Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 7835-7847 [2016-02916]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 81, Number 30 (Tuesday, February 16, 2016)]
[Notices]
[Pages 7835-7847]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-02916]


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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0026]


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 (AEA), the U.S. Nuclear Regulatory Commission (NRC) is 
publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued, and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from January 16, 2016, to February 1, 2016. The 
last biweekly notice was published on February 2, 2016.

DATES: Comments must be filed by March 17, 2016. A request for a 
hearing must be filed by April 18, 2016.

ADDRESSES: You may submit comments by any of the following methods 
(unless

[[Page 7836]]

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-2016-0026. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: 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.

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2016-0026 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-2016-0026.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section of this document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2016-0026, facility name, unit 
number(s), application date, and subject in your comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at https://www.regulations.gov, as well as entering the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

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

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

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

    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Agency 
Rules of Practice and Procedure'' in 10 CFR part 2. Interested 
person(s) should consult a current copy of 10 CFR 2.309, which is 
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The 
NRC's regulations are accessible electronically from the NRC Library on 
the NRC's Web site at https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed within 60 days, the Commission or a presiding officer designated 
by the Commission or by the Chief Administrative Judge of the Atomic 
Safety and Licensing Board Panel, will rule on the request and/or 
petition; and the Secretary or the Chief Administrative Judge of the 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's

[[Page 7837]]

right under the Act to be made a party to the proceeding; (3) the 
nature and extent of the requestor's/petitioner's property, financial, 
or other interest in the proceeding; and (4) the possible effect of any 
decision or order which may be entered in the proceeding on the 
requestor's/petitioner's interest. The petition must also set forth the 
specific contentions which the requestor/petitioner seeks to have 
litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that person's admitted 
contentions, including the opportunity to present evidence and to 
submit a cross-examination plan for cross-examination of witnesses, 
consistent with NRC regulations, policies and procedures.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of any amendment unless the Commission finds an imminent 
danger to the health or safety of the public, in which case it will 
issue an appropriate order or rule under 10 CFR part 2.
    A State, local governmental body, Federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission by April 
18, 2016. The petition must be filed in accordance with the filing 
instructions in the ``Electronic Submissions (E-Filing)'' section of 
this document, and should meet the requirements for petitions for leave 
to intervene set forth in this section, except that under Sec.  
2.309(h)(2) a State, local governmental body, or Federally-recognized 
Indian Tribe, or agency thereof does not need to address the standing 
requirements in 10 CFR 2.309(d) if the facility is located within its 
boundaries. A State, local governmental body, Federally-recognized 
Indian Tribe, or agency thereof may also have the opportunity to 
participate under 10 CFR 2.315(c).
    If a hearing is granted, any person who does not wish, or is not 
qualified, to become a party to the proceeding may, in the discretion 
of the presiding officer, be permitted to make a limited appearance 
pursuant to the provisions of 10 CFR 2.315(a). A person making a 
limited appearance may make an oral or written statement of position on 
the issues, but may not otherwise participate in the proceeding. A 
limited appearance may be made at any session of the hearing or at any 
prehearing conference, subject to the limits and conditions as may be 
imposed by the presiding officer. Persons desiring to make a limited 
appearance are requested to inform the Secretary of the Commission by 
April 18, 2016.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the

[[Page 7838]]

participant must file the document using the NRC's online, Web-based 
submission form. In order to serve documents through the Electronic 
Information Exchange System, users will be required to install a Web 
browser plug-in from the NRC's Web site. Further information on the 
Web-based submission form, including the installation of the Web 
browser plug-in, is available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at https://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, in some instances, a request to intervene will require 
including information on local residence in order to demonstrate a 
proximity assertion of interest in the proceeding. With respect to 
copyrighted works, except for limited excerpts that serve the purpose 
of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
2 (ANO-2), Pope County, Arkansas

    Date of amendment request: December 22, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15356A657.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) to add a short Allowed Outage Time (AOT) 
to restore an inoperable system for conditions under which the existing 
specifications require a plant shutdown. The proposed amendment is 
consistent with the NRC-approved Technical Specifications Task Force 
(TSTF) change traveler TSTF-426, Revision 5, ``Revise or Add Actions to 
Preclude Entry into LCO [Limiting Condition for Operation] 3.0.3--
RITSTF [Risk-Informed TSTF] Initiatives 6b & 6c.'' The availability of 
TSTF-426, Revision 5, was published in the Federal Register on May 30, 
2013 (78 FR 32476). The AOT would be added to specifications governing 
the pressurizer heaters, containment spray trains, and control room 
emergency air conditioning and ventilation systems. In addition to the 
scope of the TSTF-426 TSs revisions, the amendment would add a TS 
Action to address a single pressurizer proportional heater group having 
a capacity of less than 150 kilowatts.
    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, with NRC staff revisions provided in [brackets], 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 provides a short AOT to restore an 
inoperable system for conditions under which the existing TSs 
require a plant shutdown to begin within one hour in accordance with 
LCO 3.0.3. In addition, a new TS Action associated with Pressurizer 
proportional heater capacity for a single proportional heater group 
is proposed. Entering into TS Actions is not an initiator of any 
accident previously evaluated. As a result, the probability of an 
accident previously evaluated is not significantly increased. The 
consequences of any accident previously evaluated that may occur 
during the proposed AOTs are no different from the

