81 FR 5495 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations

NUCLEAR REGULATORY COMMISSION

Federal Register Volume 81, Issue 21 (February 2, 2016)

Page Range		5495-5505
FR Document		2016-01771
Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this regular biweekly notice. The Act requires the Commission to publish notice of any amendments issued, or proposed to be issued, and grants the Commission the authority to issue and make immediately effective any amendment to an operating license or combined license, as applicable, upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued from January 5, 2016, to January 15, 2016. The last biweekly notice was published on January 19, 2016.
Federal Register, Volume 81 Issue 21 (Tuesday, February 2, 2016)
[Federal Register Volume 81, Number 21 (Tuesday, February 2, 2016)]
[Notices]
[Pages 5495-5505]
From the Federal Register Online  [www.thefederalregister.org]
[FR Doc No: 2016-01771]


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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0019]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, 
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is 
publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued, and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from January 5, 2016, to January 15, 2016. The 
last biweekly notice was published on January 19, 2016.

DATES: Comments must be filed by March 3, 2016. A request for a hearing 
must be filed by April 4, 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 http://www.regulations.gov and search for Docket ID NRC-2016-0019. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected].
     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: Janet Burkhardt, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-1384, email: [email protected].

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2016-0019 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 http://www.regulations.gov and search for Docket ID NRC-2016-0019.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://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 [email protected]. 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-2016-0019, 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 http://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.

[[Page 5496]]

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 http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed within 60 days, the Commission or a presiding officer designated 
by the Commission or by the Chief Administrative Judge of the Atomic 
Safety and Licensing Board Panel, will rule on the request and/or 
petition; and the Secretary or the Chief Administrative Judge of the 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that person's admitted 
contentions, including the opportunity to present evidence and to 
submit a cross-examination plan for cross-examination of witnesses, 
consistent with NRC regulations, policies and procedures.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it 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 April 
4, 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 
April 4, 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

[[Page 5497]]

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 [email protected], or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://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 http://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 http://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 http://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 http://www.nrc.gov/site-help/e-submittals.html, by email to 
[email protected], 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 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, in some instances, a request to intervene will require 
including information on local residence in order to demonstrate a 
proximity assertion of interest in the proceeding. With respect to 
copyrighted works, except for limited excerpts that serve the purpose 
of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station (LSCS), Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: November 19, 2015. A publicly available 
version is in ADAMS under Accession No. ML15324A309.
    Description of amendment request: The proposed amendments would 
revise LSCS Technical Specifications

[[Page 5498]]

(TS) Section 2.1.1, ``Reactor Core SLs,'' to reflect a lower reactor 
steam dome pressure stated for Reactor Core Safety Limits (SLs) 2.1.1.1 
and 2.1.1.2. Specifically, the proposed amendment will reduce the 
reactor steam dome pressure in TS SLs 2.1.1.1 and 2.1.1.2 from 785 psig 
[pound per square inch gage] to 685 psig. This change to TS Section 
2.1.1 was identified as a result of General Electric Part 21 report 
SC05-03, ``Potential to Exceed Low Pressure Technical Specification 
Safety Limit.'' This change is valid for the NRC-approved pressure 
range pertinent to the critical power correlations applied to the fuel 
types in use at LSCS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the reactor steam dome pressure in the 
LSCS Reactor Core Safety Limits TS 2.1.1.1 and 2.1.1.2 does not 
alter the use of the analytical methods used to determine the safety 
limits that have been previously reviewed and approved by the NRC. 
The proposed change is in accordance with an NRC approved critical 
power correlation methodology, and as such, maintains required 
safety margins. The proposed change does not adversely affect 
accident initiators or precursors, nor does it alter the design 
assumptions, conditions, or configuration of the facility or the 
manner in which the plant is operated and maintained.
    The proposed change does not alter or prevent the ability of 
structures, systems, and components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits. The proposed change does 
not require any physical change to any plant SSCs nor does it 
require any change in systems or plant operations. The proposed 
change is consistent with the safety analysis assumptions and 
resultant consequences.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed reduction in the reactor dome pressure safety limit 
from 785 psig to 685 psig is a change based upon previously approved 
documents and does not involve changes to the plant hardware or its 
operating characteristics. As a result, no new failure modes are 
being introduced.
    There are no hardware changes nor are there any changes in the 
method by which any plant systems perform a safety function. No new 
accident scenarios, failure mechanisms, or limiting single failures 
are introduced as a result of the proposed change.
    The proposed change does not introduce any new accident 
precursors, nor does it involve any physical plant alterations or 
changes in the methods governing normal plant operation. Also, the 
change does not impose any new or different requirements or 
eliminate any existing requirements. The change does not alter 
assumptions made in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through the design of the 
plant structures, systems, and components, and through the 
parameters for safe operation and setpoints for the actuation of 
equipment relied upon to respond to transients and design basis 
accidents.
    Evaluation of the 10 CFR part 21 condition by General Electric 
determined that since the Minimum Critical Power Ratio improves 
during the PRFO [Pressure Regulator Failure Maximum Demand (Open)] 
transient, there is no decrease in the safety margin and therefore 
there is not a threat to fuel cladding integrity.
    The proposed change in reactor dome pressure supports the 
current safety margin, which protects the fuel cladding integrity 
during a depressurization transient, but does not change the 
requirements governing operation or availability of safety equipment 
assumed to operate to preserve the margin of safety. The change does 
not alter the behavior of plant equipment, which remains unchanged.
    The proposed change to Reactor Core Safety Limits 2.1.1.1 and 
2.1.1.2 is consistent with and within the capabilities of the 
applicable NRC approved critical power correlation for the fuel 
designs in use at LSCS, Units 1 and 2. No setpoints at which 
protective actions are initiated are altered by the proposed change.
    The proposed change does not alter the manner in which the 
safety limits are determined. This change is consistent with plant 
design and does not change the TS operability requirements; thus, 
previously evaluated accidents are not affected by this proposed 
change.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Acting Branch Chief: Justin C. Poole.

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: December 3, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15337A413.
    Description of amendment request: The amendments would revise the 
technical specification (TS) surveillance requirements (SRs) associated 
with the emergency diesel generator (EDG) fuel oil transfer system. 
Specifically, the amendments would allow for the crediting of manual 
actions, in lieu of automatic actions, without having to declare the 
EDGs inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will revise SR 3.8.1.6 by adding a note to 
allow for procedurally controlled simple manual actions associated 
with the fuel oil transfer system without having to declare the EDG 
inoperable [under] administrative control. The fuel oil transfer 
system is required to support continuous operation of standby power 
sources. The surveillance provides assurance that the fuel oil 
transfer system is OPERABLE. The fuel oil transfer system is not an 
initiator of any event previously evaluated. Therefore, the 
probability of any accident previously evaluated is not increased.
    In the event of an accident, if simple manual actions were 
necessary to restore the automatic feature of the EDG day tank fill, 
analysis shows that significant margin exists to ensure that EDG 
operability would not be adversely affected. Although the proposed 
change to allow simple manual actions could introduce additional 
potential malfunctions, such that human error could result in the 
potential to improperly realign the fuel oil transfer system during 
a DBA [design-basis accident], the improper realignment would be 
detected when the transfer of fuel oil from the storage tank to the 
day tank did not occur as expected and the error would be corrected 
prior to having a significant impact.
    The proposed change does not involve any physical changes to the 
structures, systems, or components (SSCs) in the plant. Further the 
proposed change does not alter or prevent the ability of SSCs from 
performing their intended function to mitigate the consequences of 
an event.

