[Federal Register Volume 79, Number 236 (Tuesday, December 9, 2014)]
[Notices]
[Pages 73106-73115]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-28704]


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NUCLEAR REGULATORY COMMISSION

[NRC-2014-0260]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, 
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is 
publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from November 13, 2014 to November 26, 2014. The 
last biweekly notice was published on November 25, 2014.

DATES: Comments must be filed by January 8, 2015. A request for a 
hearing must be filed by February 9, 2015.

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):
    <bullet> Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0260. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: <a href="/cdn-cgi/l/email-protection#93d0f2e1fcffbdd4f2fffff2f4fbf6e1d3fde1f0bdf4fce5"><span class="__cf_email__" data-cfemail="430022312c2f6d04222f2f22242b2631032d31206d242c35">[email protected]</span></a>. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
    <bullet> Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: 3WFN-06-A44M, 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: Mable Henderson, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-3760, email: <a href="/cdn-cgi/l/email-protection#85e8e4e7e9e0abede0ebe1e0f7f6eaebc5ebf7e6abe2eaf3"><span class="__cf_email__" data-cfemail="066b67646a63286e63686263747569684668746528616970">[email protected]</span></a>.

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2014-0260 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:
    <bullet> Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0260.
    <bullet> 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 <a href="/cdn-cgi/l/email-protection#255541570b5740564a50574640654b57460b424a53"><span class="__cf_email__" data-cfemail="106074623e6275637f65627375507e62733e777f66">[email protected]</span></a>. 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.
    <bullet> 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-2014-0260 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    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.

[[Page 73107]]

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 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 by the above date, 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 the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify 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.
    If a hearing is requested, 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.

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

[[Page 73108]]

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 
ten 10 days prior to the filing deadline, the participant should 
contact the Office of the Secretary by email at <a href="/cdn-cgi/l/email-protection#95fdf0f4e7fcfbf2bbf1faf6fef0e1d5fbe7f6bbf2fae3"><span class="__cf_email__" data-cfemail="fc94999d8e95929bd298939f979988bc928e9fd29b938a">[email protected]</span></a>, 
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 
<a href="/cdn-cgi/l/email-protection#6429372c204a3601170b11160701240a16074a030b12"><span class="__cf_email__" data-cfemail="3f726c777b116d5a4c504a4d5c5a7f514d5c11585049">[email protected]</span></a>, 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, 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.

[[Page 73109]]