[[Page 7839]]

consequences of the same accident during the existing one-hour 
allowance. As a result, the consequences of any accident previously 
evaluated are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an 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.
    No new or different accidents result from utilizing the proposed 
change. The proposed change does not involve a physical alteration 
of the plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the proposed change does not impose any new 
or different requirements. The proposed change does not alter 
assumptions made in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change increases the time the plant may operate 
without the ability to perform an assumed safety function. The 
analyses in WCAP-16125-NP-A, ``Justification for Risk-Informed 
Modifications to Selected Technical Specifications for Conditions 
Leading to Exigent plant Shutdown,'' Revision 2, August 2010, 
demonstrated that there is an acceptably small increase in risk due 
to a limited period of continued operation in these conditions and 
that this risk is balanced by avoiding the risks associated with a 
plant shutdown. As a result, the change to the margin of safety 
provided by requiring a plant shutdown within one hour is not 
significant.
    The new Pressurizer proportional heater capacity Action permits 
72 hours to restore the affect heater group to an operable status, 
consistent with the STS [Standard TSs] and consistent with TS 
requirements associated with single train inoperabilities. The 
proportional heaters are not credited in the ANO-2 accident 
analyses, but aid in Pressurizer pressure control during a loss of 
offsite power event that results in the need to perform a natural 
circulation cool down of the plant. The associated STS bases for the 
standard 72-hour AOT assumes [that] the likelihood of a loss of 
offsite power event during this time period that would require a 
demand on the proportional heaters is minimal and acknowledges the 
use of non-vital powered backup heater groups absent a loss of 
offsite power event. Note also that under emergency conditions, an 
Emergency Diesel Generator or the Alternate AC [alternating current] 
Diesel Generator (i.e., Station Blackout diesel) can be aligned to 
power any of the non-vital Pressurizer backup heater groups. As a 
result, the change to the margin of safety provided by the new 72-
hour AOT for a single proportional heater train is not significant.
    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: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Meena K. Khanna.

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: October 15, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15301A765.
    Description of amendment request: The amendments would revise the 
St. Lucie Plant, Unit Nos. 1 and 2, Renewed Facility Operating 
Licenses' licensing bases to allow the use of the commercially 
available code ``Generation of Thermal-Hydraulic Information for 
Containments (GOTHIC Version 7.2b(QA)),'' to model the containment 
response following the inadvertent actuation of the containment spray 
system during normal plant operation (referred to as the vacuum 
analysis). The amendments would also update the licensing bases to 
credit the design-basis ability of the containment vessel to withstand 
a higher external pressure differential of 1.04 pounds per square inch 
(psi) (1.05 psi for Unit No. 2), and will update Technical 
Specification 3.6.1.4 for both units to revise the allowable 
containment operating pressure range.
    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.
    This proposed amendment is related to the analysis of the 
maximum external pressure that the reactor containment building will 
experience. A proposed change to the Technical Specifications will 
limit the allowable external pressure during operation to a value 
consistent with that considered in the analysis. The analysis is 
being revised to consider containment spray pump flow higher than 
previously considered. Containment spray pumps cool and depressurize 
the containment building; therefore, higher flow impacts the 
analysis of external pressure on the containment building. The 
proposed amendment is for the use of a different analysis 
methodology using the GOTHIC computer code instead of the A-TEMPT 
and WATEMPT codes that were originally used for the Unit 1 and Unit 
2 analyses respectively. The original codes are not currently 
available. The GOTHIC code is an accepted code for similar analysis. 
The analysis performed demonstrates that in the postulated event of 
an inadvertent start of two containment spray pumps, the loading the 
reactor containment building will experience is within the design of 
the structure. With this load, the stresses experienced by the 
reactor containment building remain below the code allowable 
stresses.
    The probability of occurrence of an event that would expose the 
containment building to external pressure is not increased by the 
change in the analysis methodology used. The probability of the 
initiating event, inadvertent start of both containment spray pumps, 
is unchanged.
    The consequences of an event where the containment building is 
exposed to external pressure will not be increased as the resulting 
external pressure on the containment vessel remains within the 
design, which provides a large margin to the buckling pressure.
    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.
    This proposed amendment changes the methodology for analyzing an 
event that results in exposing the reactor containment vessel to 
external pressure. A proposed change to the Technical Specifications 
will limit the external pressure during operation to a value 
consistent with the initial condition considered in the analysis. 
The potential for a new or different kind of accident is not created 
by the use of a different analysis methodology for a previously 
defined event.
    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.
    This proposed amendment changes the methodology for analyzing an 
event that results in exposing the reactor containment building to 
external pressure. A proposed change to the Technical Specifications 
will limit the allowable external pressure during operation to a 
value consistent with the starting point considered in the analysis. 
The technical evaluation demonstrates that the use of the GOTHIC 
computer code to determine maximum containment external pressure 
will result in realistic results similar