[[Page 5499]]

    The proposed change is consistent with NRC regulatory 
requirements regarding the content of plant TS as identified in 10 
CFR 50.36. Additionally, the proposed change is consistent with 
NUREG-1433, ``Standard Technical Specifications General Electric 
BWR/4 Plants,'' in that the word `automatically' is bracketed (i.e., 
optional or as required by plant design).
    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 does not alter the physical design, safety 
limits, or safety analysis assumptions associated with the operation 
of the plant. Accordingly, the change does not introduce any new 
accident initiators, nor does it reduce or adversely affect the 
capabilities of any plant structure, system, or component to perform 
their safety function. Consequently, there are no new initiators 
that could result in a new or different kind 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 conforms to NRC regulatory guidance 
regarding the content of plant Technical Specifications. The 
proposed change does not alter the physical design, safety limits, 
or safety analysis assumptions associated with the operation of the 
plant. The proposed change has no adverse impact on current Safety 
Limits, Limiting Safety System Settings, Limiting Control Settings, 
Limiting Conditions for Operation, Surveillance Requirements, Design 
Features, or Administrative Controls.
    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: Douglas A. Broaddus.

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: December 17, 2015, and supplemented by 
letter dated January 11, 2016. Publicly-available versions are in ADAMS 
under Accession Nos. ML15351A452 and ML16011A500, respectively.
    Description of amendment request: The proposed changes, if 
approved, would amend Combined License Nos. NPF-93 and NPR-94 for 
VCSNS, Units 2 and 3, respectively. The requested amendment proposes to 
change the design of the auxiliary building Wall 11 and other changes 
to the licensing basis for the use of Category II structures, such as 
Wall 11.2 in the turbine building. The changes in the proposed 
amendment are located primarily in the VCSNS Updated Final Safety 
Analysis Report (UFSAR) Tier 2* and Tier 2 information, and also 
require conforming changes to a license condition.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not adversely affect the operation of 
any systems or equipment inside or outside the auxiliary building 
that could initiate or mitigate abnormal events, e.g., accidents, 
anticipated operational occurrences, earthquakes, floods, tornado 
missiles, and turbine missiles, or their safety or design analyses, 
evaluated in the UFSAR. The changes do not adversely affect any 
design function of the auxiliary building or the systems and 
equipment contained therein. The ability of the affected auxiliary 
building [main steam isolation valve] MSIV compartments to withstand 
the pressurization effects from the design basis pipe rupture is not 
adversely affected by the removal of the Wall 11 upper vent 
openings, because vents at these locations are not credited in the 
subcompartment pressurization analysis. MSIV compartment 
temperatures following the limiting one square foot pipe rupture 
with the vent openings removed remain acceptably within the envelope 
for environmental qualification of equipment in the compartments. 
The credit of seismic Category II Wall 11.2 as a [high energy line 
break] HELB barrier and the seismic Category II turbine building 
first bay and associated missile barriers to protect Wall 11 
openings from tornado missiles continues to provide adequate 
protection of structures, systems, and components (SSCs) required to 
safely shut down the plant, as these structures are designed to the 
same requirements as seismic Category I structures, and with the 
additional HELB loadings assumed, remain well within the applicable 
acceptance criteria.
    Therefore, the proposed activity 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 change the design function of the 
auxiliary building or of any of the systems or equipment in the 
auxiliary building or elsewhere within the Nuclear Island structure. 
These proposed changes do not introduce any new equipment or 
components that would result in a new failure mode, malfunction or 
sequence of events that could affect safety-related or nonsafety-
related equipment. This activity will not allow for a new fission 
product release path, result in a new fission product barrier 
failure mode, or create a new sequence of events that would result 
in significant fuel cladding failures.
    Therefore, this activity 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 margin of safety for the design of the auxiliary building is 
maintained through continued use of the current codes and standards 
as stated in the UFSAR and adherence to the assumptions used in the 
analyses of this structure and the events associated with this 
structure. The auxiliary building will continue to maintain a 
seismic Category I rating which preserves the current structural 
safety margins. The 3-hour fire rating requirements for the impacted 
auxiliary building walls are maintained. The Wall 11 upper vents are 
not credited in the subcompartment pressurization analysis and the 
remaining vents and pressure relief devices provide sufficient 
venting to maintain the MSIV compartment pressures below the design 
limit and design basis. The credit of turbine building Wall 11.2 as 
a HELB barrier provides protection of Wall 11 from selected dynamic 
effects, which in turn provides that essential SSCs remain protected 
from the effects of postulated HELB events. The credit of the 
seismic Category II turbine building first bay and associated 
missile barriers to provide protection of Wall 11 openings from 
tornado missiles provides sufficient protection for the essential 
SSCs located in the auxiliary building in the vicinity of Wall 11 
from the effects of external missiles. Thus the requested changes 
will not adversely affect any safety-related equipment, design code, 
function, design analysis, safety analysis input or result, or 
design/safety margin. No safety analysis or design basis acceptance 
limit/criterion is challenged or exceeded by the requested change, 
thus, no margin of safety is reduced.
    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

[[Page 5500]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Acting Branch Chief: John McKirgan.

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

    Date of amendment request: November 16, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15320A464.
    Description of amendment request: The requested amendment proposes 
to depart from Tier 2* information in the Updated Final Safety Analysis 
Report related to the construction methods used for the composite 
floors and roof of the auxiliary building.
    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 design functions of the nuclear island structures are to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the nuclear island. 
The nuclear island structures are structurally designed to meet 
seismic Category I requirements as defined in Regulatory Guide 1.29.
    The use of [American Concrete Institute (ACI)] 349 and [American 
Institute of Steel Construction (AISC)] N690 provides criteria for 
the design, qualification, fabrication, and inspection of composite 
steel beam floors and roof in the auxiliary building. These 
structures continue to meet the applicable portions of ACI 349 and 
AISC N690. The proposed change does not have an adverse impact on 
the response of the nuclear island structures to safe shutdown 
earthquake ground motions or loads due to anticipated transients or 
postulated accident conditions. The change does not impact the 
support, design, or operation of mechanical and fluid systems. There 
is no change to plant systems or the response of systems to 
postulated accident conditions. There is no change to the predicted 
radioactive releases due to normal operation or postulated accident 
conditions. The plant response to previously evaluated accidents or 
external events is not adversely affected, nor does the change 
described create any new accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change revises the description of the construction 
of composite steel beam floors and roof in the auxiliary building. 
The proposed change does not change the design function, support, 
design, or operation of mechanical and fluid systems. The proposed 
change does not result in a new failure mechanism for the pertinent 
structures or new accident precursors. As a result, the design 
function of the structures is not adversely affected by the proposed 
change.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change is consistent with ACI 349 and AISC N690. 
The design and construction of the auxiliary building floors and 
roof remain in conformance with the requirements in ACI 349 and AISC 
N690.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North Birmingham, AL 35203-2015.
    NRC Acting Branch Chief: John McKirgan.

Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna 
Steam Electric Station, Units 1 and 2 (SSES), 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. 
Publicly-available versions are in ADAMS under Package Accession Nos. 
ML15091A657, ML15296A048, and ML15296A057, and Accession No. 
ML16011A103, respectively.
    Description of amendment request: The NRC staff previously made a 
proposed determination that the amendment request dated March 19, 2015, 
involved no significant hazards consideration (80 FR 38762; July 7, 
2015). Subsequently, the supplemental letter dated October 15, 2015, 
provided additional information that expanded the scope of the 
application as originally noticed. Accordingly, this notice supersedes 
the previous notice in its entirety. The amendments would revise the 
Emergency Plan for SSES to adopt the Nuclear Energy Institute's (NEI's) 
revised emergency action level (EAL) 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). Supplemental changes in these amendments were 
discussed in a September 23, 2015, public meeting held with Susquehanna 
Nuclear, LLC. The public meeting summary was issued October 9, 2015, 
and is available in ADAMS under Accession No. ML15278A492. The 
additional information, and the changes discussed at the public 
meeting, are included in the two Susquehanna Nuclear, LLC letters dated 
October 15, 2015, and October 16, 2015. The revised Emergency Plan 
includes the appropriate plant-specific changes as a result of an 
emergency operating procedure upgrade project and corrective action in 
response to an NRC Emergency Preparedness White Finding, documented in 
NRC Inspection Report No. 05000387/2015504 and 05000388/2015504, dated 
June 22, 2015 (ADAMS Accession Nos. ML15173A297 and ML15181A332).
    On June 1, 2015, the NRC staff issued an amendment changing the 
name on the SSES license from PPL Susquehanna, LLC to Susquehanna 
Nuclear, LLC. This amendment was issued subsequent to an order issued 
on April 10, 2015, to SSES, approving an indirect license transfer.
    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 any accident previously 
evaluated?
    Response: No.
    The proposed changes to the EAL scheme to adopt the NRC-endorsed 
guidance in NEI 99-01, Revision 6, ``Development of Emergency Action 
Levels for Non-Passive Reactors,'' [and the additional plant-
specific Emergency Plan changes] do not reduce the capability to 
meet the emergency planning requirements established in 10 CFR 50.47 
and 10 CFR 50, Appendix E. The proposed changes do not reduce the 
functionality, performance, or capability of the ERO [Emergency 
Response Organization] to

[[Page 5501]]

respond in mitigating the consequences of any design basis accident.
    The probability of a reactor accident requiring implementation 
of Emergency Plan EALs has no relevance in determining whether the 
proposed changes to the EALs reduce the effectiveness of the 
Emergency Plan. As discussed in Section I.D, ``Planning Basis,'' of 
NUREG-0654, Revision 1, ``Criteria for Preparation and Evaluation of 
Radiological Emergency Response Plans and Preparedness in Support of 
Nuclear Power Plants'';
    . . . The overall objective of emergency response plans is to 
provide dose savings (and in some cases immediate life saving) for a 
spectrum of accidents that could produce offsite doses in excess of 
Protective Action Guides (PAGs). No single specific accident 
sequence should be isolated as the one for which to plan because 
each accident could have different consequences, both in nature and 
degree. Further, the range of possible selection for a planning 
basis is very large, starting with a zero point of requiring no 
planning at all because significant offsite radiological accident 
consequences are unlikely to occur, to planning for the worst 
possible accident, regardless of its extremely low likelihood. . . .
    Therefore, risk insights are not considered for any specific 
accident initiation or progression in evaluating the proposed 
changes.
    The proposed changes do not involve any physical changes to 
plant equipment or systems, nor do they alter the assumptions of any 
accident analyses. The proposed changes do not adversely affect 
accident initiators or precursors nor do they alter the design 
assumptions, conditions, and configuration or the manner in which 
the plants are operated and maintained. The proposed changes do not 
adversely affect the ability of Structures, Systems, or Components 
(SSCs) to perform their intended safety functions in mitigating the 
consequences of an initiating event within the assumed acceptance 
limits.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the EAL scheme to adopt the NRC-endorsed 
guidance in NEI 99-01, Revision 6, [and the additional plant-
specific Emergency Plan changes] do not involve any physical changes 
to plant systems or equipment. The proposed changes do not involve 
the addition of any new plant equipment. The proposed changes will 
not alter the design configuration, or method of operation of plant 
equipment beyond its normal functional capabilities. All ERO 
functions will continue to be performed as required. The proposed 
changes do not create any new credible failure mechanisms, 
malfunctions, or accident initiators.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from those that have been 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes to the EAL scheme to adopt the NRC-endorsed 
guidance in NEI 99-01, Revision 6, [and the additional plant-
specific Emergency Plan changes] do not alter or exceed a design 
basis or safety limit. There is no change being made to safety 
analysis assumptions, safety limit, or limiting safety system 
settings that would adversely affect plant safety as a result of the 
proposed changes. There are no changes to setpoints or environmental 
conditions of any SSC or the manner in which any SSC is operated. 
Margins of safety are unaffected by the proposed changes to adopt 
the NEI 99-01, Revision 6 EAL scheme guidance. The applicable 
requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E will continue 
to be met.
    Therefore, the proposed changes do not involve any reduction in 
a margin of safety.

    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: Damon D. Obie, Associate General Counsel, 
Talen Energy Supply, LLC, 835 Hamilton St., Suite 150, Allentown, PA 
18101.
    NRC Branch Chief: Douglas A. Broaddus.

Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant 
(WBN), Unit 2, Rhea County, Tennessee

    Date of amendment request: December 15, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15350A250.
    Brief description of amendment request: The amendment would revise 
the technical specification (TS) surveillance requirements (SRs) for 
the WBN, Unit 2, ice condenser lower inlet doors.
    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 ice condenser is a passive heat removal plant feature. The 
proposed amendment to the TS 3.6.12 does not change the design, 
physical features or the function of the ice condenser or the ice 
condenser doors. The ice condenser is not an accident initiator, 
thus the proposed amendment does not increase the probability of an 
accident previously evaluated.
    The ice condenser is credited in mitigating the consequences of 
postulated Design Basis Accidents (DBAs) and remains capable of 
performing its design basis functions. The proposed amendment to the 
SRs during the first cycle of WBN Unit 2 operation does not change 
the ice condenser configuration or how it behaves in the event of a 
DBA. Thus it is concluded that a significant increase in the 
consequences of an accident previously evaluated will not occur as a 
result of the proposed amendment.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed amendment does not introduce any new modes 
of plant operation, change the design function of the ice condenser 
or any other Structure System or Component (SSC), or change the mode 
of operation of the ice condenser or any other SSC. There are no new 
equipment failure modes or malfunctions created as the ice condenser 
and ice condenser lower inlet doors continue to operate in the same 
manner assumed in the accident analysis. The ice condenser is a 
passive post-accident heat removal feature that is not an accident 
initiator.
    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.
    Ice condensers have been in-service at nine nuclear units in the 
United States for many years. Operating experience has shown that an 
18-month surveillance frequency for evaluating operability is 
appropriate for the lower inlet doors. The proposed amendment to 
perform a revised schedule of lower inlet door surveillances in the 
first cycle before transitioning to the standard 18-month 
surveillance frequency does not result in a significant reduction in 
the margin of safety.
    Therefore, since there is no adverse impact of this amendment on 
the WBN Unit 2 safety analysis, there is no significant reduction in 
the margin of safety of the plant.

    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: Scott A. Vance, Associate General Counsel, 
Nuclear, Tennessee Valley Authority, 400 West Summit Hill Drive, WT 6A-
K, Knoxville, TN 37902.
    NRC Branch Chief: Benjamin G. Beasley.

