Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 66301-66314 [2016-23097]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 81, Number 187 (Tuesday, September 27, 2016)]
[Notices]
[Pages 66301-66314]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-23097]


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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0202]


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.

[[Page 66302]]

    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from August 30, 2016, to September 12, 2016. The 
last biweekly notice was published on September 13, 2016.

DATES: Comments must be filed by October 27, 2016. A request for a 
hearing must be filed by November 28, 2016.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2016-0202. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: 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: Janet.Burkhardt@nrc.gov.

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to NRC-2016-0202, facility name, unit number(s), plant 
docket number, application date, and subject when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2016-0202.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in 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-2016-0202 facility name, unit 
number(s), plant docket number, 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 will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Sec.  50.92 of title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period if 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. If the Commission 
takes 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. If the Commission makes a final no significant hazards 
consideration determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.

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

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and a petition to intervene (petition) 
with respect to issuance of the amendment to the subject facility 
operating license or combined license. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested persons should consult a 
current copy of 10 CFR 2.309, which is available at the NRC's PDR, 
located at One White Flint North, Room O1-F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852. The NRC's regulations are 
accessible electronically from the NRC Library on the NRC's Web site at 
https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a petition 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 petition; and the 
Secretary or the Chief Administrative Judge of the

[[Page 66303]]

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 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 
petitioner; (2) the nature of the petitioner's right under the Act to 
be made a party to the proceeding; (3) the nature and extent of the 
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 petitioner's interest. The petition 
must also set forth the specific contentions which the 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 
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 petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion to support 
its position on the issue. 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 petitioner 
to relief. A 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 the NRC's 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 November 28, 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 set forth in this section, except that under 10 CFR 
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. Details regarding the opportunity to 
make a limited appearance will be provided by the presiding officer if 
such sessions are scheduled.

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 (hereinafter 
``petition''), 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, as amended at 77 
FR 46562; August 3, 2012). The E-Filing process requires participants 
to submit and serve all adjudicatory documents over the internet, or in 
some cases to mail copies on electronic storage media. Participants may 
not submit paper copies of their filings unless they seek an exemption 
in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a petition (even 
in instances in which the participant, or its counsel or 
representative, already holds an NRC-issued digital ID certificate). 
Based upon this information, the Secretary will establish an electronic 
docket for the hearing in this proceeding if the Secretary has not 
already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are available on the NRC's public Web site at 
https://www.nrc.gov/site-help/e-submittals/

[[Page 66304]]

adjudicatory-sub.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 Electronic 
Filing Help Desk will not be able to offer assistance in using unlisted 
software.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a petition. 
Submissions should be in Portable Document Format (PDF). Additional 
guidance on PDF submissions is available on the NRC's public Web site 
at https://www.nrc.gov/site-help/electronic-sub-ref-mat.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 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 Electronic 
Filing Help Desk through the ``Contact Us'' link located on the NRC's 
public Web site at https://www.nrc.gov/site-help/e-submittals.html, by 
email to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m. 
and 7 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 
stating why there is good cause for not filing electronically and 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, in some instances, a petition 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.
    The Commission will issue a notice or order granting or denying a 
hearing request or intervention petition, designating the issues for 
any hearing that will be held and designating the Presiding Officer. A 
notice granting a hearing will be published in the Federal Register and 
served on the parties to the hearing.
    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.

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

    Date of amendment request: July 14, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16196A419.
    Description of amendment request: The amendment would change 
Technical Specification (TS) 5.5.6, ``Inservice Testing [IST] 
Program,'' to remove requirements duplicated in American Society of 
Mechanical Engineers (ASME) Code for Operations and Maintenance of 
Nuclear Power Plants (OM Code), Case OMN-20, ``Inservice Test 
Frequency.'' This change, thereby, will then adopt Technical 
Specification Task Force (TSTF) TSTF-545, Revision 3, ``TS Inservice 
Testing Program Removal & Clarify SR [Surveillance Requirement] Usage 
Rule Application to Section 5.5 Testing.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS Chapter 5, ``Administrative 
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating 
the ``Inservice Testing Program'' specification. Most requirements 
in the Inservice Testing Program are removed, as they are 
duplicative of requirements in the ASME OM Code, as clarified by 
Code Case OMN-20, ``Inservice Test Frequency,'' which has been 
approved for use at Columbia. The remaining requirements in the 
Section 5.5 IST Program are eliminated because the NRC has 
determined their inclusion in the TS is contrary to regulations. A 
new defined term, ``Inservice Testing Program,'' is added to the TS, 
which references the requirements of 10 CFR 50.55a(f).
    Performance of inservice testing is not an initiator to any 
accident previously evaluated. As a result, the probability of 
occurrence of an accident is not significantly affected by the 
proposed change. Inservice test frequencies under Code Case OMN-20 
are equivalent to the current testing period allowed by the TS with 
the exception that testing frequencies greater than 2 years may be 
extended by up to 6 months to facilitate test scheduling and 
consideration of plant operating conditions that may not be suitable 
for performance of the required testing. The testing frequency 
extension will not affect the ability of the components to mitigate 
any accident previously evaluated as the components are required to 
be operable during the testing period extension. Performance of 
inservice tests utilizing the