Entergy Nuclear Vermont Yankee, LLC., and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont
    Date of amendment request: June 12, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14168A302.
    Description of amendment request: The proposed amendment would 
revise the site emergency plan (SEP) and Emergency Action Level (EAL) 
scheme to reflect the reduced scope of offsite and onsite emergency 
planning and the significantly reduced spectrum of credible accidents 
that can occur for the permanently defueled condition.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the emergency plan and EAL scheme do not 
impact the function of plant structures, systems, or components 
(SSCs). The proposed changes do not affect accident initiators or 
precursors, nor does it alter design assumptions. The proposed 
changes do not prevent the ability of the on-shift staff and 
emergency response organization (ERO) to perform their intended 
functions to mitigate the consequences of any accident or event that 
will be credible in the permanently defueled condition.
    The probability of occurrence of previously evaluated accidents 
is not increased, since most previously analyzed accidents can no 
longer occur and the probability of the few remaining credible 
accidents are unaffected by the proposed amendment.
    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 reduce the scope of the emergency plan and 
EAL scheme commensurate with the hazards associated with a 
permanently shutdown and defueled facility. The proposed changes do 
not involve installation of new equipment or modification of 
existing equipment, so that no new equipment failure modes are 
introduced. Also, the proposed changes do not result in a change to 
the way that the equipment or facility is operated so that no new or 
different kinds of accident initiators are created.
    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.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed changes are 
associated with the emergency plan and EAL scheme and do not impact 
operation of the plant or its response to transients or accidents. 
The change does not affect the Technical Specifications. The 
proposed changes do not involve a change in the method of plant 
operation, and no accident analyses will be affected by the proposed 
changes. Safety analysis acceptance criteria are not affected by the 
proposed changes. The revised SEP will continue to provide the 
necessary response staff with the proposed changes.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Douglas A. Broaddus.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3 (WF3), St. Charles Parish, Louisiana
    Date of amendment request: August 28, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14241A305.
    Description of amendment request: The amendment would revise the 
10-year frequency of the Type A or Integrated Leak Rate Test (ILRT) 
that is required by Technical Specification (TS) 6.15, ``Containment 
Leakage Rate Testing Program,'' to be extended to 15 years on a 
permanent basis.
    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 amendment involves changes to the WF3 Containment 
Leakage Rate Testing Program. The proposed amendment does not 
involve a physical change to the plant or a change in the manner in 
which the plant is operated or controlled. The primary reactor 
building function is to provide an essentially leak tight barrier 
against the uncontrolled release of radioactivity to the environment 
for postulated accidents. As such, the reactor building itself and 
the testing requirements to periodically demonstrate the integrity 
of the reactor building exist to ensure the plant's ability to 
mitigate the consequences of an accident, and do not involve any 
accident precursors or initiators. Therefore, the probability of 
occurrence of an accident previously evaluated is not significantly 
increased by the proposed amendment.
    The integrity of the reactor building is subject to two (2) 
types of failure mechanisms which 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 the reactor building 
containment integrity is not degraded by plant modifications or 
maintenance activities. The design and construction requirements of 
the reactor building itself combined with the reactor building 
inspections performed in accordance with ASME [American Society for 
Mechanical Engineers Boiler and Pressure Vessel Code], Section XI, 
the Maintenance Rule and regulatory commitments serve to provide a 
high degree of assurance that the containment will not degrade in a 
manner that is detectable only by a Type A test. Based on the above, 
the proposed amendment does not involve a significant increase in 
the consequences of an accident previously evaluate.
    The proposed amendment adopts the NRC-accepted guidelines of 
[Nuclear Energy Institute (NEI) 94-01, Revision 2-A, ``Industry 
Guideline for Implementing Performance-Based Option of 10 CFR part 
50, Appendix J,'' October 2008 (ADAMS Accession No. ML100620847)] 
for development of the WF3 performance-based testing program. 
Implementation of these guidelines continues to provide adequate 
assurance that during design basis accidents, the primary 
containment and its components will limit leakage rates to less than 
values assumed in the plant safety analyses. The potential 
consequences of extending the ILRT interval to fifteen (15) years 
have been evaluated by analyzing the resulting changes in risk. The 
increase in risk in terms of person-rem per year within fifty (50) 
miles resulting from design basis accidents was estimated to be 
acceptably small and determined to be within the guidelines 
published in RG [Regulatory Guide] 1.174. Additionally, the proposed 
change maintains defense-in-depth by preserving a reasonable balance 
among prevention of core damage, prevention of containment failure, 
and consequence mitigation. WF3 has determined that the increase in 
Conditional Containment Failure Probability due to the proposed 
change would be very small. Therefore, it is

[[Page 73110]]