[[Page 5502]]

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 Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick 
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

    Date of amendment request: January 30, 2015, as supplemented by 
letter dated November 23, 2015.
    Brief description of amendments: The amendments authorized the 
upgrade of the emergency action level scheme for each unit based on the 
Nuclear Energy Institute (NEI) document NEI 99-01, Revision 6, 
``Development of Emergency Action Levels for Non-Passive Reactors,'' 
dated November 2012. NEI 99-01, Revision 6, was endorsed by the NRC by 
letter dated March 28, 2013.
    Date of issuance: January 8, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment Nos.: 268 (Unit 1) and 296 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML15344A153; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-71 and DPR-62: 
Amendments revised the Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: April 28, 2015 (80 FR 
23602). The supplemental letter dated November 23, 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 January 8, 2016.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247, and 50-
286, Indian Point Nuclear Generating Station (IP), Unit Nos. 1, 2, and 
3, and Docket No. 72-51 for IP Independent Spent Fuel Storage 
Installation (ISFSI), Westchester County, New York

    Date of application for amendments: August 20, 2013, as 
supplemented by letters dated November 21, 2013, and May 13 and July 
24, 2014.
    Brief description of amendments: The amendments modified the 
licenses to reflect a grant of Section 161A of the Atomic Energy Act, 
to authorize the licensee the authority to possess and use certain 
firearms, ammunition, and other devices such as large-capacity 
ammunition feeding devices, and to implement the NRC-approved security 
plan for IP including the general-licensed ISFSI.
    Date of issuance: January 5, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 20 days.
    Amendment Nos.: Unit 1--58, Unit 2--282, and Unit 3--259. A 
publicly-available version is in ADAMS under Package Accession No. 
ML14259A209; documents related to these amendments are listed in the 
Safety Evaluation enclosed with the amendment.
    Facility Operating License Nos. DPR-5, DPR-26, and DPR-64 and 
Special Nuclear Materials General-License: The amendments revised the 
Facility Operating Licenses including the general-licensed ISFSI.
    Date of initial notice in Federal Register: February 27, 2014 (79 
FR 11147).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 5, 2016.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
Fitzpatrick Nuclear Power Plant (Fitzpatrick), and Docket No. 72-12 for 
Fitzpatrick Independent Spent Fuel Storage Installation (ISFSI), Oswego 
County, New York

    Date of application for amendment: August 30, 2013, as supplemented 
by letters dated November 12, 2013, May 14, and July 11, 2014, and 
January 15, 2015.
    Brief description of amendment: The amendment modified the licenses 
to reflect a grant of Section 161A of the Atomic Energy Act, to 
authorize the licensee the authority to possess and use certain 
firearms, ammunition, and other devices such as large-capacity 
ammunition feeding devices, and to implement the NRC-approved security 
plan for Fitzpatrick including the general-licensed ISFSI.
    Date of issuance: January 5, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 20 days.
    Amendment No.: 310. A publicly-available version is in ADAMS under 
package Accession No. v; documents related to this amendment are listed 
in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-59 and Special Nuclear 
Materials General-License: The amendment revised the Renewed Facility 
Operating License including the general-licensed ISFSI.
    Date of initial notice in Federal Register: May 6, 2014 (79 FR 
25900).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 5, 2016.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point 
Nuclear Station (NMP), Unit 2, Oswego County, New York

    Date of amendment request: September 3, 2015.
    Brief description of amendment: The amendment changed Technical 
Specification (TS) Section 2.1.1.2, ``Reactor Core SLs [Safety 
Limits],'' to revise the cycle-specific safety limit

[[Page 5503]]

minimum critical power ratio for Cycle 16 for NMP, Unit 2.
    Date of issuance: January 5, 2016.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from the refueling outage where Global Nuclear Fuel 2 
is loaded.
    Amendment No.: 153. A publicly-available version is in ADAMS under 
Accession No. ML15341A336; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-69: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: November 3, 2015 (80 FR 
67801).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 5, 2016.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine 
Mile Point Nuclear Station, Units 1 and 2 (NMP), and Docket No. 72-1036 
for NMP Independent Spent Fuel Storage Installation (ISFSI), Oswego 
County, New York

    Date of application for amendments: August 14, 2013, as 
supplemented by letters dated September 10, 2013, and May 14, 2014.
    Brief description of amendments: The amendments modified the 
licenses to reflect a grant of Section 161A of the Atomic Energy Act, 
to authorize the licensee the authority to possess and use certain 
firearms, ammunition, and other devices such as large-capacity 
ammunition feeding devices, and to implement the NRC-approved security 
plan for NMP including the general-licensed ISFSI.
    Date of issuance: January 5, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 20 days.
    Amendment Nos.: Unit 1--220; Unit 2--154. A publicly-available 
version is in ADAMS under Package Accession No. ML14254A450; documents 
related to these amendments are listed in the Safety Evaluation 
enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-63 and NPF-69, and 
Special Nuclear Materials General-License: The amendments revised the 
Renewed Facility Operating Licenses including the general-licensed 
ISFSI.
    Date of initial notice in Federal Register: October 27, 2014 (79 FR 
63956).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 5, 2016.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear 
Power Plant (Ginna), and Docket No. 72-67 for Ginna Independent Spent 
Fuel Storage Installation (ISFSI), Wayne County, New York

    Date of application for amendment: August 14, 2013, as supplemented 
by letters dated November 4, 2013, and May 14, 2014.
    Brief description of amendment: The amendment modified the licenses 
to reflect a grant of Section 161A of the Atomic Energy Act, to 
authorize the licensee the authority to possess and use certain 
firearms, ammunition, and other devices such as large-capacity 
ammunition feeding devices, and to implement the NRC-approved security 
plan for Ginna including the general-licensed ISFSI.
    Date of issuance: January 5, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 20 days.
    Amendment No.: 120. A publicly-available version is in ADAMS under 
Package Accession No. ML14260A140; documents related to this amendment 
are listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-18 and Special Nuclear 
Materials General-License: The amendment revised the Renewed Facility 
Operating License including the general-licensed ISFSI.
    Date of initial notice in Federal Register: October 27, 2014 (79 FR 
63951).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 5, 2016.
    No significant hazards consideration comments received: No.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 (CPNPP), Somervell 
County, Texas

    Date of amendment request: January 28, 2015, as supplemented by 
letter dated July 29, 2015.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 5.5.16, ``Containment Leakage Rate Testing 
Program,'' for CPNPP, to allow an increase in the 10 CFR part 50, 
Appendix J, ``Primary Reactor Containment Leakage Testing for Water-
Cooled Power Reactors,'' Type A Integrated Leak Rate Test (ILRT) 
interval from a 10-year frequency to a maximum of 15 years and the 
extension of the containment isolation valves leakage Type C tests from 
its current 60-month frequency to 75 months in accordance with Nuclear 
Energy Institute (NEI) 94-01, Revision 3-A, ``Industry Guidance for 
Implementing Performance-Based Option of 10 CFR 50, Appendix J,'' July 
2012, and conditions and limitations specified in NEI 94-01, Revision 
2-A, ``Industry Guidance for Implementing Performance-Based Option of 
10 CFR 50, Appendix J,'' October 2008, in addition to limitations and 
conditions of NEI 94-01, Revision 3-A. The amendments also deleted the 
listing of one-time exceptions previously granted to ILRT frequencies.
    Date of issuance: December 30, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: Unit 1--165; Unit 2--165. A publicly-available 
version is in ADAMS under Accession No. ML15309A073; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: March 31, 2015 (80 FR 
17092). The supplemental letter dated July 29, 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 December 30, 2015.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, and Docket No. 72-26 for 
Diablo Canyon Independent Spent Fuel Storage Installation (ISFSI), San 
Luis Obispo County, California

    Date of application for amendments: September 24, 2013, as 
supplemented by letters dated December 18, 2013, and May 15, 2014.
    Brief description of amendments: The amendments modified the 
licenses to reflect a grant of Section 161A of the Atomic Energy Act, 
to authorize the

