Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 65807-65822 [2015-27042]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 80, Number 207 (Tuesday, October 27, 2015)]
[Notices]
[Pages 65807-65822]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-27042]


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NUCLEAR REGULATORY COMMISSION

[NRC-2015-0242]


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 September 29 to October 9, 2015. The last 
biweekly notice was published on October 13, 2015.

DATES: Comments must be filed by November 27, 2015. A request for a 
hearing must be filed by December 28, 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-2015-0242. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: OWFN-12-H08, 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: Paula Blechman, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-2242, email: Paula.Blechman@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2015-0242 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-2015-0242.
     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-2015-0242, facility name, unit 
number(s), application date, and subject in your comment submission.
    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

[[Page 65808]]

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 within 60 days, 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 set forth 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 with respect to resolution of that person's admitted 
contentions, including the opportunity to present evidence and to 
submit a cross-examination plan for cross-examination of witnesses, 
consistent with NRC regulations, policies and procedures.
    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).
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, 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

[[Page 65809]]

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.
    A State, local governmental body, federally-recognized Indian 
tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission by 
December 28, 2015. The petition must be filed in accordance with the 
filing instructions in the ``Electronic Submissions (E-Filing)'' 
section of this document, and should meet the requirements for 
petitions for leave to intervene set forth in this section, except that 
under Sec.  2.309(h)(2) a State, local governmental body, or Federally-
recognized Indian tribe, or agency thereof does not need to address the 
standing requirements in 10 CFR 2.309(d) if the facility is located 
within its boundaries. A State, local governmental body, Federally-
recognized Indian tribe, or agency thereof may also have the 
opportunity to participate under 10 CFR 2.315(c).
    If a hearing is granted, any person who does not wish, or is not 
qualified, to become a party to the proceeding may, in the discretion 
of the presiding officer, be permitted to make a limited appearance 
pursuant to the provisions of 10 CFR 2.315(a). A person making a 
limited appearance may make an oral or written statement of position on 
the issues, but may not otherwise participate in the proceeding. A 
limited appearance may be made at any session of the hearing or at any 
prehearing conference, subject to the limits and conditions as may be 
imposed by the presiding officer. Persons desiring to make a limited 
appearance are requested to inform the Secretary of the Commission by 
December 28, 2015.

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 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,

[[Page 65810]]

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, in some instances, 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. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: August 28, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15244B179.
    Description of amendment request: The amendment provides a 
temporary extension to the Completion Time for Technical Specification 
3.5.2, ``Emergency Core Cooling Systems (ECCS)--Operating,'' Required 
Action A.1. The temporary extension will be used to allow the licensee 
to effect an on-line repair of the Residual Heat Removal (RHR) pump 
motor air handling unit.
    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 ECCS provides a mitigating function, and as such, it does 
not impact the probability of an accident. The consequences of an 
accident requiring the ECCS function will continue to be mitigated 
by the operable 1B RHR system train during the extended period in 
which the 1A RHR system train is considered inoperable. Each of the 
two RHR trains are redundant, so the 1B RHR pump is capable of 
performing the necessary mitigating function.
    Additionally, engineering evaluations, as documented in the 
[Engineering Change (EC)] process, demonstrate that the 1A RHR pump 
will continue to be capable of performing its mitigating ECCS 
function using a defense-in-depth measure that establishes alternate 
forced cooling to the room.
    As such, the proposed amendment does not result in an increase 
in consequences of an accident.
    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 accident causal mechanisms are created as a result of 
this proposed license amendment request (LAR). No changes are being 
made to any SSC [structure, system, or component] that will 
introduce any new accident causal mechanisms. The defense-in-depth 
measure to install alternate forced cooling to the 1A RHR pump motor 
during the repair evolution has been analyzed and evaluated using 
the Duke Energy EC process.
    The EC concludes that the installation of alternate forced 
cooling equipment would not adversely impact other components such 
that a new or different accident scenario is created.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The performance of the fuel cladding, reactor coolant and 
containment systems will not be impacted by the proposed LAR.
    The proposed activity only impacts the amount of time that the 
1A RHR system can be considered inoperable. The amount of inoperable 
time still remains small relative to the total operating time, and 
the 1A RHR train would still be considered available (i.e., capable 
of performing its ECCS function) during the period of extended 
inoperability. However, even if the train were considered 
unavailable, the total hours of unavailability would remain bounded 
by the limits established by the Maintenance Rule program.
    Therefore, it is concluded that the proposed changes do not 
involve a significant reduction in the margin of safety.

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

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: May 19, 2015, as supplemented by letter 
dated August 20, 2015. Publicly-available versions are in ADAMS under 
Accession Nos. ML15146A056 and ML15239B290, respectively.
    Description of amendment request: The proposed amendments add a 
Reactor Protective System Nuclear Overpower--High Setpoint trip for 
three (3) reactor coolant pump operation to Technical Specification 
Table 3.3.1-1, ``Reactor Protective System Instrumentation.'' The 
existing overpower protection for three (3) reactor coolant pump 
operation is the Nuclear Overpower Flux/Flow/Imbalance trip function. 
The new setpoint provides an absolute setpoint that can be actuated 
regardless of the transient or Reactor Coolant System flow conditions 
and provides a

[[Page 65811]]