[[Page 66305]]

allowances in OMN-20 will not significantly affect the reliability 
of the tested components. As a result, the availability of the 
affected components, as well as their ability to mitigate the 
consequences of accidents previously evaluated, is not affected.
    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 design or configuration 
of the plant. The proposed change does not involve a physical 
alteration of the plant; no new or different kind of equipment will 
be installed. The proposed change does not alter the types of 
inservice testing performed. In most cases, the frequency of 
inservice testing is unchanged. However, the frequency of testing 
would not result in a new or different kind of accident from any 
previously evaluated since the testing methods are not altered.
    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 proposed change eliminates some requirements from the TS in 
lieu of requirements in the ASME Code, as modified by use of Code 
Case OMN-20. Compliance with the ASME Code is required by 10 CFR 
50.55a. The proposed change also allows inservice tests with 
frequencies greater than 2 years to be extended by 6 months to 
facilitate test scheduling and consideration of plant operating 
conditions that may not be suitable for performance of the required 
testing. The testing frequency extension will not affect the ability 
of the components to respond to an accident as the components are 
required to be operable during the testing period extension. The 
proposed change will eliminate the existing TS SR 3.0.3 allowance to 
defer performance of missed inservice tests up to the duration of 
the specified testing frequency, and instead will require an 
assessment of the missed test on equipment operability. This 
assessment will consider the effect on a margin of safety (equipment 
operability). Should the component be inoperable, the Technical 
Specifications provide actions to ensure that the margin of safety 
is protected. The proposed change also eliminates a statement that 
nothing in the ASME Code should be construed to supersede the 
requirements of any TS. The NRC has determined that statement to be 
incorrect. However, elimination of the statement will have no effect 
on plant operation or safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: July 28, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16210A528.
    Description of amendment request: The amendment would revise the 
current Columbia Emergency Plan Emergency Action Level scheme to one 
based on Nuclear Energy Institute (NEI) guidance established in NEI 99-
01, ``Development of Emergency Action Levels for Non-Passive 
Reactors,'' Revision 6, which has been endorsed by the NRC.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment affects the Columbia Generating Station 
(Columbia) Emergency Plan (EP) and associated Emergency Action 
Levels (EALs); it does not alter the Operating License or the 
Technical Specifications. The proposed amendment does not change the 
design function of any system, structure, or component and does not 
change the way the plant is maintained or operated. The proposed 
amendment does not affect any accident mitigating feature or 
increase the likelihood of malfunction for plant structures, 
systems, and components.
    The proposed amendment will not change any of the analyses 
associated with the Columbia Final Safety Analysis Report Chapter 15 
accidents because plant operation, structures, systems, components, 
accident initiators, and accident mitigation functions remain 
unchanged.
    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 affects the Columbia EP and associated 
EALs; it does not change the design function of any system, 
structure, or component and does not change the way the plant is 
operated or maintained. The proposed amendment does not create a 
credible failure mechanism, malfunction, or accident initiator not 
already considered in the design and licensing basis.
    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.
    Margin of safety is associated with the ability of the fission 
product barriers (i.e., fuel cladding, reactor coolant system 
pressure boundary, and containment structure) to limit the level of 
radiation dose to the public. The proposed amendment does not impact 
operation of the plant and no accident analyses are affected by the 
proposed amendment. The proposed amendment does not affect the 
Technical Specifications or the method of operating the plant. 
Additionally, the proposed amendment will not relax any criteria 
used to establish safety limits and will not relax any safety system 
settings. The safety analysis acceptance criteria are not affected 
by this amendment. The proposed amendment will not result in plant 
operation in a configuration outside the design basis. The proposed 
amendment does not adversely affect systems that respond to safely 
shut down the plant and to maintain the plant in a safe shutdown 
condition.
    Therefore, the proposed amendment does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: Robert J. Pascarelli.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3 (IP3), Westchester County, New York

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant (FitzPatrick), Oswego County, New York

    Date of amendment request: August 16, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16230A308.
    Description of amendment request: The amendment would transfer the 
beneficial interest in the Power Authority of the State of New York 
(PASNY) Master Decommissioning

[[Page 66306]]

Trust (Master Trust), including all rights and obligations thereunder, 
held by PASNY for IP3 and FitzPatrick to Entergy Nuclear Operations, 
Inc. (ENO). ENO also requests the NRC's consent to amendments to the 
Master Decommissioning Trust Agreement dated July 25, 1990, as amended 
(Master Trust Agreement), governing the Master Trust to facilitate this 
transfer. Finally, ENO seeks approval of license amendments to modify 
the existing trust-related license conditions to reflect the proposed 
transfer of the Master Trust to ENO and to delete other conditions so 
as to apply the requirements of 10 CFR 50.75(h)(1). ENO and Exelon 
Generation Company, LLC. (Exelon), jointly filed an application for a 
direct license transfer of FitzPatrick to Exelon on August 18, 2016. A 
separate Federal Register notice details the NRC's consideration of 
approval for the FitzPatrick 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:

