Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 70175-70190 [2016-24321]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 81, Number 196 (Tuesday, October 11, 2016)]
[Notices]
[Pages 70175-70190]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-24321]


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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0207]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of 
1954, as amended (the Act), the U.S. Nuclear Regulatory Commission 
(NRC) is publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued, and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from September 13, 2016 to September 26, 2016. 
The last biweekly notice was published on September 27, 2016.

DATES: Comments must be filed by November 10, 2016. A request for a 
hearing must be filed by December 12, 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-2016-0207. 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: Kay Goldstein, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-1506, email: Kay.Goldstein@nrc.gov.

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2016-0207, 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-2016-0207.
     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-2016-0207, facility name, unit 
number(s), application date, and subject in your comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at https://www.regulations.gov as well as entering the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.
    II. Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses and Combined Licenses and Proposed No Significant 
Hazards Consideration Determination
    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Sec.  50.92 of title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, 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

[[Page 70176]]

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 the action. 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 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 proceeding. 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) through (iii).
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of any amendment unless the Commission finds an imminent 
danger to the health or safety of the public, in which case it will 
issue an appropriate order or rule under 10 CFR part 2.
    A State, local governmental body, Federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1).
    The petition should state the nature and extent of the petitioner's 
interest in the proceeding. The petition should be submitted to the 
Commission by December 12, 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

[[Page 70177]]

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/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.

[[Page 70178]]

Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear 
Power Plant, Unit 1, New Hill, North Carolina

    Date of amendment request: May 26, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16151A001.
    Description of amendment request: The amendment would revise the 
Shearon Harris Nuclear Power Plant, Unit 1, technical specifications 
(TSs) to institute a new administrative program TS for the 
establishment, implementation, and maintenance of a Diesel Fuel Oil 
Testing Program, the specifics of which will be contained in a 
licensee-controlled document. It also relocates to this program the 
current TS surveillance requirements (SRs) for evaluating diesel fuel 
oil, along with the SRs for the draining, sediment removal, and 
cleaning of each main fuel oil storage tank at least once every 10 
years. In addition, an exception is proposed to Regulatory Guide (RG) 
1.137, Revision 1, ``Fuel Oil Systems for Standby Diesel Generators,'' 
for the allowance of performing sampling of new fuel oil offsite prior 
to its addition to the fuel oil storage tanks.
    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 institutes a new administrative program 
TS for the establishment, implementation, and maintenance of a 
Diesel Fuel Oil Testing Program. The specifics of this program will 
be contained in a licensee-controlled document. The current TS SR 
for evaluating new and stored diesel fuel oil and the cleaning of 
the fuel oil storage tanks will be relocated to this program. The 
American Society for Testing and Materials (ASTM) standard 
references pertaining to new and stored fuel oil will be relocated 
to the aforementioned program; however, requirements to perform 
testing in accordance with applicable ASTM standards are retained in 
the TS. Requirements to perform surveillances of both new and stored 
diesel fuel oil are also retained in the TS. Evaluations of future 
changes to the licensee-controlled document will be conducted 
pursuant to the requirements of 10 CFR 50.59. A more rigorous 
testing of water and sediment content is added to the ``clear and 
bright'' test used to establish the acceptability of new fuel oil 
for use prior to its addition to the fuel oil storage tanks. 
Additionally, an exception to RG 1.137 is proposed to allow for the 
performance of new fuel oil sampling offsite. These changes will not 
affect nor degrade the ability of the emergency diesel generators 
(DGs) to perform their specified safety functions as the diesel fuel 
oil continues to be properly evaluated.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems or components from 
performing their intended function to mitigate the consequences on 
an initiating event with the assumed acceptance limits. The proposed 
changes do not affect the source term, containment isolation, or 
radiological release assumptions used in evaluating the radiological 
consequences of an accident previously evaluated. Further, the 
proposed changes do not increase the types and amounts of 
radioactive effluent that may be released offsite, nor significantly 
increase individual or cumulative occupational or public radiation 
exposure.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment institutes a new administrative program 
TS for the establishment, implementation, and maintenance of a 
Diesel Fuel Oil Testing Program, of which the current TS SR for 
evaluating new and stored diesel fuel oil and the cleaning of the 
fuel oil storage tanks are relocated, including pertinent ASTM 
standard references. A more rigorous testing of water and sediment 
content is added to the ``clear and bright'' test used to establish 
the acceptability of new fuel oil for use prior to its addition to 
the fuel oil storage tanks. Additionally, an exception to RG 1.137 
is proposed to allow for the performance of new fuel oil sampling 
offsite. These changes do not alter the way any structure, system, 
or component functions and does not modify the manner in which the 
plant is operated. The requirements retained in the TS continue to 
require testing of the diesel fuel oil to ensure the proper 
functioning of the DGs.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    The proposed amendment institutes a new administrative program 
TS for the establishment, implementation, and maintenance of a 
Diesel Fuel Oil Testing Program, the specifics of which will be 
contained in a licensee-controlled document. The current TS SR for 
evaluating new and stored diesel fuel oil and the cleaning of the 
fuel oil storage tanks will be relocated to this program, along with 
the pertinent ASTM standard references. Changes to the licensee-
controlled document are performed in accordance with the provisions 
of 10 CFR 50.59, thereby providing an effective level of regulatory 
control and ensures that diesel fuel oil testing is conducted such 
that there is no significant reduction in a margin of safety.
    A more rigorous testing of water and sediment content is added 
to the ``clear and bright'' test used to establish the acceptability 
of new fuel oil for use prior to its addition to the fuel oil 
storage tanks. Additionally, an exception to RG 1.137 is proposed to 
allow for the performance of new fuel oil sampling offsite. The 
margin of safety provided by the DGs is unaffected by the proposed 
changes since there continue to be TS requirements to ensure fuel 
oil is of the appropriate quality and reliability for emergency DG 
use. The proposed changes provide the flexibility needed to improve 
fuel oil sampling and analysis methodologies, while maintaining 
sufficient controls to preserve the current margins of safety.
    Based on the above, Duke Energy concludes that the proposed 
amendment does not involve a significant hazards consideration under 
the standards set forth in 10 CFR 50.92, and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, 
Duke Energy Business Services, 550 South Tryon Street, Mail Code 
DEC45A, Charlotte, NC 28202.
    NRC Acting Branch Chief: Jeanne A. Dion.
    Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York
    Date of amendment request: August 29, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16242A332.
    Description of amendment request: The amendment would revise 
technical specification (TS) 5.5.6, Primary Containment Leak Rate 
Testing Program. These revisions would extend the Type A Primary 
Containment Integrated Leak Rate Test interval to 15 years and extend 
the Type C Local Leak Rate Test testing interval up to 75 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or

