Biweekly Notice: Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 62926-62935 [2016-21998]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 81, Number 177 (Tuesday, September 13, 2016)]
[Notices]
[Pages 62926-62935]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-21998]


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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0188]


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 August 16, 2016, to August 29, 2016. The 
last biweekly notice was published on August 30, 2016.

DATES: Comments must be filed by October 13, 2016. A request for a 
hearing must be filed by November 14, 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-2016-0188. 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: Paula Blechman, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-2242, email: Paula.Blechman@nrc.gov.

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2016-0188, 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-2016-0188.
     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-2016-0188, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

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

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

[[Page 62927]]

timely way would result, for example, in derating or shutdown of the 
facility. If the Commission takes action prior to the expiration of 
either the comment period or the notice period, it will publish in the 
Federal Register a notice of issuance. If the Commission makes a final 
no significant hazards consideration determination, any hearing will 
take place after issuance. The Commission expects that the need to take 
this action will occur very infrequently.

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

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and a petition to intervene (petition) 
with respect to issuance of the amendment to the subject facility 
operating license or combined license. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested person(s) should consult a 
current copy of 10 CFR 2.309, which is available at the NRC's PDR, 
located at One White Flint North, Room O1-F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852. The NRC's regulations are 
accessible electronically from the NRC Library on the NRC's Web site at 
https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a 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 amendment under consideration. The 
contention must be one which, if proven, would entitle the petitioner 
to relief. A petitioner who fails to satisfy these requirements with 
respect to at least one contention will not be permitted to participate 
as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that person's admitted 
contentions, including the opportunity to present evidence and to 
submit a cross-examination plan for cross-examination of witnesses, 
consistent with the NRC's regulations, policies, and procedures.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of any amendment unless the Commission finds an imminent 
danger to the health or safety of the public, in which case it will 
issue an appropriate order or rule under 10 CFR part 2.
    A State, local governmental body, Federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1).
    The petition should state the nature and extent of the petitioner's 
interest in the proceeding. The petition should be submitted to the 
Commission by November 14, 2016. The petition must be filed in 
accordance with the filing instructions in the ``Electronic Submissions 
(E-Filing)'' section of this document, and should meet the requirements 
for petitions set forth in this section, except that under 10 CFR 
2.309(h)(2) a State, local governmental body, or Federally-recognized 
Indian Tribe, or agency thereof does not need to address the standing 
requirements in 10 CFR 2.309(d) if the facility is located within its 
boundaries. A State, local governmental body, Federally-recognized 
Indian Tribe, or agency thereof may also have the opportunity to 
participate under 10 CFR 2.315(c).
    If a hearing is granted, any person who does not wish, or is not 
qualified, to become a party to the proceeding may, in the discretion 
of the presiding officer, be permitted to make a limited appearance 
pursuant to the provisions of 10 CFR 2.315(a). A person making a 
limited appearance may make an oral or written statement of position on 
the issues, but may not otherwise participate in the proceeding. A 
limited appearance may be made at any session of the hearing or at any 
prehearing conference, subject to the limits and conditions as may be 
imposed by the presiding officer. Details regarding the opportunity to 
make a limited appearance will be provided by the presiding officer if 
such sessions are scheduled.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene (hereinafter 
``petition''), and documents filed by interested governmental entities

[[Page 62928]]

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 request or 
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.

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

    Date of amendment request: June 30, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16194A342.
    Description of amendment request: The amendments would revise the 
Technical Specification (TS) requirements for the auxiliary feedwater 
(AFW) system. The proposed changes include conforming administrative 
changes to the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 62929]]


