[Federal Register Volume 79, Number 227 (Tuesday, November 25, 2014)]
[Notices]
[Pages 70209-70220]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-27630]


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NUCLEAR REGULATORY COMMISSION

[NRC-2014-0252]


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 October 30, 2014 to November 12, 2014. The 
last biweekly notice was published on November 12, 2014.

DATES: Comments must be filed by December 26, 2014. A request for a 
hearing must be filed by January 26, 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-2014-0252. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: <a href="/cdn-cgi/l/email-protection#d596b4a7bab9fb92b4b9b9b4b2bdb0a795bba7b6fbb2baa3"><span class="__cf_email__" data-cfemail="98dbf9eaf7f4b6dff9f4f4f9fff0fdead8f6eafbb6fff7ee">[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: Sandra Figueroa, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-1262, email: <a href="/cdn-cgi/l/email-protection#cdbeaca3a9bface3aba4aab8a8bfa2ac8da3bfaee3aaa2bb"><span class="__cf_email__" data-cfemail="f182909f958390df9798968494839e90b19f8392df969e87">[email protected]</span></a>.

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2014-0252 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-2014-0252.
    <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#aadaced884d8cfd9c5dfd8c9cfeac4d8c984cdc5dc"><span class="__cf_email__" data-cfemail="6e1e0a1c401c0b1d011b1c0d0b2e001c0d40090118">[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-2014-0252 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

[[Page 70210]]

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 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.

[[Page 70211]]

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 
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#ea828f8b9883848dc48e8589818f9eaa849889c48d859c"><span class="__cf_email__" data-cfemail="630b0602110a0d044d070c00080617230d11004d040c15">[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#327f617a761c6057415d47405157725c40511c555d44"><span class="__cf_email__" data-cfemail="cc819f8488e29ea9bfa3b9beafa98ca2beafe2aba3ba">[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,

[[Page 70212]]

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.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit 2, New London County, Connecticut

    Date of amendment request: March 28, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14093A027.
    Description of amendment request: The amendment would make changes 
to the Millstone Power Station, Unit 2 (MPS2) Technical Specifications 
(TSs). The proposed changes delete the TS Index and make administrative 
changes and corrections to the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes are administrative in nature. The proposed 
changes remove the TS Index and make other editorial and 
administrative corrections to the TSs. These administrative changes 
are not initiators of any accident previously evaluated, and, 
consequently, the probability and consequence of an accident 
previously evaluated is not significantly increased.
    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 are administrative in nature so no new or 
different accidents result from the proposed changes. The changes do 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed), a change in the 
method of plant operation, or new operator actions. The changes do 
not alter assumptions made in the safety analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed administrative changes do not involve a change in 
the method of plant operation, do not affect any accident analyses, 
and do not relax any safety system settings.
    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 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: Benjamin G. Beasley.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station (MPS), Unit 2, New London County, Connecticut