[[Page 7840]]

to the original analysis with the A-TEMPT and WATEMPT codes. The 
margin of safety in this analysis is maintained by assuring the 
resulting external pressure acting on the reactor containment vessel 
maintains significant margin to the buckling pressure in accordance 
with Section III of the ASME [American Society of Mechanical 
Engineers] code. For Unit 2, the original code of record limited the 
maximum external pressure to \1/3\ of the expected buckling 
pressure. The analysis of the increased external pressure for Unit 2 
has been performed in accordance with the original code of record. 
The original code of record for Unit 1 was under development at the 
time and made reference to ASME Section VIII for the analysis of 
external pressure. The rules of ASME Section VIII at that time 
limited the maximum external pressure to \1/4\ of the expected 
buckling pressure. In order to increase the allowable external 
pressure, the analysis of external pressure was performed using a 
later version of the ASME code which allows a maximum external 
pressure of \1/3\ of the buckling pressure. The later version of the 
code used for Unit 1 uses a methodology for determining the maximum 
external pressure consistent with the code used for Unit 2.
    Although the margin between the allowable external pressure and 
the expected buckling pressure for Unit 1 will be changed from a 
factor of 4 to a factor of 3, substantial margin is maintained in 
accordance with more current versions of ASME III.
    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 Boulevard, MS LAW/
JB, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Benjamin G. Beasley.

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

    Date of amendment request: December 17, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15351A165.
    Description of amendment request: The proposed change, if approved, 
would amend Combined License (COL) Nos. NPF-93 and NPF-94 for VCSNS. 
The requested amendment proposes to rename, relocate, and add radiation 
detectors to provide monitoring of the radiologically controlled area 
ventilation system (VAS) exhaust from the radiologically controlled 
areas of the auxiliary building and annex building. The changes in the 
proposed amendment are located primarily in the VCSNS Updated Final 
Safety Analysis Report (UFSAR) Tier 2 information, and involve require 
conforming changes to COL Appendix C, ``Inspections, Tests, Analyses, 
and Acceptance Criteria,'' and departing from certified AP1000 Design 
Control Document (DCD) Tier 1 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 
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 design functions of the VAS include prevention of the 
unmonitored release of airborne radioactivity to the atmosphere or 
adjacent plant areas by providing monitoring of the VAS exhaust from 
radiologically controlled areas of the auxiliary building and annex 
building, and to automatically isolate the selected building areas 
and start the containment air filtration system (VFS) upon detection 
of high radioactivity. The proposed changes to the VAS to relocate 
and add radiation detectors are acceptable as they maintain these 
design functions. These proposed changes to the VAS design as 
described in the current licensing basis do not have an adverse 
effect on any of the design functions of the systems. The proposed 
changes 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 described create any new accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes revise the VAS design as described in the 
current licensing basis to enable the system to perform required 
design functions, and are consistent with other UFSAR information. 
The proposed changes do not change the design requirements for the 
system. The relocated and new VAS radiation detectors are designed 
to the same equipment specifications, including required sensitivity 
and range, as the existing radiation detectors. The relocated and 
new VAS radiation detectors monitor the same parameters, as well as 
perform the same design functions, as the existing radiation 
detectors. The proposed changes to the system 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. The proposed changes do not result in a new 
failure mode, malfunction or sequence of events that could affect 
safety or safety-related equipment. The proposed changes do not 
allow for a new fission product release path, result in a new 
fission product barrier failure mode, or create a new sequence of 
events that would result in significant fuel cladding failures.
    Therefore, 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 do not change the codes or standards for 
the radiation detectors, or functionality of the ductwork in the 
auxiliary building and annex building. The proposed changes have no 
adverse effect on the nonsafety-related system design functions of 
the VAS for the prevention of the unmonitored release of airborne 
radioactivity to the atmosphere or adjacent plant areas by providing 
monitoring of the VAS exhaust from radiologically controlled areas 
of the auxiliary building and annex building, and to automatically 
isolate the selected building areas and start the VFS upon detection 
of high radioactivity. The proposed changes do not affect safety-
related equipment or equipment whose failure could initiate an 
accident. The proposed changes to relocate and add radiation 
detectors do not adversely interface with safety-related equipment 
or fission product barriers. Therefore, the proposed changes do not 
affect any safety-related equipment, design code, function, design 
analysis, safety analysis input or result, or design/safety margin. 
No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the requested changes, thus, 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 Avenue NW., Washington, DC 20004-2514.

[[Page 7841]]

    Acting NRC Branch Chief: John McKirgan.

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: December 22, 2015. A publicly-available 
version is in ADAMS. under Accession No.ML15356A656.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-91 and NPF-92 for VEGP, Units 3 and 4, 
respectively. The requested amendment proposes to depart from approved 
AP1000 Design Control Documents (DCD) Tier 2 information (text, tables, 
and figures) and involved Tier 2* information (as incorporated into the 
Updated Final Safety Analysis Report (UFSAR) as plant specific DCD 
information), and also involves a change to a license condition. 
Specifically, the requested amendment proposes changes to the design of 
auxiliary building Wall 11 and proposes other changes to the licensing 
basis for use of Seismic Category II structures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
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.
    Acting NRC Branch Chief: John McKirgan.