[[Page 70179]]

consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment to the TS involves the extension of the 
JAF [James A. FitzPatrick Nuclear Power Plant] Type A containment 
test interval to 15 years and the extension of the Type C test 
interval to 75 months. The current Type A test interval of 120 
months (10 years) would be extended on a permanent basis to no 
longer than 15 years from the last Type A test. The current Type C 
test interval of 60 months for selected components would be extended 
on a performance basis to no longer than 75 months. Extensions of up 
to nine months (total maximum interval of 84 months for Type C 
tests) are permissible only for non-routine emergent conditions. The 
proposed extension does not involve either a physical change to the 
plant or a change in the manner in which the plant is operated or 
controlled. The containment is designed to provide an essentially 
leak tight barrier against the uncontrolled release of radioactivity 
to the environment for postulated accidents. As such, the 
containment and the testing requirements invoked to periodically 
demonstrate the integrity of the containment exist to ensure the 
plant's ability to mitigate the consequences of an accident, and do 
not involve the prevention or identification of any precursors of an 
accident. The change in dose risk for changing the Type A test 
frequency from three-per-ten years to once-per-fifteen-years, 
measured as an increase to the total integrated plant risk for those 
accident sequences influenced by Type A testing, is 0.0087 person 
rem/year. EPRI [Electric Power Research Institute] Report No. 
1009325, Revision 2-A states that a very small population dose is 
defined as an increase of <= 1.0 person-rem per year, or <= 1% of 
the total population dose, whichever is less restrictive for the 
risk impact assessment of the extended ILRT intervals. The results 
of the risk assessment for this amendment meet these criteria. 
Moreover, the risk impact for the ILRT extension when compared to 
other severe accident risks is negligible. Therefore, this proposed 
extension does not involve a significant increase in the probability 
of an accident previously evaluated.
    As documented in NUREG-1493, Type B and C tests have identified 
a very large percentage of containment leakage paths, and the 
percentage of containment leakage paths that are detected only by 
Type A testing is very small. The JAF Type A test history supports 
this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as: (1) Activity based, 
and; (2) time based. Activity based failure mechanisms are defined 
as degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment combined with the containment inspections performed in 
accordance with ASME [American Society of Mechanical Engineers] 
Section Xl, the Maintenance Rule, and TS requirements serve to 
provide a high degree of assurance that the containment would not 
degrade in a manner that is detectable only by a Type A test. Based 
on the above, the proposed extensions do not significantly increase 
the consequences of an accident previously evaluated.
    The proposed amendment also deletes exceptions previously 
granted to allow one time extensions of the ILRT test frequency for 
JAF. These exceptions were for activities that would have already 
taken place by the time this amendment is approved; therefore, their 
deletion is solely an administrative action that has no effect on 
any component and no impact on how the unit is operated.
    Therefore, the proposed change does not result in 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 amendment to the TS involves the extension of the 
JAF Type A containment test interval to 15 years and the extension 
of the Type C test interval to 75 months. The containment and the 
testing requirements to periodically demonstrate the integrity of 
the containment exist to ensure the plant's ability to mitigate the 
consequences of an accident do not involve any accident precursors 
or initiators. The proposed change does not involve a physical 
change to the plant (i.e., no new or different type of equipment 
will be installed) or a change to the manner in which the plant is 
operated or controlled.
    The proposed amendment also deletes exceptions previously 
granted to allow one time extensions of the ILRT test frequency for 
JAF. These exceptions were for activities that would have already 
taken place by the time this amendment is approved; therefore, their 
deletion is solely an administrative action that does not result in 
any change in how the unit is operated.
    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 amendment to TS 5.5.6 involves the extension of the 
JAF Type A containment test interval to 15 years and the extension 
of the Type C test interval to 75 months for selected components. 
This amendment does not alter the manner in which safety limits, 
limiting safety system set points, or limiting conditions for 
operation are determined. The specific requirements and conditions 
of the TS Containment Leak Rate Testing Program exist to ensure that 
the degree of containment structural integrity and leak-tightness 
that is considered in the plant safety analysis is maintained. The 
overall containment leak rate limit specified by TS is maintained.
    The proposed change involves only the extension of the interval 
between Type A containment leak rate tests and Type C tests for JAF. 
The proposed surveillance interval extension is bounded by the 15-
year ILRT Interval and the 75-month Type C test interval currently 
authorized within NEI 94-01, Revision 3-A. Industry experience 
supports the conclusion that Type B and C testing detects a large 
percentage of containment leakage paths and that the percentage of 
containment leakage paths that are detected only by Type A testing 
is small. The containment inspections performed in accordance with 
ASME Section Xl, TS and the Maintenance Rule serve to provide a high 
degree of assurance that the containment would not degrade in a 
manner that is detectable only by Type A testing. The combination of 
these factors ensures that the margin of safety in the plant safety 
analysis is maintained. The design, operation, testing methods and 
acceptance criteria for Type A, B, and C containment leakage tests 
specified in applicable codes and standards would continue to be 
met, with the acceptance of this proposed change, since these are 
not affected by changes to the Type A and Type C test intervals.
    The proposed amendment also deletes exceptions previously 
granted to allow one time extensions of the ILRT test frequency for 
JAF. These exceptions were for activities that would have already 
taken place by the time this amendment is approved; therefore, their 
deletion is solely an administrative action and does not change how 
the unit is operated and maintained. Thus, there is no reduction in 
any margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: July 28, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16210A300.
    Description of amendment request: The proposed changes supports 
changes to the organization, staffing, and training requirements 
contained in

[[Page 70180]]