significant margin gain for the small steam line break accident.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment adds a high flux trip for three (3) 
Reactor Coolant Pump (RCP) Operation by modifying the existing 
Nuclear Overpower-High Setpoint function in Technical Specification 
(TS) Table 3.3.1-1 to delineate between a setpoint valid for four 
(4) RCP operation and three (3) RCP operation. TS 3.4.4 is modified 
to require the Nuclear Overpower--High Setpoint to be reset to less 
than or equal to the Allowable Value of Table 3.3.1-1 for three (3) 
RCPs operating. The proposed change provides automatic overpower 
protection when the plant is operating with three (3) RCPs. The 
existing overpower protection for three (3) RCP operation is the 
Nuclear Overpower Flux/Flow/Imbalance trip function. Providing a 
Nuclear Overpower flux setpoint provides an absolute setpoint that 
can be actuated regardless of the transient or RCS flow conditions. 
The proposed TS change does 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 significantly affected by this change. 
The proposed change does not involve changes to any structures, 
systems, or components (SSCs) that can alter the probability for 
initiating a LOCA [loss-of-coolant accident] event. This amendment 
request includes the adoption of Option A of Technical Specification 
Task Force (TSTF) TSTF-493-A, Revision 4, ``Clarify Application of 
Setpoint Methodology for LSSS [Limiting Safety System Setting] 
Functions,'' for the Nuclear Overpower--High Setpoint trip function 
of TS Table 3.3.1-1. The TS changes associated with the 
implementation of TSTF-493-A will provide additional assurance that 
the instrumentation setpoints for the Nuclear Overpower--High 
Setpoint trip function are maintained consistent with the setpoint 
methodology to ensure the required automatic trips and safety 
feature actuations occur such that the safety limits are not 
exceeded.
    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 amendment adds a high flux trip for three (3) 
Reactor Coolant Pump Operation by modifying the existing Nuclear 
Overpower-High Setpoint function in TS Table 3.3.1-1 to delineate 
between a setpoint valid for four (4) RCP operation and three (3) 
RCP operation. This proposed change and the implementation of TSTF-
493-A 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 amendment adds a high flux trip for three (3) 
Reactor Coolant Pump Operation by modifying the existing Nuclear 
Overpower-High Setpoint function in TS Table 3.3.1-1 to delineate 
between a setpoint valid for four (4) RCP operation and three (3) 
RCP operation. This proposed TS change and the implementation of 
TSTF-493-A do not involve: (1) A physical alteration of the Oconee 
Units; (2) the installation of new or different equipment; or (3) 
any impact on the fission product barriers or safety limits. The 
proposed change adds a new setpoint, which is more conservative than 
the existing high flux setpoint that initiates a protective action 
to provide protection for power excursion events initiated from 
three (3) RCP operation equivalent to that provided for four (4) RCP 
operation.
    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: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Robert J. Pascarelli.
Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington
    Date of amendment request: September 2, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15245A777.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) by adding Limiting Condition 
for Operation (LCO) 3.0.9 to address the impact of unavailable 
barriers, not explicitly addressed in the TSs but required for 
operability of supported systems in TSs. The LCO 3.0.9 establishes 
conditions under which TS systems would remain operable when required 
physical barriers are not capable of providing their safety-related 
function. Also, the proposed amendment would replace the term ``train'' 
with the term ``division'' in LCO 3.0.8 to be consistent with the 
terminology proposed in LCO 3.0.9, which is editorial in nature.
    The proposed changes to the TS are consistent with the NRC-approved 
Technical Specification Task Force (TSTF) Standard Technical 
Specification change traveler TSTF-427, ``Allowance for Non-Technical 
Specification Barrier Degradation on Supported System OPERABILITY,'' 
Revision 2 (ADAMS Accession No. ML061240055). The availability of the 
TS improvement and the model application was published in the Federal 
Register on October 3, 2006 (71 FR 58444), as part of the Consolidated 
Line Item Improvement Process (CLIIP).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee affirmed 
the applicability of the model no significant hazards consideration 
determination, which is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated

    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an unavailable barrier if risk is assessed and managed. 
The postulated initiating events which may require a functional 
barrier are limited to those with low frequencies of occurrence, and 
the overall TS system safety function would still be available for 
the majority of anticipated challenges. Therefore, the probability 
of an accident previously evaluated is not significantly increased, 
if at all. The consequences of an accident while relying on the 
allowance provided by proposed LCO 3.0.9 are no different than the 
consequences of an accident while relying on the TS required actions 
in effect without the allowance provided by proposed LCO 3.0.9. 
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, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident From any Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported

[[Page 65812]]

system TS when inoperability is due solely to an unavailable 
barrier, if risk is assessed and managed, 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 will further minimize possible concerns.
    Thus, this change does not create the possibility of a new or 
different kind of accident from an accident previously evaluated.

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

    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an unavailable 
barrier, if risk is assessed and managed. The postulated initiating 
events which may require a functional barrier are limited to those 
with low frequencies of occurrence, and the overall TS system safety 
function would still be available for the majority of anticipated 
challenges. The risk impact of the proposed TS changes was assessed 
following the three-tiered approach recommended in [NRC Regulatory 
Guide 1.177, ``An Approach for Plant-Specific Risk-Informed 
Decisionmaking: Technical Specifications,'' August 1998 (ADAMS 
Accession No. ML003740176)]. A bounding risk assessment was 
performed to justify the proposed TS changes. This application of 
LCO 3.0.9 is predicated upon the licensee's performance of a risk 
assessment and the management of plant risk. The net change to the 
margin of safety is insignificant as indicated by the anticipated 
low levels of associated risk (ICCDP [incremental conditional core 
damage probability] and ICLERP [incremental large early release 
probability]) as shown in Table 1 of Section 3.1.1 in the Safety 
Evaluation [published in the Federal Register on October 3, 2006 (71 
FR 58444)].
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the above 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 Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station (PNPS), Plymouth County, Massachusetts

    Date of amendment request: July 15, 2015. A publicly available 
version is in ADAMS under Accession No. ML15205A287.
    Description of amendment request: The amendment would revise the 
PNPS Cyber Security Plan (CSP) Implementation Schedule Milestone 8 full 
implementation date. The amendment would also revise the PNPS Facility 
Operating License No. DPR-35.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the CSP Implementation Schedule is 
administrative in nature. This change does not alter accident 
analysis assumptions, add any initiators, or affect the function of 
plant systems or the manner in which systems are operated, 
maintained, modified, tested, or inspected. The proposed change does 
not require any plant modifications which affect the performance 
capability of the structures, systems, and components relied upon to 
mitigate the consequences of postulated accidents and have no impact 
on the probability or consequences of an accident previously 
evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to the CSP Implementation Schedule is 
administrative in nature. This proposed change does not alter 
accident analysis assumptions, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected. The proposed 
change does not require any plant modifications which affect the 
performance capability of the structures, systems, and components 
relied upon to mitigate the consequences of postulated accidents and 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Plant safety margins are established through limiting conditions 
for operation, limiting safety system settings, and safety limits 
specified in the technical specifications. The proposed change to 
the CSP Implementation Schedule is administrative in nature. In 
addition, the milestone date delay for full implementation of the 
CSP has no substantive impact because other measures have been taken 
which provide adequate protection during this period of time. 
Because there is no change to established safety margins as a result 
of this change, the proposed change does not involve a significant 
reduction in a margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Benjamin G. Beasley.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: August 18, 2015. A publically-available 
version is in ADAMS under Accession No. ML15231A097.
    Description of amendment request: The proposed change would revise 
the reactor steam dome pressure specified in the technical 
specification (TS) safety limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the reactor steam dome pressure in the 
CPS [Clinton Power Station, Unit 1], DNPS [Dresdent Nuclear Power 
Station, Units 2 and 3], and QCNPS [Quad Cities Nuclear Power 
Station, Units 1 and 2] Reactor Core Safety Limits TS 2.1.1.1 and 
2.1.1.2 does not alter the use of the analytical methods used to 
determine the safety limits that have been previously reviewed and 
approved by the NRC. The proposed change is in accordance with an 
NRC approved critical power correlation