[[Page 5504]]

licensee the authority to possess and use certain firearms, ammunition, 
and other devices such as large-capacity ammunition feeding devices, 
and to implement the NRC-approved security plan for Diablo Canyon Power 
Plant and Diablo Canyon ISFSI.
    Date of issuance: January 5, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 20 days.
    Amendment Nos.: Unit 1--222; Unit 2--224, ISFSI-4. A publicly-
available version is in ADAMS under package Accession No. ML15029A249; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Facility Operating License Nos. DPR-80 and DPR-82 and Special 
Nuclear Materials License No. SNM-2511: The amendments revised the 
Facility Operating Licenses and Special Nuclear Materials License.
    Date of initial notice in Federal Register: February 18, 2015 (80 
FR 8706).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 5, 2016.
    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: September 18, 2014, and supplemented by 
letter dated May 28, 2015.
    Description of amendment: The amendment authorizes a departure from 
VCSNS, Units 2 and 3 plant-specific AP1000 Design Control Document 
(DCD) Tier 2* material contained within the VCSNS Units 2 and 3 Updated 
Final Safety Analysis Report by relocating fire area rated fire 
barriers due to changes to the layout of the switchgear rooms and 
office area in the turbine building.
    Date of issuance: December 17, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 38. A publicly-available version is in ADAMS under 
Accession No. ML15313A052; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: January 6, 2015 (80 FR 
526).
    The Commission's related evaluation of the amendment is contained 
in the Safety Evaluation dated December 17, 2015. The supplemental 
letter dated May 28, 2015, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361, 50-362, 
and 72-41, San Onofre Nuclear Generating Station, Units 2 and 3, and 
Independent Spent Fuel Storage Installation (ISFSI), San Diego County, 
California

    Date of amendment request: August 28, 2013, as supplemented by 
letters dated December 31, 2013, May 15, 2014, and February 10, 2015.
    Brief description of amendments: The conforming amendments would 
permit the security personnel at San Onofre Nuclear Generating Station 
to transfer, receive possess, transport, import, and use certain 
firearms and large capacity ammunition feeding devices not previously 
permitted to be owned or possessed under NRC authority, notwithstanding 
certain local, state, or federal firearms laws, including regulations 
that prohibit such actions.
    Date of issuance: January 5, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 20 days.
    Amendment Nos.: Unit 2-232 and Unit 3-225: A publicly-available 
version is in ADAMS under Accession No. ML15027A221; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: February 18, 2015 (80 
FR 8701). The supplemental letter dated February 10, 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 January 5, 2016.
    No significant hazards consideration comments received: Yes, 
addressed in Safety Evaluation.

Tennessee Valley Authority (TVA), Docket No. 50-296, Browns Ferry 
Nuclear Plant (BFN), Unit 3, Limestone County, Alabama

    Date of amendment request: January 27, 2015, as supplemented by 
letters dated August 13 and October 23, 2015.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) for Limiting Condition for Operation (LCO) 3.4.9, 
``RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits.'' 
The amendment also revised Note 1 of TS Surveillance Requirement 
3.4.9.1 to change the vessel pressure from less than 312 pounds per 
square inch gauge (psig) to less than 313 psig to conform to the 
modified P/T limit curves. The amendment satisfied TVA's commitment to 
submit revised BFN, Unit 3, P/T limits prior to the start of the period 
of extended operation, as discussed in NRCs Safety Evaluation Report 
dated April 2006 (ADAMS Accession No. ML061030032), related to the 
license renewal of BFN, Units 1, 2, and 3.
    Specifically, the amendment revised the current sets of TS Figures 
3.4.9-1, ``Pressure/Temperature Limits for Mechanical Heatup, Cooldown 
following Shutdown, and Reactor Critical Operations,'' and 3.4.9-2, 
``Pressure/Temperature Limits for Reactor In-Service Leak and 
Hydrostatic Testing.'' The amendment replaced the current set valid up 
to 20 effective full-power years (EFPYs) with a new set valid up to 38 
EFPYs, and replaced the current set valid up to 28 EFPYs with a new set 
valid up to 54 EFPYs.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 278. A publicly available version is in ADAMS under 
Accession No. ML15344A321; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-68: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: May 5, 2015 (80 FR 
25720). The supplemental letters dated August 13 and October 23, 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 January 7, 2016.

[[Page 5505]]

    No significant hazards consideration comments received: Yes. The 
comment received on Amendment No. 278 is addressed in the Safety 
Evaluation dated January 7, 2016.

    Dated at Rockville, Maryland, this 21st day of January 2016.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2016-01771 Filed 2-1-16; 8:45 am]
 BILLING CODE 7590-01-P
5496 Federal Register / Vol. 81, No. 21 / Tuesday, February 2, 2016 / Notices

A. Opportunity To Request a Hearing statement of the alleged facts or expert any amendment unless the Commission
and Petition for Leave To Intervene opinion which support the contention finds an imminent danger to the health
Within 60 days after the date of and on which the requestor/petitioner or safety of the public, in which case it
publication of this notice, any person(s) intends to rely in proving the contention will issue an appropriate order or rule
whose interest may be affected by this at the hearing. The requestor/petitioner under 10 CFR part 2.
must also provide references to those A State, local governmental body,
action may file a request for a hearing
specific sources and documents of Federally-recognized Indian Tribe, or
and a petition to intervene with respect
which the petitioner is aware and on agency thereof, may submit a petition to
to issuance of the amendment to the
which the requestor/petitioner intends the Commission to participate as a party
subject facility operating license or
to rely to establish those facts or expert under 10 CFR 2.309(h)(1). The petition
combined license. Requests for a
opinion. The petition must include should state the nature and extent of the
hearing and a petition for leave to
sufficient information to show that a petitioner’s interest in the proceeding.
intervene shall be filed in accordance
genuine dispute exists with the The petition should be submitted to the
with the Commission’s ‘‘Agency Rules
applicant on a material issue of law or Commission by April 4, 2016. The
of Practice and Procedure’’ in 10 CFR
fact. Contentions shall be limited to petition must be filed in accordance
part 2. Interested person(s) should with the filing instructions in the
matters within the scope of the
consult a current copy of 10 CFR 2.309, ‘‘Electronic Submissions (E-Filing)’’
amendment under consideration. The
which is available at the NRC’s PDR, section of this document, and should
contention must be one which, if
located at One White Flint North, Room meet the requirements for petitions for
proven, would entitle the requestor/
O1–F21, 11555 Rockville Pike (first leave to intervene set forth in this
petitioner to relief. A requestor/
floor), Rockville, Maryland 20852. The section, except that under § 2.309(h)(2)
petitioner who fails to satisfy these
NRC’s regulations are accessible a State, local governmental body, or
requirements with respect to at least one
electronically from the NRC Library on contention will not be permitted to Federally-recognized Indian Tribe, or
the NRC’s Web site at http:// participate as a party. agency thereof does not need to address
www.nrc.gov/reading-rm/doc- Those permitted to intervene become the standing requirements in 10 CFR
collections/cfr/. If a request for a hearing parties to the proceeding, subject to any 2.309(d) if the facility is located within
or petition for leave to intervene is filed limitations in the order granting leave to its boundaries. A State, local
within 60 days, the Commission or a intervene, and have the opportunity to governmental body, Federally-
presiding officer designated by the participate fully in the conduct of the recognized Indian Tribe, or agency
Commission or by the Chief hearing with respect to resolution of thereof may also have the opportunity to
Administrative Judge of the Atomic that person’s admitted contentions, participate under 10 CFR 2.315(c).
Safety and Licensing Board Panel, will including the opportunity to present If a hearing is granted, any person
rule on the request and/or petition; and evidence and to submit a cross- who does not wish, or is not qualified,
the Secretary or the Chief examination plan for cross-examination to become a party to the proceeding
Administrative Judge of the Atomic of witnesses, consistent with NRC may, in the discretion of the presiding
Safety and Licensing Board will issue a regulations, policies and procedures. officer, be permitted to make a limited
notice of a hearing or an appropriate Petitions for leave to intervene must appearance pursuant to the provisions
order. be filed no later than 60 days from the of 10 CFR 2.315(a). A person making a
As required by 10 CFR 2.309, a date of publication of this notice. limited appearance may make an oral or
petition for leave to intervene shall set Requests for hearing, petitions for leave written statement of position on the
forth with particularity the interest of to intervene, and motions for leave to issues, but may not otherwise
the petitioner in the proceeding, and file new or amended contentions that participate in the proceeding. A limited
how that interest may be affected by the are filed after the 60-day deadline will appearance may be made at any session
results of the proceeding. The petition not be entertained absent a of the hearing or at any prehearing
should specifically explain the reasons determination by the presiding officer conference, subject to the limits and
why intervention should be permitted that the filing demonstrates good cause conditions as may be imposed by the
with particular reference to the by satisfying the three factors in 10 CFR presiding officer. Persons desiring to
following general requirements: (1) The 2.309(c)(1)(i)–(iii). make a limited appearance are
name, address, and telephone number of If a hearing is requested, and the requested to inform the Secretary of the
the requestor or petitioner; (2) the Commission has not made a final Commission by April 4, 2016.
nature of the requestor’s/petitioner’s determination on the issue of no
right under the Act to be made a party significant hazards consideration, the B. Electronic Submissions (E-Filing)
to the proceeding; (3) the nature and Commission will make a final All documents filed in NRC
extent of the requestor’s/petitioner’s determination on the issue of no adjudicatory proceedings, including a
property, financial, or other interest in significant hazards consideration. The request for hearing, a petition for leave
the proceeding; and (4) the possible final determination will serve to decide to intervene, any motion or other
effect of any decision or order which when the hearing is held. If the final document filed in the proceeding prior
may be entered in the proceeding on the determination is that the amendment to the submission of a request for
requestor’s/petitioner’s interest. The request involves no significant hazards hearing or petition to intervene, and
petition must also set forth the specific consideration, the Commission may documents filed by interested
contentions which the requestor/ issue the amendment and make it governmental entities participating
petitioner seeks to have litigated at the immediately effective, notwithstanding under 10 CFR 2.315(c), must be filed in
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proceeding. the request for a hearing. Any hearing accordance with the NRC’s E-Filing rule
Each contention must consist of a held would take place after issuance of (72 FR 49139; August 28, 2007). The E-
specific statement of the issue of law or the amendment. If the final Filing process requires participants to
fact to be raised or controverted. In determination is that the amendment submit and serve all adjudicatory
addition, the requestor/petitioner shall request involves a significant hazards documents over the internet, or in some
provide a brief explanation of the bases consideration, then any hearing held cases to mail copies on electronic
for the contention and a concise would take place before the issuance of storage media. Participants may not