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

    Date of amendment request: December 15, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15351A023.
    Description of amendment request: The amendments would modify the 
Technical Specifications (TSs) to risk-inform the requirements 
regarding selected Required Action end states by incorporating TS Task 
Force (TSTF) traveler TSTF-423, Revision 1, ``Technical Specification 
End States, NEDC-32988-A.'' Additionally, it would modify the TS 
Required Actions with a Note prohibiting the use of limiting condition 
for operation 3.0.4.a when entering the preferred end state (Mode 3) on 
startup. The Notice of Availability for TSTF-423, Revision 1, was 
published in the Federal Register on February 18, 2011 (76 FR 9614).
    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 allows a change to certain required end 
states when the TS Completion Times for remaining in power operation 
will be exceeded. Most of the requested technical specification (TS) 
changes are to permit an end state of hot shutdown (Mode 3) rather 
than an end state of cold shutdown (Mode 4) contained in the current 
TS. The request was limited to: (1)

[[Page 7842]]

Those end states where entry into the shutdown mode is for a short 
interval, (2) entry is initiated by inoperability of a single train 
of equipment or a restriction on a plant operational parameter, 
unless otherwise stated in the applicable TS, and (3) the primary 
purpose is to correct the initiating condition and return to power 
operation as soon as is practical. Risk insights from both the 
qualitative and quantitative risk assessments were used in specific 
TS assessments.
    Such assessments are documented in Section 6 of topical report 
NEDC-32988-A, Revision 2, ``Technical Justification to Support Risk 
Informed Modification to Selected Required Action End States for BWR 
Plants.'' They provide an integrated discussion of deterministic and 
probabilistic issues, focusing on specific TSs, which are used to 
support the proposed TS end state and associated restrictions. The 
NRC staff finds that the risk insights support the conclusions of 
the specific TS assessments. Therefore, the probability of an 
accident previously evaluated is not significantly increased, if at 
all. The consequences of an accident after adopting TSTF-423 are no 
different than the consequences of an accident prior to adopting 
TSTF-423. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns.
    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 involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
If risk is assessed and managed, allowing a change to certain 
required end states when the TS Completion Times for remaining in 
power operation are exceeded (i.e., entry into hot shutdown rather 
than cold shutdown to repair equipment) will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change and the commitment by the licensee to adhere to the guidance 
in TSTF-IG-05-02, ``Implementation Guidance for TSTF-423, Revision 
1, `Technical Specifications End States, NEDC-32988-A,' '' will 
further minimize possible concerns.
    Thus, based on the above, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change allows, for some systems, entry into hot 
shutdown rather than cold shutdown to repair equipment, if risk is 
assessed and managed. The Boiling Water Reactor Owners' Group's risk 
assessment approach is comprehensive and follows NRC staff guidance 
as documented in Regulatory Guides (RG) 1.174 and 1.177. In 
addition, the analyses show that the criteria of the three-tiered 
approach for allowing TS changes are met. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in RG 1.177. A risk assessment was performed to justify 
the proposed TS changes. The net change to the margin of safety is 
insignificant.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based upon the reasoning presented above, SNC concludes that the 
requested change involves no significant hazards consideration, as 
set forth in 10 CFR 50.92(c), ``Issuance of Amendment.''

    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. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendment request: November 20, 2015, as supplemented by 
letter dated January 12, 2016. Publicly-available versions are in ADAMS 
under Accession Nos. ML15324A297 and ML16012A457, respectively.
    Description of amendment request: The proposed change would revise 
the setpoint requirements in Technical Specification (TS) 3.3.5, ``Loss 
of Power Diesel Generator Start Instrumentation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed license amendment request changes the TS 3.3.5 
requirements for loss of power diesel generator start 
instrumentation to enable elimination of manual actions for 
protection of safety-related equipment from degraded voltage 
conditions during design basis events. Elimination of these manual 
actions is required to fulfill an existing License Condition on each 
unit.
    The proposed change increases the Allowable Value (AV) for the 
4.16 kV Emergency Bus Degraded Grid Voltage Actuation function. 
Installation of new, higher precision Degraded Voltage Relays (DVRs) 
makes possible an increase in the DVR actuation setpoint 
(encompassed by the AV) to a level which provides fully automatic 
protection of safety-related equipment while minimizing the chance 
of unwanted disconnection from the preferred offsite power source, 
which is itself an analyzed condition.
    Based on the above, 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 license change request changes the TS 3.3.5 
requirements for loss of power diesel generator start 
instrumentation to enable elimination of manual actions for 
protection of safety-related equipment from degraded voltage 
conditions during design basis events. Elimination of these manual 
actions is required to fulfill an existing License Condition on each 
unit.
    The proposed changes to TS 3.3.5 do not change the methods of 
normal plant operation nor the methods of response to transient 
conditions, save that the range of automatic action provided by the 
DVRs is expanded. This change will eliminate the need for manual 
action from the degraded voltage protection scheme, as required by a 
License Condition for each unit, to achieve compliance with 10 CFR 
50.55a(h)(2) and 10 CFR part 50, Appendix A, General Design 
Criterion 17--Electric Power Systems.
    Accordingly, 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.
    Margin of safety is provided by the performance capability of 
plant equipment in preventing or mitigating challenges to fission 
product barriers under postulated operational transient and accident 
conditions. Since the proposed license amendment request changes the 
TS 3.3.5 requirements for loss of power diesel generator start 
instrumentation to enable elimination of manual actions for 
protection of safety-related equipment from degraded voltage 
conditions during design basis events, it will tend to increase the 
margin of safety by better protecting the safety-related plant 
equipment.
    Based on the above, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 7843]]

    Attorney for licensee: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, 40 Iverness Center 
Parkway, Birmingham, AL 35201.
    NRC Branch Chief: Michael T. Markley.