concluded that the proposed amendment does not significantly 
increase the consequences of an accident previously evaluated.
    Based on the above discussion, it is concluded that 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 amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 2-A, for the development of the WF3 performance-
based leakage testing program, and establishes a fifteen (15) year 
interval for the performance of the reactor building ILRT. The 
reactor building and the testing requirements to periodically 
demonstrate the integrity of the reactor building exist to ensure 
the plant's ability to mitigate the consequences of an accident, and 
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.
    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 adopts the NRC-accepted guidelines of NEI 
94-01, Revision 2-A, for the development of the WF3 performance-
based leakage testing program, and establishes a fifteen (15) year 
interval for the performance of the containment ILRT. 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 Reactor 
Building Leakage Rate Testing Program, as defined in the TS, ensure 
that the degree of the reactor building structural integrity and 
leak-tightness that is considered in the plant's safety analysis is 
maintained. The overall reactor building leakage rate limit 
specified by the TS is maintained, and the Type A, Type B, and Type 
C containment leakage tests will be performed at the frequencies 
established in accordance with the NRC-accepted guidelines of NEI 
94-01, Revision 2-A.
    Containment inspections performed in accordance with other plant 
programs serve to provide a high degree of assurance that the 
containment will not degrade in a manner that is not detectable by 
an ILRT. A risk assessment using the current WF3 risk model 
concluded that extending the ILRT test interval from ten (10) years 
to fifteen (15) years results in a very small change to the WF3 risk 
profile.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Joseph A. Aluise, Associate General 
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, (TMI-1) Dauphin County, Pennsylvania
    Date of amendment request: October 30, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14304A083.
    Description of amendment request: The amendment would change the 
TMI-1 technical specifications (TSs). Specifically, the proposed 
amendment would modify the TS Table 3.1.6.1, ``Pressure Isolation Check 
Valves between the Primary Coolant System & LPIS [Low Pressure 
Injection System],'' maximum allowable leakage limits.
    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, along 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 changes will not alter the way any structure, 
system, or component (SSC) functions, and will not alter the manner 
in which the plant is operated. In addition, the proposed amendment 
will not impact the ability of any SSC to mitigate an accident as 
currently evaluated in the UFSAR [Updated Final Safety Analysis 
Report].
    This proposed change deletes certain Reactor Coolant System 
Pressure Isolation Valve (RCS PIV) allowable leakage surveillance 
testing criteria in consideration of the safety significance and 
design capabilities of the plant and current industry testing and 
maintenance practices. The proposed change is consistent with 
Improved Standard Technical Specification (ITS) NUREG 1430, 
[``]Standard Technical Specifications, Babcock and Wilcox Plants,'' 
Revision 4, and current RCS PIV leak testing practices. The maximum 
allowable leakage rate of 5 gpm [gallons per minute] remains 
unchanged; only the leakage testing incremental testing acceptance 
criteria below the 5 gpm limit is being deleted. Since the testing 
frequency and maximum allowable leakage remains unchanged, the 
probability or consequence of an interfacing system loss-of-coolant 
accident (ISLOCA) is unaffected. There are no changes to the 
[American Society of Mechanical Engineers] ASME [Operation and 
Maintenance] OM Code leakage testing requirements and methods for 
this class of valves. Additionally, two typographical errors and one 
clerical error are being corrected which are administrative in 
nature.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed revision is not a result of changes to plant 
equipment, system design, or operating practices. The modified 
[limiting condition of operation] LCO requirement will allow some 
relaxation of the leak testing method acceptance criteria for the 
RCS PIVs, consistent with NUREG-1430. Since the functions of the 
associated systems will continue to perform without change, the 
proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
Further, the proposed changes do not introduce any new failure 
modes. Additionally, two typographical errors and one clerical error 
are being corrected which are administrative in nature.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed revision to the RCS PIV leakage testing acceptance 
criteria will not result in changes to system design or setpoints 
that are intended to ensure timely identification of plant 
conditions that could be precursors to accidents or potential 
degradation of accident mitigation systems. Since testing frequency 
and maximum allowable leakage for the RCS PIVs remain unchanged, the 
margin associated with the identification of RCS PIV degradation is 
not significantly reduced. The confidence in the ability of the 
fission product barriers (fuel cladding, RCS boundary, containment) 
to limit the level of radiation dose to the public remains the same. 
Additionally, two typographical errors and one clerical error are 
being corrected which are administrative in nature.
    Since the setpoints and design features that support the margin 
of safety are unchanged, and actions for inoperable systems continue 
to provide appropriate time limits and compensatory measures, the 
proposed changes will not significantly reduce the margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

