[Federal Register Volume 81, Number 50 (Tuesday, March 15, 2016)]
[Notices]
[Pages 13837-13849]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-05470]



[[Page 13837]]

=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[NRC-2016-0050]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

-----------------------------------------------------------------------

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 February 13, 2016, to February 29, 2016. 
The last biweekly notice was published on March 1, 2016.

DATES: Comments must be filed by April 14, 2016. A request for a 
hearing must be filed by May 16, 2016.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0050. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected].
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Lynn Ronewicz, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-1927, email: [email protected].

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2016-0050 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0050.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section of this document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2016-0050, facility name, unit 
number(s), application date, and subject in your comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at http://www.regulations.gov, as well as entering the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Sec.  50.92 of title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, 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 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

[[Page 13838]]

subject facility operating license or combined license. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested person(s) should consult a 
current copy of 10 CFR 2.309, which is available at the NRC's PDR, 
located at One White Flint North, Room O1-F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852. The NRC's regulations are 
accessible electronically from the NRC Library on the NRC's Web site at 
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a 
hearing or petition for leave to intervene is filed within 60 days, the 
Commission or a presiding officer designated by the Commission or by 
the Chief Administrative Judge of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to 
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 May 
16, 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 
May 16, 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

[[Page 13839]]

[email protected], or by telephone at 301-415-1677, to request (1) 
a digital identification (ID) certificate, which allows the participant 
(or its counsel or representative) to digitally sign documents and 
access the E-Submittal server for any proceeding in which it is 
participating; and (2) advise the Secretary that the participant will 
be submitting a request or petition for hearing (even in instances in 
which the participant, or its counsel or representative, already holds 
an NRC-issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at http://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at http://www.nrc.gov/site-help/e-submittals.html, by email to 
[email protected], or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, in some instances, a request to intervene will require 
including information on local residence in order to demonstrate a 
proximity assertion of interest in the proceeding. With respect to 
copyrighted works, except for limited excerpts that serve the purpose 
of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station (CNS), Units 1 and 2, York County, South Carolina
    Date of amendment request: January 18, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16026A048.
    Description of amendment request: The proposed amendments would 
modify the Renewed Facility Operating Licenses and Technical 
Specifications (TS) for CNS, Units 1 and 2. Specifically, the proposed 
amendments request to revise TS 5.5.2, ``Containment Leakage Rate 
Testing Program,'' to allow an increase in the existing Type A 
Integrated Leakage Rate Test (ILRT) program test interval from 10 years 
to 15

[[Page 13840]]

years in accordance with Nuclear Energy Institute (NEI) Topical Report 
NE1 94-01, Revision 3-A, ``Industry Guideline for Implementing 
Performance-Based Option of 10 CFR part 50, appendix J,'' and the 
conditions and limitations specified in NEI 94-01, Revision 2-A; 
adoption of an extension of the containment isolation valve leakage 
testing (Type C) frequency from the 60 months currently permitted by 10 
CFR part 50, appendix J, Option B, to a 75-month frequency for Type C 
leakage rate testing of selected components, in accordance with NEI 94-
01, Revision 3-A; adoption of the use of ANSI/ANS 56.8-2002, 
``Containment System Leakage Testing Requirements''; and adoption of a 
more conservative grace interval of 9 months for Type A, Type B, and 
Type C leakage tests in accordance with NEI 94-01, Revision 3-A. The 
proposed amendments also request the following administrative changes: 
Deletion of the information regarding the performance of containment 
visual inspections as required by Regulatory Position C.3, as the 
containment inspections are addressed in TS Surveillance Requirement 
3.6.1.1, deletion of the information regarding the performance of the 
next CNS, Unit 1, Type A test no later than November 13, 2015, and the 
next CNS, Unit 2, Type A test no later than February 6, 2008, as both 
Type A 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 with NRC edits in square 
brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment to the Technical Specifications (TS) 
involves the extension of the Catawba Nuclear Station (CNS) Type A 
containment integrated leak rate test interval to 15 years and the 
extension of the Type C test interval to 75 months for selected 
components. 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. The containment and the 
testing requirements invoked to periodically demonstrate the 
integrity of the containment exist to ensure the plant's ability to 
mitigate the consequences of an accident, and do not involve the 
prevention or identification of any precursors of an accident. The 
change in dose risk for changing the Type A test frequency from 
three-per-ten years to once-per-fifteen years, measured, as an 
increase to the total integrated plant risk for those accident 
sequences influenced by Type A testing, is 0.026 person-rem/year. 
EPRI Report No. 1009325, Revision 2-A states that a very small 
population dose is defined as an increase of [less than or equal to] 
1.0 person-rem per year, or [less than or equal to] 1% of the total 
population dose, whichever is less restrictive for the risk impact 
assessment of the extended ILRT intervals. Therefore, this proposed 
extension does not involve a significant increase in the probability 
of an accident previously evaluated.
    As documented in NUREG-1493, Type B and C tests have identified 
a very large percentage of containment leakage paths, and the 
percentage of containment leakage paths that are detected only by 
Type A testing is very small. The CNS Type A test history supports 
this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as: (1) Activity based, 
and; (2) time based. Activity based failure mechanisms are defined 
as degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment combined with the containment inspections performed in 
accordance with ASME Section Xl, the Maintenance Rule, and TS 
requirements serve to provide a high degree of assurance that the 
containment would not degrade in a manner that is detectable only by 
a Type A test. Based on the above, the proposed extensions do not 
significantly increase the consequences of an accident previously 
evaluated.
    The proposed amendment also deletes an exception previously 
granted to allow one-time extensions of the Unit 1 and Unit 2 ILRT 
test frequency for CNS. This exception was for activities that have 
already taken place; therefore, their deletion is solely an 
administrative action that has no effect on any component and no 
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 
CNS Type A containment integrated leak rate test interval to 15 
years and the extension of the Type C test interval to 75 months for 
selected components.
    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. The containment and the testing requirements 
to periodically demonstrate the integrity of the containment exist 
to ensure the plant's ability to mitigate the consequences of an 
accident do not involve any accident precursors or initiators. The 
proposed change does not involve a physical change to the plant 
(i.e., no new or different type of equipment will be installed) or a 
change to the manner in which the plant is operated or controlled.
    The proposed amendment also deletes an exception previously 
granted to allow one-time extensions of the Unit 1 and Unit 2 ILRT 
test frequency for CNS. This exception was for activities that have 
already taken; therefore, their deletion is solely an administrative 
action that does not result in any change in how the units are 
operated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed amendment to TS 5.5.2 involves the extension of the 
CNS Type A containment integrated leak rate test interval to 15 
years and the extension of the Type C test interval to 75 months for 
selected components. 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. 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 
CNS. The proposed surveillance interval extension is bounded by the 
15-year ILRT interval, and the 75-month Type C test interval 
currently authorized within NEI 94-01, Revision 3-A. Industry 
experience supports the conclusion that Type B and C testing detects 
a large percentage of containment leakage paths and that the 
percentage of containment leakage paths that are detected only by 
Type A testing is small. The containment inspections performed in 
accordance with ASME Section Xl, TS and the Maintenance Rule serve 
to provide a high

