' for administrative purposes. The revisions are consistent 
with Commission-approved Technical Specifications Task Force Standard 
Technical Specifications Change Traveler 510, Revision 2, ``Revision to 
Steam Generator Program Inspection Frequencies and Tube Sample 
Selection.''
    Date of issuance: October 29, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 308 and 286. A publicly-available version is in 
ADAMS under Accession No. ML14288A102; documents related to this these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: The 
amendments revised the License and TSs.
    Date of initial notice in Federal Register: July 22, 2014 (79 FR 
42547).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 29, 2014.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 16, 2012, as supplemented by 
letters dated July 12, 2013, May 30, 2014, and September 3, 2014.
    Brief description of amendments: The amendment revised Technical 
Specification (TS) 3.8.1, ``AC [Alternating Current] Sources-
Operating,'' by adding Surveillance Requirement (SR) 3.8.1.17, and 
modifying SRs 3.8.1.8, 3.8.1.11, and 3.8.2.1. The revisions are related 
to diesel generator (DG) testing duration, loading requirements, and 
frequency of DG sequencer testing.
    Date of issuance: October 21, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days after the end of the 2015 refueling outage.
    Amendment Nos.: 307 and 285. A publicly-available version is in 
ADAMS under Accession No. ML14280A522; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: The 
amendments revised the Licenses and TSs.
    Date of initial notice in Federal Register: March 4, 2013 (78 FR 
14130). The supplemental letters dated July 12, 2013, May 30, 2014, and 
September 3, 2014, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the NRC staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 21, 2014.
    No significant hazards consideration comments received: No.

[[Page 67207]]

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

    Date of amendment request: October 2, 2012, as supplemented by 
letters dated November 26, 2012, July 1, 2013, February 7, 2014, and 
October 3, 2014.
    Brief description of amendment: The amendments revised Technical 
Specification (TS) 3.8.3, ``Diesel Fuel Oil'' by removing the current 
stored diesel fuel oil numerical volume requirements from the TSs and 
replacing them with diesel generator (DG) operating time requirements 
consistent with NRC staff approved Technical Specifications Task Force 
Standard Technical Specifications Traveler 501, Revision 1, ``Relocate 
Stored Fuel Oil and Lube Oil Volume Values to Licensee Control.'' The 
amendments also revised TS 3.8.1, ``AC [alternating current] Sources-
Operating,'' by replacing the specific DG day tank fuel oil numerical 
volume requirements with the requirement to maintain greater than or 
equal to a 1-hour supply of fuel oil.
    Date of issuance: October 21, 2014.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 306 and 284. A publicly-available version is in 
ADAMS under Accession No. ML14239A491; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: The 
amendments revised the Licenses and TSs.
    Date of initial notice in Federal Register: March 4, 2013 (78 FR 
14130). The supplemental letters dated November 26, 2012, July 1, 2013, 
February 7, 2014, and October 3, 2014, provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the NRC staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 21, 2014.
    No significant hazards consideration comments received: No.

Northern States Power Company--Minnesota, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota

    Date of amendment request: October 30, 2012, as supplemented by 
letters dated May 16, 2013, June 7, 2013, March 13, 2014, and May 30, 
2014.
    Brief description of amendment: The amendment revises the Renewed 
Facility Operating License and Technical Specifications (TSs) to 
reflect fuel storage system changes; a revised criticality safety 
analysis that addresses legacy fuel types, in addition to the planned 
use of AREVA ATRIUM\TM\ 10XM fuel design; and adds a new TS 5.5.14, 
``Spent Fuel Pool Boral Monitoring Program,'' for assuring that the 
spent fuel pool storage rack neutron absorber material (Boral) 
continues to meet the minimum requirements assumed in the criticality 
safety analysis.
    Date of issuance: October 24, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 182. A publicly-available version is in ADAMS under 
Accession No. ML14197A020; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-22: This amendment 
revises the Renewed Facility Operating License and the Technical 
Specifications.
    Date of initial notice in Federal Register: June 11, 2013 (78 FR 
35063). The supplemental letters dated May 16, 2013, June 7, 2013, and 
March 13, 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 issued a revised no significant 
hazards consideration on June 24, 2014 (79 FR 35805), to consider the 
aspects of the new Boral monitoring program in TS 5.5.14 proposed in 
the May 30, 2014, supplemental letter.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 24, 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: March 24, 2014, as supplemented 
July 23, 2014.
    Brief description of amendments: The amendments revise the 
Technical Specification (TS) Reactor Core Safety Limits 2.1.1.1 and 
2.1.1.2 reactor steam dome pressure from 785 to 685 pounds per square 
inch guage (psig).
    Date of issuance: October 20, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: Unit 1--269 and Unit 2--213. A publicly-available 
version is in ADAMS under Accession No. ML14276A634; documents related 
to this these amendments are listed in the Safety Evaluation enclosed 
with the amendments.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the licenses and the Technical Specifications.
    Date of initial notice in Federal Register: June 24, 2014 (79 FR 
35806). The supplemental letter dated July 23, 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 October 20, 2014.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 31st day of October 2014.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2014-26556 Filed 11-10-14; 8:45 am]
BILLING CODE 7590-01-P