    1. Do the proposed amendments involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The requested changes delete certain license conditions 
pertaining to the decommissioning trust agreements currently in 
sections 2.Q to 2.X of the IP3 Operating License and sections 2.H to 
2.O of the FitzPatrick Operating License. In addition, conforming 
changes to 2.W and 2.X of the IP3 Operating License and 2.P and 2.Q 
of the FitzPatrick Operating License are necessary [to] reflect the 
transfer of the Master Trust from PASNY to ENO.
    The requested changes are consistent with the types of license 
amendments permitted in 10 CFR 50.75(h)(5).
    The regulations of 10 CFR 50.75(h)(4) state that ``Unless 
otherwise determined by the Commission with regard to a specific 
application, the Commission has determined that any amendment to the 
license of a utilization facility that does no more than delete 
specific license conditions relating to the terms and conditions of 
decommissioning trust agreements involves `no significant hazards 
consideration.' ''
    In addition the requested changes seek changes to the Master 
Trust agreement only to the extent that they replace PASNY, a non-
licensee, with ENO, a licensee. No other changes to the Master Trust 
agreement are contemplated.
    This request involves changes that are administrative in nature. 
No actual plant equipment or accident analyses will be affected by 
the proposed changes.
    Therefore, the proposed amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed amendments create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This request involves administrative changes to licenses that 
will be consistent with the NRC's regulations at 10 CFR 50.75(h) and 
to change the name of the entity responsible under the Master Trust 
for decommissioning from a non-licensee to a licensee.
    No actual plant equipment or accident analyses will be affected 
by the proposed changes and no failure modes not bounded by 
previously evaluated accidents will be created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed amendments involve a significant reduction in 
a margin of safety?
    Response: No.
    The request involves administrative changes to the licenses that 
will be consistent with the NRC's regulations at 10 CFR 50.75(h) and 
to change the name of the entity responsible under the Master Trust 
for decommissioning from a non-licensee to a licensee.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers to limit the level of radiation doses 
to the public. No actual plant equipment or accident analyses will 
be affected by the proposed change. Additionally, the proposed 
changes will not relax any criteria used to establish safety limits, 
will not relax any safety systems settings, or will not relax the 
bases for any limiting conditions of operation.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

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

    Date of amendment request: August 22, 2016, as supplemented by 
letter dated September 8, 2016. Publicly-available versions are in 
ADAMS under Accession Nos. ML16235A195 and ML16252A351, respectively.
    Description of amendment request: The proposed amendment would 
replace existing license condition 2.C.(4) with a new license condition 
to state that technical specification (TS) surveillance requirement 
(SR) 3.1.4.3 is not required for control rod drive 13 (CRD-13) during 
cycle 25 until the next entry into Mode 3. In addition, the condition 
would state that CRD-13 seal leakage shall be repaired prior to 
entering Mode 2, following the next Mode 3 entry, and that the reactor 
shall be shut down if CRD-13 seal leakage exceeds two gallons per 
minute. The proposed amendment also requests replacement of the 
obsolete note in TS SR 3.1.4.3 with a note to clarify that TS SR 
3.1.4.3 is not required to be performed or met for CRD-13 during cycle 
25 provided CRD-13 is administratively declared immovable, but 
trippable, and Condition D is entered for CRD-13.
    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 license amendment replaces an obsolete license 
condition concerning CRD-22 testing that applied only to operating 
cycle 21 with a new license condition to forgo the remaining two 
required surveillance tests of CRD-13 from the PNP TS surveillance 
requirement for partial movement every 92 days during cycle 25. 
Since CRD-13 remains trippable, the proposed license condition does 
not affect or create any accident initiators or precursors. As such, 
the proposed license condition does not increase the probability of 
an accident.
    The proposed license amendment does not increase the 
consequences of an accident. The ability to move a full-length 
control rod by its drive mechanism is not an initial assumption used 
in the safety analyses. The safety analyses assume full-length 
control rod insertion, except the most reactive rod, upon reactor 
trip. The surveillance requirement performed during the last 
refueling outage verified control rod drop times are within accident 
analysis assumptions. ENO [Entergy Nuclear Operations] has 
determined that CRD seal leakage does not increase the likelihood of 
an untrippable control rod. The assumptions of the safety analyses 
will be maintained, and the consequences of an accident will not be 
increased.
    Therefore, operation of the facility in accordance with the 
proposed license condition would 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?

[[Page 66307]]

    Response: No.
    The proposed license amendment does not involve a physical 
alteration of any structure, system or component (SSC) or change the 
way any SSC is operated. The proposed license condition does not 
involve operation of any required SSCs in a manner or configuration 
differently from those previously recognized or evaluated. No new 
failure mechanisms would be introduced by the requested SR interval 
extension.
    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 license amendment does not affect trippability of 
the control rod. It will have the same capability to mitigate an 
accident as it had prior to the proposed license condition.
    Therefore, the proposed amendment would 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: Jeanne Cho, Senior Counsel, Entergy Nuclear 
Operations, Inc., 440 Hamilton Ave., White Plains, NY 10601.
    NRC Branch Chief: David J. Wrona.

Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear 
Generating Station (OCNGS), Ocean County, New Jersey; and Docket No. 
50-220, Nine Mile Point Nuclear Station, Unit 1 (NMP1), Oswego County, 
New York

    Date of amendment request: August 1, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16215A128.
    Description of amendment request: The amendments would revise 
OCNGS's Technical Specification (TS) Section 2.1, ``Safety Limit--Fuel 
Cladding Integrity,'' and NMP1's TS Section 2.1.1, ``Fuel Cladding 
Integrity,'' to reduce the steam dome pressure.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with NRC edits in [brackets]:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the OCNGS TS for the reactor steam dome 
pressure in Reactor Core Safety Limits 2.1.A and 2.1.B 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. 
Additionally, the proposed change to NMP1 for the reactor steam dome 
pressure in Reactor Core Safety Limits 2.1.1.a and 2.1.1.b 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.
    Lowering the value of reactor steam dome pressure in the TS has 
no physical effect on plant equipment and therefore, no impact on 
the course of plant transients. The change is an analytical exercise 
to demonstrate the applicability of correlations and methodologies. 
There are no known operational or safety benefits.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed reduction in the reactor dome pressure safety limit 
from 800 psia [pounds per square inch absolute] to 700 psia 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 changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in 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 GE [General 
Electric] determined that since the MCPR [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.A and 
2.1.B is consistent with and within the capabilities of the 
applicable NRC approved critical power correlation for the fuel 
designs in use at OCNGS. Additionally, the proposed change to 
Reactor Core Safety Limits 2.1.1.a and 2.1.1.b is consistent with 
and within the capabilities of the NRC approved critical power 
correlation for the fuel designs in use at NMP1. 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Acting Branch Chief: Shaun M. Anderson.

Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment request: July 21, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16208A076.
    Description of amendment request: The proposed changes are 
consistent

[[Page 66308]]

with the NRC-approved Technical Specifications Task Force (TSTF) 
Traveler, TSTF-545, Revision 3, ``TS [Technical Specification] 
Inservice Testing [IST] Program Removal & Clarify SR [Surveillance 
Requirement] Usage Rule Application to Section 5.5 Testing.'' The 
proposed change would revise the TSs to eliminate the Section 5.5.6, 
``Inservice Testing Program.'' A new defined term, ``INSERVICE TESTING 
PROGRAM,'' would be added to the TS Definitions section. TS SRs that 
currently refer to the Inservice Testing Program from Section 5.5.6 
would be revised to refer to the new defined term, ``INSERVICE TESTING 
PROGRAM.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS Chapter 5, ``Administrative 
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating 
the ``Inservice Testing Program'' specification. Most requirements 
in the IST Program are removed, as they are duplicative of 
requirements in the ASME [American Society of Mechanical Engineers] 
OM [Operation and Maintenance] Code, as clarified by Code Case OMN-
20, ``Inservice Test Frequency.'' The remaining requirements in the 
Section 5.5.6 IST Program are eliminated because the NRC has 
determined their inclusion in the TS is contrary to regulations. A 
new defined term, ``Inservice Testing Program,'' is added to the TS, 
which references the requirements of 10 CFR 50.55a(f).
    Performance of IST is not an initiator to any accident 
previously evaluated. As a result, the probability of occurrence of 
an accident is not significantly affected by the proposed change. 
Inservice test frequencies under Code Case OMN-20 are equivalent to 
the current testing period allowed by the TS with the exception that 
testing frequencies greater than 2 years may be extended by up to 6 
months to facilitate test scheduling and consideration of plant 
operating conditions that may not be suitable for performance of the 
required testing. The testing frequency extension will not affect 
the ability of the components to mitigate any accident previously 
evaluated as the components are required to be operable during the 
testing period extension. Performance of inservice tests utilizing 
the allowances in OMN-20 will not significantly affect the 
reliability of the tested components. As a result, the availability 
of the affected components, as well as their ability to mitigate the 
consequences of accidents previously evaluated, is not affected.
    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 design or configuration 
of the plant. The proposed change does not involve a physical 
alteration of the plant; no new or different kind of equipment will 
be installed. The proposed change does not alter the types of 
inservice testing performed. In most cases, the frequency of IST is 
unchanged. However, the frequency of testing would not result in a 
new or different kind of accident from any previously evaluated 
since the testing methods are not altered.
    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 proposed change eliminates some requirements from the TS in 
lieu of requirements in the ASME Code, as modified by use of Code 
Case OMN-20. Compliance with the ASME Code is required by 10 CFR 
50.55a. The proposed change also allows inservice tests with 
frequencies greater than 2 years to be extended by 6 months to 
facilitate test scheduling and consideration of plant operating 
conditions that may not be suitable for performance of the required 
testing. The testing frequency extension will not affect the ability 
of the components to respond to an accident as the components are 
required to be operable during the testing period extension. The 
proposed change will eliminate the existing TS SR 3.0.3 allowance to 
defer performance of missed inservice tests up to the duration of 
the specified testing frequency, and instead will require an 
assessment of the missed test on equipment operability. This 
assessment will consider the effect on a margin of safety (equipment 
operability). Should the component be inoperable, the TS provide 
actions to ensure that the margin of safety is protected. The 
proposed change also eliminates a statement that nothing in the ASME 
Code should be construed to supersede the requirements of any TS. 
The NRC has determined that statement to be incorrect. However, 
elimination of the statement will have no effect on plant operation 
or safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Robert B. Haemer, Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: David J. Wrona.

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

    Date of amendment request: August 11, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16224B122.
    Description of amendment request: The amendment request proposes 
changes to plant-specific Tier 2 information incorporated into the 
Updated Final Safety Analysis Report (UFSAR), and involves changes to 
combined license Appendix C (and corresponding plant-specific Tier 1 
information). The proposed changes are to information identifying the 
frontal face area and screen surface area for the In-Containment 
Refueling Water Storage Tank (IRWST) screens, the location and 
dimensions of the protective plate located above the containment 
recirculation (CR) screens, and increasing the maximum Normal Residual 
Heat Removal System flowrate through the IRWST and CR screens. Pursuant 
to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of 
the design as certified in the 10 CFR part 52, appendix D, design 
certification rule is also requested for the plant-specific Design 
Control Document Tier 1 material departures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with the NRC staff's edits in 
square brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the location and dimensions of the 
protective plate continues to provide sufficient space surrounding 
the containment recirculation screens for debris to settle before 
reaching the screens as confirmed by an evaluation demonstrating 
that the protective plate continues to fulfill its design function 
of preventing debris from reaching the screens. In addition, the 
increase to the minimum IRWST screen size reinforces the ability of 
the screens to perform their design function with the increased 
[Residual Heat Removal System (RNS)] maximum flowrate proposed. The 
proposed changes do not adversely affect any accident initiating 
component, and thus the probabilities of the accidents previously 
evaluated are not affected. The affected equipment does not 
adversely affect the ability of equipment to contain radioactive 
material. Because the proposed change does not affect a release path 
or increase the