[[Page 13841]]

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 an exception previously 
granted to allow one-time extensions of the Unit 1 and Unit 2 ILRT 
test frequency for CNS. This exception was for activities that have 
already 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: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street-EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Michael T. Markley.
Duke Energy Progress, Inc., Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina
    Date of amendment request: November 19, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15323A085.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to allow the extension of the 
Type A containment test interval to 15 years and the extension of the 
Type B and Type C test intervals for selected components to 120 months 
and 75 months, respectively. The proposed amendment also deletes from 
the TSs an already implemented one-time extension of the Type A test 
frequency.
    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 H. B. Robinson Steam Electric Plant 
Unit No. 2 (HBRSEP2) Type A containment test interval to 15 years, 
the extension of the Type B test intervals to 120 months for 
selected components, and the extension of the Type C test interval 
to 75 months for selected components. 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 B test interval of each reactor shutdown for refueling 
but in no case at intervals greater than 2 years would be extended 
on a performance basis to no longer than 120 months. The current 
Type C test interval of each reactor shutdown for refueling but in 
no case at intervals greater than 2 years 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 extensions do 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. The containment and the 
testing requirements invoked to periodically demonstrate the 
integrity of the containment exist to ensure the plant's ability to 
mitigate the consequences of an accident, and do not involve the 
prevention or identification of any precursors of an accident. The 
change in dose risk for changing the Type A test frequency from 
three-per-ten years to once-per-fifteen years, measured, as an 
increase to the total integrated plant risk for those accident 
sequences influenced by Type A testing, is 0.020 person-rem 
[roentgen equivalent man]/year. The Electric Power Research 
Institute (EPRI) Report No. 1009325, Revision 2-A, states that a 
very small population dose is defined as an increase of <=1.0 
person-rem per year, or <=1% of the total population dose, whichever 
is less restrictive for the risk impact assessment of the extended 
integrated leak rate test (ILRT) intervals. Therefore, this proposed 
extension does not involve a significant increase in the probability 
of an accident previously evaluated.
    As documented in NUREG-1493, Type B and C tests have identified 
a very large percentage of containment leakage paths, and the 
percentage of containment leakage paths that are detected only by 
Type A testing is very small. The HBRSEP2 Type A test history 
supports this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as: (1) Activity based, 
and (2) time based. Activity based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment combined with the containment inspections performed in 
accordance with the American Society of Mechanical Engineers (ASME) 
Section XI, the Maintenance Rule, and TS requirements serve to 
provide a high degree of assurance that the containment would not 
degrade in a manner that is detectable only by a Type A test. Based 
on the above, the proposed extensions do not significantly increase 
the consequences of an accident previously evaluated.
    The proposed amendment also deletes an exception previously 
granted to allow one-time extension of the ILRT test frequency for 
HBRSEP2. This exception was for an activity that has already taken 
place so the deletion is solely an administrative action that has no 
effect on any component and no impact on how the unit is operated.
    Therefore, the proposed change does not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment to the TS involves the extension of the 
HBRSEP2 Type A containment test interval to 15 years, the Type B 
test interval to 120 months for selected components and the 
extension of the Type C test interval to 75 months for selected 
components. The containment and the testing requirements to 
periodically demonstrate the integrity of the containment exist to 
ensure the plant's ability to mitigate the consequences of an 
accident do not involve any accident precursors or initiators. The 
proposed change does not involve a physical change to the plant 
(i.e., no new or different type of equipment will be installed) or a 
change to the manner in which the plant is operated or controlled.
    The proposed amendment also deletes an exception previously 
granted to allow one-time extension of the ILRT test frequency for 
HBRSEP2. This exception was for an activity that has already taken 
place so the deletion is solely an administrative action that has no 
effect on any component and no impact on how the unit is operated.
    Therefore, the proposed change does not 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 5.5.16 involves the extension of 
the HBRSEP2 Type A containment test interval to 15 years, the Type B 
test interval to 120 months for selected components 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

