Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 21593-21607 [2016-08323]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 81, Number 70 (Tuesday, April 12, 2016)]
[Notices]
[Pages 21593-21607]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-08323]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[NRC-2016-0073]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

-----------------------------------------------------------------------

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

[[Page 21594]]

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 March 15, 2016, to March 28, 2016. The last 
biweekly notice was published on March 29, 2016.

DATES: Comments must be filed by May 12, 2016. A request for a hearing 
must be filed by June 13, 2016.

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-2016-0073. 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: Lynn Ronewicz, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-1927, email: Lynn.Ronewicz@nrc.gov.

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2016-0073 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-2016-0073.
     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 of this document.
     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-2016-0073, 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 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; (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

[[Page 21595]]

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) through (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 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 June 
13, 2016. 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 
June 13, 2016.

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.

[[Page 21596]]

    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at https://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, 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) through (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 Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick 
Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County, North 
Carolina; Docket No. 50-261, H. B. Robinson Steam Electric Plant (RNP) 
Unit No. 2, Darlington County, South Carolina; and Docket No. 50-400, 
Shearon Harris Nuclear Power Plant (HNP), Unit 1, Wake and Chatham 
Counties, North Carolina

    Date of amendment request: February 1, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16040A077.
    Description of amendment request: The amendments would change the 
licensee's name from Duke Energy Progress, Inc. to Duke Energy 
Progress, LLC.
    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 do not involve a significant increase in 
the probability of any accident previously evaluated because no 
accident initiators or assumptions are affected. The proposed 
conversion and name change is administrative in nature and has no 
direct effect on any plant system, plant

[[Page 21597]]

personnel qualifications, or the operation and maintenance of BSEP, 
RNP, and HNP.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated because the 
proposed name change is administrative in nature and does not 
involve new failure mechanisms, malfunctions, or accident 
initiators. The proposed changes have no direct effect on any plant 
system, plant personnel qualifications, or operation and maintenance 
of BSEP, RNP, and HNP.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes will not involve a significant reduction in 
the margin of safety because the proposed changes do not involve 
changes to the initial conditions contributing to accident severity 
or consequences, or reduce response or mitigation capabilities. The 
proposed name change is administrative in nature and has no direct 
effect on any plant system, plant personnel qualifications, or 
operation and maintenance of BSEP, RNP, and HNP.

    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 St., M/C DEC45A, Charlotte, NC 
28202.
    NRC Branch Chief: Benjamin G. Beasley.

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

    Date of amendment request: October 29, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15307A293.
    Description of amendment request: The amendment proposes to modify 
Technical Specification (TS) 5.5.13, ``Primary Containment Leakage Rate 
Testing Program,'' by incorporating Nuclear Energy Institute (NEI) 
topical report 94-01, Revision 3-A, as the implementation document for 
the RBS performance-based containment leakage rate testing program. 
Based on the guidance in NEI 94-01, Revision 3-A, the proposed change 
would allow the RBS Type A Test (Integrated Leak Rate Test) frequency 
to be extended from 10 to 15 years, and the Type C Tests (Local Leak 
Rate Tests) frequency to be extended from 60 to 75 months. 
Additionally, the amendment proposes to modify Surveillance Requirement 
(SR) 3.6.5.1.3 to extend the frequency of the Drywell Bypass Test from 
10 to 15 years and to revise its allowed extension per SR 3.0.2 from 12 
to 9 months.
    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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review 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 incorporates NEI topical report 94-01, 
Revision 3-A, into TS 5.5.13 as the basis for the RBS containment 
leakage rate testing program, which would allow for extensions to 
the frequencies of the Type A and Type C Tests. The proposed 
amendment also requests an extension to the Drywell Bypass Test 
frequency. The proposed changes do not involve any physical changes 
to the plant or any changes in the normal operation or control of 
the plant. In its license amendment request, the licensee identified 
the loss-of-coolant accident (LOCA) inside containment and the fuel 
handling accident (FHA) as the previously evaluated accidents in the 
Updated Safety Analysis Report that could potentially be impacted by 
the change. Changing the frequency of containment leakage rate 
testing has no impact upon the likelihood of a LOCA or of an FHA. 
Therefore, the probability of occurrence of an accident previously 
evaluated is not significantly increased by the proposed amendment.
    The guidelines in NEI 94-01, Revision 3-A, provide a framework 
for a licensee's containment leakage rate testing program, the 
purpose of which is to ensure that the primary containment limits 
the uncontrolled release of radioactivity to the environment during 
a design-basis accident. As part of its amendment request, the 
licensee evaluated the potential consequences of extending the test 
intervals and determined that the change in risk was estimated to be 
acceptably small and within the guidelines, as published in 
Regulatory Guide 1.174. The proposed amendment does not change the 
overall containment leakage rate limit specified by the TSs. 
Therefore, it is concluded that the proposed amendment does not 
significantly increase the consequences of an accident previously 
evaluated.
    Based on the above discussion, the proposed amendment does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve any physical changes to the 
plant or any changes in the normal operation or control of the 
plant. The proposed changes do not create any new accident 
precursors or initiators, and do not change any existing accident 
precursors or initiators, as described in the RBS safety analyses.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, for the development of the RBS performance-
based leakage rate testing program, to allow for frequency 
extensions for the Type A and Type C Tests. The proposed amendment 
also requests an extension to the Drywell Bypass Test frequency. The 
proposed changes do not alter the manner in which safety limits, 
limiting safety system setpoints, or limiting conditions for 
operation are determined. The specific requirements and conditions 
of the containment leakage rate testing program, as defined in the 
TSs, ensure that the primary containment will continue to provide a 
leaktight barrier to the uncontrolled release of radioactivity to 
the environment during a design-basis accident. The proposed 
amendment does not change the overall containment leakage rate limit 
specified by the TSs. Additionally, the proposed amendment does not 
include any changes to the Containment Inservice Inspection Plan at 
RBS, which serves to provide a high degree of assurance that the 
containment will not degrade in a manner that is not detectable by 
the Type A Test.
    Based on the above discussion, 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 
its review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, LA 70113.
    NRC Branch Chief: Meena K. Khanna.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station (PNPS), Plymouth County, Massachusetts