[[Page 65813]]

methodology, and as such, maintains required safety margins. The 
proposed change does not adversely affect accident initiators or 
precursors, nor does it alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained.
    The proposed change does not alter or prevent the ability of 
structures, systems, and components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits. The proposed change does 
not require any physical change to any plant SSCs nor does it 
require any change in systems or plant operations. The proposed 
change is consistent with the safety analysis assumptions and 
resultant consequences.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    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 reduction in the reactor dome pressure safety limit 
from 785 psig [pounds per square inch gauge] to 685 psig is a change 
based upon previously approved documents and does not involve 
changes to the plant hardware or its operating characteristics. As a 
result, no new failure modes are being introduced. There are no 
hardware changes nor are there any changes in the method by which 
any plant systems perform a safety function. No new accident 
scenarios, failure mechanisms, or limiting single failures are 
introduced as a result of the proposed change.
    The proposed change does not introduce any new accident 
precursors, nor does it involve any physical plant alterations or 
changes in the methods governing normal plant operation. Also, the 
change does not impose any new or different requirements or 
eliminate any existing requirements. The change does not alter 
assumptions made in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through the design of the 
plant structures, systems, and components, and through the 
parameters for safe operation and setpoints for the actuation of 
equipment relied upon to respond to transients and design basis 
accidents. Evaluation of the 10 CFR part 21 condition by General 
Electric determined that since the Minimum Critical Power Ratio 
improves during the PRFO [pressure regulator failure maximum demand 
(open)] transient, there is no decrease in the safety margin and 
therefore there is not a threat to fuel cladding integrity. The 
proposed change in reactor dome pressure supports the current safety 
margin, which protects the fuel cladding integrity during a 
depressurization transient, but does not change the requirements 
governing operation or availability of safety equipment assumed to 
operate to preserve the margin of safety. The change does not alter 
the behavior of plant equipment, which remains unchanged.
    The proposed change to Reactor Core Safety Limits 2.1.1.1 and 
2.1.1.2 is consistent with and within the capabilities of the 
applicable NRC approved critical power correlation for the fuel 
designs in use at CPS, DNPS, and QCNPS. No setpoints at which 
protective actions are initiated are altered by the proposed change. 
The proposed change does not alter the manner in which the safety 
limits are determined. This change is consistent with plant design 
and does not change the TS operability requirements; thus, 
previously evaluated accidents are not affected by this proposed 
change.
    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 
requested amendments involve 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.

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

    Date of amendment request: July 2, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15198A153.
    Description of amendment request: The amendments would revise the 
technical specifications (TSs) related to communications and 
manipulator crane requirements. The licensee requested that these 
requirements be relocated to the Updated Final Safety Analysis Report 
(UFSAR) and related procedures and be controlled in accordance with 10 
CFR 50.59, ``Changes, tests, and experiments.''
    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 changes remove [from the TSs] the current necessity 
of establishing and maintaining communications between the control 
room and the refueling station and the minimum load capacities and 
load limit controls required for the manipulator crane limits and 
relocate the requirements to the UFSAR and related procedures, which 
will have no impact on any safety related structures, systems or 
components. Once relocated to the UFSAR and related procedures, 
changes to establishing and maintaining communications between the 
control room and the refueling station and the minimum load 
capacities and load limit controls required for the manipulator 
crane limits will be controlled in accordance with 10 CFR 50.59.
    The probability of occurrence of a previously evaluated accident 
is not increased because these changes do not introduce any new 
potential accident initiating conditions. The consequences of 
accidents previously evaluated in the UFSAR are not affected because 
the ability of the components to perform their required functions is 
not affected.
    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 remove [from the TSs] the current necessity 
of establishing and maintaining communications between the control 
room and the refueling station and the minimum load capacities and 
load limit controls required for the manipulator crane limits and 
relocate the requirements to the UFSAR and related procedures, which 
will have no impact on any safety related structures, systems or 
components. Once relocated to the UFSAR and related procedures, 
changes to establishing and maintaining communications between the 
control room and the refueling station and the minimum load 
capacities and load limit controls required for the manipulator 
crane limits will be controlled in accordance with 10 CFR 50.59.
    The proposed changes do not introduce new modes of plant 
operation and do not involve physical modifications to the plant (no 
new or different type of equipment will be installed). There are no 
changes in the method by which any safety related plant structure, 
system, or component (SSC) performs its specified safety function. 
As such, the plant conditions for which the design basis accident 
analyses were performed remain valid.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of the proposed change. There will be no adverse effect or 
challenges imposed on any SSC as a result of the proposed changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is related to confidence in the ability of the 
fission product barriers to

[[Page 65814]]

perform their accident mitigation functions. The proposed changes 
remove [from the TSs] the current necessity of establishing and 
maintaining communications between the control room and the 
refueling station and the minimum load capacities and load limit 
controls required for the manipulator crane limits and relocate the 
requirements to the UFSAR and related procedures, which will have no 
impact on any safety related structures, systems or components. Once 
relocated to the UFSAR and related procedures, changes to 
establishing and maintaining communications between the control room 
and the refueling station and the minimum load capacities and load 
limit controls required for the manipulator crane limits will be 
controlled in accordance with 10 CFR 50.59. The proposed changes do 
not physically alter any SSC. There will be no effect on those SSCs 
necessary to assure the accomplishment of protection functions. 
There will be no impact on the overpower limit, departure from 
nucleate boiling ratio (DNBR) limits, loss of cooling accident peak 
cladding temperature (LOCA PCT), or any other margin of safety. The 
applicable radiological dose consequence acceptance criteria will 
continue to be met.
    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 S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB, 
Juno Beach, FL 33408-0420.
    NRC Branch Chief: Shana R. Helton.