STP Nuclear Operating Company (STPNOC), Docket Nos. 50-498 and 50-499, 
South Texas Project (STP), Units 1 and 2, Matagorda County, Texas

    Date of amendment request: June 19, 2013, as supplemented by 
letters dated October 3, October 31, November 13, November 21, and 
December 23, 2013 (two letters); January 9, February 13, February 27, 
March 17, March 18, May 15, May 22, June 25, and July 15, 2014; and 
March 10, March 25, and August 20, 2015. For the convenience of the 
reader, the ADAMS accession numbers of the amendment request, 
supplements, and additional documents (if publicly available) are 
provided below in a table in the ``Availability of Documents'' section.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TSs) and licensing basis for Facility 
Operating License Nos. NPF-76 and NPF-80, for STP, Units 1 and 2, as 
documented in the Updated Final Safety Analysis Report (UFSAR). The 
changes incorporate use of both a deterministic and a risk-informed 
approach to address safety issues discussed in Generic Safety Issue 
(GSI)-191, ``Assessment of Debris Accumulation on PWR [Pressurized-
Water Reactor] Sump Performance,'' and to close Generic Letter (GL) 
2004-02, ``Potential Impact of Debris Blockage on Emergency 
Recirculation during Design Basis Accidents at Pressurized-Water 
Reactors,'' dated September 13, 2004 (ADAMS Accession No. ML042360586), 
for STP.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes are a methodology change for assessment of 
debris effects that adds the results of a risk-informed evaluation 
to the STP licensing basis, changes to the [emergency core cooling 
system (ECCS)] and [containment spray system (CSS)] TS to extend the 
required completion time for potential [loss-of-coolant accident 
(LOCA)] debris related effects and associated administrative TS 
changes. The methodology change concludes that the ECCS and CSS will 
have sufficient defense-in-depth and safety margin and will operate 
with high probability following a LOCA when considering the impacts 
and effects of debris accumulation on containment emergency sump 
strainers in recirculation mode, as well as core flow blockage due 
to in-vessel effects, following loss of coolant accidents. The 
methodology change also supports the changes to the TS.
    There is no significant increase in the probability of an 
accident previously evaluated. The proposed changes address 
mitigation of loss of coolant accidents and have no effect on the 
probability of the occurrence of a loss of coolant accident. The 
proposed methodology and TS changes do not implement any physical 
changes to the facility or any [structures, systems, and components 
(SSCs)], and do not implement any changes in plant operation that 
could lead to a different kind of accident.
    The proposed changes do not involve a significant increase in 
the consequences of an accident previously evaluated. The 
methodology change confirms that required SSCs supported by the 
containment sumps will perform their safety functions with a high 
probability, as required, and does not alter or prevent the ability 
of SSCs to perform their intended function to mitigate the 
consequences of an accident previously evaluated within the 
acceptance limits. The safety analysis acceptance criteria in the 
UFSAR continue to be met for the proposed methodology change. The 
evaluation of the changes determined that containment integrity will 
be maintained. The dose consequences were considered in the 
assessment and quantitative evaluation of the effects on dose using 
input from the risk-informed approach shows the increase in dose 
consequences is small.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any the accident 
previously evaluated in the UFSAR.
    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 are a methodology change for assessment of 
debris effects from LOCAs that are already evaluated in the STP 
UFSAR, an extension of TS required completion time for potential 
LOCA debris related effects on ECCS and CSS, and associated 
administrative changes to the TS. No new or different kind accident 
is being evaluated. None of the changes install or remove any plant 
equipment, or alter the design, physical configuration, or mode of 
operation of any plant structure, system or component. The proposed 
changes do not introduce any new failure mechanisms or malfunctions 
that can initiate an accident. The proposed changes do not introduce 
failure modes, accident initiators, or equipment malfunctions that 
would cause a new or different kind of accident.
    Therefore, the proposed changes do not create the possibility 
for a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes are a methodology change for assessment of 
debris effects from LOCAs that are already evaluated in the STP 
UFSAR, an extension of TS required completion time for potential 
LOCA debris related effects on ECCS and CSS, and associated 
administrative changes to the TS. The effects from a full spectrum 
of LOCAs, including double-ended guillotine breaks for all piping 
sizes up to and including the largest pipe in the reactor coolant 
system, are analyzed. Appropriate redundancy and consideration of 
loss of offsite power and worst case single failure are retained, 
such that defense-in-depth is maintained.
    Application of the risk-informed methodology showed that the 
increase in risk from the contribution of debris effects is very 
small as defined by [NRC Regulatory Guide (RG) 1.174, ``An Approach 
for Using Probabilistic Risk Assessment in Risk-Informed Decisions 
on Plant-Specific Changes to the Licensing Basis''] and that there 
is adequate defense in depth and safety margin. Consequently, STP 
determined that the risk-informed method demonstrates the 
containment sumps will continue to support the ability of safety 
related components to perform their design functions when the 
effects of debris are considered. The proposed change does not alter 
the manner in which safety limits are determined or acceptance 
criteria associated with a safety limit. The proposed change does 
not implement any changes to plant operation, and does not 
significantly affect SSCs that respond to safely shutdown the plant 
and to maintain the plant in a safe shutdown condition. The proposed 
change does not significantly affect the existing safety margins in 
the barriers for the release of radioactivity. There are no changes 
to any of the safety analyses in the UFSAR.
    Defense in depth and safety margin was extensively evaluated for 
the methodology change and the associated TS changes. The evaluation 
determined that there is substantial defense in depth and safety 
margin that provide a high level of confidence that the calculated 
risk for the methodology and TS changes is conservative and that the 
actual risk is likely much lower.
    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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.