Section 5.0 of the technical specifications (TSs) after the license no 
longer authorizes operation of the reactor or placement or retention of 
fuel in the reactor pressure vessel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes would not take effect until CPS has 
permanently ceased operation and entered a permanently defueled 
condition. The proposed changes would revise the CPS TS by deleting 
or modifying certain portions of the TS administrative controls 
described in Section 5.0 of the TS that are no longer applicable to 
a permanently shutdown and defueled facility.
    The proposed changes do not involve any physical changes to 
plant structures, systems, and components (SSCs) or the manner in 
which SSCs are operated, maintained, modified, tested, or inspected. 
The proposed changes do not involve a change to any safety limits, 
limiting safety system settings, limiting control settings, limiting 
conditions for operation, surveillance requirements, or design 
features.
    The deletion and modification of provisions of the facility 
administrative controls do not affect the design of SSCs necessary 
for safe storage of spent irradiated fuel or the methods used for 
handling and storage of such fuel in the Spent Fuel Pool (SFP). The 
proposed changes are administrative in nature and do not affect any 
accidents applicable to the safe management of spent irradiated fuel 
or the permanently shutdown and defueled condition of the reactor.
    In a permanently defueled condition, the only credible accidents 
are the Fuel Handling Accident (FHA), Postulated Radioactive 
Releases Due to Liquid Radwaste Tank Failures, and Cask Drop 
Accident. Other accidents such as Loss of Coolant Accident, Loss of 
Feedwater, and Reactivity and Power Distribution Anomalies will no 
longer be applicable to a permanently defueled reactor plant.
    The probability of occurrence of previously evaluated accidents 
is not increased, since extended operation in a permanently defueled 
condition will be the only operation allowed, and therefore, bounded 
by the existing analyses. Additionally, the occurrence of postulated 
accidents associated with reactor operation is no longer credible in 
a permanently defueled reactor. This significantly reduces the scope 
of applicable accidents.
    The proposed changes in the administrative controls do not 
affect the ability to successfully respond to previously evaluated 
accidents and do not affect radiological assumptions used in the 
evaluations. The proposed changes narrow the focus of nuclear safety 
concerns to those associated with safely maintaining spent nuclear 
fuel. These changes remove the implication that CPS can return to 
operation once the final certification required by 10 CFR 
50.82(a)(1)(ii) is submitted to the NRC. Any event involving safe 
storage of spent irradiated fuel or the methods used for handling 
and storage of such fuel in the SFP would evolve slowly enough that 
no immediate response would be required to protect the health and 
safety of the public or station personnel. Adequate communications 
capability is provided to allow facility personnel to safely manage 
storage and handling of irradiated fuel. As a result, no changes to 
radiological release parameters are involved. There is no effect on 
the type or amount of radiation released, and there is no effect on 
predicted offsite doses in the event of an accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to delete and/or modify certain TS 
administrative controls have no impact on facility SSCs affecting 
the safe storage of spent irradiated fuel, or on the methods of 
operation of such SSCs, or on the handling and storage of spent 
irradiated fuel itself. The proposed changes do not result in 
different or more adverse failure modes or accidents than previously 
evaluated because the reactor will be permanently shut down and 
defueled and CPS will no longer be authorized to operate the 
reactor.
    The proposed changes will continue to require proper control and 
monitoring of safety significant parameters and activities. The 
proposed changes do not result in any new mechanisms that could 
initiate damage to the remaining relevant safety barriers in support 
of maintaining the plant in a permanently shutdown and defueled 
condition (e.g., fuel cladding and SFP cooling). Since extended 
operation in a defueled condition will be the only operation 
allowed, and therefore bounded by the existing analyses, such a 
condition does not create the possibility of a new or different kind 
of accident.
    The proposed changes do not alter the protection system design 
or create new failure modes. The proposed changes do not involve a 
physical alteration of the plant, and no new or different kind of 
equipment will be installed. Consequently, there are no new 
initiators that could result in a new or different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes involve deleting and/or modifying certain 
TS administrative controls once the CPS facility has been 
permanently shutdown and defueled. As specified in 10 CFR 
50.82(a)(2), the 10 CFR 50 license for CPS will no longer authorize 
operation of the reactor or emplacement or retention of fuel into 
the reactor vessel following submittal of the certifications 
required by 10 CFR 50.82(a)(1). As a result, the occurrence of 
certain design basis postulated accidents are no longer considered 
credible when the reactor is permanently defueled. The only 
remaining credible accidents are the FHA, the Postulated Radioactive 
Releases Due to Liquid Radwaste Tank Failures, and the Cask Drop 
Accident. The FHA is the limiting Chapter 15 dose event for CPS in 
its decommissioned state.
    The proposed changes do not adversely affect the inputs or 
assumptions of any of the design basis analyses that impact the FHA. 
The proposed changes are limited to those portions of the TS 
administrative controls that are not related to the safe storage and 
maintenance of spent irradiated fuel.
    These proposed changes do not directly involve any physical 
equipment limits or parameters. The requirements that are proposed 
to be revised and/or deleted from the CPS TS are not credited in the 
existing accident analysis for the remaining applicable postulated 
accidents; therefore, they do not contribute to the margin of safety 
associated with the accident analysis. Certain postulated DBAs 
[design-basis accidents] involving the reactor are no longer 
possible because the reactor will be permanently shut down and 
defueled and CPS will no longer be authorized to operate the 
reactor.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear,. 4300 Winfield Road, Warrenville, IL 60555.
    Acting NRC Branch Chief: G. Edward Miller.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center, Linn County, Iowa
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
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
No. 1, Rockingham County, New Hampshire
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

[[Page 70181]]

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

    Date of amendment request: July 28, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16214A276.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) consistent with Technical Specifications 
Task Force Traveler 545, Revision 3, ``TS Inservice Testing [IST] 
Program Removal & Clarify SR [Surveillance Requirement] Usage Rule 
Application to Section 5.5 Testing'' (ADAMS Accession No. ML15294A555).
    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 eliminates Technical Specifications (TS) 
Section 5.5.6 and 5.5.7, ``Inservice Testing Program,'' for Duane 
Arnold and Point Beach, respectively, and eliminates TS Section 
6.8.4.i, ``Inservice Testing Program'' for St. Lucie Units 1 and 2. 
The proposed change eliminates the requirements regarding [IST] from 
TS 4.0.5 in the Seabrook and Turkey Point TS. Most requirements in 
the [IST] 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, ``Inservice Test Frequency.'' The remaining requirements 
related to the 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 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 [IST] 
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 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 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William Blair, Managing Attorney--Nuclear, 
Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-
0420.
    Acting NRC Branch Chief: Jeanne A. Dion.