    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change requires that three AFW steam supplies must 
be operable. The AFW system is not an initiator of any accident 
previously evaluated; therefore, the probability of occurrence of an 
accident is not significantly affected by the proposed changes. The 
change does not involve a significant increase in the consequences 
of an accident because three steam supplies are needed to ensure 
that the AFW system can perform its specified function for all 
postulated events in the presence of a single failure. The proposed 
changes do not adversely impact the ability of the AFW system to 
mitigate the consequences of accidents previously evaluated because 
the proposed change reduces the allowable out of service time for a 
single inoperable steam supply.
    Therefore, the proposed change 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 requires that three steam supplies are 
available to ensure that the AFW system can perform its specified 
function in the presence of a single failure. As such, the proposed 
change adds a more restrictive requirement than currently exists 
because the LCO [limiting condition for operation] can no longer be 
met with only two AFW steam supplies operable.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
The proposed change does not create any new failure modes for 
existing equipment or any new limiting single failures. 
Additionally, the proposed change does not involve a change in the 
methods governing normal plant operation, and all safety functions 
will continue to perform as previously assumed in the accident 
analyses. Thus, the proposed change does not adversely affect the 
design function or operation of any plant structures, systems, or 
components.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The proposed change does not challenge the performance or integrity 
of any safety-related 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change adds a more restrictive requirement than 
currently exists because the LCO can no longer be met with only two 
AFW steam supplies operable. The proposed change will not adversely 
affect the operation of plant equipment or the function of equipment 
assumed in the accident analysis. The proposed amendment does not 
involve changes to any safety analyses assumptions, safety limits, 
or limiting safety system settings. The change does not adversely 
impact plant operating margins or the reliability of equipment 
credited in the safety analyses.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB, 
Juno Beach, FL 33408-0420.
    NRC Acting Branch Chief: Tracy J. Orf.

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

    Date of amendment request: June 2, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16154A226.
    Description of amendment request: The proposed changes revise the 
Combined Licenses (COL) concerning the design details of the safety-
related passive core cooling system (PXS), the nonsafety-related normal 
residual heat removal system (RNS), and the nonsafety-related 
containment air filtration system (VFS). The amendment request proposes 
changes to the Updated Final Safety Analysis Report (UFSAR) in the form 
of departures from the plant-specific Design Control Document (DCD) 
Tier 2 information, and involves changes to related plant-specific DCD 
Tier 1 information, with corresponding changes to the associated COL 
Appendix C information. Because, this proposed change requires a 
departure from Tier 1 information in the Westinghouse Advanced Passive 
1000 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 changes do not affect the operation of any systems 
or equipment that initiate an analyzed accident or alter any 
structures, systems, and components (SSCs) accident initiator or 
initiating sequence of events. The proposed changes result from 
identifying PSX, RNS, and VFS piping lines required to be described 
in the licensing basis as ASME [American Society of Mechanical 
Engineers] Code Section III, evaluated to meet the LBB [leak-before-
break] design criteria, or designed to withstand combined normal and 
seismic design basis loads without a loss of functional capability. 
Neither planned or inadvertent operation nor failure of the PXS, 
RNS, or VFS is an accident initiator or part of an initiating 
sequence of events for an accident previously evaluated. Therefore, 
the probabilities of the accidents evaluated in the UFSAR are not 
affected.
    The proposed changes do not have an adverse impact on the 
ability of the PXS, RNS, or VFS to perform their design functions. 
The design of the PXS, RNS, and VFS continues to meet the same 
regulatory acceptance criteria, codes, and standards as required by 
the UFSAR. In addition, the changes ensure that the capabilities of 
the PXS, RNS, and VFS to mitigate the consequences of an accident 
meet the applicable regulatory acceptance criteria, and there is no 
adverse effect on any safety-related SSC or function used to 
mitigate an accident. The changes do not affect the prevention and 
mitigation of other abnormal events, e.g., anticipated operational 
occurrences, earthquakes, floods and turbine missiles, or their 
safety or design analyses. Therefore, the consequences of the 
accidents evaluated in the UFSAR are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The 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 result from 
identifying PXS, RNS, and VFS piping lines required to be described 
in the licensing basis as ASME Code Section III, evaluated to meet 
the LBB design criteria, or designed to withstand combined normal 
and seismic design basis loads without a loss of functional 
capability. These proposed changes do not adversely affect any other 
PXS, RNS, VFS, or SSC design functions or methods of operation in a 
manner that results in a new failure mode, malfunction, or sequence 
of events that affect safety-related or nonsafety-related equipment. 
Therefore, this activity does not allow for a new fission product 
release path, result in a new fission product barrier failure mode, 
or create a new sequence of events that results in significant fuel 
cladding failures.