Availability of Documents

    For further details with respect to this action, see the 
application for license amendment dated June 19, 2013, listed below in 
the table, in addition to supplements, requests for additional 
information responses, and other relevant documents.

[[Page 7844]]



------------------------------------------------------------------------
               Title                      Date       ADAMS Accession No.
------------------------------------------------------------------------
SECY-12-0093, ``Closure Options for      07/09/2012  ML121320270
 Generic Safety Issue-191,
 Assessment of Debris Accumulation
 on Pressurized-Water Reactor Sump
 Performance.''
STP Pilot Submittal and Request for      01/31/2013  ML13043A013
 Exemption for a Risk-Informed
 Approach to Resolve Generic Safety
 Issue (GSI)-191.
NRC Letter to STPNOC, ``South Texas      04/01/2013  ML13066A519
 Project, Units 1 and 2--
 Supplemental Information Needed
 for Acceptance of Requested
 Licensing Action Re: Request for
 Exemption for a Risk-Informed
 Approach to Resolve Generic Safety
 Issue 191''.
Revised STP Pilot Submittal and          06/19/2013  ML131750250
 Requests for Exemptions and                         (package)
 License Amendment for a Risk-
 Informed approach to Resolving
 Generic Safety Issue (GSI)-191.
NRC Letter to STPNOC, ``South Texas      08/13/2013  ML13214A031
 Project, Units 1 and 2--Acceptance
 of Requests for Exemptions and
 License Amendment Request for
 Approval of a Risk-Informed
 Approach to Resolve Generic Safety
 Issue GSI-191''.
Corrections to Information Provided      10/03/2013  ML13295A222
 in Revised STP Pilot Submittal and
 Requests for Exemptions and
 License Amendment for a Risk-
 Informed Approach to Resolving
 Generic Safety Issue (GSI)-191.
Submittal of GSI-191 Chemical            10/31/2013  ML13323A673
 Effects Test Reports.                               (package)
Supplement 1 to Revised STP Pilot        11/13/2013  ML13323A128
 Submittal and Requests for                          (package)
 Exemptions and License Amendment
 for a Risk-Informed Approach to
 Resolving Generic Safety Issue
 (GSI)-191.
Supplement 1 to Revised STP Pilot        11/21/2013  ML13338A165
 Submittal for a Risk-Informed
 Approach to Resolving Generic
 Safety Issue (GSI)-191 to
 Supersede and Replace the Revised
 Pilot Submittal.
Response to STP-GSI-191-EMCB-RAI-1.      12/23/2013  ML14015A312
Response to NRC Request for              12/23/2013  ML14015A311
 Reference Document for STP Risk-
 Informed GSI-191 Application.
Response to Request for Additional       01/09/2014  ML14029A533
 Information re Use of RELAP5 in
 Analyses for Risk-Informed GSI-191
 Licensing Application.
Submittal of CASA Grande Code and        02/13/2014  ML14052A110
 Analyses for STP's Risk-Informed                    (package, portions
 GSI-191 Licensing Application.                       redacted)
Submittal of GSI-191 Chemical            02/27/2014  ML14072A075
 Effects Test Reports.                               (package)
Response to NRC Accident Dose            03/17/2014  ML14086A383
 Branch Request for Additional                       (package)
 Information Regarding STP Risk-
 Informed GSI-191 Application.
Submittal of CASA Grande Source          03/18/2014  (proprietary,
 Code for STP's Risk-Informed GSI-                   non-public)
 191 Licensing Application.
Second Submittal of CASA Grande          05/15/2014  ML14149A354
 Source Code for STP's Risk-
 Informed GSI-191 Licensing
 Application.
First Set of Responses to April,         05/22/2014  ML14149A439
 2014, Requests for Additional                       (package)
 Information Regarding STP Risk-
 Informed GSl-191 Licensing
 Application--Revised.
Second Set of Responses to April,        06/25/2014  ML14178A467
 2014, Requests for Additional                        (package)
 Information Regarding STP Risk-
 Informed GSI-191 Licensing
 Application.
Third Set of Responses to April,         07/15/2014  ML14202A045
 2014, Requests for Additional
 Information Regarding STP Risk-
 Informed GSI-191 Licensing
 Application.
Submittal of Updated CASA Grande         03/10/2015  ML15072A092
 Input for STP's Risk-Informed GSI-
 191 Licensing Application.
Description of Revised Risk-             03/25/2015  ML15091A440
 Informed Methodology and Responses
 to Round 2 Requests for Additional
 Information Regarding STP Risk-
 Informed GSI-191 Licensing
 Application.
Supplement 2 to STP Pilot Submittal      08/20/2015  ML15246A125
 and Requests for Exemptions and
 License Amendment for a Risk-
 Informed Approach to Address
 Generic Safety Issue (GSI)-191 and
 Respond to Generic Letter (GL)
 2004[dash]02.
------------------------------------------------------------------------