[[Page 73111]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Esquire, Exelon 
Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Meena Khanna.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, 
California
    Date of amendment request: November 25, 2013. A publicly-available 
version is in ADAMS under Accession No. ML13330A557.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs) to permit the use of Risk-
Informed Completion Times (CTs) in accordance with Technical 
Specification Task Force (TSTF) traveler, TSTF-505, Revision 1, 
``Provide Risk-Informed Extended Completion Times--RITSTF [Risked-
Informed TSTF] Initiative 4b.'' The proposed amendment would, in part, 
modify selected Required Actions to permit extending the CTs in 
accordance with a new TS-required risk-informed completion time (RICT) 
program. The availability of the model safety evaluation for TSTF-505 
was published by the NRC staff in the Federal Register on March 15, 
2012 (77 FR 15399,) for referencing in license amendment applications.
    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 permits the extension of CTs provided the 
associated risk is assessed and managed in accordance with the NRC[-
]approved Risk Informed Completion Time (RICT) Program. The proposed 
change does not involve a significant increase in the probability of 
an accident previously evaluated because the change involves no 
change to the plant or its modes of operation. The proposed change 
does not increase the consequences of an accident [previously 
evaluated] because the design basis mitigation function of the 
affected systems is not changed and the consequences of an accident 
during the extended CT are no different from those during the 
existing CT.
    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] 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not change the design, configuration, 
or method of operation of the plant. The proposed change does not 
involve a physical alteration of the plant (no new or different kind 
of equipment will be installed).
    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 permit[s] the extension of CTs provided risk 
is assessed and managed in accordance with the NRC[-]approved RICT 
Program. The proposed change implements a risk-informed 
configuration management program to assure that adequate margins of 
safety are maintained. Application of these new specifications and 
the configuration management program considers cumulative effects of 
multiple systems or components being out of service and does so more 
effectively than the current TS.
    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 
requested amendment involve no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: Michael T. Markley.
South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028, 
Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, 
South Carolina
    Date of amendment request: September 11, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14254A371.
    Description of amendment request: The proposed changes would revise 
the Combined Licenses by clarifying the position on design diversity, 
specifically human diversity, as related to the Component Interface 
Module (CIM) and Diverse Actuation System (DAS) design.
    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 requested amendment proposes changes to licensing basis 
documents to clarify the position on the human diversity aspects of 
design diversity as related to the Component Interface Module (CIM) 
and Diverse Actuation System (DAS) design processes. A review 
confirmed that the clarified position on human diversity would not 
change the CIM or DAS design. The requested changes to information 
presented in the Tier 2* and Tier 2 supporting documentation clarify 
the level of human diversity applied. The change continues to comply 
with the regulatory guidance in NUREG/CR-6303 regarding credible 
defenses against a postulated Common Cause Failure (CCF) of the 
Plant Monitoring and Safety System. The proposed change does not 
affect the plant itself. The change does not affect prevention and 
mitigation of abnormal events, e.g., accidents, anticipated 
operational occurrences, earthquakes, floods and turbine missiles, 
or their safety or design analyses. No safety-related structure, 
system, or component (SSC) or function is adversely affected. The 
change does not involve nor interface with any SSC accident 
initiator or initiating sequence of events, and thus, the 
probabilities of the accidents evaluated in the Updated Final Safety 
Analysis Report (UFSAR) are not affected. This activity will not 
allow for a new fission product release path, nor will it result in 
a new fission product barrier failure mode, nor create a new 
sequence of events that would result in significant fuel cladding 
failures. Because the proposed changes do not change any safety-
related SSC or function credited in the mitigation of an accident, 
the consequences of the accidents evaluated in the UFSAR are not 
affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes clarify the position on human diversity and 
show that the CIM/DAS diversity meets the regulatory guidance in 
NUREG/CR-6303. The clarified descriptions do not affect the plant 
itself. Therefore, the proposed changes do not affect any safety-
related equipment itself, nor do they affect equipment whose failure 
could initiate an accident or a failure of a fission product 
barrier. No analysis is adversely affected by the proposed changes. 
No system or design function or equipment qualification would be 
adversely affected by the proposed changes. Furthermore, the 
proposed changes do not result in a new failure mode, malfunction, 
or sequence of events that could affect safety or safety-related 
equipment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different