    Date of amendment request: April 11, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14112A072.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to add the Framatome-ANP 
(AREVA) topical report for the M5[supreg] (hereafter referred as M5) 
fuel rod cladding material to TS 6.9.1.8.b, ``Core Operating Limits 
Report.'' The M5 fuel rod cladding material was approved by the NRC in 
Topical Report BAW-10240(P)(A), Revision 0, ``Incorporation of 
M5<SUP>TM</SUP> Properties in FramatomeANP Approved Methods.''
    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 TS change is: (1) Adding BAW-10240(P)(A) to the 
list of approved methodologies for determining core operating limits 
at MPS2.
    The proposed change to TS 6.9.1.8.b permits the use of the 
appropriate methodology to analyze accidents for cores containing 
fuel with M5 cladding to ensure that the plant continues to meet 
applicable design criteria and safety analysis acceptance criteria. 
The proposed change to the list of NRC-approved methodologies listed 
in TS 6.9.1.8.b has no impact on plant operation and configuration. 
The list of methodologies in TS 6.9.1.8.b does not impact either the 
initiation of an accident or the mitigation of its consequences.
    The NRC has previously approved use of M5 fuel rod cladding 
material provided that licensees ensure compliance with the 
conditions set forth in the NRC SE [Safety Evaluation] for topical 
report BAW-10240(P)(A). Confirmation that these conditions are 
satisfied is performed under 10 CFR 50.59 as part of the normal core 
reloads process.
    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 addition of topical report BAW-10240(P)(A) to the 
list of NRC approved methodologies listed in TS 6.9.1.8.b, has no 
impact on any plant configuration or system performance. There is no 
change to the parameters within which the plant is normally 
operated, and thus, the possibility of a new or different type of 
accident is not created.
    Therefore, the addition of BAW-10240(P)(A) to TS 6.9.1.8.b does 
not create the possibility of a new or different kind of accident or 
malfunction from those previously evaluated within the FSAR.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to the list of NRC-approved methodologies 
listed in TS 6.9.1.8.b has no impact on any plant configuration or 
system performance. Topical report BAW-10240(P)(A) has been reviewed 
and approved by the NRC for use with M5 fuel rod cladding. Approved 
methodologies will be used to ensure that the plant continues to 
meet applicable design criteria and safety analysis acceptance 
criteria.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: Benjamin G. Beasley.

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power 
Station, Unit 3, New London County, Connecticut

    Date of amendment request: November 6, 2013. A publicly-available 
version is in ADAMS under Accession No. ML13322A415.
    Description of amendment request: The amendment would revise 
Technical

[[Page 70213]]

Specification (TS) 3/4.5.4, ``Refueling Water Storage Tank,'' and TS 3/
4.6.2.1, ``Depressurization and Cooling Systems, Containment Quench 
Spray System [QSS],'' to provide additional operational margin for 
control of the Refueling Water Storage Tank (RWST) temperature.
    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 affects the allowable limit for RWST 
temperature. Since the RWST is a passive component used as a water 
supply for ECCS [Emergency Core Cooling System] and QSS that operate 
only following an accident, the proposed change cannot cause an 
accident or affect the probability of any accident.
    Evaluations have been performed to address the impact of raising 
the maximum RWST temperature on the performance of the ECCS and QSS. 
The evaluations demonstrate that NPSH [Net Positive Suction Head] 
margin would be maintained for the ECCS and QSS pumps that take 
suction from the RWST following a Safety Injection Actuation Signal 
or a Containment Depressurization Actuation Signal. Pipe and 
component stress limits continue to be met at the higher RWST 
temperature. Thus, it is concluded that the ECCS and QSS will 
continue to meet the design basis requirements.
    The FSAR [Final Safety Analysis Report] Chapter 15 accident 
analyses and Chapter 6 containment analyses were performed assuming 
an RWST temperature that bounds the proposed technical specification 
change. Thus, the proposed change has no significant impact on the 
consequences of an accident as documented in the current analysis of 
record.
    Changing the ACTION statement to include the wording ``the 
next'' is administrative and editorial in nature. This proposed 
change does not alter the effective technical content of the ACTION 
statement.
    Thus, it is concluded that the proposed changes do not involve a 
significant increase in the probability or consequences of any 
analyzed accident.
    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 only increases the allowable range for the 
RWST temperature. As such, it cannot initiate a transient or 
accident. Evaluations have been performed that demonstrate that the 
ECCS and QSS systems will have adequate NPSH and the design bases 
will be met.
    Thus, the proposed change cannot create the possibility of a new 
or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Evaluations have been performed that demonstrate that the ECCS 
and QSS pumps will maintain NPSH margin when taking suction from the 
RWST at the higher temperature limit. The mechanical component 
stress requirements will continue to be met at the higher 
temperature.
    Thus, the ECCS and QSS will continue to operate as required to 
mitigate a design basis accident.
    The accident analyses were performed with assumed RWST 
temperatures that bound this proposed change. The containment 
analysis and accident analyses demonstrate that the design basis 
requirements are met.
    Thus, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
    NRC Branch Chief: Benjamin G. Beasley.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power 
Station (MPS), Unit 3, New London County, Connecticut
    Date of amendment request: March 28, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14093A026.
    Description of amendment request: The amendment would make the 
following changes to the MPS3 Technical Specifications (TSs):