    Date of amendment request: January 14, 2016. A publicly available 
version is in ADAMS under Accession No. ML16021A459.
    Description of amendment request: The amendment would revise the 
PNPS Emergency Plan to decrease the Emergency Response Organization 
(ERO) staff training requirements

[[Page 21598]]

identified for the ``on-site'' Chemistry Technician.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, along with NRC edits in square 
brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed training requirements change has no effect on 
normal plant operation or on any accident initiator. The change 
affects the response to radiological emergencies addressed in the 
SEP [site emergency plan]. The ability of the emergency response 
organization to respond adequately to radiological emergencies has 
been evaluated. Changes in the training provided to the on-shift 
organization, such as the reassignment of key on-shift emergency 
personnel to perform related RP [radiation protection] functions, 
provide assurance of an effective emergency response without 
competing or conflicting duties. An analysis was also performed on 
the effect of the proposed change on the timeliness of performing 
major tasks for the major functional areas of the SEP. The analysis 
concluded that the reduction in training requirements for the ``on-
shift'' Chemistry Technician to support the initial RP support tasks 
does not affect the ability to perform the required RP Technician or 
Chemistry Technician tasks.
    Therefore, the change in ERO staff training does not increase 
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 affects the training requirements for the 
``on-shift'' Chemistry Technician and for supplementing onsite 
personnel in response to a radiological emergency. It has been 
evaluated and determined not to significantly affect the ability to 
perform required or related functions. It has no effect on the plant 
design or on the normal operation of the plant and does not affect 
how the plant is physically operated under emergency conditions. The 
reduction in ERO training requirements for the ``on shift'' 
Chemistry Technician in the SEP does not affect the plant operating 
procedures which are performed by plant staff during all plant 
conditions.
    No new or different accidents are postulated to occur and there 
are no changes in any of the accidents previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not affect plant design or method of 
operation. 10 CFR 50.47(b) and 10 CFR 50 Appendix E establish 
emergency planning standards and requirements that require adequate 
staffing, satisfactory performance of key functional areas and 
critical tasks, and timely augmentation of the response capability. 
Since the SEP was originally developed, there have been improvements 
in the technology used to support the SEP functions and in the 
capabilities of onsite personnel. A functional analysis was 
performed on the effect of the proposed change on the timeliness of 
performing major tasks for the functional areas of the SEP. The 
analysis concluded that a reduction in training requirements for the 
``on-shift'' Chemistry Technician would not significantly affect the 
ability to perform the required SEP tasks. Thus, the proposed change 
has been determined not to adversely affect the ability to meet the 
emergency planning standards as described in 10 CFR 50.47(b) and 
requirements in 10 CFR 50 Appendix E.
    The proposed ERO staff training change does not involve a 
reduction in any margin of safety. The proposed change is consistent 
with the original and current ERO staffing levels implemented at 
PNPS.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Travis L. Tate.

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

    Date of amendment request: February 4, 2016. A publicly available 
version is in ADAMS under Accession No. ML16035A227.
    Description of amendment request: The amendments would add 
Surveillance Requirement (SR) 3.5.2.10 to the list of applicable SRs 
shown in SR 3.5.3.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff corrections 
shown in [brackets]:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The proposed LAR [license amendment request] is purely an 
administrative change; therefore, the probability of any accident 
previously evaluated is not significantly increased. The systems and 
components required by the TS [technical specifications] for which 
SR 3.5.2.10 is applicable, continue to be operable and capable of 
performing any mitigation function assumed in the accident analysis. 
As a result, the consequences of any accident previously evaluated 
are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an[y] 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 LAR is purely an administrative change. The 
proposed change to add SR 3.5.2.10 to the list of applicable 
surveillances in SR 3.5.3.1 does not create a new or different kind 
of accident [than] previously evaluated.
    The change does not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. In addition, 
the change does not impose any new or different requirements. The 
change does not alter assumptions made in the safety analysis. The 
proposed change is consistent with the safety analysis assumptions 
and current plant operating practice.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed LAR is purely an administrative change to add SR 
3.5.2.10 to the list of applicable surveillances in SR 3.5.3.1.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the Final Safety Analysis Report and Bases to TS). 
Similarly, there is no impact to safety analysis acceptance criteria 
as described in the plant licensing basis.
    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: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL 
60555.
    NRC Branch Chief: Travis L. Tate.

[[Page 21599]]

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

    Date of amendment request: January 29, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16029A418.
    Description of amendment request: The amendment would revise the 
post-loss-of-coolant-accident (post-LOCA) drawdown time for secondary 
containment from 12 to 19 minutes as described in the CPS Updated 
Safety Analysis Report and technical specification bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change results in additional heat added to 
Secondary Containment and the resultant increase in the time to 
achieve and maintain the required negative pressure in Secondary 
Containment following a LOCA. Neither the additional heat load from 
DCS [dry-cask storage] activities, nor the resultant increase in the 
time to achieve and maintain the required negative pressure in 
Secondary Containment affect any initiator or precursor of any 
accident previously evaluated. Therefore, the proposed change does 
not involve a significant increase in the probability of an accident 
previously evaluated.
    The proposed change results in an increase in the post-LOCA 
radiological dose to a Control Room occupant. However, the resultant 
post-LOCA Control Room dose remains within the regulatory limits of 
10 CFR 50.67 and GDC [General Design Criterion] 19. Therefore, the 
proposed change does not involve a significant increase in the 
consequences of an accident previously evaluated.
    In summary, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not alter the design function or 
operation of Secondary Containment or the Standby Gas Treatment 
system [SGTS], or the ability of each to perform its design 
function. EGC has evaluated the post-LOCA pressure response of 
Secondary Containment assuming the higher heat load, utilizing the 
design basis short-term pressure response analysis. The results of 
this analysis validated that SGTS will achieve and maintain the 
required negative pressure in Secondary Containment within the 
specified timeframe. The proposed change does not alter the safety 
limits, or safety analysis associated with the operation of the 
plant. Accordingly, the change does not introduce any new accident 
initiators. Rather, this proposed change is the result of an 
evaluation of the Control Room doses following the most limiting 
LOCA that can occur at CPS. The proposed change does not introduce 
any new modes of plant operation. As a result, no new failure modes 
are introduced.
    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 revised post-LOCA dose consequences to a Control Room 
occupant were calculated in accordance with the requirements of 10 
CFR 50.67, Regulatory Guide 1.183, and SRP [Standard Review Plan] 
15.0.1 and are consistent with the post-LOCA dose calculations 
approved by the NRC in Amendment No. 167 to the CPS Facility 
Operating License NPF-62.
    The margin of safety is considered to be that provided by 
meeting the applicable regulatory limits. The additional heat load 
that is added to Secondary Containment during DCS activities, 
leading to an increase in Secondary Containment drawdown time 
results in an increase in Control Room dose following the LOCA 
design basis accident. However, since the Control Room dose 
following the design basis accident remains within the regulatory 
limits, there is not a significant reduction in a margin of safety.
    Therefore, operation of CPS in accordance with the proposed 
change will 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: Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    Acting NRC Branch Chief: Justin C. Poole.

FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-
346, Davis-Besse Nuclear Power Station (DBNPS), Unit No. 1, Ottawa 
County, Ohio

    Date of amendment request: December 16, 2015, as supplemented by 
letters dated February 2 and March 7, 2016. Publicly-available versions 
are in ADAMS under Accession Nos. ML15350A314, ML16033A085, and 
ML16067A195.
    Description of amendment request: The amendment would allow the 
licensee to transition the current fire protection program at DBNPS to 
a performance-based, risk-informed fire protection program consistent 
with 10 CFR, Section 50.48(c), ``National Fire Protection Association 
Standard NFPA 805.'' The 2001 Edition of NFPA 805, ``Performance-Based 
Standard for Fire Protection for Light Water Reactor Electric 
Generating Plants,'' is incorporated by reference into 10 CFR 50.48(c), 
with exceptions, modifications, and supplementation. The amendment 
would also allow the licensee to make changes to the DBNPS fire 
protection program without prior NRC approval, provided that specified 
conditions are met. The proposed amendment would change the facility 
operating license, technical specifications, and design basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Operation of DBNPS in accordance with the proposed amendment 
does not increase the probability or consequences of accidents 
previously evaluated. The Updated Final Safety Analysis Report 
(UFSAR) documents the analyses of design basis accidents (DBAs) at 
DBNPS. The proposed amendment does not affect accident initiators, 
nor does it alter design assumptions, conditions, or configurations 
of the facility that would increase the probability of accidents 
previously evaluated. Further, the changes to be made for fire 
hazard protection and mitigation do not adversely affect the ability 
of SSCs [structures, systems, and components] to perform their 
design functions for accident mitigation, nor do they affect the 
postulated initiators or assumed failure modes for accidents 
described and evaluated in the UFSAR. SSCs required to shut down the 
reactor safely and to maintain it in a safe and stable condition 
will remain capable of performing their design functions.
    The purpose of the proposed amendment is to permit DBNPS to 
adopt a new fire protection licensing basis, which complies with the 
requirements of 10 CFR 50.48(a) and 10 CFR 50.48(c) and the guidance 
in [Regulatory Guide] RG 1.205, Revision 1. The NRC considers that 
NFPA 805 provides an acceptable methodology and performance criteria 
for licensees to identify fire protection requirements that are an 
acceptable alternative to the 10 CFR 50, Appendix R required fire 
protection features (69 FR 33536, June 16, 2004). Engineering 
analyses, which may include engineering evaluations, probabilistic 
safety assessments, and fire modeling calculations, have been 
performed to demonstrate that the

[[Page 21600]]

performance-based requirements of NFPA 805 have been satisfied.
    NFPA 805, taken as a whole, provides an acceptable alternative 
for satisfying General Design Criterion 3 (GDC 3) of Appendix A to 
10 CFR 50, meets the underlying intent of the NRC's existing fire 
protection regulations and guidance, and provides for DID [defense-
in-depth]. The goals, performance objectives, and performance 
criteria specified in Chapter 1 of the standard ensure that, if 
there are any increases in CDF [core damage frequency] or risk, the 
increase will be small and consistent with the intent of the 
Commission's Safety Goal Policy.
    Based on this, the implementation of the proposed amendment does 
not increase the probability of any accident previously evaluated. 
Equipment required to mitigate an accident remains capable of 
performing the assumed function(s). The proposed amendment will not 
affect the source term, containment isolation, or radiological 
release assumptions used in evaluating the radiological consequences 
of any accident previously evaluated. The applicable radiological 
dose criteria will continue to be met. Therefore, the consequences 
of any accident previously evaluated are not significantly increased 
with the implementation of the proposed amendment.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Operation of DBNPS in accordance with the proposed amendment 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated. The proposed change 
does not alter the requirements or functions for systems required 
during accident conditions. Implementation of the new fire 
protection licensing basis that complies with the requirements of 10 
CFR 50.48(a) and 10 CFR 50.58(c) and the guidance in RG 1.205, 
Revision 1, will not result in new or different accidents.
    The proposed amendment does not adversely affect accident 
initiators or alter design assumptions, conditions, or 
configurations of the facility. The proposed amendment does not 
adversely affect the ability of SSCs to perform their design 
function. SSCs required to maintain the plant in a safe and stable 
condition remain capable of performing their design functions.
    The proposed amendment does not introduce new or different 
accident initiators, nor does it alter design assumptions, 
conditions, or configurations of the facility. The proposed 
amendment does not adversely affect the ability of SSCs to perform 
their design function. SSCs required to safely shutdown the reactor 
and maintain it in a safe and stable condition remain capable of 
performing their design functions.
    The purpose of the proposed amendment is to permit DBNPS to 
adopt a new fire protection licensing basis that complies with the 
requirements of 10 CFR 50.48(a) and 10 CFR 50.48(c) and the guidance 
in Regulatory Guide 1.205, Revision 1. The NRC considers that NFPA 
805 provides an acceptable methodology and appropriate performance 
criteria for licensees to identify fire protection systems and 
features that are an acceptable alternative to the 10 CFR 50, 
Appendix R required fire protection features (69 FR [Federal 
Register] 33536, June 16, 2004).
    The requirements of NFPA 805 address only fire protection and 
the impacts of fire on the plant that have previously been 
evaluated. Based on this, implementation of the proposed amendment 
would not create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated. No new 
accident scenarios, transient precursors, failure mechanisms, or 
limiting single failures will be introduced as a result of this 
amendment. There will be no adverse effect or challenges imposed on 
any safety-related system as a result of this amendment. Therefore, 
the possibility of a new or different kind of accident from any kind 
of accident previously evaluated is not created with the 
implementation of this amendment.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Operation of DBNPS in accordance with the proposed amendment 
does not involve a significant reduction in the margin of safety. 
The proposed amendment does not alter the manner in that safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not affected by this change. The proposed amendment does not 
adversely affect existing plant safety margins or the reliability of 
equipment assumed to mitigate accidents in the UFSAR. The proposed 
amendment does not adversely affect the ability of SSCs to perform 
their design function. SSCs required to safely shut down the reactor 
and to maintain it in a safe and stable condition remain capable of 
performing their design functions.
    The purpose of the proposed amendment is to permit FENOC to 
adopt a new fire protection licensing basis which complies with the 
requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c) and the guidance 
in RG 1.205, Revision 1. The NRC considers that NFPA 805 provides an 
acceptable methodology and performance criteria for licensees to 
identify fire protection systems and features that are an acceptable 
alternative to the 10 CFR 50 Appendix R required fire protection 
features (69 FR 33536, June 16, 2004). Engineering analyses, which 
may include engineering evaluations, probabilistic safety 
assessments, and fire modeling calculations, have been performed to 
demonstrate that the performance-based requirements of NFPA 805 do 
not result in a significant reduction in the margin of safety.
    The proposed changes are evaluated to ensure that risk and 
safety margins are kept within acceptable limits. Therefore, the 
transition to NFPA 805 does not involve a significant reduction in 
the margin of safety.