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

    Date of amendment request: July 28, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16210A030.
    Description of amendment request: The proposed amendment would 
eliminate technical specification (TS), Section 5.5.5, ``Inservice 
Testing [IST] Program,'' to remove requirements duplicated in American 
Society of Mechanical Engineers (ASME) Code for Operation and 
Maintenance of Nuclear Power Plants (OM Code), Case OMN-20, ``Inservice 
Test Frequency.'' A new defined term, ``Inservice Testing Program,'' is 
added to TS Section 1.1, ``Definitions.'' The proposed change to the TS 
is consistent with TSTF-545, Revision 3, ``TS Inservice Testing Program 
Removal & Clarify SR [surveillance requirement] Usage Rule Application 
to Section 5.5 Testing.'' TS SRs that currently refer to the IST 
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 OM Code, as clarified by Code Case OMN-20, 
``Inservice Test Frequency.'' 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 the 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

[[Page 70182]]

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 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 frequency, and will instead 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: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: David J. Wrona.

South Carolina Electric & Gas Company and South Carolina Public Service 
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear 
Station, Units 2 and 3, Fairfield County, South Carolina

    Date of amendment request: August 12, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16225A437.
    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 (RNS) 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:

    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 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 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 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

[[Page 70183]]

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

South Carolina Electric & Gas Company and South Carolina Public Service 
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear 
Station, Units 2 and 3, Fairfield, South Carolina

    Date of amendment request: September 8, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16252A200.
    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 component's (SSC's) 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: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius, LLC, 1111 Pennsylvania NW., Washington, DC 20004-2514.
    NRC Branch Chief: Jennifer Dixon-Herrity.

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

    Date of amendment request: August 29, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16243A463.
    Description of amendment request: The amendment would remove the 
administrative controls associated with the Limiting Condition for 
Operation (LCO) of Technical Specification (TS) 3.5.4, ``Refueling 
Water Storage Tank.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with NRC staff edits in square 
brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change removes an administrative note added by 
Amendment No. 192. The administrative control applied by Amendment 
No. 192 was issued to prevent or reduce the risk for drainage of the 
Reactor Water Storage Tank (RWST) when aligned to the non-safety, 
non-seismic purification system. The station has implemented a 
modification that qualifies the interconnection of the RWST to the 
purification system. The installed design prevents the RWST being 
drained below the current Technical Specifications minimum volume 
requirement due to a failure in the non-safety purification system. 
The RWST will continue to perform its safety function and the 
overall system performance has not been affected [by] this proposed 
amendment. Assumptions previously made in evaluating the 
consequences of the accident are not altered, and the consequences 
of the accident are not increased. Therefore, the proposed change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The Purification 
Loop supports the Spent Fuel System and is not credited for safe 
shutdown of the plant or accident mitigation. Therefore, the 
proposed change has insignificant impact on the probability and 
consequences of an accident previously evaluated. A combination of 
design and administrative controls ensure that the Purification Loop 
maintains RWST boron concentration and water volume requirements 
whenever the contents of the RWST are processed through the system. 
The RWST is operated under System Operating Procedure for the Spent 
Fuel Cooling System and is protected by maintaining the isolation 
valve for the lower return line locked closed in modes 1 through 4.
    2. Does the proposed change create the possibility of a new or 
different kind of

[[Page 70184]]

accident from any accident previously evaluated?
    Response: No.
    The proposed change does not introduce a new or different 
accident previously evaluated. The station implemented a qualified 
design that prevents the RWST from being drained below the current 
TS 3.5.4.a minimum volume requirement. The proposed change does not 
alter the design requirements of the RWST or any Structure, System 
or Component or its function during accident conditions. The changes 
do not alter assumptions made in the safety analysis and the current 
TS LCO are maintained. The Purification Loop supports the Spent Fuel 
System and is not credited for safe shutdown of the plant or 
accident mitigation. The proposed change removes a note added by 
Amendment No. 192 that applied an administrative control to manage 
the risk of a postulated RWST drainage scenario by the purification 
system.
    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 removes a note added by Amendment No. 192. 
The proposed change does not alter the safety limits, limiting 
safety system settings or limiting conditions for operation of the 
RWST. The modification preserved the current licensing and design 
bases of the RWST, therefore the margin of safety for the RWST are 
not affected. The proposed changes do not adversely affect systems 
that respond to safely shutdown the plant and to maintain the plant 
in a safe shutdown condition. The Purification Loop supports the 
Spent Fuel System and is not credited for safe shutdown of the plant 
or accident mitigation.
    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: Kathryn M. Sutton, Morgan, Lewis & Bockius 
LLP, 1111 Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Inc. (SNC); Georgia Power Company; 
Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; 
City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear 
Plant (HNP), Unit No. 2, Appling County, Georgia

    Date of amendment request: August 29, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16245A257.
    Description of amendment request: The amendment would revise the 
values for the reactor core Safety Limit 2.1.1.2 for Minimum Critical 
Power Ratios for both single and dual recirculation loop operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with NRC staff edits in 
brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Safety Limit Minimum Critical Power Ratio (SLMCPR) ensures 
that 99.9% of the fuel rods in the core will not be susceptible to 
boiling transition during normal operation or the most limiting 
postulated design-basis transient event. The new SLMCPR values 
preserve the existing margin to the onset of transition boiling; 
therefore, the probability of fuel damage is not increased as a 
result of this proposed change. The determination of the revised HNP 
Unit 2 SLMCPRs has been performed using NRC-approved methods of 
evaluation. These plant-specific calculations are performed each 
operating cycle and may require changes for future cycles. The 
revised SLMCPR values do not change the method of operating the 
plant; therefore, they have no effect on the probability of an 
accident, initiating event, or transient:
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes result only from a specific analysis for 
the HNP Unit 2 core reload design. These changes do not involve any 
new or different methods for operating the facility. No new 
initiating events or transients result from these changes.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The new SLMCPRs have been calculated using NRC-approved methods 
of evaluation with plant and cycle-specific input values for the 
fuel and core design for the upcoming cycle of operation. The SLMCPR 
values ensure that 99.9% of the fuel rods in the core will not be 
susceptible to boiling transition during normal operation or the 
most limiting postulated design-basis transient event. The operating 
MCPR limit is set appropriately above the safety limit value to 
ensure adequate margin when the cycle-specific transients are 
evaluated. Accordingly, the margin of safety is maintained with the 
revised values.