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

    Date of amendment request: August 18, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15246A445.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) Section 5.5.12, ``Primary 
Containment Leakage Rate Testing Program,'' by replacing the reference 
to the NRC Regulatory Guide 1.163, ``Performance-Based Containment 
Leak-Test Program,'' with a reference to the Nuclear Energy Institute 
(NEI) topical report NEI 94-01, Revision 3-A, ``Industry Guideline for 
Implementing Performance-Based Option of 10 CFR part 50, appendix J,'' 
and conditions and limitations specified in NEI 94-01, Revision 2-A, as 
the implementation document used by DAEC to implement the performance-
based containment leakage rate testing program.
    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 amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, ``Industry Guideline for Implementing 
Performance-Based Option of 10 CFR part 50, Appendix J,'' for 
development of the DAEC performance-based containment testing 
program. NEI 94-01 allows, based on risk and performance, an 
extension of Type A and Type C containment leak test intervals. 
Implementation of these guidelines continues to provide adequate 
assurance that during design basis accidents, the primary 
containment and its components will limit leakage rates to less than 
the values assumed in the plant safety analyses.
    The findings of the DAEC risk assessment confirm the general 
findings of previous studies that the risk impact with extending the 
containment leak rate is small. Per the guidance provided in 
Regulatory Guide 1.174, an extension of the leak test interval in 
accordance with NEI 94-01, Revision 3-A results in an estimated 
change within the very small change region.
    Since the change is implementing a performance-based containment 
testing program, the proposed amendment does not involve either a 
physical change to the plant or a change in the manner in which the 
plant is operated or controlled. The requirement for containment 
leakage rate acceptance will not be changed by this amendment. 
Therefore, the containment will continue to perform its design 
function as a barrier to fission product releases.
    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 change to implement a performance-based containment 
testing program, associated with integrated leakage rate test 
frequency, does not change the design or operation of structures, 
systems, or components of the plant.
    The proposed changes would continue to ensure containment 
integrity and would ensure operation within the bounds of existing 
accident analyses. There are no accident initiators created or 
affected by these changes. Therefore, the proposed changes will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    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.
    Margin of safety is related to confidence in the ability of the 
fission product barriers (fuel cladding, reactor coolant system, and 
primary containment) to perform their design functions during and 
following postulated accidents. The proposed change to implement a 
performance-based containment testing program, associated with 
integrated leakage rate test frequency, does not affect plant 
operations, design functions, or any analysis that verifies the 
capability of a structure, system, or component of the plant to 
perform a design function. In addition, this change does not affect 
safety limits, limiting safety system setpoints, or limiting 
conditions for operation.
    The specific requirements and conditions of the TS Primary 
Containment Leakage Rate Testing Program exist to ensure that the 
degree of containment structural integrity and leak-tightness that 
is considered in the plant safety analysis is maintained. The 
overall containment leak rate limit specified by TS is maintained. 
This ensures that the margin of safety in the plant safety analysis 
is maintained. The design, operation, testing methods and acceptance 
criteria for Type A, B, and C containment leakage tests specified in 
applicable codes and standards would continue to be met, with the 
acceptance of this proposed change, since these are not affected by 
implementation of a performance-based containment testing program.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. James Petro, P.O. Box 14000 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, Units 2 and 3, Fairfield County, 
South Carolina

    Date of amendment request: May 4, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15124A911.
    Description of amendment request: The proposed amendment and 
exemption identify portions of the licensing basis that would more 
appropriately be classified as Tier 2, specifically the Tier 2* 
information on Fire Area Figures 9A-1, 9A-2, 9A-3, 9A-4, 9A-5, and 9A-
201 in the Virgil C. Summer Nuclear Station Units 2 and

[[Page 65815]]

3 Updated Final Safety Analysis Report. With the reclassification, 
prior NRC approval would continue to be required for any safety 
significant changes to the Fire Area Figures because any revisions to 
that information would follow the Tier 2 change process provided in 10 
CFR part 52, appendix D, Section VIII.B.5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would reclassify Fire Area Figures Tier 
2* information. The proposed amendment does not modify the design, 
construction, or operation of any plant structures, systems, or 
components (SSCs), nor does it change any procedures or method of 
control for any SSCs. Because the proposed amendment does not change 
the design, construction, or operation of any SSCs, it does not 
adversely affect any design function as described in the Updated 
Final Safety Analysis Report.
    Therefore, the proposed amendment does not affect the 
probability of an accident previously evaluated. Similarly, because 
the proposed amendment does not alter the design or operation of the 
nuclear plant or any plant SSCs, the proposed amendment does not 
represent a change to the radiological effects of an accident, and 
therefore, does not involve an increase in the 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 amendment would reclassify Fire Area Figures Tier 
2* information. The proposed amendment is not a modification, 
addition to, or removal of any plant SSCs. Furthermore, the proposed 
amendment is not a change to procedures or method of control of the 
nuclear plant or any plant SSCs. The only impact of this activity is 
the reclassification of information in the Updated Final Safety 
Analysis Report.
    Because the proposed amendment only reclassifies information and 
does not change the design, construction, or operation of the 
nuclear plant or any plant operations, the 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 amendment would reclassify Fire Area Figures Tier 
2* information. The proposed amendment is not a modification, 
addition to, or removal of any plant SSCs. Furthermore, the proposed 
amendment is not a change to procedures or method of control of the 
nuclear plant or any plant SSCs. The only impact of this activity is 
the reclassification of information in the Updated Final Safety 
Analysis Report.
    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. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendment request: August 31, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15261A673.
    Description of amendment request: The proposed change would 
eliminate the current requirement to perform the Residual Heat Removal 
(RHR) autoclosure interlock Surveillance Requirement (SR) 3.4.14.2 and 
revise Action Condition C to eliminate the RHR autoclosure interlock 
from the Action Condition.
    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 two motor-operated gate valves located in each RHR System 
suction line are normally-closed to maintain the low pressure RHR 
System (design pressure of 600 psig) isolated from the high pressure 
[reactor coolant system] RCS (normal operating pressure of 2235 
psig). An [autoclosure interlock] ACI was provided to isolate the 
low pressure RHR System from the RCS when the pressure increases 
above the ACI setpoint. However, spurious ACI actuation has resulted 
in RHR System isolation and subsequent loss of decay heat removal 
capability. The removal of the ACI feature will preclude this 
inadvertent isolation, thus increasing the likelihood that RHR will 
be available to remove decay heat. The addition of a control room 
alarm to alert the operator that a suction/isolation valve(s) is not 
fully closed when the RCS pressure is above the alarm setpoint and 
administrative procedures will ensure that the RHR System will be 
isolated from the RCS, if the RCS pressure increases above the alarm 
setpoint, which will decrease the likelihood of an interfacing 
system [Loss-of-Coolant Accident] LOCA. Therefore, the performance 
of the RHR System would not be adversely affected by the ACI 
deletion and the RHR suction isolation valve alarm installation.
    The RHR ACI provides automatic closure to the RHR System suction 
isolation valves on high RCS pressure; however, rapid overpressure 
protection of the RHR System is provided by the RHR relief valves 
and not by the slow acting suction isolation valves. This RHR System 
overpressure protection is not affected by the removal of the ACI, 
this feature also serves to decrease the likelihood of an 
interfacing system LOCA. Thus, the RHR System integrity will not be 
affected by the removal of the ACI feature. In addition, the removal 
of the ACI feature does not adversely affect any fission barrier, 
alter any assumptions made in the radiological consequences 
evaluations, or affect the mitigation of radiological consequences.
    The impact of ACI removal on RHR shutdown cooling, low 
temperature overpressure protection, and interfacing system LOCA 
initiating event frequency was assessed. For each of these areas 
that were assessed, it was concluded that the removal of ACI and the 
accompanying plant changes provides a benefit to plant safety.
    With the deletion of the ACI, there is no longer any potential 
for spurious automatic closure of a RHR System suction isolation 
valve resulting in inadvertent RHR System isolation and loss of 
shutdown cooling.
    Therefore, it is concluded 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.
    The removal of the RHR System ACI, and corresponding TS 
requirements, does not result in the initiation of any accident nor 
create any new credible limiting single failures.
    The removal of the ACI eliminates the potential for spurious 
circuitry actuation causing isolation of the RHR system. 
Furthermore, the addition of an alarm to alert the operator that a 
suction valve is not fully closed when RCS pressure is above the 
alarm setpoint reduces the likelihood that the RHR system will be 
exposed to high pressure conditions. These modifications and the 
resulting elimination of the ACI TS Surveillance Requirement will 
not result in the RHR system being operated in any unanalyzed modes, 
either during normal or accident conditions. Also, the AHA system 
will continue to be maintained and surveilled as it is currently.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
The proposed change does not