[[Page 13842]]

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, Type B tests and Type C 
tests for HBRSEP2. The proposed surveillance interval extension is 
bounded by the 15-year ILRT interval, the 120-month Type B interval 
and the 75-month Type C test interval currently authorized within 
NEI 94-01, Revision 3-A. Industry experience supports the conclusion 
that Types 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, TS and the 
Maintenance Rule serve to provide a high degree of assurance that 
the containment would not degrade in a manner that is detectable 
only by Type A testing. The combination of these factors ensures 
that the margin of safety in the plant safety analysis is 
maintained. The design, operation, testing methods and acceptance 
criteria for Types 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, Type B and Type C test intervals.
    The proposed amendment also deletes an exception previously 
granted to allow one-time extension of the ILRT test frequency for 
HBRSEP2. This exception was for an activity that has already taken 
place so the deletion is solely an administrative action that has no 
effect on any component and no impact on how the unit is operated.
    Therefore, the proposed change does not 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: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, 
Charlotte, NC 28202.
    NRC Branch Chief: Benjamin G. Beasley.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station (LGS), Units 1 and 2, Montgomery County, 
Pennsylvania
    Date of amendment request: January 15, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16015A316.
    Description of amendment request: The amendments would reduce the 
reactor vessel steam dome pressure associated with the Technical 
Specification (TS) Safety Limits (SLs) specified in TS 2.1.1 and TS 
2.1.2. The amendments would also revise the setpoint and allowable 
value for the main steam line low pressure isolation function in TS 
Table 3.3.2-2. The proposed changes address a 10 CFR part 21 issue 
concerning the potential to violate the SLs limits during a pressure 
regulator failure maximum demand (open) transient.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because decreasing the reactor vessel steam dome pressure in TS 
Safety Limits 2.1.1 and 2.1.2 for reactor thermal power ranges and 
increasing the trip setpoint and allowable value for the main steam 
line low pressure isolation effectively expands the validity range 
for GEXL critical power correlation and the calculation of the 
minimum critical power ratio. The critical power ratio rises during 
the pressure reduction following the scram that terminates the 
Pressure Regulator Failure Maximum Demand (Open) (PRFO) transient. 
The reduction in the reactor vessel steam dome pressure value in the 
SL and the increase in the trip setpoint and the allowable value for 
the main steam line low pressure isolation provides adequate margin 
to accommodate the pressure reduction during the PRFO transient 
within the revised TS limit.
    The proposed changes do not alter the use of the analytical 
methods used to determine the safety limits that have been 
previously reviewed and approved by the NRC. The proposed changes 
are in accordance with an NRC approved critical power correlation 
methodology and do not adversely affect accident initiators or 
precursors.
    The proposed changes do not alter or prevent the ability of 
structures, systems, and components from performing their intended 
function to mitigate the consequences of an initiating event within 
the applicable acceptance limits. The proposed changes are 
consistent with the safety analysis and resultant consequences.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed reduction in the reactor vessel steam dome 
pressure value in the safety limit in conjunction with the increase 
in the trip setpoint and the allowable value for the main steam line 
low pressure isolation reflects a wider range of applicability for 
the GEXL critical power correlation which is approved by the NRC for 
both GE14 and GNF2 fuel types in [the] LGS reactor cores.
    In addition, no new failure modes are being introduced. There 
are no changes in the method by which any plant systems perform a 
safety function. No new accident scenarios, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
changes.
    The proposed changes do not introduce any new accident 
precursors, nor do they involve any changes in the methods governing 
normal plant operation. The proposed changes do not alter the 
outcome of the safety analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through the design of the 
plant structures, systems, and components, and through the 
parameters for safe operation and setpoints for the actuation of 
equipment relied upon to respond to transients and design basis 
accidents. Evaluation of the 10 CFR part 21 condition by General 
Electric determined that, since the critical power ratio improves 
during the PRFO transient, there is no impact on the fuel safety 
margin, and therefore, there is no challenge to fuel cladding 
integrity. The proposed changes do not change the requirements 
governing operation or availability of safety equipment assumed to 
operate to preserve the margin of safety.
    The proposed changes are consistent with the applicable NRC 
approved critical power correlation for the fuel designs in use at 
LGS. The proposed changes do not alter the manner in which the 
safety limits are determined.
    The reduction in value of the reactor vessel steam dome pressure 
safety limit and the increase in the trip setpoint and allowable 
value for the main steam line low pressure isolation provides 
adequate margin to accommodate the pressure reduction during the 
PRFO transient within the revised TS limit.
    Therefore, the proposed changes do not involve a significant 
reduction in any 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 13843]]