[[Page 73112]]

kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to information presented in referenced 
licensing basis documents clarify the position regarding human 
diversity and do not affect the plant itself. The proposed changes 
do not adversely affect the design, construction, or operation of 
any plant SSCs, including any equipment whose failure could initiate 
an accident or a failure of a fission product barrier. No analysis 
is adversely affected by the proposed changes. Furthermore, no 
system function, design function, or equipment qualification will be 
adversely affected by the changes.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence J. Burkhart.
South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028, 
Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, 
South Carolina
    Date of amendment request: October 23, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14296A758.
    Description of amendment request: The proposed changes would revise 
the Combined Licenses (COLs) changing the description and scope of the 
Initial Test Program. Because, this proposed change requires a 
departure from Tier 1 information in the Westinghouse Advanced Passive 
1000 Design Control Document (DCD), the licensee also requested an 
exemption from the requirements of the Generic DCD Tier 1 in accordance 
with 10 CFR 52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment is related to the conduct of the Initial 
Test Program. The proposed changes are made in compliance with the 
applicable regulatory guides, are only related to the general 
aspects of how the program is executed and do not change any 
technical content for preoperational or startup tests. No changes 
are made to any design aspect of the plant.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment is related to the conduct of the Initial 
Test Program. The proposed changes are made in compliance with the 
applicable regulatory guides, are only related to the general 
aspects of how the program is executed and do not change any 
technical content for preoperational or startup tests. These changes 
do not affect the design or analyzed operation of any system.
    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 amendment is related to the conduct of the Initial 
Test Program. The proposed changes are made in compliance with the 
applicable regulatory guides, are only related to the general 
aspects of how the program is executed and do not change any 
technical content for preoperational or startup tests. No safety 
analysis or design basis acceptance limit/criterion is challenged or 
exceeded by the proposed changes, thus no margin of safety is 
reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company, Inc. Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, 
Georgia
    Date of amendment request: October 16, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14290A139.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-91 and NPF-92 for the VEGP, Units 3 and 4. 
The requested amendment proposes changes to revise the VEGP Updated 
Final Safety Analysis Report (UFSAR), involving Tier 1 and associated 
Tier 2 departures to add or delete piping line numbers of existing 
piping lines, or updating the functional capability classification of 
existing process flow lines.
    Because this proposed change requires a departure from Tier 1 
information in the Westinghouse Advanced Passive 1000 design control 
document (DCD), the licensee also requested an exemption from the 
requirements of the Generic DCD Tier 1 in accordance with 10 CFR 
52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The COL Appendix C Tables and corresponding plant-specific Tier 
1 Tables proposed changes involve updating piping line name/number 
or functional capability requirements. These changes do not affect 
any system design function. Adding or updating information for 
existing ASME Section III piping does not involve (i.e., cannot 
affect) any accident initiating event or component failure, thus, 
the probabilities of the accidents previously evaluated are not 
affected. The maximum allowable leakage rate specified in the 
Technical Specifications is unchanged and radiological material 
release source terms are not affected, thus, the radiological 
releases in the accident analyses are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The COL Appendix C Tables and corresponding plant-specific Tier 
1 Tables proposed changes to update piping line name/number or 
functional capability requirements do not adversely affect the 
design or quality of any structure, system, or component. Adding or 
updating ASME Section III piping line information for existing 
process piping lines to a licensing table does not create a new 
fault or sequence of events that could result in a radioactive 
material release.
    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.