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

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

    Date of amendment request: March 11, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16076A433.
    Description of amendment request: The amendment would adopt 
Technical Specification (TS) Task Force (TSTF) Change Traveler TSTF-
535, Revision 0, ``Revise Shutdown Margin [SDM] Definition to Address 
Advanced Fuel Designs.'' The SDM (i.e., the amount of reactivity by 
which the reactor is subcritical), is calculated under the conservative 
conditions that the reactor is Xenon free, the most reactive control 
rod is outside the reactor core, and the moderator temperature produces 
the maximum reactivity. For standard fuel designs, maximum reactivity 
occurs at a moderator temperature of 68 degrees Fahrenheit ([deg]F), 
which is reflected in the temperature specified in the TSs. New, 
advanced boiling water reactor fuel designs can have a higher 
reactivity at moderator shutdown temperatures above 68[emsp14][deg]F. 
Therefore, the proposed amendment, consistent with TSTF-535, Revision 
0, seeks to modify the TSs to require the SDM to be calculated at 
whatever temperature produces the maximum reactivity (i.e., 
temperatures at or above 68[emsp14][deg]F). The availability of this TS 
improvement was announced in the Federal Register (FR) published on 
February 26, 2013 (78 FR 13100), as part of the Consolidated Line Item 
Improvement Process, and has been requested with no variations or 
deviations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the definition of SDM. SDM is not an 
initiator to any accident previously evaluated. Accordingly, the 
proposed change to the definition of SDM has no effect on the 
probability of any

[[Page 21601]]

accident previously evaluated. SDM is an assumption in the analysis 
of some previously evaluated accidents and inadequate SDM could lead 
to an increase in consequences for those accidents. However, the 
proposed change revises the SDM definition to ensure that the 
correct SDM is determined for all fuel types at all times during the 
fuel cycle. As a result, the proposed change does not adversely 
affect the consequences of any accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the definition of SDM. The change 
does not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operations. The change does not alter 
assumptions made in the safety analysis regarding SDM.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the definition of SDM. The proposed 
change does not alter the manner in which safety limits, limiting 
safety system settings or limiting conditions for operation are 
determined. The proposed change ensures that the SDM assumed in 
determining safety limits, limiting safety system settings or 
limiting conditions for operation is correct for all Boiling Water 
Reactor fuel types at all times during the fuel cycle.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, P.O. Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: Meena K. Khanna.