    (a) Delete TS index pages i through xix.
    (b) Replace the first sub-letter under TS Table 4.3-2 Item 4, 
Steam Line Isolation--``Manual Initiation,'' which currently appears 
as sub-letter ``d'' on Page 3/4 3-37, with sub-letter ``a.''
    (c) Revise TS 6.3.2, Facility Staff Qualifications, from:
    ``If the operations manager does not hold a senior reactor 
operator license for Millstone Unit No. 3, then the operations 
manager shall have held a senior reactor operator license at a 
pressurized water reactor, and the assistant operations manager 
shall hold a senior reactor operator license for Millstone Unit No. 
3.''
    to:
    ``The operations manager or at least one operations middle 
manager shall hold a senior reactor operator license for Millstone 
Unit No. 3.''
    (d) Replace the term ``SORC'' in paragraph b of the ``Licensee 
initiated changes to the REMODCM,'' described in TS 6.13 with the 
term ``FSRC.''

    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 are administrative in nature. The proposed 
changes remove the TS Index and make other editorial and 
administrative corrections to the TSs. These administrative changes 
are not initiators of any accident previously evaluated, and, 
consequently, the probability and consequences of an accident 
previously evaluated is not significantly increased.
    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 are administrative in nature so no new or 
different accidents result from the proposed changes. The changes do 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed), a change in the 
method of plant operation, or new operator actions. The changes do 
not alter assumptions made in the safety analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed administrative changes do not involve a change in 
the method of plant operation, do not affect any accident analyses, 
and do not relax any safety system settings.
    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 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
    NRC Branch Chief: Benjamin G. Beasley.

[[Page 70214]]

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, Claiborne 
County, Mississippi

    Date of amendment request: August 1, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14216A383.
    Description of amendment request: The proposed amendment would 
revise the allowable values in Technical Specification (TS) Table 
3.3.5.1-1, ``Emergency Core Cooling Instrumentation,'' Functions 4.c 
and 5.c; Table 3.3.6.3-1, ``RHR [Residual Heat Removal] Containment 
Spray System Instrumentation,'' Function 4; and Table 3.3.8.1-1, ``Loss 
of Power Instrumentation,'' Functions 1.c, 2.c, and 2.e. Revisions of 
setpoint calculations supporting the above tables, identified that the 
allowable values in the above functions are non-conservative. The 
licensee has noted that while the allowable values are non-
conservative, the setpoints remain conservative.
    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 TS allowable value changes involve changes in the 
margin between the allowable values and the setpoints. The proposed 
TS changes do not change the trip setpoints. The proposed TS changes 
do not degrade the performance of, or increase the challenges to, 
any safety systems assumed to function in the accident analysis. The 
proposed TS changes do not impact the usefulness of the 
[surveillance requirements] SRs in evaluating the operability of 
required systems and components, or the way in which the 
surveillances are performed. In addition, the [* * *] trip setpoints 
for the associated TRM [Technical Requirements Manual] functions are 
not considered an initiator of any analyzed accident, nor does a 
revision to the allowable value introduce any accident initiators. 
Therefore, the proposed change does not involve a significant 
increase in the probability of an accident previously evaluated.
    The consequences of a previously evaluated accident are not 
significantly increased. The proposed change does not affect the 
performance of any equipment credited to mitigate the radiological 
consequences of an accident. Evaluation of the proposed TS changes 
demonstrated that the availability of credited equipment is not 
significantly affected because of the reduction in margin between 
the allowable values and the trip setpoints.
    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 TS changes involves changes in allowable value 
settings to correct non-conservative values. The proposed TS changes 
do not introduce any failure mechanisms of a different type than 
those previously evaluated, since there are no physical changes 
being made to the facility.
    No new or different equipment is being installed. No installed 
equipment is being operated in a different manner. As a result, no 
new failure modes are being introduced. The way surveillance tests 
are performed remains unchanged.
    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 TS change involves changes in the allowable value 
settings to correct non-conservative values. The impact of the 
change on system availability is not significant, based on the 
frequency of the testing being unchanged, the existence of redundant 
systems and equipment, and overall system reliability. The proposed 
change does not significantly impact the condition or performance of 
structures, systems, and components relied upon for accident 
mitigation. The proposed change does not result in any hardware 
changes or in any changes to the analytical limits assumed in 
accident analyses. Existing operating margin between plant 
conditions and actual plant setpoints is not significantly reduced 
due to these changes. The proposed change does not impact any safety 
analysis assumptions or results.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Douglas A. Broaddus.