[[Page 73113]]

    The COL Appendix C Tables and corresponding plant-specific Tier 
1 Tables proposed changes involve updating piping line name/number 
or functional capability requirements information for new/existing 
process piping lines. Adding or updating the ASME Section III piping 
line name/number or functional capability requirements in the tables 
would not affect any radioactive material barrier. No safety 
analysis or design basis acceptance limit/criterion is challenged or 
exceeded by the proposed changes, thus, no margin of safety is 
reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

III. Previously Published Notices of Consideration of Issuance of 
Amendments to Facility Operating Licenses and Combined Licenses, 
Proposed No Significant Hazards Consideration Determination, and 
Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit 1, Washington County, Nebraska
    Date of amendment request: November 7, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14311A158.
    Brief description of amendment request: The proposed amendment 
revises a limited number of Technical Specification Surveillance 
Requirements by adding a note or footnote permitting a one-time 
extension from a refueling frequency (i.e., at least once per 18 
months) to a maximum of 28 months. These surveillance requirements 
include (1) manual containment isolation actuation, (2) manual 
recirculation actuation and recirculation actuation logic, (3) steam 
generator level calibration, (4) visual examination of the high-
efficiency particulate air and charcoal filters in the containment 
recirculating air cooling and filtering system, (5) emergency diesel 
generators, and (6) residual heat removal system integrity. An 
extension is necessary because these tests will expire before the next 
refueling outage begins on April 11, 2015.
    Date of publication of individual notice in Federal Register: 
November 17, 2014 (79 FR 68487).
    Expiration date of individual notice: December 17, 2014 (public 
comments); January 17, 2015 (hearing requests).

IV. 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.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
1, Pope County, Arkansas
    Date of amendment request: January 28, 2013, as supplemented by 
letter dated August 28, 2013.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) requirements related to direct current (DC) 
electrical systems as specified in TS Limiting Condition for Operation 
(LCO) 3.8.4, ``DC Sources--Operating,'' LCO 3.8.5, ``DC Sources--
Shutdown,'' and LCO 3.8.6, ``Battery Parameters.'' A new ``Battery 
Monitoring and Maintenance Program'' is now required under TS Section 
5.5, ``Administrative Controls--Programs and Manuals.'' These changes 
are consistent with the NRC-approved Technical Specifications Task 
Force (TSTF) Traveler TSTF-500, Revision 2, ``DC Electrical Rewrite--
Update to TSTF-360.'' The availability of this TS improvement was 
announced in the Federal Register on September 1, 2011 (76 FR 54510).
    Date of issuance: November 24, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 250. A publicly-available version is in ADAMS under 
Accession No. ML14254A133; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the Technical Specifications/license.
    Date of initial notice in Federal Register: April 30, 2013 (78 FR 
25313). The supplemental letter dated August 28, 2013, 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 November 24, 2014.
    No significant hazards consideration comments received: No.

[[Page 73114]]