    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, 40 Iverness Center 
Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: July 29, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16211A436.
    Description of amendment request: The amendment request proposes to 
add to License Condition 2.D.(1) of the VEGP Units 3 and 4 combined 
licenses an Interim Amendment Request process for changes during 
construction when emergent conditions are present.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff'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 amendment would add an Interim Amendment Request 
process to Condition 2.0.(1) of the Vogtle 3 and 4 COLs [combined 
licenses] to allow construction to continue, at SNC's [Southern 
Nuclear Operating Company] own risk, in emergent conditions, where a 
non-conforming condition that has little or no safety significance 
is discovered and the work activity cannot be adjusted. The Interim 
Amendment Request process would require SNC to submit a Nuclear 
Construction Safety Assessment which (1) identifies the proposed 
change; (2) evaluates whether emergent conditions are present; (3) 
evaluates whether the change would result in any material decrease 
in safety; and (4) evaluates whether continued construction would 
make the non-conforming condition irreversible. Only if the 
continued construction would have no material decrease in safety 
would the NRC issue a determination that construction could continue 
pending SNC's initiation of the COL-ISG-025 PAR [preliminary 
amendment request]/LAR [license amendment request] process. The 
requirement to include a Nuclear Construction Safety Assessment 
ensures that the proposed amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. If the continued construction would result a 
material decrease in safety, then continued construction would not 
be authorized.

[[Page 70185]]

    The proposed amendment does not modify the design, construction, 
or operation of any plant structures, systems, or components (SSCs), 
nor does it change any procedures or method of control for any SSCs. 
Because the proposed amendment does not change the design, 
construction, or operation of any SSCs, it does not adversely affect 
any design function as described in the Updated Final Safety 
Analysis Report.
    The proposed amendment does not affect the probability of an 
accident previously evaluated. Similarly, because the proposed 
amendment does not alter the design or operation of the nuclear 
plant or any plant SSCs, the proposed amendment does not represent a 
change to the radiological effects of an accident, and therefore, 
does not involve an increase in the consequences of an accident 
previously evaluated.
    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 would add an Interim Amendment Request 
process to Condition 2.0.(1) of the Vogtle 3 and 4 COLs to allow 
construction to continue, at SNC's own risk, in emergent conditions, 
where a non-conforming condition that has little or no safety 
significance is discovered and the work activity cannot be adjusted. 
The Interim Amendment Request process would require SNC to submit a 
Nuclear Construction Safety Assessment which (1) identifies the 
proposed change; (2) evaluates whether emergent conditions are 
present; (3) evaluates whether the change would result in any 
material decrease in safety; and (4) evaluates whether continued 
construction would make the non-conforming condition irreversible. 
Only if the continued construction would have no material decrease 
in safety would NRC issue a determination that construction could 
continue pending SNC's initiation of the COL-ISG-025 PAR/LAR 
process.
    The proposed amendment is not a modification, addition to, or 
removal of any plant SSCs. Furthermore, the proposed amendment is 
not a change to procedures or method of control of the nuclear plant 
or any plant SSCs. The proposed amendment only adds a new screening 
process and does not change the design, construction, or operation 
of the nuclear plant or any plant operations.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment would add an Interim Amendment Request 
process to Condition 2.0.(1) of the Vogtle 3 and 4 COLs to allow 
construction to continue, at SNC's own risk, in emergent conditions, 
where a non-conforming condition that has little or no safety 
significance is discovered and the work activity cannot be adjusted. 
The Interim Amendment Request process would require SNC to submit a 
Nuclear Construction Safety Assessment which (1) identifies the 
proposed change; (2) evaluates whether emergent conditions are 
present; (3) evaluates whether the change would result in any 
material decrease in safety; and (4) evaluates whether continued 
construction would make the non-conforming condition irreversible. 
Only if the continued construction would have no material decrease 
in safety would the NRC issue determination that construction could 
continue pending SNC's initiation of the COL-ISG-025 PAR/LAR 
process.
    The proposed amendment is not a modification, addition to, or 
removal of any plant SSCs. Furthermore, the proposed amendment is 
not a change to procedures or method of control of the nuclear plant 
or any plant SSCs. The proposed amendment does not alter any design 
function or safety analysis. Consequently, no safety analysis or 
design basis acceptance limit/criterion is challenged or exceeded by 
the proposed amendment, thus the margin of safety is not reduced. 
The only impact of this activity is the addition of an Interim 
Amendment Request process.
    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, Inc. Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant Units 3 and 4, Burke County, Georgia

    Date of amendment request: September 9, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16253A412.
    Description of amendment request: The amendment request proposes 
changes to update the Protection and Safety Monitoring System (PMS) 
design, specifically the description of the roles of the Qualified Data 
Processing System (QDPS) and the safety displays. The proposed changes 
add Main Control Room (MCR) safety-related display divisions A and D to 
plant-specific Tier 1 (and associated COL Appendix C) and the Updated 
Final Safety Analysis Report (UFSAR), and correct the name of the QDPS 
in the UFSAR by referring to the QDPS as a system, rather than a 
subsystem. Because, this proposed change requires a departure from Tier 
1 information in 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 proposed change to the roles of the qualified data 
processing system (QDPS) and safety-related displays, as well as the 
change to add Division A and Division D of the main control room 
(MCR) safety-related displays to the listing of PMS equipment, as 
identified in Combined License (COL) Appendix C (and plant-specific 
Tier 1) Table 2.5.2-1 and Updated Final Safety Analysis Report 
(UFSAR) Table 3.11-1 and 3l.6-2 do not alter any accident initiating 
component/system failure or event, thus the probabilities of the 
accidents previously evaluated are not affected.
    The proposed changes do not adversely affect safety-related 
equipment or a radioactive material barrier, and this activity dos 
not involve the containment of radioactive material.
    The radioactive material source terms and release paths used in 
the safety analysis are unchanged, thus the radiological releases in 
the UFSAR accident analysis are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change to the roles of the QDPS and safety-related 
displays, as well as the change to add Division A and Division D of 
the MCR safety-related displays to the listing of PMS equipment, as 
identified in COL Appendix C (and plant-specific Tier 1) Table 
2.5.2-1 and UFSAR Table 3.11-1 and 3l.6-2 does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The proposed changes do not alter the design 
or capability of any sensors which provide input to the QDPS. The 
functionality of the QDPS to process the input obtained from sensors 
into data to be sent to the safety displays is not affected by the 
proposed changes. The proposed changes do not affect any functions 
performed by the safety displays, nor do the proposed changes affect 
the capability of the safety displays to display the data received 
from the QDPS.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