Exelon Generation Company, LLC (EGC), Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Calvert County, 
Maryland

    Date of amendment request: September 18, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14265A219.
    Description of amendments request: The amendment(s) would revise 
Technical Specification (TS) 5.5.16, ``Containment Leakage Rate Testing 
Program'' to allow for permanent extensions of Type A Integrated Leak 
Rate Testing (ILRT) and Type C Leak Rate Testing frequencies to 15 
years and 75 months, respectively.
    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 to the TS involves the extension of the 
Calvert Cliffs Unit 1 and 2 Type A containment test interval to 15 
years and the extension of the Type C test interval to 75 months. 
The current Type A test interval of 120 months (10 years) would be 
extended on a permanent basis to no longer than 15 years from the 
last Type A test. The current Type C test interval of 60 months for 
selected components would be extended on a performance basis to no 
longer than 75 months. Extensions of up to nine months (total 
maximum interval of 84 months for Type C tests) are permissible only 
for non-routine emergent conditions. The proposed extension does not 
involve either a physical change to the plant or a change in the 
manner in which the plant is operated or controlled. The containment 
is designed to provide an essentially leak tight barrier against the 
uncontrolled release of radioactivity to the environment for 
postulated accidents. As such, the containment and the testing 
requirements invoked to periodically demonstrate the integrity of 
the containment exist to ensure the plant's ability to mitigate the 
consequences of an accident, and do not involve the prevention or 
identification of any precursors of an accident. 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 Calvert Cliffs Unit 1 and 2 Type A 
test history supports this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as (1) activity based and 
(2) time based. Activity based failure

[[Page 70215]]