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5498 Federal Register / Vol. 81, No. 21 / Tuesday, February 2, 2016 / Notices

(TS) Section 2.1.1, ‘‘Reactor Core SLs,’’ any plant systems perform a safety function. Exelon Nuclear, 4300 Winfield Road,
to reflect a lower reactor steam dome No new accident scenarios, failure Warrenville, IL 60555.
pressure stated for Reactor Core Safety mechanisms, or limiting single failures are NRC Acting Branch Chief: Justin C.
Limits (SLs) 2.1.1.1 and 2.1.1.2. introduced as a result of the proposed Poole.
change.
Specifically, the proposed amendment The proposed change does not introduce Exelon Generation Company, LLC, and
will reduce the reactor steam dome any new accident precursors, nor does it PSEG Nuclear LLC, Docket Nos. 50–277
pressure in TS SLs 2.1.1.1 and 2.1.1.2 involve any physical plant alterations or and 50–278, Peach Bottom Atomic
from 785 psig [pound per square inch changes in the methods governing normal Power Station, Units 2 and 3, York and
gage] to 685 psig. This change to TS plant operation. Also, the change does not Lancaster Counties, Pennsylvania
Section 2.1.1 was identified as a result impose any new or different requirements or
of General Electric Part 21 report SC05– eliminate any existing requirements. The Date of amendment request:
03, ‘‘Potential to Exceed Low Pressure change does not alter assumptions made in December 3, 2015. A publicly-available
Technical Specification Safety Limit.’’ the safety analysis. version is in ADAMS under Accession
This change is valid for the NRC- Therefore, the proposed change does not No. ML15337A413.
create the possibility of a new or different Description of amendment request:
approved pressure range pertinent to the kind of accident from any previously
critical power correlations applied to The amendments would revise the
evaluated.
the fuel types in use at LSCS. technical specification (TS) surveillance
3. Does the proposed change involve a
Basis for proposed no significant significant reduction in a margin of safety? requirements (SRs) associated with the
hazards consideration determination: Response: No. emergency diesel generator (EDG) fuel
As required by 10 CFR 50.91(a), the The margin of safety is established through oil transfer system. Specifically, the
licensee has provided its analysis of the the design of the plant structures, systems, amendments would allow for the
issue of no significant hazards and components, and through the parameters crediting of manual actions, in lieu of
consideration, which is presented for safe operation and setpoints for the automatic actions, without having to
below: actuation of equipment relied upon to declare the EDGs inoperable.
respond to transients and design basis Basis for proposed no significant
1. Does the proposed change involve a accidents.
significant increase in the probability or
hazards consideration determination:
Evaluation of the 10 CFR part 21 condition
consequences of an accident previously by General Electric determined that since the
As required by 10 CFR 50.91(a), the
evaluated? Minimum Critical Power Ratio improves licensee has provided its analysis of the
Response: No. during the PRFO [Pressure Regulator Failure issue of no significant hazards
The proposed change to the reactor steam Maximum Demand (Open)] transient, there is consideration, which is presented
dome pressure in the LSCS Reactor Core no decrease in the safety margin and below:
Safety Limits TS 2.1.1.1 and 2.1.1.2 does not therefore there is not a threat to fuel cladding
alter the use of the analytical methods used 1. Does the proposed change involve a
integrity. significant increase in the probability or
to determine the safety limits that have been The proposed change in reactor dome
previously reviewed and approved by the consequences of an accident previously
pressure supports the current safety margin, evaluated?
NRC. The proposed change is in accordance which protects the fuel cladding integrity
with an NRC approved critical power Response: No.
during a depressurization transient, but does The proposed change will revise SR 3.8.1.6
correlation methodology, and as such, not change the requirements governing
maintains required safety margins. The by adding a note to allow for procedurally
operation or availability of safety equipment controlled simple manual actions associated
proposed change does not adversely affect
assumed to operate to preserve the margin of with the fuel oil transfer system without
accident initiators or precursors, nor does it
safety. The change does not alter the behavior having to declare the EDG inoperable [under]
alter the design assumptions, conditions, or
of plant equipment, which remains administrative control. The fuel oil transfer
configuration of the facility or the manner in
which the plant is operated and maintained. unchanged. system is required to support continuous
The proposed change does not alter or The proposed change to Reactor Core operation of standby power sources. The
prevent the ability of structures, systems, and Safety Limits 2.1.1.1 and 2.1.1.2 is consistent surveillance provides assurance that the fuel
components (SSCs) from performing their with and within the capabilities of the oil transfer system is OPERABLE. The fuel oil
intended function to mitigate the applicable NRC approved critical power transfer system is not an initiator of any
consequences of an initiating event within correlation for the fuel designs in use at event previously evaluated. Therefore, the
the assumed acceptance limits. The proposed LSCS, Units 1 and 2. No setpoints at which probability of any accident previously
change does not require any physical change protective actions are initiated are altered by evaluated is not increased.
to any plant SSCs nor does it require any the proposed change. In the event of an accident, if simple
change in systems or plant operations. The The proposed change does not alter the manual actions were necessary to restore the
proposed change is consistent with the safety manner in which the safety limits are automatic feature of the EDG day tank fill,
analysis assumptions and resultant determined. This change is consistent with analysis shows that significant margin exists
consequences. plant design and does not change the TS to ensure that EDG operability would not be
Therefore, the proposed change does not operability requirements; thus, previously adversely affected. Although the proposed
involve a significant increase in the evaluated accidents are not affected by this change to allow simple manual actions could
probability or consequences of an accident proposed change. introduce additional potential malfunctions,
previously evaluated. Therefore, the proposed changes do not such that human error could result in the
2. Does the proposed change create the involve a significant reduction in a margin of potential to improperly realign the fuel oil
possibility of a new or different kind of safety. transfer system during a DBA [design-basis
accident from any accident previously accident], the improper realignment would
evaluated? The NRC staff has reviewed the be detected when the transfer of fuel oil from
Response: No. licensee’s analysis and, based on this the storage tank to the day tank did not occur
The proposed reduction in the reactor review, it appears that the three as expected and the error would be corrected
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dome pressure safety limit from 785 psig to standards of 10 CFR 50.92(c) are prior to having a significant impact.
685 psig is a change based upon previously satisfied. Therefore, the NRC staff The proposed change does not involve any
approved documents and does not involve proposes to determine that the physical changes to the structures, systems,
changes to the plant hardware or its or components (SSCs) in the plant. Further
requested amendments involve no the proposed change does not alter or prevent
operating characteristics. As a result, no new
failure modes are being introduced. significant hazards consideration. the ability of SSCs from performing their
There are no hardware changes nor are Attorney for licensee: Bradley J. intended function to mitigate the
there any changes in the method by which Fewell, Associate General Counsel, consequences of an event.