[[Page 62930]]

    Therefore, the requested 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 existing safety margins. The 
proposed changes ensure that PXS, RNS, and VFS design requirements 
and design functions are met. The proposed changes maintain existing 
safety margin through continued application of the existing 
requirements of the UFSAR, while adding additional design features 
to ensure the PXS, RNS, and VFS perform the design functions 
required to meet the existing safety margins. Therefore, the 
proposed changes satisfy the same design functions in accordance 
with the same codes and standards as stated in the UFSAR. These 
changes do not adversely affect any design code, function, design 
analysis, safety analysis input or result, or design/safety margin. 
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 requested amendment does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius, LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Acting Branch Chief: Jennifer Dixon-Herrity.

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

    Date of amendment request: July 1, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16188A268.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 5.5.12, ``Primary Containment Leakage Rate 
Testing Program,'' by (1) increasing the existing Type A integrated 
leak test program test interval from 10 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'' (ADAMS Accession No. ML12221A202), and the 
conditions and limitations specified in NEI 94-01, Revision 2-A (ADAMS 
Accession No. ML100620847); (2) extending the containment isolation 
valve leakage test (Type C) frequency from 60 months to 75 months in 
accordance with NEI 94-01, Revision 3-A; (3) adopting the use of the 
American National Standards Institute/American Nuclear Society 56.8-
2002, ``Containment System Leakage Testing Requirements;'' and (4) 
adopting a more conservative grace interval of 9 months for Type A, 
Type B, and Type C leakage test in accordance with NEI 94-01, Revision 
3-A.
    Additionally, the amendment would also delete the information from 
TS 5.5.12 regarding the performance of Type A tests for Unit Nos. 1 and 
2 in 2008 and 2010, respectively, on the basis that both tests have 
already occurred.
    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 to the Technical Specifications (TS) 
involves the extension of the Edwin I. Hatch Nuclear Plant (HNP), 
Units 1 and 2 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 Type 
A test frequency from three in ten years to one in fifteen years, 
measured as an increase in the total integrated plant dose risk for 
those accident sequences influenced by Type A testing, is 9.90E-03 
person-rem/yr using the Electric Power Research Institute (EPRI) 
guidance values, and drops to 1.96E-03 person-rem/yr using the EPRI 
Expert Elicitation values. Therefore, this proposed extension does 
not involve a significant increase in the probability of an accident 
previously evaluated.
    In addition, as documented in NUREG-1493, ``Performance-Based 
Containment Leak-Test Program,'' Types 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 HNP, Units 1 and 2 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 leakage rate test (LLRT) 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 American Society of Mechanical Engineers (ASME) 
Section XI, 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 onetime extensions of the 
ILRT [integrated leak rate test] test frequency for both Units 1 and 
2. These exceptions were for activities that have already taken 
place; therefore, their deletion is solely an administrative action 
that has no effect on any component and no physical impact on how 
the units are 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 
HNP, Unit 1 and 2 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. The proposed change 
does not involve a physical change to the plant (i.e., no new or 
different type of equipment will be installed) nor does it alter the 
design, configuration, or change the manner in which the plant is 
operated or controlled beyond the standard functional capabilities 
of the equipment.
    The proposed amendment also deletes exceptions previously 
granted to allow one-time extensions of the ILRT test frequency for 
both Units 1 and 2. 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 units are operated.

[[Page 62931]]