[[Page 66309]]

expected dose rates, the potential radiological releases in the 
UFSAR accident analyses are unaffected.
    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 activity to change the location and dimensions of 
the protective plate above the containment recirculation screens, to 
change the minimum IRWST screen size, and to increase the maximum 
RNS flowrate through the IRWST and CR screens does not alter the 
method in which safety functions are accomplished. The analyses 
demonstrate that the screens are able to perform their functions in 
a similar manner and perform adequately in response to an accident, 
and no new failure modes are introduced 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 to the design does not change any of the 
codes or standards to which the IRWST screens, containment 
recirculation screens, and containment recirculation screen 
protective plate are designed as documented in the UFSAR. The 
containment recirculation screen protective plate continues to 
prevent debris from reaching the CR screens, and the IRWST and CR 
screens maintain their ability to block debris while at the proposed 
increase in RNS maximum flowrate.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes.
    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: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.

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

    Date of amendment request: August 23, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16236A265.
    Description of amendment request: The amendment request proposes 
changes to the Fire Pump Head and Diesel Fuel Day Tank. Because, this 
proposed change requires a departure from Tier 1 information in the 
Westinghouse Electric Company's AP1000 Design Control Document (DCD), 
the licensee also requested an exemption from the requirements of the 
Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
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 increase in head pressure by the proposed change to the fire 
protection system (FPS) motor-driven and diesel-driven fire pumps 
maintains compliance with National Fire Protection Association 
(NFPA) Standard NFPA-14, Standard for the Installation of Standpipe, 
Private Hydrants, and Hose Systems, 2000 Edition, requirements by 
providing adequate pressure in the standpipe and automatic sprinkler 
system to maintain the ability to fight and/or contain a postulated 
fire. The proposed change to the diesel-driven fire pump fuel day 
tank volume maintains the availability of the diesel-driven fire 
pump for service upon failure of the electric motor-driven fire pump 
or a loss of offsite power by providing a fuel day tank that is 
reserved exclusively for the diesel-driven pump and meets the 
minimum capacity requirements of NFPA 20, Standard for the 
Installation of Stationary Pumps for Fire Protection, 1999 Edition. 
These changes do not affect the operation of any systems or 
equipment that initiate an analyzed accident or alter any 
structures, systems, and [components (SSCs)] accident initiator or 
initiating sequence of events.
    These changes have no adverse impact on the support, design, or 
operation of mechanical and fluid systems. The response of systems 
to postulated accident conditions is not adversely affected by the 
proposed changes. There is no change to the predicted radioactive 
releases due to normal operation or postulated accident conditions. 
Consequently, the plant response to previously evaluated accidents 
is not impacted, nor does the proposed change 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 changes do not affect the operation of any systems 
or equipment that may initiate a new or different kind of accident, 
or alter any SSC such that a new accident initiator or initiating 
sequence of events is created. The proposed changes to the fire pump 
performance specifications and fire pump fuel day tank volume do not 
affect any safety-related equipment, nor do they add any new 
interface to safety-related SSCs. No system or design function or 
equipment qualification is affected by this change. The changes do 
not introduce a new failure mode, malfunction, or sequence of events 
that could affect safety or safety-related equipment.
    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 changes maintain compliance with the applicable 
Codes and Standards, thereby maintaining the margin of safety 
associated with these SSCs. The proposed changes do not alter any 
applicable design codes, code compliance, design function, or safety 
analysis. Consequently, no safety analysis or design basis 
acceptance limit/criterion is challenged or exceeded by the proposed 
change, thus the margin of safety is not reduced.
    Because no safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by these changes, no margin of 
safety is reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

Southern Nuclear Operating Company, Inc., Docket Nos. 50-321 and 50-
366, Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County, 
Georgia

    Date of amendment request: July 28, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16214A252.
    Description of amendment request: The amendments would revise the 
technical specifications (TSs) at the Edwin I. Hatch Nuclear Plant, 
Units 1 and 2, to eliminate the ``lnservice Testing Program'' from TS 
5.5, ``Programs and Manuals,'' and add a new defined term, ``INSERVICE 
TESTING PROGRAM,'' to TS 1.1, ``Definitions.'' This request is 
submitted in accordance with Technical

[[Page 66310]]

Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, ``TS 
lnservice Testing Program Removal & Clarify SR [Surveillance 
Requirement] Usage Rule Application to Section 5.5 Testing.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS Chapter 5, ``Administrative 
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating 
the ``lnservice Testing Program'' specification. Most requirements 
in the lnservice Testing Program are removed, as they are 
duplicative of requirements in the ASME OM [American Society of 
Mechanical Engineers Operation and Maintenance] Code, as clarified 
by Code Case OMN-20, ``lnservice Test Frequency.'' The remaining 
requirements in the Section 5.5 IST [Inservice Testing] Program are 
eliminated because the NRC has determined their inclusion in the TS 
is contrary to regulations. A new defined term, ``INSERVICE TESTING 
PROGRAM,'' is added to the TS, which references the requirements of 
10 CFR 50.55a(f).
    Performance of inservice testing is not an initiator to any 
accident previously evaluated. As a result, the probability of 
occurrence of an accident is not significantly affected by the 
proposed change. lnservice test frequencies under Code Case OMN-20 
are equivalent to the current testing period allowed by the TS with 
the exception that testing frequencies greater than 2 years may be 
extended by up to 6 months to facilitate test scheduling and 
consideration of plant operating conditions that may not be suitable 
for performance of the required testing. The testing frequency 
extension will not affect the ability of the components to mitigate 
any accident previously evaluated as the components are required to 
be operable during the testing period extension. Performance of 
inservice tests utilizing the allowances in OMN-20 will not 
significantly affect the reliability of the tested components. As a 
result, the availability of the affected components, as well as 
their ability to mitigate the consequences of accidents previously 
evaluated, is not affected.
    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 design or configuration 
of the plant. The proposed change does not involve a physical 
alteration of the plant; no new or different kind of equipment will 
be installed. The proposed change does not alter the types of 
inservice testing performed. In most cases, the frequency of 
inservice testing is unchanged. However, the frequency of testing 
would not result in a new or different kind of accident from any 
previously evaluated since the testing methods are not altered.
    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 eliminates some requirements from the TS in 
lieu of requirements in the ASME Code, as modified by use of Code 
Case OMN-20. Compliance with the ASME Code is required by 10 CFR 
50.55a. The proposed change also allows inservice tests with 
frequencies greater than 2 years to be extended by 6 months to 
facilitate test scheduling and consideration of plant operating 
conditions that may not be suitable for performance of the required 
testing. The testing frequency extension will not affect the ability 
of the components to respond to an accident as the components are 
required to be operable during the testing period extension.
    The proposed change will eliminate the existing TS SR 3.0.3 
allowance to defer performance of missed inservice tests up to the 
duration of the specified testing frequency, and instead will 
require an assessment of the missed test on equipment operability. 
This assessment will consider the effect on a margin of safety 
(equipment operability). Should the component be inoperable, the 
Technical Specifications provide actions to ensure that the margin 
of safety is protected. The proposed change also eliminates a 
statement that nothing in the ASME Code should be construed to 
supersede the requirements of any TS. The NRC has determined that 
statement to be incorrect. However, elimination of the statement 
will have no effect on plant operation or safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, Inc., 40 Inverness Center 
Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendment request: July 28, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16214A252.
    Description of amendment request: The amendments would revise the 
technical specifications (TSs) at the Joseph M. Farley Nuclear Plant, 
Units 1 and 2, to eliminate the ``lnservice Testing Program'' from TS 
5.5, ``Programs and Manuals,'' and add a new defined term, ``INSERVICE 
TESTING PROGRAM,'' to TS 1.1, ``Definitions.'' This request is 
submitted in accordance with Technical Specifications Task Force (TSTF) 
Traveler TSTF-545, Revision 3, ``TS lnservice Testing Program Removal & 
Clarify SR [Surveillance Requirement] Usage Rule Application to Section 
5.5 Testing.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS Chapter 5, ``Administrative 
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating 
the ``lnservice Testing Program'' specification. Most requirements 
in the lnservice Testing Program are removed, as they are 
duplicative of requirements in the ASME OM Code [American Society of 
Mechanical Engineers Operation and Maintenance Code], as clarified 
by Code Case OMN-20, ``lnservice Test Frequency.'' The remaining 
requirements in the Section 5.5 IST [Inservice Testing] Program are 
eliminated because the NRC has determined their inclusion in the TS 
is contrary to regulations. A new defined term, ``INSERVICE TESTING 
PROGRAM,'' is added to the TS, which references the requirements of 
10 CFR 50.55a(f).
    Performance of inservice testing is not an initiator to any 
accident previously evaluated. As a result, the probability of 
occurrence of an accident is not significantly affected by the 
proposed change. lnservice test frequencies under Code Case OMN-20 
are equivalent to the current testing period allowed by the TS with 
the exception that testing frequencies greater than 2 years may be 
extended by up to 6 months to facilitate test scheduling and 
consideration of plant operating conditions that may not be suitable 
for performance of the required testing. The testing frequency 
extension will not affect the ability of the components to mitigate 
any accident previously evaluated as the

[[Page 66311]]