    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: Douglas A. Broaddus.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station (DBNPS), Unit No. 1, Ottawa County, 
Ohio
    Date of amendment request: February 17, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16049A513.
    Description of amendment request: The licensee proposes to change 
the emergency plan for DBNPS, Unit No. 1, by revising the emergency 
action level (EAL) scheme based on the Nuclear Energy institute's 
(NEl's) guidance in NEI 99-01, Revision 6, ``Development of Emergency 
Action Levels for Non-Passive Reactors.'' The NEI 99-01, Revision 6, 
was endorsed by the NRC by letter dated March 28, 2013 (ADAMS Accession 
No. ML12346A463).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to DBNPS's EAL scheme to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not involve any 
physical changes to plant systems or equipment. The proposed changes 
do not alter any of the requirements of the technical 
specifications. The proposed changes do not modify any plant 
equipment and do not impact any failure modes that could lead to an 
accident. Additionally, the proposed changes do not impact the 
ability of structures, systems, or components (SSCs) to perform 
their intended safety functions in mitigating the consequences of an 
initiating event within the assumed acceptance limits.
    Therefore, the proposed 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 changes to DBNPS's EAL scheme to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not involve any 
physical changes to plant systems or equipment. The proposed changes 
do not involve the addition of any new plant equipment. The proposed 
changes will not alter the design configuration, or method of 
operation of plant equipment beyond its normal functional 
capabilities. DBNPS functions will continue to be performed as 
required. The proposed changes do not create any new credible 
failure mechanisms, malfunctions, or accident initiators.
    Therefore, the proposed 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 changes to DBNPS's EAL scheme to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not involve any 
physical changes to plant systems or equipment. Margins of safety 
are unaffected by the proposed changes. There are no changes being 
made to safety analysis assumptions, safety limits, or limiting 
safety system settings that would adversely affect plant safety as a 
result of the proposed EAL scheme change. The proposed change does 
not affect the technical specifications. There are no changes to 
environmental conditions of any of the SSC or the manner in which 
any SSC is operated. The applicable requirements of 10 CFR 50.47 and 
10 CFR part 50, appendix E will continue to be met.
    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: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    Acting NRC Branch Chief: Justin C. Poole.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
    Date of amendment request: January 29, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16034A032.
    Description of amendment request: The proposed amendment would 
modify technical specification (TS) requirements to address Generic 
Letter 2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, 
Decay Heat Removal, and Containment Spray Systems,'' as described in 
the Technical Specification Task Force (TSTF) Traveler TSTF-523, 
Revision 2, ``Generic Letter 2008-01, Managing Gas Accumulation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises or adds SRs [Surveillance 
Requirements] that require verification that the ECCS [Emergency 
Core Cooling System], RHR [Residual Heat Removal] System, and the 
Containment Spray (CTS) System are not rendered inoperable due to 
accumulated gas and to provide allowances which permit performance 
of the revised verification. Gas accumulation in the subject systems 
is not an initiator of any accident previously evaluated. As a 
result, the probability of any accident previously evaluated is not 
significantly increased. The proposed SRs ensure that the subject 
systems continue to be capable to perform their assumed safety 
function and are not rendered inoperable due to gas accumulation. 
Thus, 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.
    The proposed change revises or adds SRs that require 
verification that the ECCS, the RHR System, and the CTS System are 
not rendered inoperable due to accumulated gas and to provide 
allowances which permit performance of the revised verification. 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 that could initiate an accident. The proposed change 
does not alter assumptions made in the safety analysis and is 
consistent with the safety analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, the RHR System, and the CTS System are 
not rendered inoperable due to accumulated gas and to provide 
allowances which permit performance of the revised verification. The 
proposed change adds new requirements to manage gas accumulation in 
order to ensure the subject systems are capable of performing their 
assumed safety functions. The proposed SRs are more comprehensive 
than the current SRs and will ensure that the assumptions of the 
safety analysis are protected. The proposed change does not 
adversely affect