[[Page 65816]]

challenge the performance or integrity of any safety-related system.
    Therefore, it is concluded that the proposed changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Removal of the ACI interlock, and its corresponding TS 
Surveillance Requirement, does not alter or prevent any plant 
response such that the margin of safety to any applicable acceptance 
criteria is significantly decreased. In fact, the addition of a 
control room alarm that identifies that the suction valve is not 
fully open, together with the existing overpressure alarm, ensures 
that the margin of safety to an AHA overpressure condition is not 
significantly reduced.
    Furthermore, the actuation of safety-related components and the 
response of plant systems to accident scenarios are not affected, 
and thus will remain as assumed in the safety analysis.
    Therefore, the proposed change will not adversely affect the 
operation or safety function of equipment assumed in the safety 
analysis.
    For the reasons noted above, it is concluded that the proposed 
change does not involves 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: Leigh D. Perry, SVP & General Counsel of 
Operations and Nuclear, Southern Nuclear Operating Company, 40 Iverness 
Center Parkway, Birmingham, AL 35201.
    NRC Branch Chief: Robert J. Pascarelli.

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: September 1, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15244A602).
    Description of amendment request: The proposed change would amend 
Combined License (COL) Nos. NPF-91 and NPF-92 for the VEGP Units 3 and 
4. The requested amendment proposes to revise the VEGP Units 3 and 4 
plant-specific emergency planning inspections, tests, analyses, and 
acceptance criteria in Appendix C of the VEGP Units 3 and 4 COLs, to 
remove the copy of Updated Final Safety Analysis Report (UFSAR) Table 
7.5-1, ``Post-Accident Monitoring System,'' from Appendix C of the VEGP 
Units 3 and 4 COLs.
    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 VEGP [Units] 3 and 4 emergency planning inspections, tests, 
analyses, and acceptance criteria (ITAAC) provide assurance that the 
facility has been constructed and will be operated in conformity 
with the license, the provisions of the Act, and the Commission's 
rules and regulations. The proposed change to remove the copy of 
UFSAR Table 7.5-1 from Appendix C of the VEGP [Units 3 and 4] COLs 
does not affect the design of a system, structure, or component 
(SSC) used to meet the design bases of the nuclear plant. Nor do the 
changes affect the construction or operation of the nuclear plant 
itself, so there is no change to the probability or consequences of 
an accident previously evaluated. Removing the copy of UFSAR Table 
7.5-1 from Appendix C of the COLs does not affect prevention and 
mitigation of abnormal events, e.g., accidents, anticipated 
operational occurrences, earthquakes, floods and turbine missiles, 
or their safety or design analyses. No safety-related SSC or 
function is adversely affected. The changes do not involve nor 
interface with any SSC accident initiator or initiating sequence of 
events, and thus, the probabilities of the accidents evaluated in 
the UFSAR are not affected. Because the changes do not involve any 
safety-related SSC or function used to mitigate an accident, the 
consequences of the accidents evaluated in the UFSAR are not 
affected.
    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 VEGP [Units] 3 and 4 emergency planning ITAAC provide 
assurance that the facility has been constructed and will be 
operated in conformity with the license, the provisions of the Act, 
and the Commission's rules and regulations. The changes do not 
affect the design of an SSC used to meet the design bases of the 
nuclear plant, nor do the changes affect the construction or 
operation of the nuclear plant. Consequently, there is no new or 
different kind of accident from any accident previously evaluated. 
The changes do not affect safety-related equipment, nor do they 
affect equipment which, if it failed, could initiate an accident or 
a failure of a fission product barrier. In addition, the changes do 
not result in a new failure mode, malfunction or sequence of events 
that could affect safety or safety-related equipment.
    No analysis is adversely affected. No system or design function 
or equipment qualification is adversely affected by the changes. 
This activity will not allow for a new fission product release path, 
result in a new fission product barrier failure mode, nor create a 
new sequence of events that would result in significant fuel 
cladding failures.
    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 VEGP [Units] 3 and 4 emergency planning ITAAC provide 
assurance that the facility has been constructed and will be 
operated in conformity with the license, the provisions of the Act, 
and the Commission's rules and regulations. The changes do not 
affect the assessments or the plant itself. The changes do not 
adversely interface with safety-related equipment or fission product 
barriers. No safety analysis, design basis limit or acceptance 
criterion are challenged or exceeded by the proposed change.
    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: Mr. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence J. Burkhart.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424, 50-425, 
52-025, 52-026, Vogtle Electric Generating Plant, Units 1, 2, 3, and 4, 
Burke County, Georgia and Southern Nuclear Operating Company, Inc. 
(SNC), Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, 
Units 1 and 2, Houston County, Alabama, Docket Nos. 50-321 and 50-366, 
Edwin I. Hatch Nuclear Plant, Units 1 and 2, City of Dalton, GA