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5500 Federal Register / Vol. 81, No. 21 / Tuesday, February 2, 2016 / Notices

satisfied. Therefore, the NRC staff probability or consequences of an accident noticed. Accordingly, this notice
proposes to determine that the previously evaluated. supersedes the previous notice in its
amendment request involves no 2. Does the proposed amendment create entirety. The amendments would revise
the possibility of a new or different kind of
significant hazards consideration. the Emergency Plan for SSES to adopt
accident from any accident previously
Attorney for licensee: Ms. Kathryn M. evaluated? the Nuclear Energy Institute’s (NEI’s)
Sutton, Morgan, Lewis & Bockius LLC, Response: No. revised emergency action level (EAL)
1111 Pennsylvania Avenue NW., The proposed change revises the scheme described in NEI 99–01,
Washington, DC 20004–2514. description of the construction of composite Revision 6, ‘‘Development of Emergency
NRC Acting Branch Chief: John steel beam floors and roof in the auxiliary Action Levels for Non-Passive Reactors’’
McKirgan. building. The proposed change does not (ADAMS Accession No. ML12326A805),
change the design function, support, design, which was endorsed by the NRC as
Southern Nuclear Operating Company, or operation of mechanical and fluid systems.
Inc., Docket Nos. 52–025 and 52–026, documented in NRC letter dated March
The proposed change does not result in a
Vogtle Electric Generating Plant, Units 3 new failure mechanism for the pertinent 28, 2013 (ADAMS Accession No.
and 4, Burke County, Georgia structures or new accident precursors. As a ML12346A463). Supplemental changes
result, the design function of the structures in these amendments were discussed in
Date of amendment request: is not adversely affected by the proposed a September 23, 2015, public meeting
November 16, 2015. A publicly- change. held with Susquehanna Nuclear, LLC.
available version is in ADAMS under Therefore, the proposed amendment does The public meeting summary was
Accession No. ML15320A464. not create the possibility of a new or different issued October 9, 2015, and is available
Description of amendment request: kind of accident from any accident
in ADAMS under Accession No.
The requested amendment proposes to previously evaluated.
3. Does the proposed amendment involve ML15278A492. The additional
depart from Tier 2* information in the information, and the changes discussed
Updated Final Safety Analysis Report a significant reduction in a margin of safety?
Response: No. at the public meeting, are included in
related to the construction methods the two Susquehanna Nuclear, LLC
The proposed change is consistent with
used for the composite floors and roof ACI 349 and AISC N690. The design and letters dated October 15, 2015, and
of the auxiliary building. construction of the auxiliary building floors October 16, 2015. The revised
Basis for proposed no significant and roof remain in conformance with the Emergency Plan includes the
hazards consideration determination: requirements in ACI 349 and AISC N690.
appropriate plant-specific changes as a
As required by 10 CFR 50.91(a), the Therefore, the proposed amendment does
not involve a significant reduction in a result of an emergency operating
licensee has provided its analysis of the
margin of safety. procedure upgrade project and
issue of no significant hazards
corrective action in response to an NRC
consideration, which is presented The NRC staff has reviewed the Emergency Preparedness White Finding,
below: licensee’s analysis and, based on this documented in NRC Inspection Report
1. Does the proposed amendment involve review, it appears that the three No. 05000387/2015504 and 05000388/
a significant increase in the probability or standards of 10 CFR 50.92(c) are 2015504, dated June 22, 2015 (ADAMS
consequences of an accident previously satisfied. Therefore, the NRC staff
evaluated?
Accession Nos. ML15173A297 and
proposes to determine that the ML15181A332).
Response: No. amendment request involves no
The design functions of the nuclear island On June 1, 2015, the NRC staff issued
structures are to provide support, protection,
significant hazards consideration. an amendment changing the name on
and separation for the seismic Category I Attorney for licensee: Mr. M. Stanford the SSES license from PPL
mechanical and electrical equipment located Blanton, Balch & Bingham LLP, 1710 Susquehanna, LLC to Susquehanna
in the nuclear island. The nuclear island Sixth Avenue North Birmingham, AL Nuclear, LLC. This amendment was
structures are structurally designed to meet 35203–2015. issued subsequent to an order issued on
seismic Category I requirements as defined in NRC Acting Branch Chief: John
Regulatory Guide 1.29.
April 10, 2015, to SSES, approving an
McKirgan.
The use of [American Concrete Institute indirect license transfer.
(ACI)] 349 and [American Institute of Steel Susquehanna Nuclear, LLC, Docket Nos. Basis for proposed no significant
Construction (AISC)] N690 provides criteria 50–387 and 50–388, Susquehanna hazards consideration determination:
for the design, qualification, fabrication, and Steam Electric Station, Units 1 and 2 As required by 10 CFR 50.91(a), the
inspection of composite steel beam floors and (SSES), Luzerne County, Pennsylvania licensee has provided its analysis of the
roof in the auxiliary building. These issue of no significant hazards
structures continue to meet the applicable Date of amendment request: March
19, 2015, as supplemented by letters consideration, which is presented
portions of ACI 349 and AISC N690. The
proposed change does not have an adverse dated October 15, 2015, October 16, below, along with NRC edits in square
impact on the response of the nuclear island 2015, and January 8, 2016. Publicly- brackets:
structures to safe shutdown earthquake available versions are in ADAMS under 1. Does the proposed change involve a
ground motions or loads due to anticipated Package Accession Nos. ML15091A657, significant increase in the probability or
transients or postulated accident conditions. consequences of any accident previously
The change does not impact the support, ML15296A048, and ML15296A057, and
evaluated?
design, or operation of mechanical and fluid Accession No. ML16011A103, Response: No.
systems. There is no change to plant systems respectively. The proposed changes to the EAL scheme
or the response of systems to postulated Description of amendment request: to adopt the NRC-endorsed guidance in NEI
accident conditions. There is no change to The NRC staff previously made a 99–01, Revision 6, ‘‘Development of
the predicted radioactive releases due to proposed determination that the Emergency Action Levels for Non-Passive
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normal operation or postulated accident amendment request dated March 19, Reactors,’’ [and the additional plant-specific
conditions. The plant response to previously 2015, involved no significant hazards Emergency Plan changes] do not reduce the
evaluated accidents or external events is not capability to meet the emergency planning
adversely affected, nor does the change
consideration (80 FR 38762; July 7,
requirements established in 10 CFR 50.47
described create any new accident 2015). Subsequently, the supplemental and 10 CFR 50, Appendix E. The proposed
precursors. letter dated October 15, 2015, provided changes do not reduce the functionality,
Therefore, the proposed amendment does additional information that expanded performance, or capability of the ERO
not involve a significant increase in the the scope of the application as originally [Emergency Response Organization] to