[[Page 13844]]

any current plant safety margins or the reliability of the equipment 
assumed in the safety analysis. Therefore, there are no changes 
being made to any safety analysis assumptions, safety limits or 
limiting safety system settings that would adversely affect plant 
safety as a result of the proposed change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: David J. Wrona.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California
    Date of amendment request: January 21, 2106. A publicly-available 
version is in ADAMS under Accession No. ML16021A067.
    Description of amendment request: The amendments would revise or 
add Surveillance Requirements to verify that the system locations 
susceptible to gas accumulation are sufficiently filled with water and 
to provide allowances, which permit performance of the verification. 
The amendments would revise Technical Specification (TS) 3.4.6, ``RCS 
[Reactor Coolant System] Loops--MODE 4''; TS 3.4.7, ``RCS Loops--MODE 
5, Loops Filled''; TS 3.4.8, ``RCS Loops--MODE 5, Loops Not Filled''; 
TS 3.5.2, ``ECCS [Emergency Core Cooling System]--Operating''; TS 
3.6.6, ``Containment Spray and Cooling Systems''; TS 3.9.5, ``RHR 
[Residual Heat Removal] and Coolant Circulation--High Water Level''; 
and TS 3.9.6, ``RHR and Containment Circulation--Low Water Level.'' The 
proposed amendments would modify TS requirements to address Generic 
Letter 2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, 
Decay Heat Removal, and Containment Spray Systems,'' as described in 
Technical Specification Task Force TSTF-523, Revision 2, ``Generic 
Letter 2008-01, Managing Gas Accumulation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises or adds Surveillance Requirement(s) 
(SRs) that require verification that the Emergency Core Cooling 
System (ECCS), the Residual Heat Removal (RHR) System, and the 
Containment Spray (CS) System are not rendered inoperable due to 
accumulated gas and to provide allowances which permit performance 
of the revised verification. Gas accumulation in the subject systems 
is not an initiator of any accident previously evaluated. As a 
result, the probability of any accident previously evaluated is not 
significantly increased. The proposed SRs ensure that the subject 
systems continue to be capable to perform their assumed safety 
function and are not rendered inoperable due to gas accumulation. 
Thus, 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 accident from any accident previously evaluated?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, RHR System, and CS System are not 
rendered inoperable due to accumulated gas and to provide allowances 
which permit performance of the revised verification. 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 
that could initiate an accident. The proposed change does not alter 
assumptions made in the safety analysis and is consistent with the 
safety analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, the RHR System, and the CS System are 
not rendered inoperable due to accumulated gas, and to provide 
allowances which permit performance of the revised verification. The 
proposed change adds new requirements to manage gas accumulation in 
order to ensure the subject systems are capable of performing their 
assumed safety functions. The proposed SRs are more comprehensive 
than the current SRs, and will ensure that the assumptions of the 
safety analysis are protected. The proposed change does not 
adversely affect any current plant safety margins or the reliability 
of the equipment assumed in the safety analysis. Therefore, there 
are no changes being made to any safety analysis assumptions, safety 
limits, or limiting safety system settings that would adversely 
affect plant safety as a result of the proposed change.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, CA 94120.
    NRC Branch Chief: Robert J. Pascarelli.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, 
Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield 
County, South Carolina
    Date of amendment request: January 19, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16019A403.
    Description of amendment request: The requested amendment proposes 
to depart from Tier 2* information in the Updated Final Safety Analysis 
Report (which includes the plant-specific design control document Tier 
2 information) related to the construction methods used for the 
composite floors and roof of the auxiliary building.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The design functions of the nuclear island structures are to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the nuclear island. 
The nuclear island structures are structurally designed to meet 
seismic Category I requirements as defined in Regulatory Guide 1.29.
    The use of ACI 349 and AISC N690 provides criteria for the 
design, qualification, fabrication, and inspection of composite 
steel beam floors and roof in the auxiliary building. These 
structures continue to meet the applicable portions of ACI 349 and 
AISC N690. The proposed change does not have an adverse impact on 
the response of the nuclear island structures to safe shutdown 
earthquake ground motions or loads due to

[[Page 13845]]