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

    Date of amendment request: December 16, 2015, as supplemented by 
letter dated March 7, 2016. Publicly-available versions are in ADAMS 
under Accession Nos. ML15356A048 and ML16069A021, respectively.
    Description of amendment request: The licensee proposes to revise 
TS 3/4.3.1, ``Reactor Trip System Instrumentation,'' and TS 3/4.3.2, 
``Engineered Safety Feature Actuation System Instrumentation,'' to 
implement the Allowed Outage Time, Bypass Test Time, and Surveillance 
Frequency changes approved by the NRC in WCAP-15376-P-A, Rev. 1, 
``Risk-Informed Assessment of the Reactor Trip System (RTS) and 
Engineered Safety Features Actuation System (ESFAS) Surveillance Test 
Intervals and Reactor Trip Breaker Test and Completion Times.''
    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 overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed. The same reactor trip system (RTS) 
and engineered safety feature actuation system (ESFAS) 
instrumentation will continue to be used. The protection systems 
will continue to function in a manner consistent with the plant 
design basis. These changes to the Technical Specifications do not 
result in a condition where the design, material, and construction 
standards that were applicable prior to the change are altered.
    The proposed changes will not modify any system interfaces. The 
proposed changes will not affect the probability of any event 
initiators. There will be no degradation in the performance of or an 
increase in the number of challenges imposed on safety-related 
equipment assumed to function during an accident situation. There 
will be no change to normal plant operating parameters or accident 
mitigation performance. The proposed changes will not alter any 
assumptions or change any mitigation actions in the radiological 
consequence evaluations in the Final Safety Analysis Report (FSAR).
    The determination that the results of the proposed changes are 
acceptable was established in the NRC Safety Evaluation prepared for 
WCAP-1 5376-P-A (issued by letter dated December 20, 2002 
[ML023540534]). Implementation of the proposed changes will result 
in an insignificant risk impact. Applicability of these conclusions 
has been verified through plant-specific reviews and implementation 
of the generic analysis results in accordance with the NRC Safety 
Evaluation conditions.
    The proposed changes to the Completion Times, bypass test times, 
and Surveillance Frequencies reduce the potential for inadvertent 
reactor trips and spurious engineered safety feature (ESF) 
actuations, and therefore do not increase the probability of any 
accident previously evaluated. The proposed changes do not change 
the response of the plant to any accidents and have an insignificant 
impact on the reliability of the RTS and ESFAS signals. The RTS and 
ESFAS instrumentation will remain highly reliable and the proposed 
changes will not result in a significant increase in the risk of 
plant operation. This is demonstrated by showing that the impact on 
plant safety as measured by the increase in core damage frequency 
(CDF) is less than 1.0E-06 per year and the increase in large early 
release frequency (LERF) is less than 1.0E-07 per year. In addition, 
for the Completion Time changes, the incremental conditional core 
damage probabilities (ICCDP) and incremental conditional large early 
release probabilities (ICLERP) are less than 5.0E-07 and 5.0E-08, 
respectively. These changes meet the acceptance criteria in 
Regulatory Guides 1.174 and 1.177. Therefore, since the RTS and 
ESFAS instrumentation will continue to perform their functions with 
high reliability as originally assumed, and the risk impact as 
measured by the [Delta]CDF, [Delta]LERF, ICCDP, and ICLERP risk 
metrics is within the acceptance criteria of existing regulatory 
guidance, there will not be a significant increase in the 
consequences of any accidents.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do 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 changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
The proposed changes are consistent with safety analysis assumptions 
and resultant consequences.
    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.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. The proposed changes will not affect the normal method of 
plant operation. No performance requirements will be affected or 
eliminated.
    The proposed changes will not result in physical alteration to 
any plant system nor will there be any change in the method by which 
any safety-related plant system performs its safety function. The 
proposed changes do not include any changes to the instrumentation 
setpoints or changes to the accident analysis assumptions.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting

[[Page 21602]]

single failures are introduced as a result of these changes. There 
will be no adverse effect or challenges imposed on any safety-
related system as a result of these 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.
    The proposed changes do not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Analysis 
Limit (SAL). There will be no effect on the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions.
    The redundancy of RTS and ESFAS is maintained, and diversity 
with regard to the signals that provide reactor trip and ESF 
actuation is also maintained. All signals credited as primary or 
secondary, and all operator actions credited in the accident 
analyses will remain the same. The proposed changes will not result 
in plant operation in a configuration outside the design basis. The 
calculated impact on risk is insignificant and meets the acceptance 
criteria contained in Regulatory Guides 1.174 and 1.177. Although 
there was no attempt to quantify any positive human factors benefit 
due to increased Completion Times and bypass test times, it is 
expected that there would be a net benefit due to a reduced 
potential for spurious reactor trips and actuations associated with 
testing.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety, as follows:
    (a) Reduced testing should result in fewer inadvertent reactor 
trips, less frequent actuation of ESFAS components, less frequent 
distraction of operations personnel without significantly affecting 
RTS and ESFAS reliability.
    (b) The Completion Time extensions for the reactor trip breakers 
should provide additional time to complete test and maintenance 
activities while at power, potentially reducing the number of forced 
outages related to compliance with reactor trip breaker Completion 
Times, and provide consistency with the Completion Times for the 
logic trains.
    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: J. Hagood Hamilton, Jr., South Carolina 
Electric & Gas Company, P.O. Box 764, Columbia, SC 29218.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: February 15, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16046A009.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-91 and NPF-92 for the VEGP Units 3 and 4. The 
requested amendment proposes changes to the Updated Final Safety 
Analysis Report (UFSAR) in the form of departures from the incorporated 
plant-specific Design Control Document Tier 2 information and involves 
related changes to the associated plant-specific Tier 2* information. 
Specifically, the proposed departures consist of changes to UFSAR text 
and tables, and information incorporated by reference into the UFSAR 
related to updates to WCAP-16096, ``Software Program Manual for Common 
QTM Systems,'' and WCAP-16097, ``Common Qualified Platform 
Topical Report.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    WCAP-16096 (Common Q Software Program Manual) was updated to 
Revision 4 to reference later NRC endorsed regulatory guides and 
standards and update the requirements for the software design and 
development processes for the Common Q portion of the AP1000 
Protection and Safety Monitoring System (PMS). WCAP-16097 (Common Q 
Topical Report) was updated to Revision 3 to describe new Common Q 
components and standards currently used for the AP1000 PMS 
implementation of the Common Q platform. These two WCAPs have been 
reviewed and approved by the NRC in Safety Evaluations dated 
February 7, 2013. WCAP-15927 was updated to reference the newest 
revisions of WCAP-16096 and WCAP-16097 and for editorial 
corrections. The proposed activity adopts the updated versions as 
incorporated by reference documents into the UFSAR. Other proposed 
document changes support the implementation of the updated versions 
of WCAP-16096, WCAP-16097, and WCAP-15927.
    The Common Q platform is an acceptable platform for nuclear 
safety-related applications. The Common Q system meets the 
requirements of 10 CFR part 50, Appendix A, General Design Criteria 
(Criteria 1, 2, 4, 13, 19, 20, 21, 22, 23, 24, and 25), the 
Institute of Electrical and Electronics Engineers Standard 603-1991 
for the design of safety-related reactor protection systems, 
engineered safety features systems and other plant systems, and the 
guidelines of Regulatory Guide 1.152 and supporting industry 
standards for the design of digital systems.
    Because the Common Q platform and the PMS implementation of the 
Common Q platform meet the criteria in the applicable General Design 
Criteria, the revisions to these documents do not affect the 
prevention and mitigation of abnormal events, such as accidents, 
anticipated operational occurrences, earthquakes, floods and turbine 
missiles, or their safety or design analyses as described in the 
licensing basis. The incorporation of the updated documents does not 
adversely affect the interface with any structure, system, or 
component accident initiator or initiating sequence of events. Thus, 
the probabilities of the accidents previously evaluated in the UFSAR 
are not affected.
    Therefore, the proposed amendment does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to adopt the updated WCAP-16096, WCAP-
16097, and WCAP-15927 into the UFSAR do not adversely affect the 
design or operation of safety-related equipment or equipment whose 
failure could initiate an accident beyond what is already described 
in the licensing basis. These changes do not adversely affect 
fission product barriers. No safety analysis or design basis 
acceptance limit/criterion is challenged or exceeded by the 
requested change.
    Therefore, this activity does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to adopt the updated WCAP-16096, WCAP-
16097, and WCAP-15927 into the UFSAR do not adversely affect the 
design, construction, or operation of any plant SSCs, including any 
equipment whose failure could initiate an accident or a failure of a 
fission product barrier. No analysis is adversely affected by the 
proposed changes. Furthermore, no system function, design function, 
or equipment qualification will be adversely affected by the 
changes.
    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.