[[Page 70186]]

    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    There is no safety-related structure, system or component (SSC) 
or function adversely affected by the proposed change to the roles 
of the QDPS and safety-related displays, nor by the change to add 
Division A and Division D of the MCR safety-related displays to the 
listing of Protection and Safety Monitoring System (PMS) equipment. 
The proposed changes do not alter the mechanisms by which system 
components are actuated or controlled. Because no safety analysis or 
design basis acceptance limit/criterion is challenged or exceeded by 
the proposed 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. 52-025 and 52-
026, Vogtle Electric Generating Plant Units 3 and 4, Burke County, 
Georgia

    Date of amendment request: September 9, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16253A204.
    Description of amendment request: The amendment request proposes 
changes to revise plant-specific Tier 1, plant-specific Tier 2, and 
combined license (COL) Appendix C information concerning the details of 
the Class 1E direct current and uninterruptible power supply system 
(IDS), specifically adding seven Class 1E fuse panels to the IDS 
design. These proposed changes provide electrical isolation between the 
non-Class 1E IDS battery monitors and their respective Class 1E battery 
banks. 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 with NRC staff 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 revise plant-specific Tier 1, COL 
Appendix C, and [Updated Final Safety Analysis Report (UFSAR)] 
information concerning details of the IDS, specifically the addition 
of seven Class 1E fuse isolation panels at the interconnection of 
the non-Class 1E IDS battery monitors and Class 1E IDS circuits, are 
necessary to conform to Regulatory Guide 1.75 Rev. 2 (consistent 
with UFSAR Appendix 1A exceptions) and IEEE 384-1981 to prevent a 
fault on non-Class 1E circuits or equipment from degrading the 
operation of Class 1E IDS circuits and equipment below an acceptable 
level. The proposed changes do not adversely affect the design 
functions of the IDS, including the Class 1E battery banks and the 
battery monitors.
    These proposed changes to revise plant-specific Tier 1, COL 
Appendix C, and UFSAR information concerning details of the IDS, 
specifically the addition of seven Class 1E fuse isolation panels at 
the interconnection of the non-Class 1E IDS battery monitors and 
Class 1E IDS circuits as described in the current licensing basis do 
not have an adverse effect on any of the design functions of any 
plant systems. The proposed changes do not adversely affect any 
plant electrical system and do not affect the support, design, or 
operation of mechanical and fluid systems required to mitigate the 
consequences of an accident. There is no change to plant systems or 
the response of systems to postulated accident conditions. There is 
no change to the predicted radioactive releases due to postulated 
accident conditions. The plant response to previously evaluated 
accidents or external events is not adversely affected, nor do the 
proposed changes 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 to revise plant-specific Tier 1, COL 
Appendix C, and UFSAR information concerning details of the IDS, 
specifically the addition of seven Class 1E fuse isolation panels at 
the interconnection of the non-Class 1E IDS battery monitors and 
Class 1E IDS circuits, are necessary to conform to Regulatory Guide 
1.75 Rev. 2 (consistent with UFSAR Appendix 1A exceptions) and IEEE 
384-1981 to prevent a fault on non-Class 1E circuits or equipment 
from degrading the operation of Class 1E IDS circuits and equipment 
below an acceptable level. The proposed changes do not adversely 
affect any plant electrical system and do not adversely affect the 
design function, support, design, or operation of mechanical and 
fluid systems. The proposed changes do not result in a new failure 
mechanism or introduce any new accident precursors. No design 
function described in the UFSAR is adversely affected by the 
proposed changes.
    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.
    There is no safety-related [structure, system, and component 
(SSC)] or function adversely affected by the proposed change to add 
IDS fuse isolation panels to non-Class 1E IDS battery monitors and 
Class 1E IDS circuits. No safety analysis or design basis acceptance 
limit/criterion is challenged or exceeded by the proposed changes 
and no margin or 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. 52-025 and 52-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, 
Georgia

    Date of amendment request: September 13, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16257A711.
    Description of amendment request: The amendment request proposes 
changes to the Updated Final Safety Analysis Report (UFSAR) in the form 
of departures from the incorporated plant-specific Design Control 
Document Tier 2* information. The proposed departure consists of 
changes to Tier 2* information in the UFSAR to change the provided 
minimum reinforcement area in the column line 7.3 wall from elevation 
82'-6'' to elevation 100'-0''.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    As indicated in the UFSAR Subsection 3H.5.1.2, the wall at 
column line 7.3 is a shear wall that connects the shield building 
and the nuclear island exterior wall at column line I. Deviations 
were identified in

[[Page 70187]]