anticipated transients or postulated accident conditions. The change 
does not impact the support, design, or operation of mechanical and 
fluid systems. There is no change to plant systems or the response 
of systems to postulated accident conditions. There is no change to 
the predicted radioactive releases due to normal operation or 
postulated accident conditions. The plant response to previously 
evaluated accidents or external events is not adversely affected, 
nor does the change described create any new accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change revises the description of the construction 
of composite steel beam floors and roof in the auxiliary building. 
The proposed change does not change the design function, support, 
design, or operation of mechanical and fluid systems. The proposed 
change does not result in a new failure mechanism for the pertinent 
structures or new accident precursors. As a result, the design 
function of the structures is not adversely affected by the proposed 
change.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change is consistent with ACI 349 and AISC N690. 
The design and construction of the auxiliary building floors and 
roof remain in conformance with the requirements in ACI 349 and AISC 
N690.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 
1, 2, and 3, Maricopa County, Arizona
    Date of amendment request: February 27, 2015, as supplemented by 
letter dated January 19, 2016.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 1.3, ``Completion Times''; TS 3.7.5, ``Auxiliary 
Feedwater (AFW) System''; TS 3.8.1, ``AC [Alternating Current] 
Sources--Operating''; and TS 3.8.9, ``Distribution Systems--
Operating''; to remove the second Completion Times. The amendment also 
revised Example 1.3-3 in TS 1.3, ``Completion Times,'' by adding a 
discussion of administrative controls to combinations of conditions to 
ensure that the Completion Times for those conditions are not 
inappropriately extended.
    The changes are consistent with the NRC-approved Technical 
Specification Task Force (TSTF) Traveler TSTF-439-A, Revision 2, 
``Eliminate Second Completion Times Limiting Time From Discovery of 
Failure to Meet an LCO [Limiting Condition of Operation],'' dated June 
20, 2005.
    Date of issuance: February 19, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: Unit 1--197; Unit 2--197; Unit 3--197. A publicly-
available version is in ADAMS under Accession No. ML16004A013; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: 
The amendments revised the Operating Licenses and TSs.
    Date of initial notice in Federal Register: May 12, 2015 (80 FR 
27195). The supplement dated January 19, 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 February 19, 2016.
    No significant hazards consideration comments received: No.
Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick 
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina
    Date of amendment request: February 19, 2015, as supplemented by 
letter dated November 5, 2015.
    Description of amendment request: The amendments revised (1) 
technical specifications (TSs) by replacing AREVA Topical Report ANP-
10298PA, ``ACE/ATRIUM 10XM Critical Power Correlation,'' Revision 0, 
March 2010, with Revision 1, March 2014, of the same topical report; 
and (2) Appendix B, ``Additional Conditions,'' by removing the license 
condition issued by Amendment Nos. 262 and 290 for Units 1 and Unit 2, 
respectively.
    Date of issuance: February 9, 2016.
    Effective date: Once approved, the Unit 1 amendment shall be 
implemented prior to start-up. from the 2016 Unit 1 refueling outage, 
and the Unit 2 amendment shall be implemented prior to start-up from 
the 2017 Unit 2 refueling outage.
    Amendment Nos.: 269 and 297. A publicly-available version is in 
ADAMS under Accession No. ML16019A029;

[[Page 13846]]

documents related to these amendments are listed in the Safety 
Evaluation (SE) enclosed with the amendments.
    Facility Operating License Nos. DPR-71, and DPR-62: Amendments 
revised the renewed facility operating licenses and TSs.
    Date of initial notice in Federal Register: April 28, 2015 (80 FR 
23603). The supplemental letter dated November 5, 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 an SE dated February 9, 2016.
    No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station (CGS), 
Benton County, Washington
    Date of amendment request: September 2, 2015.
    Brief description of amendment: The amendment revised the Technical 
Specification (TS) requirements for unavailable barriers by adding 
Limiting Condition for Operation (LCO) 3.0.9. The LCO allows a delay 
time for entering a supported system TS, when the inoperability is 
solely due to an unavailable barrier, if the risk is assessed and 
managed. The change is consistent with NRC-approved Technical 
Specification Task Force (TSTF) Standard Technical Specification (STS) 
Change TSTF-427, Revision 2, ``Allowance for Non Technical 
Specification Barrier Degradation on Supported System OPERABILITY'' 
(ADAMS Accession No. ML061240055). The availability of this TS 
improvement was published in the Federal Register on October 3, 2006 
(71 FR 58444), as part of the Consolidated Line Item Improvement 
Process.
    Additionally, LCO 3.0.8 has been revised to replace the term 
``train'' with ``division'' to be consistent with CGS's TS definition 
of ``OPERABLE-OPERABILITY'' and the terminology used in Section 1.3, 
``Completion Times,'' of the CGS TS.
    Date of issuance: February 16, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 237. A publicly-available version is in ADAMS under 
Accession No. ML16020A031; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-21: The amendment 
revised the Facility Operating License and TSs.
    Date of initial notice in Federal Register: October 27, 2015 (80 FR 
65811).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 16, 2016.
    No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One (ANO), Units 1 and 2, Pope County, Arkansas
    Date of amendment request: May 20, 2015.
    Brief description of amendments: The amendments revised the full 
implementation date (Milestone 8) of the ANO, Units 1 and 2, Cyber 
Security Plan, and revised the associated physical protection license 
conditions for each renewed facility operating license.
    Date of issuance: February 24, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--255; Unit 2--303. A publicly-available 
version is in ADAMS under Accession No. ML16027A109; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-51 and NPF-6: The 
amendments revised the renewed facility operating licenses.
    Date of initial notice in Federal Register: June 23, 2015 (80 FR 
35982).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 24, 2016.
    No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2, Westchester County, New York
    Date of amendment request: December 9, 2014, as supplemented by two 
letters dated May 20, 2015, and letters dated June 8, 2015, and June 
29, 2015.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 5.5.14, ``Containment Leakage Rate Testing 
Program,'' to extend the frequency of the containment integrated leak 
rate test from once every 10 years to once every 15 years on a 
permanent basis.
    Date of issuance: February 23, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 283. A publicly-available version is in ADAMS under 
Accession No. ML15349A794; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. DPR-26: The amendment revised the 
Facility Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: March 17, 2015 (80 FR 
13905). The supplemental letters dated May 20, 2015; June 8, 2015; and 
June 29, 2015, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the 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 February 23, 2016.
    No significant hazards consideration comments received: Yes. The 
comments submitted by the State of New York on November 20, 2015, are 
addressed in the NRC staff's Safety Evaluation dated February 23, 2016.
Entergy Operations, Inc.; System Energy Resources, Inc.; South 
Mississippi Electric Power Association; and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne 
County, Mississippi
    Date of amendment request: May 27, 2015, as supplemented by letters 
dated October 28, 2015, and December 10, 2015.
    Brief description of amendment: The amendment revised the GGNS 
Technical Specifications (TSs) to allow for a permanent extension of 
the Type C leakage rate testing frequency and reduction of the Type B 
and Type C grace intervals that are required by GGNS TS 5.5.12, ``10 
CFR part 50, appendix J, Testing Program,'' by including a reference to 
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,'' dated July 2012. In addition, the amendment 
changed Surveillance Requirement (SR) 3.6.5.1.1 by deleting the 
information regarding the performance of the last Type A test that has 
already occurred. This amendment