[[Page 21603]]

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

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

    Date of amendment request: March 11, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16071A333.
    Description of amendment request: The amendments would revise the 
Technical Specifications to add a new condition to extend the allowed 
completion time to restore one Essential Raw Cooling Water train to 
OPERABLE status from 72 hours to 7 days for planned maintenance, when 
the opposite unit is defueled or in Mode 6, following defueling under 
certain restrictions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    1. Does the proposed amendment involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    Response: No.
    The proposed change adds new Condition A to Technical 
Specification (TS) 3.7.8, Essential Raw Cooling Water (ERCW) System 
for Sequoyah Nuclear Plant (SQN) Units 1 and 2. The proposed change 
will extend the allowed completion time to restore ERCW System train 
to OPERABLE status from 72 hours to 7 days for planned maintenance 
when the opposite unit is defueled or in mode 6 following defueled 
with refueling water cavity level >= [greater than or equal to] 23 
ft. above top of reactor vessel flange and UHS [ultimate heat sink] 
Temperature is <= [less than or equal to 79 degrees F. This change 
does not result in any physical changes to plant safety-related 
structures, systems, or components (SSCs). The UHS and associated 
ERCW system function is to remove plant system heat loads during 
normal and accident conditions. As such, the UHS and ERCW system are 
not design basis accident initiators, but instead perform accident 
mitigation functions by serving as the heat sink for safety-related 
equipment to ensure the conditions and assumptions credited in the 
accident analyses are preserved. During operation under the proposed 
change with one ERCW train inoperable, the other ERCW train will 
continue to perform the design function of the ERCW system. 
Therefore, the proposed change does not involve a significant 
increase in the probability of an accident previously evaluated.
    Accordingly, as demonstrated by TVA design heat transfer and 
flow modeling calculations, operation with one ERCW System 
inoperable for 7 days for planned maintenance when the opposite unit 
is defueled or in mode 6 following defueled with refueling water 
cavity level >= 23 ft. above top of reactor vessel flange, the fuel 
cladding, Reactor Coolant System (RCS) pressure boundary, and 
containment integrity limits are not challenged during worst-case 
post-accident conditions. Accordingly, the conclusions of the 
accident analyses will remain as previously evaluated such that 
there will be no significant increase in the post-accident dose 
consequences.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve any physical changes to 
plant safety related SSCs or alter the modes of plant operation in a 
manner that is outside the bounds of the current UHS and ERCW system 
design heat transfer and flow modeling analyses. The proposed change 
to add new Condition A to TS 3.7.8, ERCW System, which would extend 
the allowed completion time to restore ERCW System train to OPERABLE 
status from 72 hours to 7 days for planned maintenance when the 
opposite unit is defueled or in mode 6 following defueled with 
refueling water cavity level >= 23 ft. above top of reactor vessel 
flange. Thus, although the specified ERCW system alignments result 
in reduced heat transfer flow capability, the plant's overall 
ability to reject heat to the UHS during normal operation, normal 
shutdown, and hypothetical worst-case accident conditions will not 
be significantly affected by this proposed change. Because the 
safety and design requirements continue to be met and the integrity 
of the RCS pressure boundary is not challenged, no new credible 
failure mechanisms, malfunctions, or accident initiators are 
created, and there will be no effect on the accident mitigating 
systems in a manner that would significantly degrade the plant's 
response to an accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to add new Condition A to TS 3.7.8, ERCW 
System, which would extend the allowed completion time to restore 
ERCW System train to OPERABLE status from 72 hours to 7 days for 
planned maintenance when the opposite unit is defueled or in mode 6 
following defueled with refueling water cavity level >= 23 ft. above 
top of reactor vessel flange. As demonstrated by TVA design basis 
heat transfer and flow modeling calculations, the design limits for 
fuel cladding, RCS pressure boundary, and containment integrity are 
not exceeded under both normal and post-accident conditions. As 
required, these calculations include evaluation of the worst-case 
combination of meteorology and operational parameters, and establish 
adequate margins to account for measurement and instrument 
uncertainties. While operating margins have been reduced by the 
proposed change in order to support necessary maintenance 
activities, the current limiting design basis accidents remain 
applicable and the analyses conclusions remain bounding such that 
the accident safety margins are maintained. Accordingly, the 
proposed change will not significantly degrade the margin of safety 
of any SSCs that rely on the UHS and ERCW system for heat removal to 
perform their safety related functions.
    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Branch Chief: Benjamin G. Beasley.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: January 27, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16033A470.
    Description of amendment request: The amendment would revise the 
Technical Specifications to allow the use of Optimized ZIRLO\TM\ as an 
approved fuel rod cladding.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change would allow the use of Optimized ZIRLO\TM\ 
clad nuclear fuel in the reactor. The NRC approved topical report 
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, addresses Optimized 
ZIRLO\TM\ and demonstrates that Optimized ZIRLO\TM\ has essentially 
the same properties as currently licensed ZIRLO[supreg]. The fuel 
cladding itself is not an accident initiator and does not affect 
accident probability. Use of Optimized ZIRLO\TM\ fuel cladding will