    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 5.5.12 involves the extension of 
the HNP, Units 1 and 2 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 HNP, 
Units 1 and 2. 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, ``Industry Guideline 
for Implementing Performance-Based Option of 10 CFR part 50, 
Appendix J,'' 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 
XI, and TS 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 
both HNP Units 1 and 2. These exceptions were for activities that 
have taken place; therefore, their deletion is solely an 
administrative action and does not change how the units are 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.
    Attorney for licensee: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, Inc., 40 Inverness Center 
Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: July 25, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16207A496.
    Description of amendment request: The amendment request proposes 
changes to plant-specific Tier 1 information, with corresponding 
changes to the associated combined license (COL) Appendix C 
information, and involves associated Tier 2 information in the Updated 
Final Safety Analysis Report (UFSAR). 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. Specifically, the requested amendment proposes 
clarifications to a plant-specific Tier 1 (and COL Appendix C) table 
and a UFSAR table in regard to the inspections of the excore source, 
intermediate, and power range detectors.
    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 Nuclear Regulatory 
Commission (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 change to specify the inspection of the excore 
source, intermediate, and power range detectors is done to verify 
that aluminum surfaces are contained in stainless steel or titanium, 
and avoids the introduction of aluminum into the post-loss of 
coolant accident (LOCA) containment environment due to detector 
materials. The proposed change does not alter any safety related 
functions. The materials of construction are compatible with the 
post [-]LOCA conditions inside containment and will not 
significantly contribute to hydrogen generation or chemical 
precipitates. The change does not affect the operation of any 
systems or equipment that initiate an analyzed accident or alter any 
structures, systems, and components (SSC) accident initiator or 
initiating sequence of events.
    The change does not impact the support, design, or operation of 
mechanical and fluid systems. There is no change to plant systems or 
the response of systems to postulated accident conditions. There is 
no change to the predicted radioactive releases due to normal 
operation or postulated accident conditions. Consequently, the plant 
response to previously evaluated accidents or external events is not 
adversely affected, 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 change does 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 change to specify the 
inspection of the excore source, intermediate, and power range 
detectors is done to verify that aluminum surfaces are contained in 
stainless steel or titanium, and avoids the introduction of aluminum 
into the post[-]LOCA containment environment due to detector 
materials. In addition, the proposed change to the [inspection, 
test, analysis, and acceptance criteria (ITAAC)] verified materials 
of construction does not alter the design function of the excore 
detectors. The detector canning materials of construction are 
compatible with the post-LOCA containment environment and do not 
contribute a significant amount of hydrogen or chemical 
precipitates. The change to the ITAAC aligns the inspection with the 
Tier 2 design feature. Consequently, because the excore detectors 
functions are unchanged, there are no adverse effects on accidents 
previously evaluated in the UFSAR.
    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 specify the inspection of the excore 
source, intermediate, and power range detectors is done to verify 
that aluminum surfaces are contained in stainless steel or titanium, 
and avoids the introduction of aluminum into the post-LOCA 
containment environment, does not alter any safety-related 
equipment, 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.
    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

[[Page 62932]]

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 Acting Branch Chief: Jennifer Dixon-Herrity.

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

    Date of amendment request: June 14, 2016, as revised on August 12, 
2016. A publicly-available version is in ADAMS under Accession Nos. 
ML16166A409 and ML16225A655, respectively.
    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* and associated Tier 2 information. Specifically, the 
proposed departures consist of changes to the UFSAR to revise the 
details of the structural design of auxiliary building floors.
    A biweekly Federal Register notice was published on August 2, 2016, 
providing an opportunity to comment, request a hearing, and petition 
for leave to intervene for a License Amendment Request (LAR) for the 
VEGP combined licenses. Since that time, the licensee has submitted a 
revision to the original LAR, dated August 12, 2016, that increases the 
scope of the LAR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The design functions of the auxiliary building floors are to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the auxiliary 
building. The auxiliary building is a seismic Category I structure 
and is designed for dead, live, thermal, pressure, safe shutdown 
earthquake loads, and loads due to postulated pipe breaks. The 
proposed changes to UFSAR descriptions and figures are intended to 
address changes in the detail design of floors in the auxiliary 
building. The thickness and strength of the auxiliary building 
floors are not reduced. As a result, the design function of the 
auxiliary building structure is not adversely affected by the 
proposed changes. 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 
changes described create any new accident precursors. Therefore, the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The changes to UFSAR descriptions are proposed to address 
changes in the detail design of floors in the auxiliary building. 
The thickness, geometry, and strength of the structures are not 
adversely altered. The concrete and reinforcement materials are not 
altered. The properties of the concrete are not altered. The changes 
to the design details of the auxiliary building structure do not 
create any new accident precursors. As a result, the design function 
of the auxiliary building structure is not 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.
    The criteria and requirements of American Concrete Institute 
(ACI) 349 and American Institute of Steel Construction (AISC) N690 
provide a margin of safety to structural failure. The design of the 
auxiliary building structure conforms to criteria and requirements 
in ACI 349 and AISC N690 and therefore maintains the margin of 
safety. Analysis of the connection design confirms that code 
provisions are appropriate to the floor to wall connection. The 
proposed changes to the UFSAR address changes in the detail design 
of floors in the auxiliary building. The proposed changes also 
incorporate the requirements for development and anchoring of headed 
reinforcement which were previously approved. There is no change to 
design requirements of the auxiliary building structure. There is no 
change to the method of evaluation from that used in the design 
basis calculations. There is not a significant change to the in 
structure response spectra. Therefore, the proposed amendment does 
not result in 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 Acting Branch Chief: Jennifer Dixon-Herrity.