[[Page 13847]]

does not alter the Type A testing frequencies nor any other 
requirements as specified in the existing GGNS TS.
    Date of issuance: February 17, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No: 209. A publicly-available version is in ADAMS under 
Accession No. ML16011A247; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-29: The amendment revised the 
Facility Operating License and TSs.
    Date of initial notice in Federal Register: September 29, 2015 (80 
FR 58516). The supplemental letters dated October 28, 2015, and 
December 10, 2015, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 17, 2016.
    No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit No. 1, Lake County, Ohio
    Date of amendment request: March 25, 2014, as supplemented by 
letters dated October 7, 2014, and August 24, 2015.
    Brief description of amendment: The amendment modifies the 
Technical Specifications (TSs) by relocating certain surveillance 
frequencies to a licensee-controlled program, the Surveillance 
Frequency Control Program, using probabilistic risk guidelines 
contained in NRC-approved Nuclear Energy Institute (NEI) 04-10, 
Revision 1, ``Risk-Informed Technical Specifications Initiative 5b, 
Risk-Informed Method for Control of Surveillance Frequencies.'' The 
changes are consistent with the approved Technical Specification Task 
Force (TSTF) Traveler TSTF-425, Revision 3, ``Relocate Surveillance 
Frequencies to Licensee Control-RITSTF Initiative 5b.''
    Date of issuance: February 23, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 171. A publicly-available version is in ADAMS under 
Accession No. ML15307A349; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-58: Amendment revised the 
Facility Operating License and TSs.
    Date of initial notice in Federal Register: September 16, 2014 (79 
FR 55512). The supplemental letters dated October 7, 2014, and August 
24, 2015, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 23, 2016.
    No significant hazards consideration comments received: No.
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: October 12, 2015.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) related to facility staff qualifications 
for licensed operators.
    Date of issuance: February 25, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos: 268 and 263. A publicly-available version is in 
ADAMS under Accession No. ML16008B072; documents related to these 
amendments are listed in the Safety Evaluation (SE) enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: December 22, 2015 (80 
FR 79620).
    The Commission's related evaluation of the amendments is contained 
in an SE dated February 25, 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: August 20, 2015, as supplemented by 
letter dated January 27, 2016.
    Brief description of amendment: The amendment made administrative 
changes to update personnel and committee titles in the Technical 
Specifications (TSs), deleted outdated or completed additional actions 
contained in Appendix B, Additional Conditions, of the license, and 
relocated the definition of Process Control Program from the TSs to the 
Updated Safety Analysis Report.
    Date of issuance: February 23, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 286. A publicly-available version is in ADAMS under 
Accession No. ML15307A013; 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, TSs, and Appendix B to the license.
    Date of initial notice in Federal Register: October 13, 2015 (80 FR 
61486). The supplemental letter dated January 27, 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 February 23, 2016.
    No significant hazards consideration comments received: No.
Pacific Gas and Electric Company (PG&E), Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo 
County, California
    Date of amendment request: February 25, 2015, as supplemented by 
letter dated July 8, 2015.
    Brief description of amendments: The amendments incorporated into 
the licensing basis an analysis of pressurizer reaching a water-solid 
(filled) condition associated with the main feedwater pipe rupture 
accident summarized in the Updated Final Safety Analysis Report 
(UFSAR), Section 15.4.2.2. Further, the amendments involved the 
addition of time critical operator actions and modifications of the 
PG&E Design Class I backup nitrogen accumulators, which are credited in 
the new pressurizer filling analysis.
    Date of issuance: February 19, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days following PG&E implementation of Design Class 1 backup 
nitrogen accumulator modifications, planned for the nineteenth 
refueling outage 2R19 for Unit No. 2.