components are required to be operable during the testing period 
extension. Performance of inservice tests utilizing the allowances 
in OMN-20 will not significantly affect the reliability of the 
tested components. As a result, the availability of the affected 
components, as well as their ability to mitigate the consequences of 
accidents previously evaluated, is not affected.
    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 design or configuration 
of the plant. The proposed change does not involve a physical 
alteration of the plant; no new or different kind of equipment will 
be installed. The proposed change does not alter the types of 
inservice testing performed. In most cases, the frequency of 
inservice testing is unchanged. However, the frequency of testing 
would not result in a new or different kind of accident from any 
previously evaluated since the testing methods are not altered.
    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 eliminates some requirements from the TS in 
lieu of requirements in the ASME Code, as modified by use of Code 
Case OMN-20. Compliance with the ASME Code is required by 10 CFR 
50.55a. The proposed change also allows inservice tests with 
frequencies greater than 2 years to be extended by 6 months to 
facilitate test scheduling and consideration of plant operating 
conditions that may not be suitable for performance of the required 
testing. The testing frequency extension will not affect the ability 
of the components to respond to an accident as the components are 
required to be operable during the testing period extension.
    The proposed change will eliminate the existing TS SR 3.0.3 
allowance to defer performance of missed in service tests up to the 
duration of the specified testing frequency, and instead will 
require an assessment of the missed test on equipment operability. 
This assessment will consider the effect on a margin of safety 
(equipment operability). Should the component be inoperable, the 
Technical Specifications provide actions to ensure that the margin 
of safety is protected. The proposed change also eliminates a 
statement that nothing in the ASME Code should be construed to 
supersede the requirements of any TS. The NRC has determined that 
statement to be incorrect. However, elimination of the statement 
will have no effect on plant operation or safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, Inc., 40 Iverness Center 
Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: July 28, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16214A252.
    Description of amendment request: The amendments would revise the 
technical specifications (TSs) at the Vogtle Electric Generating Plant, 
Units 1 and 2, to eliminate the ``lnservice Testing Program'' from the 
TS 5.5, ``Programs and Manuals,'' section and to add a new defined 
term, ``INSERVICE TESTING PROGRAM,'' to the TS 1.1, ``Definitions,'' 
section. This request is submitted in accordance with Technical 
Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, ``TS 
lnservice Testing Program Removal & Clarify SR Usage Rule Application 
to Section 5.5 Testing.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS Chapter 5, ``Administrative 
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating 
the ``lnservice Testing Program'' specification. Most requirements 
in the lnservice Testing Program are removed, as they are 
duplicative of requirements in the ASME OM [American Society of 
Mechanical Engineers Operation and Maintenance] Code, as clarified 
by Code Case OMN-20, ``lnservice Test Frequency.'' The remaining 
requirements in the Section 5.5 IST [Inservice Testing] Program are 
eliminated because the NRC has determined their inclusion in the TS 
is contrary to regulations. A new defined term, ``INSERVICE TESTING 
PROGRAM,'' is added to the TS, which references the requirements of 
10 CFR 50.55a(f).
    Performance of inservice testing is not an initiator to any 
accident previously evaluated. As a result, the probability of 
occurrence of an accident is not significantly affected by the 
proposed change. lnservice test frequencies under Code Case OMN-20 
are equivalent to the current testing period allowed by the TS with 
the exception that testing frequencies greater than 2 years may be 
extended by up to 6 months to facilitate test scheduling and 
consideration of plant operating conditions that may not be suitable 
for performance of the required testing. The testing frequency 
extension will not affect the ability of the components to mitigate 
any accident previously evaluated as the components are required to 
be operable during the testing period extension. Performance of 
inservice tests utilizing the allowances in OMN-20 will not 
significantly affect the reliability of the tested components. As a 
result, the availability of the affected components, as well as 
their ability to mitigate the consequences of accidents previously 
evaluated, is not affected.
    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 design or configuration 
of the plant. The proposed change does not involve a physical 
alteration of the plant; no new or different kind of equipment will 
be installed. The proposed change does not alter the types of 
inservice testing performed. In most cases, the frequency of 
inservice testing is unchanged. However, the frequency of testing 
would not result in a new or different kind of accident from any 
previously evaluated since the testing methods are not altered.
    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 eliminates some requirements from the TS in 
lieu of requirements in the ASME Code, as modified by use of Code 
Case OMN-20. Compliance with the ASME Code is required by 10 CFR 
50.55a. The proposed change also allows inservice tests with 
frequencies greater than 2 years to be extended by 6 months to 
facilitate test scheduling and consideration of plant operating 
conditions that may not be suitable for performance of the required 
testing. The testing frequency extension will not affect the ability 
of the components to respond to an accident as the components are 
required to be operable during the testing period extension.
    The proposed change will eliminate the existing TS SR 3.0.3 
allowance to defer performance of missed in service tests up to the 
duration of the specified testing frequency, and instead will 
require an

[[Page 66312]]

assessment of the missed test on equipment operability. This 
assessment will consider the effect on a margin of safety (equipment 
operability). Should the component be inoperable, the Technical 
Specifications provide actions to ensure that the margin of safety 
is protected. The proposed change also eliminates a statement that 
nothing in the ASME Code should be construed to supersede the 
requirements of any TS. The NRC has determined that statement to be 
incorrect. However, elimination of the statement will have no effect 
on plant operation or safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, Inc., 40 Inverness Center 
Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Michael T. Markley.