[[Page 21604]]

continue to meet the 10 CFR 50.46 acceptance criteria and, 
therefore, will not increase the consequences of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Use of Optimized ZIRLO\TM\ clad fuel will not result in changes 
in the operation or configuration of the facility. Topical Report 
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, demonstrated that the 
material properties of Optimized ZIRLO\TM\ are similar to those of 
standard ZIRLO[supreg]. Therefore, Optimized ZIRLOTM fuel 
rod cladding will perform similarly to those fabricated from 
standard ZIRLO[supreg], thus precluding the possibility of the fuel 
cladding becoming an accident initiator and causing a new or 
different type of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will not involve a significant reduction in 
the margin of safety because it has been demonstrated that the 
material properties of the Optimized ZIRLO\TM\ are not significantly 
different from those of standard ZIRLO[supreg]. Optimized ZIRLO\TM\ 
is expected to perform similarly to standard ZIRLO[supreg] for all 
normal operating and accident scenarios, including both loss-of-
coolant accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, 
where the slight difference in Optimized ZIRLO\TM\ material 
properties relative to standard ZIRLO[supreg] could have some impact 
on the overall accident scenario, plant-specific LOCA analyses using 
Optimized ZIRLO\TM\ properties will demonstrate that the acceptance 
criteria of 10 CFR 50.46 have been satisfied.
    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: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
    NRC Branch Chief: Robert J. Pascarelli.

III. Notice of Issuance of Amendments to Facility Operating Licenses 
and Combined Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.

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

    Date of amendment request: June 30, 2015, as supplemented by 
letters dated August 11, 2015; September 24, 2015; October 8, 2015; 
December 7, 2015; February 10, 2016; and February 25, 2016.
    Brief description of amendments: The amendments revised selected 
Technical Specification Completion Times to support repair activity 
associated with the Nuclear Service Water System, Train `A'.
    Date of issuance: March 16, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 282 and 261. A publicly-available version is in 
ADAMS under Accession No. ML15306A141; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Facility Operating License Nos. NPF-9 and NPF-18: Amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: August 20, 2015 (80 FR 
50663). The supplemental letters dated August 11, 2015; September 24, 
2015; October 8, 2015; December 7, 2015; February 10, 2016; and 
February 25, 2016, 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 March 16, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 24, 2015.
    Brief description of amendment request: The amendment changed the 
VY Cyber Security Plan Implementation Schedule Milestone 8 full 
implementation date of June 30, 2016, to December 15, 2017. The 
amendment also revised the existing Renewed Facility Operating License 
Security Plan license condition.
    Date of issuance: March 14, 2016.
    Effective date: As of the date of issuance, and shall be 
implemented by June 30, 2015.
    Amendment No.: 265. A publicly-available version is in ADAMS under 
Accession No. ML16014A169; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-28: The amendment 
revised the Facility Operating License.
    Date of initial notice in Federal Register: September 8, 2015 (80 
FR 53900).
    The Commission's related evaluation of this amendment is contained 
in the Safety Evaluation dated March 14, 2016.
    No significant hazards consideration comments received: No.

[[Page 21605]]

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

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

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

    Date of amendment request: July 29, 2015, as supplemented by letter 
dated November 4, 2015.
    Brief description of amendments: The amendments revised the 
emergency plan definition of annual training frequency to ``once per 
calendar year not to exceed 18 months between training sessions.''
    Date of issuance: March 17, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 316/294; 221/155; and 121. A publicly-available 
version is in ADAMS under Accession No. ML15352A164; documents related 
to these amendments are listed in the safety evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-53, DPR-69, DPR-63, 
NPF-69, and DPR-18: The amendments revised the emergency plans.
    Date of initial notice in Federal Register: December 8, 2015 (80 FR 
76320).
    The Commission's related evaluation of the amendments is contained 
in a safety evaluation dated March 17, 2016.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating 
Station (LGS), Unit 1, Montgomery County, Pennsylvania

    Date of amendment request: November 19, 2015.
    Brief description of amendment: The amendment revised the technical 
specifications (TSs) related to the safety limit minimum critical power 
ratios. The changes resulted from a cycle-specific analysis performed 
to support the operation of LGS, Unit 1, in the upcoming Cycle 17.
    Date of issuance: March 15, 2016.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from the spring 2016 refueling outage.
    Amendment No.: 221. A publicly-available version is in ADAMS under 
Accession No. ML16041A021; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-39: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: January 5, 2016 (81 FR 
275).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 15, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 4, 2014, as supplemented by 
letters dated January 29, February 6, April 28, July 6, September 4, 
October 1, and October 26, 2015, and January 15, 2016.
    Brief description of amendments: The amendments changed the 
Technical Specifications (TSs) and Renewed Facility Operating Licenses 
(RFOLs) to allow plant operation from the currently licensed Maximum 
Extended Load Line Limit Analysis (MELLLA) domain to plant operation in 
the expanded MELLLA Plus (MELLLA+) domain under the previously approved 
extended power uprate conditions of 3,951 megawatts thermal rated core 
thermal power. The expanded MELLLA+ operating domain increases 
operating flexibility by allowing control of reactivity at maximum 
power by changing flow rather than by control rod insertion and 
withdrawal.
    Date of issuance: March 21, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 1 year of issuance.
    Amendments Nos.: 305 and 309. A publicly-available version is in 
ADAMS under Accession No. s; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    RFOL Nos. DPR-44 and DPR-56: The amendments revised the RFOLs and 
TSs.
    Date of initial notice in Federal Register: December 2, 2014 (79 FR 
71454). The supplemental letters dated January 29, February 6, April 
28, July 6, September 4, October 1, and October 26, 2015, and January 
15, 2016, 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 March 21, 2016.
    No significant hazards consideration comments received: Yes.