[[Page 13848]]

    Amendment Nos.: Unit 1--223; Unit 2--225. A publicly-available 
version is in ADAMS under Accession No. ML16032A006; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses and UFSAR.
    Date of initial notice in Federal Register: April 28, 2015 (80 FR 
23605). The supplemental letter dated July 8, 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 19, 2016
    No significant hazards consideration comments received: No.
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: May 12, 2015, as supplemented by letters 
dated September 15, 2015; November 25, 2015; and January 28, 2016.
    Brief description of amendments: The amendments revised and added 
Surveillance Requirements to verify that the system locations 
susceptible to gas accumulation are sufficiently filled with water and 
to provide allowances that permit performance of the verification. The 
changes are consistent with Technical Specification Trask Force 
Traveler (TSTF)-523, Revision 2, ``Generic Letter 2008-01, Managing Gas 
Accumulation.''
    Date of issuance: February 26, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: Unit 1--200, Unit 2--196. A publicly-available 
version is in ADAMS under Accession No. ML15345A131, documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-2 and NPF-8: The amendments 
revised the Renewed Facility Operating Licenses and Technical 
Specifications.
    Date of initial notice in Federal Register: June 23, 2015 (80 FR 
35982). The supplemental letters dated September 15, 2015; November 25, 
2015; and January 28, 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 February 26, 2016.
    No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, 
Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield 
County, South Carolina
    Date of amendment request: May 18, 2015.
    Description of amendment: The amendment authorizes changes to the 
VCSNS, Units 2 and 3 Updated Final Safety Analysis Report by revising 
the Radiation Emergency Plan to expand the plume exposure pathway 
emergency planning zone (EPZ) boundary. The Evacuation Time Estimates 
Study and Alert and Notification System Design Report have also been 
revised to encompass the expanded EPZ boundary.
    Date of issuance: February 5, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 41. A publicly-available version is in ADAMS under 
Accession No. ML15292A404; documents related to this amendment are 
listed in a Safety Evaluation enclosed with the amendment.
    Facility Combined Licenses Nos. NPF-93 and NPF-94: Amendment 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: September 29, 2015 (80 
FR 585120).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 5, 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: August 21, 2015, as supplemented by 
letters dated September 17, 2015, and September 22, 2015.
    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* and associated Tier 2 information. The changes are to 
demonstrate that the capacity of mechanical couplers welded to 
structural steel embed plates required by American Concrete Institute 
(ACI) 349-01, ``Code Requirements for Nuclear Safety Related Concrete 
Structures,'' is satisfied using American Institute of Steel 
Construction (AISC) N690-1994, ``Specification for the Design, 
Fabrication, and Erection of Steel Safety-Related Structures for 
Nuclear Facilities,'' analysis and testing provisions.
    Date of issuance: November 5, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 40. A publicly-available version is in ADAMS under 
Accession No. ML15287A031; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: September 3, 2015 (80 
FR 53340). The supplemental letters dated September 17, 2015, and 
September 22, 2015, provided additional information that did not change 
the scope or the conclusions of the no significant hazards 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 5, 2015.
    No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri
    Date of amendment request: May 8, 2015, as supplemented by letter 
dated November 9, 2015.
    Brief description of amendment: The amendment revised Technical 
Specifications (TSs) 2.1.1.1 and 5.6.5 to adopt the NRC-approved 
methodologies of Westinghouse Commercial Atomic Power reports (WCAP)-
14483-A, ``Generic Method for Expanded Core Operating Limits Report,'' 
and WCAP-14565-P-A, Addendum 2-P-A, ``VIPRE-1 Modeling and 
Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic 
Safety Analysis,'' respectively. The change in TS 2.1.1.1 would provide 
the departure from nucleate boiling ratio in a form that reduces the 
need for cycle-specific license amendments, and the change in TS 5.6.5 
adds an NRC-approved methodology for determining core operating limits.

[[Page 13849]]

    Date of issuance: February 29, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 216. A publicly-available version is in ADAMS under 
Accession No. ML16020A516; 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 and TSs.
    Date of initial notice in Federal Register: July 7, 2015 (80 FR 
38763). The supplemental letter dated November 9, 2015, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 29, 2016.
    No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket No. 50-339, North Anna 
Power Station, Unit No. 2, Louisa County, Virginia
    Date of amendment request: May 22, 2015. As supplemented by letter 
dated October 13, 2015.
    Brief description of amendment: The amendment revised the Technical 
Specification (TS) 3.8.1, ``AC Sources-Operating,'' to remove the 
limitation in Note 1 that the surveillance is only applicable to Unit 
1. Revised Surveillance Requirement (SR) 3.8.1.8 is applicable to both 
units.
    Date of issuance: February 22, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 260. A publicly-available version is in ADAMS under 
Accession No. ML16013A444. Documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-7: Amendment revised the 
Facility Operating License and Technical Specification.
    Date of initial notice in Federal Register: July 21, 2015 (80 FR 
43131). The supplement letter dated October 13, 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.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 22, 2016.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 2nd day of March 2016.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2016-05470 Filed 3-14-16; 8:45 am]
 BILLING CODE 7590-01-P