Exelon Generation Company, LLC, Docket No. 50-373, LaSalle County 
Station (LSCS), Unit 1, LaSalle County, Illinois
    Date of amendment request: December 20, 2013, as supplemented by 
letters dated February 26, 2014, September 11, 2014 (2 letters), and 
October 14, 2014.
    Brief description of amendment: The amendment revised the LSCS, 
Unit 1, pressure and temperature curves, Figures 3.4.11-1 through 
3.4.11-3, in Technical Specification 3.4.11, ``RCS [Reactor Coolant 
System] Pressure and Temperature (P/T) Limits.''
    Date of issuance: November 25, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 210. A publicly-available version is in ADAMS under 
Accession No. ML14288A151; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-11: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 5, 2014 (79 FR 
45490). The supplemental letters dated September 11, 2014 (2 letters) 
and October 14, 2014, 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 November 25, 2014.
    No significant hazards consideration comments received: No.
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: November 6, 2013, supplemented by 
letters dated June 13, 2014, and August 15, 2014.
    Brief description of amendments: The amendments revised the 
Technical Specification 3.6.13, Divider Barrier Integrity, Surveillance 
Requirement 3.6.13.5 for the divider barrier seal inspection for the 
Donald C. Cook Nuclear Plant, Units 1 and 2.
    Date of issuance: November 20, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 324 for Unit 1 and 307 for Unit 2.
    Renewed Facility Operating License Nos. DPR-58 and DPR-74: The 
amendments revise the Facility Operating Licenses and Technical 
Specifications.
    Date of initial notice in Federal Register: February 19, 2014 (79 
FR 9496). The supplemental letters dated June 13, 2014, and August 15, 
2014, 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 November 20, 2014.
    No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota
    Date of amendment request: March 11, 2013, as supplemented by 
letter dated July 3, 2014.
    Brief description of amendment: The amendment changes the reactor 
steam dome pressure value specified in technical specification (TS) 
2.1.1, ``Reactor Core SLs [Safety Limits],'' from 785 pounds per square 
inch gauge (psig) to 686 psig. This change resolves a 10 CFR part 21, 
``Reporting of Defects and Noncompliance,'' condition concerning a 
potential to momentarily violate the safety limit specified in TS 
2.1.1.1 during a pressure regulator failure maximum demand (open) 
transient.
    Date of issuance: November 25, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 185. A publicly-available version is in ADAMS under 
Accession No. ML14281A318; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-22: This amendment 
revises the Renewed Facility Operating License and the Technical 
Specifications.
    Date of initial notice in Federal Register: June 11, 2013 (78 FR 
35064). The supplemental letter dated July 3, 2014, 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 November 25, 2014.
    No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota
    Date of amendment request: May 23, 2013, as supplemented by letter 
dated March 25, June 26, and October 20, 2014.
    Brief description of amendment: The amendments revised Technical 
Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR),'' to 
reference and allow use of Westinghouse report WCAP-16045-P-A, 
``Qualification of the Two-Dimensional Transport Code PARAGON'' and 
WCAP-16045-P-A, Addendum 1-A, ``Qualification of the NEXUS Nuclear Data 
Methodology,'' to determine core operating limits.
    Date of issuance: November 19, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: Unit 1-211; Unit 2-199. A publicly-available 
version is in ADAMS under Accession No. ML14296A666; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-42, and DPR-60: These 
amendments revised the Renewed Facility Operating License and the 
Technical Specifications.
    Date of initial notice in Federal Register: August 20, 2013 (78 FR 
51229). The supplements dated March 25, June 26, and October 20, 2014, 
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 November 19, 2014.
    No significant hazards consideration comments received: No.

[[Page 73115]]

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: July 2, 2013, and revised by letters 
dated February 14, and June 20, 2014, and supplemented by letters dated 
August 28 and October 14, 2014.
    Brief description of amendment: The amendment revises the design of 
connections between reinforced concrete and steel plate concrete 
composite construction included in the VEGP, Units 3 and 4 updated 
Final Safety Analysis Report (UFSAR) and changes to the Technical 
Report, ``APP-GW-GLR-602, AP1000 Shield Building Design Details for 
Select Wall and RC/SC Connections,'' (prepared by Westinghouse Electric 
Company and reviewed by the NRC as part of the design certification 
rule). This Technical Report is incorporated by reference in the VEGP, 
Units 3 and 4 UFSAR.
    Date of issuance: November 21, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 26. A publicly-available version is in ADAMS under 
Accession No. ML14322A275; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: September 3, 2013 (78 
FR 54287).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 21, 2014.
    No significant hazards consideration comments received: No.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: September 3, 2013 (78 
FR 54287).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 21, 2014.
    No significant hazards consideration comments received: No.


    Dated at Rockville, Maryland, this 1st day of December 2014.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2014-28704 Filed 12-8-14; 8:45 am]
BILLING CODE 7590-01-P


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