mechanisms are defined as degradation due to system and/or component 
modifications or maintenance. Local leak rate test requirements and 
administrative controls such as configuration management and 
procedural requirements for system restoration ensure that 
containment integrity is not degraded by plant modifications or 
maintenance activities. The design and construction requirements of 
the containment combined with the containment inspections performed 
in accordance with ASME [American Society of Mechanical Engineers] 
Section Xl, the Maintenance Rule, and TS requirements serve to 
provide a high degree of assurance that the containment would not 
degrade in a manner that is detectable only by a Type A test. Based 
on the above, the proposed extension does not significantly increase 
the consequences of an accident previously evaluated.
    The proposed amendment also deletes exceptions previously 
granted to allow one time extensions of the ILRT test frequency for 
both Units 1 and 2 and exceptions from conducting post modification 
ILRT following replacement of the Units 1 and 2 Steam Generators. 
These exceptions were for things that have already taken place so 
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 
Calvert Cliffs Unit 1 and 2 Type A containment test interval to 15 
years and the extension of the Type C test interval to 75 months. 
The containment and the testing requirements to periodically 
demonstrate the integrity of the containment exist to ensure the 
plant's ability to mitigate the consequences of an accident do not 
involve any accident precursors or initiators. The proposed change 
does not involve a physical change to the plant (i.e., no new or 
different type of equipment will be installed) or a change to the 
manner in which the plant is operated or controlled.
    The proposed amendment also deletes exceptions previously 
granted to allow one time extensions of the ILRT test frequency for 
both Units 1 and 2 and exceptions from conducting post modification 
ILRT following replacement of the Units 1 and 2 Steam Generators. 
These exceptions were for things that have already taken place so 
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 a 
margin of safety?
    Response: No.
    The proposed amendment to TS 5.5.16 involves the extension of 
the Calvert Cliffs Unit 1 and 2 Type A containment test interval to 
15 years and the extension of the Type C test interval to 75 months 
for selected components. This amendment does not alter the manner in 
which safety limits, limiting safety system set points, or limiting 
conditions for operation are determined. The specific requirements 
and conditions of the TS Containment Leak Rate Testing Program exist 
to ensure that the degree of containment structural integrity and 
leak-tightness that is considered in the plant safety analysis is 
maintained. The overall containment leak rate limit specified by TS 
is maintained.
    The proposed change involves only the extension of the interval 
between Type A containment leak rate tests and Type C tests for 
Calvert Cliffs Unit 1 and 2. The proposed surveillance interval 
extension is bounded by the 15 year ILRT Interval and the 75 month 
Type C test interval currently authorized within NEI [Nuclear Energy 
Institute] 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 XI 
and TS serve to provide a high degree of assurance that the 
containment would not degrade in a manner that is detectable only by 
Type A testing. The combination of these factors ensures that the 
margin of safety in the plant safety analysis is maintained. The 
design, operation, testing methods and acceptance criteria for Type 
A, B, and C containment leakage tests specified in applicable codes 
and standards would continue to be met, with the acceptance of this 
proposed change, since these are not affected by changes to the Type 
A and Type C test intervals.
    The proposed amendment also deletes exceptions previously 
granted to allow one time extensions of the ILRT test frequency for 
both Units 1 and 2 and exceptions from conducting post modification 
ILRT following replacement of the Units 1 and 2 Steam Generators. 
These exceptions were for things that have already taken place so 
their deletion is 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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200 
Exelon Way, Kennett Square, PA 19348.
    NRC Branch Chief: Benjamin G. Beasley.

Exelon Generation Company, LLC (EGC), Docket Nos. 50-220 and 50-410, 
Nine Mile Point Nuclear Station, Units 1 and 2 (NMP1 And NMP2), Oswego 
County, New York

    Date of amendment request: September 11, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14254A007.
    Description of amendment request: The proposed amendments would 
revise the licensed operator training and qualification education and 
experience eligibility requirements specified in NMP1 Technical 
Specification (TS) 6.3.1 and NMP2 TS 5.3.1 to the eligibility 
requirements specified in this License Amendment Request. The proposed 
eligibility requirements correspond to the eligibility requirements 
contained in the current National Academy for Nuclear Training (NANT) 
Academy Document, ACAD 10-001, ``Guidelines for Initial Training and 
Qualification of Licensed Operators,'' dated February 2010.
    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 NRC considered the impact of previously evaluated accidents 
during the rulemaking process, and by promulgation of the revised 10 
CFR 55 rule, determined that this impact remains acceptable when 
licensees have an accredited licensed operator training program 
which is based on a Systems Approach to Training (SAT). EGC 
maintains an Institute of Nuclear Power Operations (INPO) National 
Academy for Nuclear Training (NANT) accredited program which is 
based on a SAT. The NRC has concluded in Regulatory Information 
Summary (RIS) 2001-01, ``Eligibility of Operator License 
Applicants,'' and NUREG-1021, ``Operator Licensing Examination 
Standards For Power Reactors,'' that standards and guidelines 
applied by INPO in their accredited training programs are equivalent 
to those put forth by or endorsed by the NRC. Therefore, maintaining 
an INPO accredited SAT-based licensed operator training program is 
equivalent to maintaining an NRC approved licensed operator training 
program which conforms to applicable NRC Regulatory Guidelines or 
NRC endorsed industry standards. The proposed changes conform to 
NANT ACAD 10-001 licensed operator education and experience 
eligibility requirements.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 70216]]