the constructed wall from the design requirements. The wall was 
repaired in accordance with American Concrete Institute (ACI) 349-
01. This change impacts UFSAR Table 3H.5-5. For the south face of 
the Vogtle Unit 3 column line 7.3 wall, the provided minimum steel 
for wall section 11 for the vertical reinforcement from the wall 
segment of elevation 82'-6'' to 100'-0'' is decreased from 3.12 
in\2\/ft to 3.08 in\2\/ft. The change of the provided versus 
required vertical reinforcing steel does not change the performance 
of the affected portion of the auxiliary building for postulated 
loads. The criteria and requirements of ACI 349-01 provide a margin 
of safety to structural failure. The design of the auxiliary 
building structure conforms to criteria and requirements in ACI 349-
01 and therefore maintains the margin of safety. This change does 
not involve any accident initiating components or events, thus 
leaving the probabilities of an accident unaltered. The reduced 
margin does not adversely affect any safety-related structures or 
equipment nor does the reduced margin reduce the effectiveness of a 
radioactive material barrier. Thus, the proposed change would not 
affect any safety-related accident mitigating function served by the 
containment internal structures.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The reduction of the provided versus required vertical 
reinforcing steel does not change the performance of the affected 
portion of the auxiliary building. As demonstrated by the continued 
conformance to the applicable codes and standards governing the 
design of the structures, the wall withstands the same effects as 
previously evaluated. There is no change to the design function of 
the wall, and no new failure mechanisms are identified as the same 
types of accidents are presented to the wall before and after the 
change.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change of the provided versus required vertical 
reinforcing steel, identified in UFSAR Table 3H.5-5, is not a 
significant reduction in the margin of safety. For the south face of 
the Vogtle Unit 3 column line 7.3 wall, the provided minimum steel 
for wall section 11 for the vertical reinforcement from the wall 
segment of elevation 82'-6'' to 100'-0'' is decreased from 3.12 
in\2\/ft to 3.08 in\2\/ft. The change of the provided versus 
required vertical reinforcing steel does not change the performance 
of the affected portion of the auxiliary building for postulated 
loads. The criteria and requirements of ACI 349-01 provide a margin 
of safety to structural failure. The design of the auxiliary 
building structure conforms to criteria and requirements in ACI 349-
01 and therefore maintains the margin of safety. The reduction in 
margin does not alter any design function, design analysis, or 
safety analysis input or result, and sufficient margin exists to 
justify departure from the Tier 2 * requirements for the wall. As 
such, because the system continues to respond to design basis 
accidents in the same manner as before without any changes to the 
expected response of the structure, no safety analysis or design 
basis acceptance limit/criterion is challenged or exceeded by the 
proposed changes. Accordingly, no significant safety margin is 
reduced by the change.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of amendment request: August 12, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16225A663.
    Description of amendment request: The amendments would modify the 
Technical Specifications (TSs) for Units 1, 2, and 3 by revising TS 
4.3.1.2, ``Fuel Storage Criticality,'' to preclude the placement of 
fuel in the new fuel storage vaults. This TS change would remove the 
existing TS 4.3.1.2 criticality criteria wording in its entirety, and 
replaces it with language that specifically restricts the placement of 
fuel in the new fuel storage vaults.
    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 does not change the fuel handling 
processes, the fuel handling equipment, or require alteration of the 
plant fuel storage systems. The amendment places a restriction on 
use of the new fuel storage vaults, requiring that new fuel be 
placed only in the spent fuel pool racks. Because no changes to fuel 
handling equipment, fuel storage systems, or fuel handling processes 
are involved, the proposed amendment does not increase the 
probability or consequences of a fuel handling accident.
    Therefore, the proposed change does not increase the probability 
or consequences of a previously evaluated accident.
    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 modification to the Technical Specifications does 
not require changes to the plant hardware or alter the operating 
characteristics of any plant system. As a result, no new failure 
modes are being introduced. Therefore, the change does not introduce 
a new or different kind of accident from those previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to TS 4.3.1.2 ensures that the criticality 
margins of safety for fuel storage are maintained, by excluding the 
new fuel storage vault as an approved fuel storage location. The 
change restricts the storage of new fuel to the spent fuel pool 
racks, which are fully analyzed from a criticality standpoint. The 
change does not physically alter the fuel storage systems, or modify 
fuel storage requirements in such a way as to degrade the margins of 
criticality 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: Sherry A. Quirk, General Counsel, Tennessee 
Valley Authority, 400 West Summit Hill Dr., WT 6A, Knoxville, TN 37902.
    NRC Acting Branch Chief: Jeanne A. Dion.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: May 10, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16134A069.
    Description of amendment request: The amendments would extend the 
Surry Power Station, Unit Nos. 1 and 2, Technical Specification 3.2, 
``Chemical and Volume Control System,'' paragraph E requirements for 
primary grade water

[[Page 70188]]

(PG) lockout from being applicable in Refueling Shutdown and Cold 
Shutdown to being applicable in Refueling Shutdown, Cold Shutdown, 
Intermediate Shutdown, and Hot Shutdown (except during the approach to 
critical and within 1 hour following reactor shutdown from reactor 
critical or power operation).
    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 conservatively imposes additional 
operational controls on the highest capacity flow path of PG to the 
Reactor Coolant System (RCS). These controls are currently credited 
in the boron dilution analysis in Refueling Shutdown and Cold 
Shutdown modes. The proposed change extends these controls into 
Intermediate and Hot Shutdown modes. As such, the change will 
provide defense against rapid reactivity insertions due to boron 
dilution events and reduce the probability of boron dilution events. 
The proposed change will have no impact on normal operating plant 
releases and will not increase the predicted radiological 
consequences of accidents postulated in the UFSAR [Updated Final 
Safety Analysis Report]. The proposed change makes no physical 
modifications and does not change plant design.
    Therefore, neither the probability of occurrence nor the 
consequences of any accident previously evaluated is significantly 
increased.
    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 is an extension of existing operational 
controls on PG flow to the RCS to include additional operating 
modes. The change precludes high flow rate boron dilutions in 
Intermediate and Hot Shutdown modes similar to the current TS 
requirement in Refueling and Cold Shutdown modes. It does not affect 
the operation of the emergency boration function of the Chemical and 
Volume Control System (CVCS).
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously analyzed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change provides defense against rapid reactivity 
insertions to potential boron dilution events in shutdown operating 
modes and reduces the probability of boron dilution events. As such, 
it increases the margin of safety for the boron dilution event.
    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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
    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.

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: January 18, 2016, as supplemented by 
letter dated June 20, 2016.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 5.5.2, ``Containment Leakage Rate Testing Program,'' 
to allow (1) an increase in the existing Type A Integrated Leakage Rate 
Testing Program test interval from 10 years to 15 years, in accordance 
with Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision 
3-A, ``Industry Guideline for Implementing Performance-Based Option of 
10 CFR part 50, appendix J,'' and the conditions and limitations 
specified in NEI 94-01, Revision 2-A; (2) adoption of an extension of 
the containment isolation valve leakage testing (Type C) frequency from 
the 60 months currently permitted by 10 CFR part 50, appendix J, Option 
B, to a 75-month frequency for Type C leakage rate testing of selected 
components, in accordance with NEI 94-01, Revision 3-A; (3) adoption of 
the use of American National Standards Institute/American Nuclear 
Society (ANSI/ANS)-56.8-2002, ``Containment System Leakage Testing 
Requirements''; and (4) adoption of a more conservative grace interval 
of 9 months for Type A, Type B, and Type C leakage tests, in accordance 
with NEI 94-01, Revision 3-A.
    The amendments also made the following administrative changes: (1) 
Deletion of the information regarding the performance of containment 
visual inspections as required by Regulatory Position C.3, as the 
containment inspections are addressed in TS Surveillance Requirement 
3.6.1.1, and (2) deletion of the information regarding the performance 
of the next Catawba Nuclear Station, Unit 1, Type A test no later than 
November 13, 2015, and the next Catawba Nuclear Station, Unit 2, Type A 
test no later than February 6, 2008, as both Type A tests have already 
occurred.
    Date of issuance: September 12, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: 286 (Unit 1) and 282 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16229A113; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments

[[Page 70189]]

revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: March 15, 2016 (81 FR 
13839). The supplemental letter dated June 20, 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 12, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 18, 2016, as supplemented by 
letter dated June 30, 2016.
    Brief description of amendment: The amendments modified Technical 
Specification (TS) 5.5.2, ``Containment Leakage Rate Testing Program,'' 
for a one-time extension to the 10-year frequency of the integrated 
leakage rate test (ILRT) or Type A test. This revision extends the 
period from 10 years to 10.5 years between successive tests, changing 
the performance of the next ILRT from fall 2017 to spring 2019 for Unit 
1 and from spring 2017 to fall 2018 for Unit 2.
    Date of issuance: September 26, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 290 (Unit 1) and 269 (Unit 2). A publicly available 
version is in ADAMS under Accession No. ML16236A053; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: May 10, 2016 (81 FR 
28894). The supplemental letter dated June 30, 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 26, 2016.
    No significant hazards consideration comments received: No.

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

Duke Energy Progress, Inc., Docket No. 50-261; H. B. Robinson Steam 
Electric Plant Unit No. 2 (RNP), Darlington County, South Carolina

Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear 
Power Plant, Unit 1, (HNP), Wake and Chatham Counties, North Carolina

    Date of amendment request: February 1, 2016.
    Description of amendment request: The amendments revised the 
licensee's name from Duke Energy Progress, Inc. to Duke Energy 
Progress, LLC.
    Date of issuance: September 13, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 271 and 299 (BSEP); 152 (HNP); 246 (RNP). A 
publicly-available version is in ADAMS under Accession No. ML16217A118; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-71, DPR-62 (BSEP), NPF-
63 (HNP), and NFP-23 (RNP): Amendments revised the Renewed Facility 
Operating Licenses.
    Date of initial notice in Federal Register: April 12, 2016 (81 FR 
21596).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 13, 2016.
    No significant hazards consideration comments received: No.

Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear 
Power Plant (HNP), Unit 1, Wake and Chatham Counties, North Carolina

    Date of amendment request: October 29, 2015, as supplemented by 
letters dated, February 16, 2016, August 8 and 26, 2016, and September 
8 and 16, 2016.
    Brief description of amendment: The amendment revised Technical 
Specifications to allow the `A' Emergency Service Water (ESW) pump to 
be inoperable for 14 days to allow for the replacement of the `A' Train 
ESW pump. The amendment is applicable on a one-time basis.
    Date of issuance: September 16, 2016.
    Effective date: As of the date of issuance and shall be implemented 
by October 29, 2016.
    Amendment No.: 153. A publicly-available version is in ADAMS under 
Accession No. ML16253A059; documents related to this amendment are 
listed in the Safety Evaluation (SE) enclosed with the amendment.
    Renewed Facility Operating License No. NPF-63: Amendment revised 
the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 5, 2016 (81 FR 
260). The supplemental letters dated February 16, 2016, August 8 and 
26, 2016, and September 8 and 16, 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 an SE dated September 16, 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 application for amendment: March 18, 2016.
    Brief description of amendment: The amendments revised the 
technical specifications (TSs) on a change to the method of calculating 
core reactivity for the purpose of performing the Reactivity Anomalies 
surveillance.
    Date of issuance: September 15, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: Unit 1-224 and Unit 2-158. A publicly-available 
version is in ADAMS under Accession No. ML16188A029; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License No. DPR-63 and NPF-69: The 
amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: May 10, 2016 (81 FR 
28897).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 15, 2016.
    No significant hazards consideration comments received: No.

[[Page 70190]]

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant (PNPP), Unit No. 1, Lake County, Ohio

    Date of amendment request: October 29, 2015, as supplemented by 
letter dated April 22, 2016.
    Brief description of amendment: The amendment revised the PNPP 
emergency action level (EAL) scheme to one based on the Nuclear Energy 
Institute (NEI) guidance in NEI 99-01, Revision 6, ``Development of 
Emergency Action Levels for Non-Passive Reactors.''
    Date of issuance: September 14, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment No.: 173. A publicly-available version is in ADAMS under 
Accession No. ML16158A331; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-58: The amendment revised the 
Facility Operating License to authorize revision to the PNPP emergency 
plan.
    Date of initial notice in Federal Register: December 22, 2015 (80 
FR 79620). The supplemental letter dated 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 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 14, 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: January 19, 2016, as supplemented by a 
letter dated May 6, 2016.
    Brief description of amendments: The amendments revised the 
Operating Licenses' licensing basis to allow elimination of the end-of-
cycle moderator temperature coefficient (MTC) surveillance test as 
supported by NRC-Approved Topical Report CE NPSD-91 1-A and Amendment 
1-A, ``Analysis of Moderator Temperature Coefficients in Support of a 
Change in the Technical Specification End of Cycle Negative MTC 
Limit,'' and St. Lucie specific supporting information. The amendments 
also add NRC-approved Westinghouse PARAGON Topical Report WCAP-16045-P-
A, Revision 0, ``Qualification of the Two-Dimensional Transport Code 
PARAGON,'' to the Technical Specification list of Core Operating Limits 
Report methodologies.
    Date of issuance: September 19, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 235 and 185. A publicly-available version is in 
ADAMS under Accession No. ML16183A138; documents related to these 
amendments are listed in the Safety Evaluation (SE) enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 2016 (81 FR 
17506). The supplemental letter dated May 6, 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 an SE dated September 19, 2016.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 28th 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. 2016-24321 Filed 10-7-16; 8:45 am]
 BILLING CODE 7590-01-P