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: September 11, 2014, as supplemented by 
letters dated October 15, 2014, and December 18, 2014.
    Description of amendment: The amendments revised the Updated Final 
Safety Analysis Report by clarifying how human diversity was applied 
during the design process for the Component Interface Module and 
Diverse Actuation System.
    Date of issuance: July 17, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 28. A publicly-available version is in ADAMS under 
Accession No. ML15176A703; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined License Nos. NPF-93 and NPF-94: Amendments 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: December 9, 2014 (79 FR 
73111). The supplemental letters dated October 15, 2014, and December 
18, 2014, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 17, 2015.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 10, 2015.
    Brief description of amendment: The amendments revised the VCSNS 
Units 2 and 3 Updated Final Safety Analysis Report (UFSAR) by revising 
the references to human factors-related plans. The UFSAR-referenced 
plans are the Human Factors Engineering Design Verification plan, Task 
Support Verification plan, and the Integrated

[[Page 21606]]

System Validation plan. The UFSAR references to those plans required an 
update to the latest version of those plans due to changes within the 
plans. The amendments involved changes to the approved VCSNS Units 2 
and 3 UFSAR Tier 2* information, as defined in 10 CFR part 52, appendix 
D, section II.F.
    Date of issuance: September 23, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 33. A publicly-available version is in ADAMS under 
Accession No. ML15189A363; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined License Nos. NPF-93 and NPF-94: Amendments 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: March 31, 2015 (80 FR 
17094). The supplemental letter dated March 24, 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 23, 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: August 24, 2015.
    Brief description of amendment: The amendments authorized changes 
to the VCSNS Units 2 and 3 Updated Final Safety Analysis Report Tier 2 
and Tier 2* information to revise the seismic Category I and II 
structures containing mechanical couplers welded to structural steel 
utilizing combined partial joint penetration weld with fillet weld 
reinforcement with fillet welds satisfying the minimum size 
requirements for C2/C3J couplers to demonstrate the capacity required 
by code is established by appropriate analyses and testing.
    Date of issuance: November 12, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 36. A publicly-available version is in ADAMS under 
Accession No. ML15301A100; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined License Nos. NPF-93 and NPF-94: Amendments 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: September 3, 2015 (80 
FR 53336). The supplemental letters dated September 23, 2015, and 
October 1, 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 November 12, 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: October 22, 2015.
    Brief description of amendment: The amendments authorized changes 
to the VCSNS Combined Licenses (COLs). Specifically, the changes were 
to VCSNS Units 2 and 3 COLs, Appendix A, Technical Specifications, 
Section 5.0, ``Administrative Controls,'' by revising the title ``Shift 
Supervisor'' to ``Shift Manager.''
    Date of issuance: February 29, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 42. A publicly-available version is in ADAMS under 
Accession No. ML16042A476; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined License Nos. NPF-93 and NPF-94: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: November 24, 2015 (80 
FR 73242).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 29, 2016.
    No significant hazards consideration comments received: No.

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 12, 2015, as supplemented by letters 
dated September 21, 2015; November 25, 2015; and January 28, 2016.
    Brief description of amendments: The amendments revised and added 
Surveillance Requirements to verify that the system locations 
susceptible to gas accumulation are sufficiently filled with water and 
to provide allowances that permit performance of the verification. The 
changes are consistent with TSTF-523, Revision 2, ``Generic Letter 
2008-01, Managing Gas Accumulation.''
    Date of issuance: March 21, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: 178 (Unit 1) and 159 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16063A475, documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-2 and NPF-8: The amendments 
revised the Renewed Facility Operating Licenses and Technical 
Specifications.
    Date of initial notice in Federal Register: June 23, 2015 (80 FR 
35984). The supplemental letters dated September 21, 2015; November 25, 
2015; and January 28, 2016, 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 March 21, 2016.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia

    Date of amendment request: April 2, 2015, as supplemented by 
letters dated November 12, 2015, and February 9, 2016.
    Brief description of amendments: The amendments revised the 
technical specifications (TSs) as necessary to relocate the pressure 
and temperature (P-T or P/T) limit curves and associated references to 
a pressure and temperature limits report (PTLR). Specifically, the 
request modified Section 1.0, ``Definitions''; Limiting Conditions for 
Operation and Surveillance Requirement Applicability Section 3.4.9, 
``RCS Pressure and Temperature

[[Page 21607]]

(P/T) Limits''; and Section 5.0, ``Administrative Controls,'' of the 
TSs for both units to delete reference to the P-T curves and to include 
reference to the unit-specific PTLRs. The amendments also implemented 
new P-T limits for both units.
    Date of issuance: March 23, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 277 and 221. A publicly-available version is in 
ADAMS under Accession No. ML16062A099; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: July 7, 2015 (80 FR 
38760). The supplemental letters dated November 12, 2015, and February 
9, 2016, 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 March 23, 2016.
    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, Luzerne County, Pennsylvania

    Date of amendment request: March 19, 2015, as supplemented by 
letters dated October 15, 2015; October 16, 2015; and January 8, 2016. 
A publicly-available version is in ADAMS under Accession Nos. 
ML15091A657, ML15296A048, ML15296A057, and ML16011A103, respectively.
    Brief description of amendments: The amendments revised the 
Emergency Plan for the Susquehanna Steam Electric Station (SSES) to 
adopt the Nuclear Energy Institute's (NEI's) revised Emergency Action 
Level scheme described in NEI 99-01, Revision 6, ``Development of 
Emergency Action Levels for Non-Passive Reactors'' (ADAMS Accession No. 
ML12326A805), which was endorsed by the NRC as documented in NRC letter 
dated March 28, 2013 (ADAMS Accession No. ML12346A463). This request 
was submitted by PPL Susquehanna, LLC; however, on June 1, 2015 (ADAMS 
Accession No. ML15054A066), the NRC staff issued an amendment changing 
the name on the SESS license from PPL Susquehanna, LLC to Susquehanna 
Nuclear, LLC. This amendment was issued subsequent to an order issued 
on April 10, 2015 (ADAMS Accession No. ML15058A073), to SSES, approving 
an indirect license transfer of the SESS license to Talen Energy 
Corporation.
    Date of issuance: March 28, 2016.
    Effective date: As of the date of issuance and shall be implemented 
on or before December 31, 2016.
    Amendment Nos.: 265 (Unit 1) and 246 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16062A216; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: July 7, 2015 (80 FR 
38762). The supplemental letters dated October 15, 2015; October 16, 
2015; and January 8, 2016, provided additional information that 
clarified the application and 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 NRC staff published a subsequent notice 
in the Federal Register on February 2, 2016 (81 FR 5500).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 28, 2016.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 1st day of April 2016.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2016-08323 Filed 4-11-16; 8:45 am]
 BILLING CODE 7590-01-P