    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 involves changes to the licensed operator 
training programs, which are administrative in nature. The EGC 
licensed operator training programs have been accredited by the 
National Nuclear Accrediting Board (NNAB) and are based on a SAT, 
which the NRC has previously found to be acceptable.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed TS changes are administrative in nature. The 
proposed TS changes do not affect plant design, hardware, system 
operation, or procedures for accident mitigation systems. The 
proposed changes do not significantly impact the performance or 
proficiency requirements for licensed operators. As a result, the 
ability of the plant to respond to and mitigate accidents is 
unchanged by the proposed TS 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: Donald P. Ferraro, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Suite 305, 
Kennett Square, PA 19348.
    NRC Branch Chief: Benjamin G. Beasley.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Perry, OH

    Date of amendment request: September 12, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14255A150.
    Description of amendment request: The proposed amendment modifies 
the Technical Specification (TS) definition of SHUTDOWN MARGIN (SDM) to 
require determination of SDM at the temperature that represents the 
most reactive state throughout the operating cycle. The proposed 
changes are intended to be consistent with the approved Technical 
Specification Task Force (TSTF) Traveler TSTF-535, Revision 0.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the definition of Shutdown Margin 
(SDM). SDM is not an initiator to any accident previously evaluated. 
Accordingly, the proposed change to the definition of SDM has no 
effect on the probability of any accident previously evaluated. SDM 
is an assumption in the analysis of some previously evaluated 
accidents and inadequate SDM could lead to an increase in 
consequences for those accidents. However, the proposed change 
revises the SDM definition to ensure that the correct SDM is 
determined for all BWR [boiling-water reactor] fuel types at all 
times during the fuel cycle.
    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 previously evaluated?
    Response: No.
    The proposed change revises the definition of SDM. The change 
does not involve a physical alteration of the plant that is, no new 
or different type of equipment will be installed or a change in the 
methods governing normal plant operations. The change does not alter 
assumptions made in the safety analysis regarding SDM.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises the definition of SDM. The proposed 
change does not alter the manner in which safety limits, limiting 
safety system settings or limiting conditions for operation are 
determined. The proposed change ensures that the SDM assumed in 
determining safety limits, limiting safety system settings or 
limiting conditions for operation is correct for all BWR fuel types 
at all times during the fuel cycle.
    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.
    NRC Branch Chief: Travis L. Tate.

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

    Date of amendment request: August 29, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14252A230.
    Description of amendment request: The amendments would modify the 
Technical Specifications (TSs) by removing TS 3/4.4.7, ``Chemistry,'' 
which provides limits on the oxygen, chloride, and fluoride content in 
the reactor coolant system to minimize corrosion. The licensee 
requested that these requirements be relocated to the Updated Final 
Safety Analysis Report (UFSAR) and related procedures and be controlled 
in accordance with 10 CFR 50.59, ``Changes, tests, and experiments.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration, 
which is presented as follows:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change acts to remove current Reactor Coolant 
System (RCS) chemistry limits and monitoring requirements from the 
TS and relocate the requirements to the UFSAR and related 
procedures. Monitoring and maintaining RCS chemistry minimizes the 
potential for corrosion of RCS piping and components. Corrosion 
effects are considered a long-term impact on RCS structural 
integrity. Because RCS chemistry will continue to be monitored and 
controlled, removing the current TS requirements and relocating the 
requirements to the UFSAR and related procedures will not present an 
adverse impact to the RCS and subsequently, will not impact the 
probability or consequences of an accident previously evaluated. 
Furthermore, once relocated to the UFSAR and related procedures, 
changes to RCS chemistry limits and monitoring requirements will be 
controlled in accordance with 10 CFR 50.59.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 acts to remove current Reactor Coolant 
System (RCS) chemistry limits and monitoring requirements from the 
TS and relocate the requirements to the UFSAR and related 
procedures. The proposed change does not introduce new modes of 
plant operation and