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

    Date of amendment request: June 7, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16159A403.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) to allow use of component cooling system 
(CCS) pump 2B-B to support CCS Train 1B operability when the normally-
aligned CCS pump is inoperable. The amendment would provide increased 
flexibility for maintaining CCS operability.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    Response: No.
    The proposed change to allow the use of the CCS pump 2B-B to 
support Train 1B operability does not result in any physical changes 
to plant safety-related structures, systems, or components (SSCs). 
The CCS functions to remove plant system heat loads during normal, 
shutdown, and accident conditions. The CCS will continue to perform 
this function with equipment qualified to the same standards. The 
CCS is not an accident initiator, but instead performs accident 
mitigation functions by serving as the heat sink for safety-related 
equipment, ensuring the conditions and assumptions credited in the 
accident analyses are preserved. Therefore, the proposed change does 
not involve a significant increase in the probability of an accident 
previously evaluated.
    The purpose of this change is to modify the CCS TS to allow the 
use of CCS pump 2B-B to replace CCS pump C-S in supporting Train 1B 
operability. The proposed change provides assurance that the minimum 
conditions necessary for the CCS to perform its heat removal safety 
function are maintained. Accordingly, operation as specified by the 
addition of the Notes and the additional surveillance requirement 
will provide the necessary assurance that fuel cladding, reactor 
coolant system pressure boundary, and containment integrity limits 
are not challenged during worst-case pos[t]-accident conditions. CCS 
pump C-S and pump 2B-B are identical pumps with identical controls 
except that the CCS pump 2B-B does not receive an automatic start 
signal from a Unit 1 Safety Injection (SI) actuation signal. To 
compensate for the lack

[[Page 62933]]

of the SI actuation signal, CCS pump 2B-B is required to be in 
operation to support Unit 1 operation when substituting for CSS pump 
C-S. With the CCS pump 2B-B in operation, the pump will continue to 
operate following a SI actuation signal. Accordingly, the 
conclusions of the accident analyses will remain as previously 
evaluated such that there will be no significant increase in the 
consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve any physical changes to 
plant safety-related SSCs or alter the modes of plant operation in a 
manner that will change the design function or operation of the CCS. 
The proposed additional limits on CCS alignment and CCS pump 2B-B 
operation provide assurance that the conditions and assumptions 
credited in the accident analyses are preserved. Thus, the plant's 
overall ability to reject heat to the ultimate heat sink during 
normal operation, normal shutdown, and worst-case accident 
conditions will not be significantly affected by this proposed 
change. Because the safety and design requirements continue to be 
met and the integrity of the reactor coolant system pressure 
boundary is not challenged, no new credible failure mechanisms, 
malfunctions, or accident initiators are created, and there will be 
no effect on the accident mitigating systems in a manner that would 
significantly degrade the plant's response to an accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change modifies the CCS TS to maintain the CCS 
Train 1B operable while aligned with CCS pump 2B-B. With CCS pump 
2B-B in operation when aligned to CCS Train 1B, CCS pump 2B-B will 
operate to provide the CCS accident mitigation function if a 
postulated accident occurs. CCS pumps C-S and 2B-B are identical 
pumps and will perform the same function with this change, resulting 
in essentially no change in the safety margin before the change to 
the safety margin after the change. Accordingly, the proposed change 
will not significantly reduce the margin of safety of any SSCs that 
rely on the CCS for heat removal to perform their safety-related 
functions.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Sherry A. Quirk, Executive Vice 
President and General Counsel, Tennessee Valley Authority, 400 West 
Summit Hill Drive, 6A Tower West, Knoxville, TN 37902.
    NRC Acting Branch Chief: Tracy J. Orf.