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Federal Register / Vol. 81, No. 21 / Tuesday, February 2, 2016 / Notices 5501

respond in mitigating the consequences of 3. Does the proposed change involve a The ice condenser is credited in mitigating
any design basis accident. significant reduction in a margin of safety? the consequences of postulated Design Basis
The probability of a reactor accident Response: No. Accidents (DBAs) and remains capable of
requiring implementation of Emergency Plan The proposed changes to the EAL scheme performing its design basis functions. The
EALs has no relevance in determining to adopt the NRC-endorsed guidance in NEI proposed amendment to the SRs during the
whether the proposed changes to the EALs 99–01, Revision 6, [and the additional plant- first cycle of WBN Unit 2 operation does not
reduce the effectiveness of the Emergency specific Emergency Plan changes] do not change the ice condenser configuration or
Plan. As discussed in Section I.D, ‘‘Planning alter or exceed a design basis or safety limit. how it behaves in the event of a DBA. Thus
Basis,’’ of NUREG–0654, Revision 1, ‘‘Criteria There is no change being made to safety it is concluded that a significant increase in
for Preparation and Evaluation of analysis assumptions, safety limit, or limiting the consequences of an accident previously
Radiological Emergency Response Plans and safety system settings that would adversely evaluated will not occur as a result of the
Preparedness in Support of Nuclear Power affect plant safety as a result of the proposed proposed amendment.
Plants’’; changes. There are no changes to setpoints or Therefore, the proposed amendment does
. . . The overall objective of emergency environmental conditions of any SSC or the not involve a significant increase in the
response plans is to provide dose savings manner in which any SSC is operated. probability or consequences of an accident
(and in some cases immediate life saving) for Margins of safety are unaffected by the previously evaluated.
a spectrum of accidents that could produce proposed changes to adopt the NEI 99–01, 2. Does the proposed amendment create
offsite doses in excess of Protective Action Revision 6 EAL scheme guidance. The the possibility of a new or different kind of
Guides (PAGs). No single specific accident applicable requirements of 10 CFR 50.47 and accident from any accident previously
sequence should be isolated as the one for 10 CFR 50, Appendix E will continue to be evaluated?
which to plan because each accident could met. Response: No.
have different consequences, both in nature Therefore, the proposed changes do not The proposed amendment does not create
and degree. Further, the range of possible involve any reduction in a margin of safety. the possibility of a new or different kind of
selection for a planning basis is very large, accident from any accident previously
The NRC staff has reviewed the evaluated. The proposed amendment does
starting with a zero point of requiring no
planning at all because significant offsite licensee’s analysis and, based on this not introduce any new modes of plant
radiological accident consequences are review, it appears that the three operation, change the design function of the
unlikely to occur, to planning for the worst standards of 10 CFR 50.92(c) are ice condenser or any other Structure System
possible accident, regardless of its extremely satisfied. Therefore, the NRC staff or Component (SSC), or change the mode of
low likelihood. . . . proposes to determine that the operation of the ice condenser or any other
Therefore, risk insights are not considered SSC. There are no new equipment failure
amendment request involves no modes or malfunctions created as the ice
for any specific accident initiation or significant hazards consideration. condenser and ice condenser lower inlet
progression in evaluating the proposed Attorney for licensee: Damon D. Obie, doors continue to operate in the same
changes. Associate General Counsel, Talen manner assumed in the accident analysis.
The proposed changes do not involve any
Energy Supply, LLC, 835 Hamilton St., The ice condenser is a passive post-accident
physical changes to plant equipment or heat removal feature that is not an accident
systems, nor do they alter the assumptions of Suite 150, Allentown, PA 18101.
NRC Branch Chief: Douglas A. initiator.
any accident analyses. The proposed changes Therefore, the proposed amendment does
do not adversely affect accident initiators or Broaddus.
not create the possibility of a new or different
precursors nor do they alter the design Tennessee Valley Authority, Docket No. kind of accident from any previously
assumptions, conditions, and configuration evaluated.
50–391, Watts Bar Nuclear Plant (WBN),
or the manner in which the plants are 3. Does the proposed amendment involve
operated and maintained. The proposed Unit 2, Rhea County, Tennessee
a significant reduction in a margin of safety?
changes do not adversely affect the ability of Date of amendment request: Response: No.
Structures, Systems, or Components (SSCs) December 15, 2015. A publicly-available Ice condensers have been in-service at nine
to perform their intended safety functions in version is in ADAMS under Accession nuclear units in the United States for many
mitigating the consequences of an initiating No. ML15350A250. years. Operating experience has shown that
event within the assumed acceptance limits. Brief description of amendment an 18-month surveillance frequency for
Therefore, the proposed changes do not evaluating operability is appropriate for the
involve a significant increase in the
request: The amendment would revise
the technical specification (TS) lower inlet doors. The proposed amendment
probability or consequences of an accident to perform a revised schedule of lower inlet
previously evaluated. surveillance requirements (SRs) for the door surveillances in the first cycle before
2. Does the proposed change create the WBN, Unit 2, ice condenser lower inlet transitioning to the standard 18-month
possibility of a new or different kind of doors. surveillance frequency does not result in a
accident from any accident previously Basis for proposed no significant significant reduction in the margin of safety.
evaluated? hazards consideration determination: Therefore, since there is no adverse impact
Response: No. As required by 10 CFR 50.91(a), the of this amendment on the WBN Unit 2 safety
The proposed changes to the EAL scheme licensee has provided its analysis of the analysis, there is no significant reduction in
to adopt the NRC-endorsed guidance in NEI issue of no significant hazards the margin of safety of the plant.
99–01, Revision 6, [and the additional plant-
specific Emergency Plan changes] do not
consideration, which is presented The NRC staff has reviewed the
involve any physical changes to plant below: licensee’s analysis and, based on this
systems or equipment. The proposed changes 1. Does the proposed amendment involve review, it appears that the three
do not involve the addition of any new plant a significant increase in the probability or standards of 10 CFR 50.92(c) are
equipment. The proposed changes will not consequence of an accident previously satisfied. Therefore, the NRC staff
alter the design configuration, or method of evaluated? proposes to determine that the
operation of plant equipment beyond its Response: No. amendment request involves no
normal functional capabilities. All ERO The ice condenser is a passive heat significant hazards consideration.
wgreen on DSK2VPTVN1PROD with NOTICES

functions will continue to be performed as removal plant feature. The proposed Attorney for licensee: Scott A. Vance,
required. The proposed changes do not create amendment to the TS 3.6.12 does not change
any new credible failure mechanisms, the design, physical features or the function
Associate General Counsel, Nuclear,
malfunctions, or accident initiators. of the ice condenser or the ice condenser Tennessee Valley Authority, 400 West
Therefore, the proposed changes do not doors. The ice condenser is not an accident Summit Hill Drive, WT 6A–K,
create the possibility of a new or different initiator, thus the proposed amendment does Knoxville, TN 37902.
kind of accident from those that have been not increase the probability of an accident NRC Branch Chief: Benjamin G.
previously evaluated. previously evaluated. Beasley.

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Document Created: 2016-02-02 00:32:15
Document Modified: 2016-02-02 00:32:15
Category		Regulatory Information
Collection		Federal Register
sudoc Class		AE 2.7: GS 4.107: AE 2.106:
Publisher		Office of the Federal Register, National Archives and Records Administration
Section		Notices
Action		Biweekly notice.
Dates		Comments must be filed by March 3, 2016. A request for a hearing must be filed by April 4, 2016.
Contact		Janet Burkhardt, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-1384, email: [email protected]
FR Citation		81 FR 5495