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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: October 9, 2015, as 
supplemented by letter dated May 12, 2016.
    Brief description of amendments: The amendments approve a revision 
to the emergency action levels from a scheme based on Nuclear Energy 
Institute (NEI) 99-01, Revision 5, ``Methodology for Development of 
Emergency Action Levels,'' to a scheme provided in the subsequent 
Revision 6 of NEI 99-01.
    Date of issuance: September 8, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 365 days from the date of issuance.
    Amendment Nos.: Unit 1--198; Unit 2--198; Unit 3--198. A publicly-
available version is in ADAMS under Accession No. ML16180A109; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: 
The amendments revised the Operating Licenses.
    Date of initial notice in Federal Register: December 8, 2015 (80 FR 
76318). The supplemental letter dated May 12, 2016, provided additional 
information that clarified the application, incorporated recent 
emergency preparedness frequently asked questions, did not expand the 
scope of the application as originally noticed, and did not change the 
staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 8, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 24, 2013, as supplemented by 
letters dated February 9, March 11, April 13, July 6, and August 13, 
2015; and February 24 and April 22, 2016.
    Brief description of amendments: These amendments modify the 
operating licenses and technical specifications (TSs) to incorporate a 
new fire protection licensing basis in accordance with 10 CFR 50.48(c). 
The amendments authorize the transition of the licensee's fire 
protection program to a risk-informed, performance-based program based 
on the 2001 Edition of National Fire Protection Association Standard 
805, ``Performance-Based Standard for Fire Protection for Light Water 
Reactor Electric Generating Plants.''
    Date of issuance: August 30, 2016.
    Effective date: As of the date of issuance and shall be implemented 
in accordance with the schedule contained in the revised paragraph 2.E. 
and page 12 of Appendix C, Additional Conditions to the Renewed 
Facility Operating Licenses.
    Amendment Nos.: 318 and 296. A publicly-available version is in 
ADAMS under Accession No. ML16175A359; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: August 5, 2014 (79 FR 
45488). The supplemental letters dated February 9, March 11, April 13, 
July 6, and August 13, 2015; and February 24 and April 22, 2016, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the NRC staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 30, 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, Oswego County, New York

    Date of amendment request: October 8, 2015, as supplemented by 
letter dated April 7, 2016.

[[Page 66313]]

    Brief description of amendment: The amendments modified the 
technical specifications (TSs) to allow for brief, inadvertent 
simultaneous opening of redundant secondary containment personnel 
access doors during brief entry and exit conditions.
    Date of issuance: August 31, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 223 (Unit 1) and 157 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16197A486; 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: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: January 5, 2016 (81 FR 
262). The supplemental letter dated April 7, 2016, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 31, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 31, 2015, as supplemented by 
letters dated April 20 and July 15, 2016.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) consistent with Technical Specification 
Task Force Traveler 422, Revision 2, ``Change in Technical 
Specifications End States (CE NPSD-1186).''
    Date of issuance: August 30, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 234 and 184. A publicly-available version is in 
ADAMS under Accession No. ML16210A374; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: November 24, 2015 (80 
FR 73237). The supplemental letters dated April 20 and July 15, 2016, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 30, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 18, 2016.
    Brief description of amendment: The amendment revised the DAEC 
technical specifications (TSs) Section 2.1.1, ``Reactor Core [Safety 
Limits],'' to change the Safety Limit Minimum Critical Power Ratio 
(SLMCPR) for two recirculation loop operation and for single 
recirculation loop operation. The changes reflected the cycle-specific 
analysis. The amendment also removed an outdated historical footnote 
from TS Table 3.3.5.1-1.
    Date of issuance: September 12, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 297. A publicly-available version is in ADAMS under 
Accession No. ML16211A514; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-49: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 5, 2016 (81 FR 
43665).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 12, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 18, 2015, as supplemented by 
letters dated January 29, April 14, and May 31, 2016.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 5.5.12, ``Primary Containment Leakage Rate Testing 
Program,'' to state that the program shall be in accordance with 
Nuclear Energy Institute (NEI) 94-01, Revision 3-A, ``Industry 
Guideline for Implementing Performance-Based Option of 10 CFR part 50, 
appendix J.''
    Date of issuance: August 30, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 296. A publicly-available version is in ADAMS under 
Accession No. ML16210A008; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendments.
    Renewed Facility Operating License No. DPR-49: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 2015 (80 FR 
65814). The supplemental letters dated January 29, April 14, and May 
31, 2016, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 30, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 26, 2013, as supplemented by 
letters dated September 16, 2013, July 29, August 28, September 25, 
November 14, December 19, 2014; January 16, May 12, August 26, 2015; 
and February 22, April 7, and May 3, 2016.
    Brief description of amendments: The amendments authorized the 
transition of the Point Beach fire protection program to a risk-
informed, performance-based program based on National Fire Protection 
Association Standard 805 (NFPA 805), ``Performance-Based Standard for 
Fire Protection for Light Water Reactor Electric Generating Plants,'' 
2001 Edition, in accordance with 10 CFR 50.48(c).
    Date of issuance: September 8, 2016.
    Effective date: As of the date of issuance and shall be implemented 
as described in the Transition License Conditions.
    Amendment Nos.: 256 and 260. A publicly-available version is in 
ADAMS under Accession No. ML16196A093;

[[Page 66314]]

documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revised the Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: July 8, 2014 (79 FR 
28580). The supplemental letters dated September 16, 2013, July 29, 
August 28, September 25, November 14, December 19, 2014; January 16, 
May 12, August 26, 2015; and February 22, April 7, and May 3, 2016, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 8, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 15, 2015, as supplemented by 
letters dated May 4, 2016, and June 1, 2016.
    Brief description of amendment: The amendment revised the Technical 
Specifications to allow implementation of the F* (F-star) alternate 
repair criterion for steam generator tubes.
    Date of issuance: September 6, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 2. A publicly-available version is in ADAMS under 
Accession No. ML16203A365; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-96: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: February 16, 2016 (81 
FR 7844). The supplemental letters dated May 4, 2016, and June 1, 2016, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 6, 2016.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 19th day of September 2016.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2016-23097 Filed 9-26-16; 8:45 am]
 BILLING CODE 7590-01-P