    Attorney for licensee: Steve Frantz, Esq., Morgan, Lewis & Bockius, 
1111 Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: December 15, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15362A023.
    Description of amendment request: The amendment would revise 
Technical Specifications (TSs) 3.4.17, ``Steam Generator (SG) Tube 
Integrity''; 5.7.2.12, ``Steam Generator (SG) Program''; and 5.9.9, 
``Steam Generator Tube Inspection Report,'' to exclude portions of the 
SG tubes below the top of the tube sheet from needing to be plugged.
    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.
    Allowing the use of an alternate repair criteria as proposed in 
this amendment request does not involve a significant increase in 
the probability or consequence of an accident previously evaluated.
    The presence of the tubesheet enhances the tube integrity in the 
region of the hardroll by precluding tube deformation beyond its 
initial expanded outside diameter. The resistance to both tube 
rupture and tube collapse is strengthened by the presence of the 
tubesheet in that region. Hardrolling of the tube into the tubesheet 
results in an interference fit between the tube and the tubesheet. 
Tube rupture cannot occur because the contact between the tube and 
tubesheet does not permit sufficient movement of tube material. In a 
similar manner, the tubesheet does not permit sufficient movement of 
tube material to permit buckling collapse of the tube during 
postulated loss-of-coolant-accident (LOCA) loadings.
    The type of degradation for which the F* [the length of 
mechanical expansion required to prevent pullout for all normal 
operating and postulated accident conditions] has been developed 
(cracking with a circumferential orientation) can theoretically lead 
to a postulated tube rupture event, provided that the postulated 
through-wall circumferential crack exists near the top of the 
tubesheet. An evaluation including analysis and testing has

[[Page 7845]]

been performed to determine the resistive strength of roll expanded 
tubes within the tubesheet. That evaluation provides the basis for 
the acceptance criteria for tube degradation subject to the F* 
criterion.
    The F* length of roll expansion is sufficient to preclude tube 
pullout from tube degradation located below the F* distance, 
regardless of the extent of the tube degradation. The existing 
technical specification leakage rate requirements and accident 
analysis assumptions remain unchanged in the unlikely event that 
significant leakage from this region does occur. As noted above, 
tube rupture and pullout are not expected for tubes using the ARC 
[alternative repair criterion]. Any leakage out of the tube from 
within the tubesheet at any elevation in the tubesheet is fully 
bounded by the existing Main Steam Line Break (MSLB) analysis 
included in the WBN Unit 2 Final Safety Analysis Report (FSAR).
    Therefore, the proposed ARC does not adversely impact any other 
previously evaluated design basis accident.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Implementation of the proposed ARC does not introduce any 
significant changes to the plant design basis. Use of the criterion 
does not provide a mechanism to result in an accident initiated 
outside of the region of the tubesheet expansion. A hypothetical 
accident as a result of any tube degradation in the expanded portion 
of the tube would be bounded by the existing tube rupture accident 
analysis. Tube bundle structural integrity and leak tightness are 
expected to be maintained.
    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 use of the ARC has been demonstrated to maintain the 
integrity of the tube bundle commensurate with the requirements of 
Regulatory Guide 1.121, ``Bases for Plugging Degraded PWR 
[Pressurized-Water Reactor] Steam Generator Tubes,'' for indications 
in the free span of tubes and the primary to secondary pressure 
boundary under normal and postulated accident conditions. Acceptable 
tube degradation for the F* criterion is any degradation indication 
in the tubesheet region, more than the F* distance below either the 
bottom of the transition between the roll expansion and the 
unexpanded tube, or the top of the tubesheet, whichever is lower. 
The safety factors used in the verification of the strength of the 
degraded tube are consistent with the safety factors in the American 
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel 
Code used in SG design. The F* distance has been verified by testing 
to be greater than the length of roll expansion required to preclude 
both tube pullout and significant leakage during normal and 
postulated accident conditions. Resistance to tube pullout is based 
upon the primary to secondary pressure differential as it acts on 
the surface area of the tube, which includes the tube wall cross-
section, in addition to the inside diameter-based area of the tube. 
The leak testing acceptance criteria are based on the primary to 
secondary leakage limit in the technical specifications and the 
leakage assumptions used in the UFSAR [Updated FSAR] accident 
analyses. Implementation of the ARC will decrease the number of 
tubes which must be taken out of service with tube plugs. Plugs 
reduce the RCS flow margin; thus, implementation of the ARC will 
maintain the margin of flow that would otherwise be reduced in the 
event of increased plugging.
    Based on the above, it is concluded that the proposed change 
does not result in a significant reduction in or a loss of margin 
with respect to plant safety as defined in the FSAR or the bases of 
the WBN Unit 2 technical specifications.

    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: Ralph E. Rodgers, General Counsel, Tennessee 
Valley Authority, 400 West Summit Hill Dr., 6A West Tower, Knoxville, 
TN 37902.
    NRC Branch Chief: Benjamin G. Beasley.

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 (AEA), 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.