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

    Date of amendment request: May 16, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16137A572.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TS) to correct an administrative error 
regarding the steam generator (SG) narrow range (NR) level specified in 
Surveillance Requirement (SR) 3.4.6.3.
    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 WBN Unit 2 TS SR 3.4.6.3 is being amended due [to] an 
administrative error that was incorporated into the initial 
submittal of the WBN Unit 2 Revision 0 TS. The impact of this 
amendment will not affect how plant equipment is operated or 
maintained. This proposed amendment corrects an error in the 
required SG NR level minimum value, while in Mode 4, from 32% to 6%. 
The purpose for this SR is to ensure the steam generator u-tubes are 
covered with water on the secondary side of the tubes. For the WBN 
Unit 2 SGs the lower SG NR level tap is above the top of the U-
tubes. Therefore, a 6% NR level ensures the U-tubes are covered with 
water. There are no changes to the physical plant or analytical 
methods.
    The proposed amendment does not impact any accident initiators, 
analyzed events, or assumed mitigation of accident or transient 
events. The proposed changes do not involve the addition or removal 
of any equipment or any design changes to the facility. The proposed 
changes do not affect any design functions, or analyses that verify 
the capability of structures, systems, and components (SSCs) to 
perform a design function. The proposed changes do not change any of 
the accidents previously evaluated in the Final Safety Analysis 
Report (FSAR). The proposed changes do not affect SSCs, operating 
procedures, and administrative controls that have the function of 
preventing or mitigating any of these accidents.
    Therefore, the proposed change 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.
    No actual plant equipment or accident analyses will be affected 
by the proposed amendment. The proposed amendment will not change 
the design function of any SSCs or result in any new failure 
mechanisms, malfunctions, or accident initiators not considered in 
the design and licensing bases. The proposed amendment does not 
impact any accident initiators, analyzed events, or assumed 
mitigation of accident or transient events.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment does not involve any physical changes to 
the plant or alter the manner in which plant systems are intended to 
be operated, maintained, modified, tested, or inspected. The 
proposed amendment does not alter the manner in which safety limits, 
limiting safety system settings or limiting conditions for operation 
are determined. The safety analysis acceptance criteria are not 
affected by this change. The proposed amendment will not result in 
plant operation in a configuration outside the design basis. The 
proposed amendment does not adversely affect systems that respond to 
safely shutdown the plant and to maintain the plant in a safe 
shutdown condition.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not affected by this change. The proposed change will not result 
in plant operation in a configuration outside the design basis.
    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. Sherry A. Quirk, Executive Vice 
President and General Counsel, Tennessee Valley Authority, 400 West 
Summit Hill Drive, 6A Tower West, Knoxville, Tennessee 37902.
    NRC Acting Branch Chief: Tracy J. Orf.

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

    During the period since publication of the last biweekly notice, 
the

[[Page 62934]]

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.

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

    Date of application for amendment: January 29, 2016, as 
supplemented by letters dated June 2 and 10, 2016.
    Brief description of amendment: The proposed amendment approved the 
post- loss-of-coolant-accident drawdown time for secondary containment 
from 12 to 19 minutes as described in the CPS Updated Safety Analysis 
Report and technical specification bases.
    Date of issuance: August 17, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 210. A publicly-available version is in ADAMS under 
Accession No. ML16217A332; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-62: The amendment revised the 
Updated Safety Analysis Report.
    Date of initial notice in Federal Register: April 12, 2016 (81 FR 
21599). The supplemental letters dated June 2 and 10, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 17, 2016.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 19, 2015, as 
supplemented by letter dated April 11, 2016.
    Brief description of amendments: The amendments revised LSCS 
technical specifications (TS), Section 2.1.1, ``Reactor Core SLs 
[safety limits],'' to reflect a lower reactor steam dome pressure 
stated for reactor core SLs, Sections 2.1.1.1 and 2.1.1.2. 
Specifically, the amendment reduced the reactor steam dome pressure in 
TS SLs, Sections 2.1.1.1 and 2.1.1.2, from 785 psig [pound per square 
inch gage] to 700 psia [pound per square inch absolute]. This change to 
TS, Section 2.1.1, was identified as a result of 10 CFR part 21, 
General Electric report SC05-03, ``Potential to Exceed Low Pressure 
Technical Specification Safety Limit.'' This change is valid for the 
NRC-approved pressure range pertinent to the critical power 
correlations applied to the fuel types in use at LSCS.
    Effective date: As of the date of issuance and the amendment shall 
be implemented for LSCS, Unit 1, within 30 days of issuance of the 
amendment. Also, the amendment shall be implemented for LSCS, Unit 2, 
prior to startup following refueling outage L2R16 in February 2017.
    Amendment Nos.: 220 for NPF-11, Unit 1, and 206 for NPF-18, Unit 2. 
The publicly-available version of documents related to these amendments 
are listed in the Safety Evaluation enclosed with the amendments in 
ADAMS under Accession No. ML16155A110.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Licenses and Technical Specifications.
    Date of initial notice in Federal Register: February 2, 2016 (81 FR 
5497). The supplemental letter dated April 11, 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 August 23, 2016.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook 
Nuclear Plant (CNP), Unit 2, Berrien County, Michigan