    Date of amendment request: August 31, 2015. A publicly-available 
version is in ADAMS under Accession Package No. ML15246A045.
    Description of amendment request: The amendments request NRC 
approval of a standard emergency plan for all Southern Nuclear 
Operating Company, Inc., sites and site-specific annexes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 65817]]


    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes have no effect on normal plant operation or 
on any accident initiator or precursors, and do not impact the 
function of plant structures, systems, or components (SSCs). The 
proposed changes do not alter or prevent the ability of the 
emergency response organization to perform its intended functions to 
mitigate the consequences of an accident or event. The ability of 
the emergency response organization to respond adequately to 
radiological emergencies has been demonstrated as acceptable through 
a staffing analysis as required by 10 CFR 50 Appendix E.IV.A.9.
    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 will not change the design function or 
operation of SSCs. The changes do not impact the accident analysis. 
The changes do not involve a physical alteration of the plant, a 
change in the method of plant operation, or new operator actions. 
The proposed changes do not introduce failure modes that could 
result in a new accident, and the changes do not alter assumptions 
made in the safety analysis. As demonstrated by the SNC staffing 
analysis performed in accordance with 10 CFR 50 Appendix E.IV.A.9, 
the proposed changes do not alter or prevent the ability of the 
emergency response organization to perform its intended functions to 
mitigate the consequences of an accident or event.
    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 changes involve a significant reduction in 
a margin of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed changes are 
associated with the Emergency Plan and do not impact operation of 
the plant or its response to transients or accidents. The changes do 
not affect the Technical Specifications. The changes do not involve 
a change in the method of plant operation, and no accident analyses 
will be affected by the proposed changes. Safety analysis acceptance 
criteria are not affected. The Standard Emergency Plan will continue 
to provide the necessary response staff for emergencies as 
demonstrated by staffing and functional analyses including the 
necessary timeliness of performing major tasks for the functional 
areas of the Emergency Plan. The proposed changes do not adversely 
affect SNC's ability to meet the requirements of 10 CFR 50 Appendix 
E and the emergency planning standards of 10 CFR 50.47.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Leigh D. Perry, SVP & General Counsel of 
Operations and Nuclear, Southern Nuclear Operating Company, 40 Iverness 
Center Parkway, Birmingham, AL 35201.
    NRC Branch Chief: Robert J. Pascarelli.

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.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3 (WF3) St. Charles Parish, Louisiana

    Date of amendment request: July 2, 2015, as supplemented by letter 
dated August 14, 2015. Publicly-available versions are in ADAMS under 
Accession Nos. ML15197A106 and ML15226A346, respectively.
    Brief description of amendment: This notice is being reissued in 
its entirety to remove information that was inadvertently included in 
the notice published in the Federal Register on September 29, 2015 (80 
FR 58520), for WF3. The proposed amendment will modify the Technical 
Specification (TS) 3.1.3.4, ``Control Element Assembly [CEA] Drop 
Time'' and Final Safety Analysis Report, Chapter 15, ``Accident 
Analyses.'' The proposed amendment would change TS 3.1.3.4 to revise 
the arithmetic average of all CEA drop times to be less than or equal 
to 3.5 seconds.
    Date of publication of individual notice in the Federal Register: 
September 8, 2015 (80 FR 53892).
    Expiration date of individual notice: October 8, 2015 (public 
comments); and November 9, 2015 (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 Commissions 
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.

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

    Date of amendment request: October 21, 2014, as supplemented by 
letters dated June 18, and July 28, 2015.
    Description of amendment: The amendment revised the emergency 
action level scheme for Fermi 2 based

[[Page 65818]]

on the Nuclear Energy Institute (NEI) 99-01, Revision 6, ``Development 
of Emergency Action Levels for Non-Passive Reactors,'' dated November 
2012.
    Date of issuance: September 29, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 202. A publicly-available version is in ADAMS under 
Accession No. ML15233A084; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-43: Amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: December 23, 2014 (79 
FR 77045). The supplemental letters dated June 18, and July 28, 2015, 
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 September 29, 2015.
    No significant hazards consideration comments received: None.

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

    Date of amendment request: June 10, 2014, as supplemented by 
letters dated October 9, and December 31, 2014, and January 30, 2015.
    Brief description of amendment: By order dated August 14, 2015, as 
published in the Federal Register on August 24, 2015 (80 FR 51329), the 
NRC approved a direct license transfer for Facility Operating License 
No. NPF-47 for the River Bend Station, Unit 1. This amendment reflects 
the direct transfer of the license to Entergy Louisiana, LLC.
    Date of issuance: October 1, 2015.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 189. A publicly-available version of the amendment 
and the order are in ADAMS under Accession Nos. ML15265A116 and 
ML15146A410, respectively; documents related to this amendment are 
listed in the safety evaluation (SE) enclosed with the order dated 
August 14, 2015. Subsequent to the issuance of the order, the licensee 
submitted a letter dated September 23, 2015 (ADAMS Accession No. 
ML15268A338). This letter provided additional notifications of 
regulatory approvals and the closing transaction date, as was required 
by the order.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: August 24, 2015 (80 FR 
51329). The supplements dated October 9, and December 31, 2014, January 
30, and September 23, 2015, contained clarifying information, did not 
expand the application beyond the scope of the notice as originally 
published in the Federal Register, and did not affect the applicability 
of the generic no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in an SE dated August 14, 2015.
    Comments received: Yes. The comments received on the license 
transfer request are addressed in the SE dated August 14, 2015.

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: March 28, 2014, as supplemented by 
letters dated April 24, June 9, June 11, and August 13, 2014; and May 
4, 2015.
    Brief description of amendment: The amendment revised the renewed 
facility operating license and the associated technical specifications 
to be consistent with the permanent cessation of reactor operations and 
permanent defueling of the reactor.
    Date of issuance: October 7, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 263. A publicly-available version is in ADAMS under 
Accession No. ML15117A551; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-28: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: February 17, 2015 (80 
FR 8358). The supplemental letters dated April 24, June 9, June 11, and 
August 13, 2014; and May 4, 2015, 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 this amendment is contained 
in a Safety Evaluation dated October 7, 2015.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of amendment request: July 23, 2015, as supplemented by 
letters dated July 28, 2015, and August 25, 2015.
    Brief description of amendment: The amendment modified the 
technical specifications (TSs) to allow for the temporary operation of 
the borated water storage tank (BWST) under administrative and design 
controls while connected to seismic Class II piping. This change would 
support necessary cleanup and surveillance activities associated with 
the TMI Fall 2015 Refueling Outage and Fuel Cycle 21 operation.
    Date of issuance: October 1, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 7 days.
    Amendment No.: 289. A publicly-available version is in ADAMS under 
Accession No. ML15225A158; documents related to this amendment are 
listed in the safety evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-50: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of Initial Notice in Federal Register: August 7, 2015 (80 FR 
47529). The supplemental letter dated August 25, 2015, 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 October 1, 2015.
    No significant hazards consideration comments received: No.

Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: December 5, 2014.
    Brief description of amendments: The amendments revised Technical 
Specifications (TSs) to adopt Technical Specification Task Force 
Traveler 439, Revision 2, ``Eliminate Second

[[Page 65819]]

Completion Times Limiting Time from Discovery of Failure to Meet an LCO 
[Limiting Condition for Operation].'' The second completion times 
associated with TS 3.6.2.1, ``Containment Spray and Cooling Systems,'' 
were deleted.
    Date of Issuance: October 5, 2015.
    Effective Date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos. 228 and 178. A publicly-available version is in 
ADAMS under Accession No. ML15251A094; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the TSs.
    Date of initial notice in Federal Register: February 3, 2015 (80 FR 
5801).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 5, 2015.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 7, 2014.
    Brief description of amendments: The amendments revised the 
scheduled completion date for Milestone 8 of the Cyber Security Plan 
implementation schedule and License Condition 3.E in Renewed Facility 
Operating License Nos. DPR-31 and DPR-41.
    Date of issuance: September 28, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 266 and 261. The amendments are in ADAMS under 
Accession No. ML15233A379; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: January 6, 2015 (80 FR 
535).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 28, 2015.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 13, 2015.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs). The amendment added a note to TS Surveillance 
Requirement 4.4.1.3.4, which requires verification that residual heat 
removal loop operations susceptible to gas accumulation are 
sufficiently filled with water in accordance with the Surveillance 
Frequency Control Program.
    Date of issuance: October 6, 2015.
    Effective date: As of its date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 150. A publicly-available version is in ADAMS under 
Accession No. ML15231A144; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-86: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 4, 2015 (80 FR 
46350).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 6, 2015.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant (DCPP), Unit Nos. 1 and 2, San Luis Obispo 
County, California

    Date of amendment request: October 17, 2014, as supplemented by 
letter dated February 19, 2015.
    Brief description of amendments: The amendments revised the DCPP 
Cyber Security Plan (CSP) Milestone h full implementation schedule as 
set forth in the CSP implementation schedule.
    Date of issuance: September 30, 2015.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance. All subsequent changes to the 
NRC-approved CSP implementation schedule as approved by the NRC staff 
with this license amendment will require prior NRC approval pursuant to 
10 CFR 50.90.
    Amendment Nos.: Unit 1--220; Unit 2--222. A publicly-available 
version is in ADAMS under Accession No. ML15245A542; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: April 7, 2015 (80 FR 
18659). The supplemental letter dated February 19, 2015, 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 September 30, 2015.
    No significant hazards consideration comments received: No.

South Carolina Electric & 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: May 26, 2015, as supplemented by letter 
dated May 28, 2015 and as revised by letters dated June 9, and June 29, 
2015.
    Description of amendment: The amendment authorized changes to the 
VCSNS Units 2 and 3 Updated Final Safety Analysis Report on the 
applicability of the American Institute of Steel Construction (AISC) 
N690-1994, ``Specification for the Design, Fabrication and Erection of 
Steel Safety-Related Structures for Nuclear Facilities,'' to allow use 
of the American Welding Society (AWS) D1.1-2000, ``Structural Welding 
Code-Steel,'' in lieu of the AWS D1.1-1992 edition identified in AISC 
N690-1994.
    Date of issuance: September 1, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 30. A publicly-available version is in ADAMS under 
Accession No. ML15224A750; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: June 8, 2015 (80 FR 
32413). However, the June 29, 2015, letter revised the application 
including the No Significant Hazard Determination. Therefore, the staff 
issued a revised notice on July 9, 2015, (80 FR 39450).
    The Commission's related evaluation of the amendment is contained 
in the Safety Evaluation dated September 1, 2015.
    No significant hazards consideration comments received: Yes. The 
comments were addressed in the Safety Evaluation.

[[Page 65820]]

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

    Date of amendment request: November 12, 2014, as supplemented by 
letter dated August 27, 2015.
    Brief description of amendments: The amendments revised the 
completion date for Milestone 8, full implementation, of the Cyber 
Security Plan from December 31, 2015, to December 31, 2017.
    Date of issuance: October 1, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: Unit 2-231; Unit 3-224. A publicly-available 
version is in ADAMS under Accession No. ML15209A935; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: April 7, 2015 (80 FR 
18659). The supplemental letter dated August 27, 2015, 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 1, 2015.
    No significant hazards consideration comments received: No.

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: May 26, 2015, as supplemented by letter 
dated May 28, 2015, and as revised by letters dated June 9, and June 
29, 2015.
    Brief description of amendment: The license amendment revised the 
Combined Licenses (COLs) by revising the VEGP Units 3 and 4 Updated 
Final Safety Analysis Report on the applicability of the American 
Institute of Steel Construction (AISC) N690-1994, ``Specification for 
the Design, Fabrication and Erection of Steel Safety-Related Structures 
for Nuclear Facilities,'' to allow use of a newer version of the 
American Welding Society (AWS) D1.1-200, ``Structural Welding Code-
Steel,'' in lieu of the AWS D1.1-1992 edition identified in AISC N690-
1994. The use of AWS D1.1-2000 applies to future and installed 
structural welding.
    Date of issuance: August 31, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 37. A publicly-available version is in ADAMS under 
Accession No. ML15215A288; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: June 9, 2015 (80 FR 
32624). A revised notice was issued on July 9, 2015 (80 FR 39454) as 
the June 29, 2015, letter revised the scope of the amendment request 
and the licensee revised the original no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 31, 2015.
    No significant hazards consideration comments received: Yes. The 
comments were addressed in the Safety Evaluation.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of amendment request: May 19, 2015.
    Brief description of amendments: The amendments revised the minimum 
indicated nitrogen cover pressure required per the Vogtle Electric 
Generating Plant Technical Specifications (TS) Surveillance Requirement 
3.5.1.3 from the current requirement of 626 pounds per square inch 
gauge (psig) back to the previous requirement of 617 psig based on 
installation of updated instrumentation.
    Date of issuance: October 5, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 177 and 158. A publicly-available version is in 
ADAMS under Accession No. ML15222A753; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. NPF-68 and NPF-81: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: July 21, 2015 (80 FR 
43129).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 5, 2015.
    No significant hazards consideration comments received: No.

Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna 
Steam Electric Station, Units 1 and 2 (SSES-1 and 2), Luzerne County, 
Pennsylvania

    Date of amendment request: August 11, 2014, as supplemented by 
letters dated April 6, 2015, and July 16, 2015.
    Brief description of amendments: The amendments changed SSES-1 and 
2, Technical Specification (TS) 3.4.10, ``RCS [Reactor Coolant System] 
Pressure and Temperature (P/T) Limits,'' specifically revising the P/T 
Limits curves. The revision provides P/T Limits curves that extend into 
the vacuum region (e.g., below 0 pounds per square inch gauge) to 
mitigate the risk of a level transient during startup, account for 
updated surveillance material and fluence data for the reactor vessel 
beltline materials, and replace the current 35.7 and 30.2 effective 
full power year (EFPY) P/T Limits curves for SSES-1 and 2, 
respectively, with new curves that are valid for 40 EFPY. This license 
amendment request was submitted by PPL Susquehanna, LLC; however, on 
June 1, 2015, the NRC staff issued an amendment changing the name on 
the SSES license from PPL Susquehanna, LLC to Susquehanna Nuclear, LLC 
(ADAMS Accession No. ML15054A066). These amendments were issued 
subsequent to an order issued on April 10, 2015, to SSES, approving an 
indirect license transfer of the SSES license to Talen Energy 
Corporation (ADAMS Accession No. ML15058A073).
    Date of issuance: September 30, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 263 (Unit 1) and 244 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML15243A140; documents related 
to these amendments are listed in the safety evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-14 and NPF-22: Amendments 
revised the Facility Operating License and TSs.
    Date of initial notice in Federal Register: November 25, 2014 (79 
FR 70217). The supplemental letters dated April 6, 2015, and July 16, 
2015, provided additional information that clarified the application, 
expanded the scope of the application as originally noticed, and 
changed the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register. As 
such, the staff published a subsequent

[[Page 65821]]

notice in the Federal Register on July 30, 2015 (80 FR 45559).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 30, 2015.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear 
Plant, Unit 1, Limestone County, Alabama

    Date of amendment request: March 9, 2015, as supplemented by letter 
dated August 19, 2015.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.1.4, ``Control Rod Scram Times,'' based on 
Technical Specification Task Force Change Traveler-460, Revision 0, 
``Control Rod Scram Time Testing Frequency,'' revising the frequency of 
Surveillance Requirement 3.1.4.2 regarding control rod scram time 
testing from ``120 days cumulative operation in MODE 1'' to ``200 days 
cumulative operation in MODE 1.'' Implementation of this amendment will 
also include incorporation of the revised acceptance criterion value of 
7.5 percent for ``slow'' control rods into the TS Bases.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 289. A publicly-available version is in ADAMS under 
Accession No. ML15251A540; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-33: Amendment revised 
the Facility Operating License and TSs.
    Date of initial notice in Federal Register: June 9, 2015 (80 FR 
32629). The supplemental letter dated August 19, 2015, 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 September 29, 2015.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: November 22, 2013, as supplemented by 
letters dated December 16, 2014; June 19, 2015; July 24, 2015; August 
5, 2015; and August 31, 2015.
    Brief description of amendments: The amendments converted the 
current technical specifications to the improved technical 
specifications (ITSs) and relocate certain requirements to other 
licensee-controlled documents. The ITSs are based on NUREG-1431, Rev. 
3.0, ``Standard Technical Specifications, Westinghouse Plants,'' Rev. 
3.0; ``NRC Final Policy Statement on Technical Specification 
Improvements for Nuclear Power Reactors,'' dated July 22, 1993 (58 FR 
39132); and 10 CFR 50.36, ``Technical Specifications.'' Technical 
Specification Task Force changes were also incorporated. The purpose of 
the conversion is to provide clearer and more readily understandable 
requirements in the technical specifications for SQN to ensure safe 
operation. In addition, the amendments include a number of issues that 
were considered beyond the scope of NUREG-1431.
    Date of issuance: September 30, 2015.
    Effective date: As of its date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 334--Unit 1 and 327--Unit 2. A publicly-available 
version is in ADAMS under Accession No. ML15238B499; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. DPR-77 and DPR-79. The amendments 
revised the TSs.
    Date of initial notice in Federal Register: June 24, 2014 (79 FR 
35807). The supplemental letters dated December 16, 2014, June 19, July 
24, August 5, and August 31, 2015, 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 September 30, 2015.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 1, 2013, as supplemented by 
letters dated April 21, 2014, January 29, 2015, and June 12, 2015.
    Brief description of amendment: The amendment revised the Limiting 
Condition for Operation for the Alternating Current Sources--Operating 
in Technical Specification 3.8.1 to provide additional time to restore 
an inoperable offsite circuit, modify Surveillance Requirements, and 
modify the current licensing basis, as described in the Updated Final 
Safety Analysis Report for the available maintenance feeder for the 
Common Station Service Transformers A and B.
    Date of issuance: September 29, 2015.
    Effective date: As of the date of issuance and shall be implemented 
after the issuance of the Facility Operating License for Unit 2.
    Amendment No.: 103. A publicly-available version is in ADAMS under 
Accession No. ML15225A094; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-90: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 29, 2013 (78 FR 
64547). The supplemental letters dated April 21, 2014, January 29 and 
June 12, 2015, provided additional information that expanded the scope 
of the application as originally noticed. A notice published in the 
Federal Register on August 28, 2015, supersedes the original notice in 
its entirety to update the expanded scope of the amendment description 
and include the staff's proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 29, 2015.
    No significant hazards consideration determination comments 
received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: October 2, 2014, as supplemented 
by letters dated July 6, July 16, and August 31, 2015.
    Brief description of amendment: The amendment adopted the NRC-
endorsed Nuclear Energy Institute (NEI) 99-01, Revision 6, 
``Methodology for the Development of Emergency Action Levels for Non-
Passive Reactors.''
    Date of issuance: October 7, 2015.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 212. A publicly-available version is in ADAMS under 
Accession No. ML15251A493; documents related to this amendment

[[Page 65822]]

are listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-30: The amendment 
revised the Emergency Action Level Technical Bases Document.
    Date of initial notice in Federal Register: February 3, 2015 (80 FR 
5813). The supplemental letters dated July 6, July 16, and August 31, 
2015, 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 October 7, 2015.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 19th day of October 2015.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2015-27042 Filed 10-26-15; 8:45 am]
BILLING CODE 7590-01-P