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, 
Millstone Power Station, Unit No. 2 (MPS2) and Unit No. 3 (MPS3), New 
London County, Connecticut

    Date of amendment request: January 15, 2015, as supplemented by 
letters dated April 15, July 16, July 30, November 2, and December 1, 
2015.
    Brief description of amendment: The amendments revised the MPS2 and 
MPS3 Technical Specifications (TSs) to adopt NRC-approved Technical 
Specifications Task Force (TSTF) Standard Technical Specifications 
(STS) Change Traveler TSTF-523, Revision 2, ``Generic Letter 2008-01, 
Managing Gas Accumulation.''
    Date of issuance: January 29, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 325 and 267. A publicly-available version is in 
ADAMS under Accession No. ML16011A400; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-65 and NPF-49: 
Amendments revised the Renewed Operating License and TSs.
    Date of initial notice in Federal Register: July 21, 2015 (80 FR 
43126). The supplemental letter dated April 15, 2015, was published 
with the January 15, 2015, application, in the initial FR notice. The 
supplemental letters dated July 16, July 30, November 2, and December 
1, 2015, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a

[[Page 7846]]

Safety Evaluation dated January 29, 2016.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: February 2, 2015, as supplemented by 
letters dated August 11, 2015, and October 20, 2015.
    Brief description of amendments: The amendments modified the 
technical specifications (TSs) to allow for brief, inadvertent, 
simultaneous opening of redundant secondary containment personnel 
access doors during normal entry and exit conditions.
    Date of issuance: January 28, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 220 and 182. A publicly-available version is in 
ADAMS under Accession No. ML15356A140; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. NPF-39 and NPF-85: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: April 14, 2015 (80 FR 
20022). The supplemental letters dated August 11, 2015, and October 20, 
2015, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 28, 2016.
    No significant hazards consideration comments received: Yes.

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

    Date of amendment request: February 23, 2015, as supplemented by 
letters dated August 12, 2015, and October 20, 2015.
    Brief description of amendments: The amendments modified the 
technical specifications (TSs) to allow for brief, inadvertent, 
simultaneous opening of redundant secondary containment personnel 
access doors during normal entry and exit conditions.
    Date of issuance: February 1, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendments Nos.: 303 and 307. A publicly-available version is in 
ADAMS under Accession No. ML15350A179; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the Renewed Facility Operating Licenses and the TSs.
    Date of initial notice in Federal Register: April 14, 2015 (80 FR 
20023). The supplemental letters dated August 12, 2015, and October 20, 
2015, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 1, 2016.
    No significant hazards consideration comments received: Yes.

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

    Date of amendment request: January 15, 2015, as supplemented by 
letters dated May 4, 2015, June 9, 2015, and January 12, 2016.
    Brief description of amendment: The amendment revised the technical 
specifications (TSs) to add a limiting condition for operation, 
applicability, required actions, completion times, and surveillance 
requirements for the residual heat removal containment spray and 
associated interlock permissive instrumentation. A new TS Section 
3.6.1.9, ``Residual Heat Removal (RHR) Containment Spray,'' has been 
added to reflect the reliance on containment spray to maintain the 
drywell within design temperature limits during a small steam line 
break. In addition, the ``Drywell Pressure--High'' function that serves 
as an interlock permissive to allow RHR containment spray mode 
alignment has been relocated from the Technical Requirements Manual to 
TS 3.3.5.1, ``Emergency Core Cooling System (ECCS) Instrumentation.''
    Date of issuance: January 22, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 253. A publicly-available version is in ADAMS under 
Accession No. ML15343A301; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-46: The amendment 
revised the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 17, 2015 (80 FR 
13910). The supplemental letters dated May 4, 2015, June 9, 2015, and 
January 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 NRC staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 22, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 26, 2013, as supplemented by 
letter dated June 3, 2015.
    Brief description of amendment: The amendments are to Combined 
License Nos. NPF-93 and NPF-94 for VCSNS, Units 2 and 3. The amendments 
authorized changes to the VCSNS, Units 2 and 3, Updated Final Safety 
Analysis Report to revise the details of the effective thermal 
conductivity resulting from the oxidation of the inorganic zinc 
component of the containment vessel coating system.
    Date of issuance: October 9, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 34. A publicly-available version is in ADAMS under 
Accession No.
    ML15272A417; documents related to these amendments are listed in 
the Safety Evaluation enclosed with the amendments.
    Facility Combined Licenses No. NPF-93 and NPF-94: Amendments 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: February 19, 2014 (79 
FR 9490). The supplemental letter dated June 3, 2015, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.

[[Page 7847]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 9, 2015.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch Nuclear 
Plant (HNP), Unit No. 1, Appling County, Georgia

    Date of application for amendment: September 1, 2015.
    Brief description of amendments: The amendment revised the 
Technical Specification value of the Safety Limit Minimum Critical 
Power Ratio to support operation in the next fuel cycle.
    Date of issuance: January 29, 2016.
    Effective date: As of the date of issuance and shall be implemented 
prior to reactor startup following the HNP, Unit 1, spring 2016, 
refueling outage.
    Amendment No.: 275. A publicly-available version is in ADAMS under 
Accession No. ML15342A398; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-57: Amendment revised 
the license and the Technical Specifications.
    Date of initial notice in Federal Register: November 3, 2015 (80 FR 
67802).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 29, 2016.
    No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: April 29, 2015.
    Brief description of amendment: The amendment revised the Cyber 
Security Plan Implementation Milestone 8 completion date and the 
physical protection license condition.
    Date of issuance: January 28, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 214. A publicly-available version is in ADAMS under 
Accession No. ML15328A059; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-30: The amendment 
revised the Operating License.
    Date of initial notice in Federal Register: July 7, 2015 (80 FR 
38778).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 28, 2016.
    No significant hazards consideration comments received: No.


    Dated at Rockville, Maryland, this 8th day of February 2016.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
 Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2016-02916 Filed 2-12-16; 8:45 am]
 BILLING CODE 7590-01-P