    Date of amendment request: October 19, 2015, as supplemented by 
letters dated January 21, 2016, and April 18, 2016.
    Brief description of amendment: The amendment revised the CNP, Unit 
2, technical specification (TS) requirements for the Engineered Safety 
Feature Actuation System Instrumentation by adding a new Condition for 
inoperable required channels for main feedwater pump trips, and by 
adding a footnote to the Applicable Mode column of TS Table 3.3.2-1 to 
reflect the new Condition.
    Date of issuance: August 19, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment No.: 313. A publicly-available version is in ADAMS under 
Accession No. ML16216A181; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-74: The amendment 
revises the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: December 22, 2015 (80 
FR 79621). The supplemental letters dated January 21, 2016, and April 
18, 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

[[Page 62935]]

Safety Evaluation dated August 19, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 6, 2015, as supplemented by 
letter dated April 12, 2016.
    Brief description of amendment: The amendment revised the value of 
reactor steam dome pressure specified within the Reactor Core Safety 
Limits Technical Specification (TS) 2.1.1. This resolved a 10 CFR part 
21, condition concerning a potential to momentarily violate Reactor 
Core Safety Limit (TSs 2.1.1.1 and 2.1.1.2) during a pressure regulator 
failure maximum demand (Open) transient.
    Date of issuance: August 18, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 295. A publicly-available version is in ADAMS under 
Accession No. ML16153A091; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-49: The amendment 
revised the License and TSs.
    Date of initial notice in Federal Register: November 10, 2015 (80 
FR 69713). The supplemental letter dated April 12, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 18, 2016.
    No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: September 11, 2015, as supplemented by 
letter dated April 8, 2016.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) by removing the current stored diesel fuel oil and 
lube oil numerical volume requirements from the TSs and replacing them 
with emergency diesel generator operating time requirements consistent 
with NRC-approved Technical Specifications Task Force (TSTF) Traveler 
TSTF-501, Revision 1, ``Relocate Stored Fuel Oil and Lube Oil Volume 
Values to Licensee Control,'' including plant-specific variances.
    Date of issuance: August 19, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 289. A publicly-available version is in ADAMS under 
Accession No. ML16182A363; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the License and TSs.
    Date of initial notice in Federal Register: November 24, 2015 (80 
FR 73239). The supplemental letter dated April 8, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated August 19, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 16, 2015, as supplemented by 
letter dated February 12, 2016.
    Brief description of amendment: The amendment authorized changes to 
the VEGP Units 3 and 4 Updated Final Safety Analysis Report in the form 
of departures from the incorporated plant-specific Design Control 
Document Tier 2* information. The proposed changes are related to 
changes to construction methods and construction sequence used for the 
composite floors and roof of the auxiliary building.
    Date of issuance: June 29, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 49. A publicly-available version is in ADAMS under 
Accession No. ML16146A734; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Combined License Nos. NPF-91 and NPF-92: The amendment 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: February 2, 2016 (81 FR 
5495). The supplemental letter dated February 12, 2016, provided 
additional information that did not expand the scope of the amendment 
request and did not change the NRC staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in the Safety Evaluation dated June 29, 2016.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 31st day of August 2016.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2016-21998 Filed 9-12-16; 8:45 am]
 BILLING CODE 7590-01-P