[[Page 70217]]

it does not involve physical modifications to the plant (no new or 
different type of equipment will be installed). There are no changes 
in the method by which any safety related plant structure, system, 
or component (SSC) performs its specified safety function. As such, 
the plant conditions for which the design basis accident analyses 
were performed remain valid.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of the proposed change. There will be no adverse effect or 
challenges imposed on any SSC as a result of the proposed change.
    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 
the margin of safety?
    Response: No.
    Margin of safety is related to confidence in the ability of the 
fission product barriers to perform their accident mitigation 
functions. The proposed change acts to remove current Reactor 
Coolant System (RCS) chemistry limits and monitoring requirements 
from the TS and relocate the requirements to the UFSAR and related 
procedures. The proposed change will maintain limits on RCS 
chemistry parameters and will continue to provide associated 
monitoring requirements. The proposed change does not physically 
alter any SSC. There will be no effect on those SSCs necessary to 
assure the accomplishment of protection functions. There will be no 
impact on the overpower limit, departure from nucleate boiling ratio 
(DNBR) limits, loss of cooling accident peak cladding temperature 
(LOCA PCT), or any other margin of safety. The applicable 
radiological dose consequence acceptance criteria will continue to 
be met.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station (SSES), Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: August 11, 2014. A publicly-available 
version is available in ADAMS under Accession No. ML14223A780.
    Description of amendment request: The amendment proposes changes to 
SSES, Units 1 and 2, Technical Specification (TS) 3.4.10, ``RCS 
[Reactor Coolant System] Pressure and Temperature (P/T) Limits,'' which 
includes revisions to the P/T Limits curves. The primary effect of the 
revision is to provide P/T Limits curves that extend into the vacuum 
region to mitigate the risk of a level transient during startup and 
shutdown.
    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 change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes request that the P/T limits curves in TS 
3.4.10, ``RCS Pressure and Temperature (P/T) Limits'' be revised by 
extending each of the P/T Limits curves below 0 psig to allow 
operation with the RPV [reactor pressure vessel] at a vacuum.
    The P/T curves are used as operational limits during heatup or 
cooldown maneuvering, when pressure and temperature indications are 
monitored and compared to the applicable curve to determine that 
operation is within the allowable region. The P/T curves provide 
assurance that station operation is consistent with previously 
evaluated accidents.
    Thus, the probability of an accident or the radiological 
consequences of an accident previously evaluated are not 
significantly increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not change the response of any plant 
equipment to transient conditions. The proposed changes do not 
introduce any new equipment, modes of system operation, or failure 
mechanisms.
    Therefore, there are no new types of failures or new or 
different kinds of accidents or transients that could be created by 
these changes. The proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The consequences of a previously evaluated accident are not 
increased by these proposed changes, since the Loss of Coolant 
Accident analyzed in the FSAR [Final Safety Analysis Report] assumes 
a complete break of the reactor coolant pressure boundary. The 
proposed changes to the P/T Limits curves do not change this 
assumption.
    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 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Meena K. Khanna.

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.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: March 18, 2014.

[[Page 70218]]

    Brief description of amendment: The amendment adopts Technical 
Specification (TS) Task Force (TSTF) change traveler TSTF-535, Revision 
0, ``Revise Shutdown Margin Definition to Address Advanced Fuel 
Designs,'' at Columbia Generating Station. The notice of availability 
of TSTF-535, Revision 0, was published in the Federal Register on 
February 26, 2013 (78 FR 13100).
    Date of issuance: November 12, 2014.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 228. A publicly-available version is in ADAMS under 
Accession No. ML14290A360; 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 Technical Specifications.
    Date of initial notice in Federal Register: July 22, 2014 (79 FR 
42544).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 12, 2014.
    No significant hazards consideration comments received: No.

Dominion Energy Kewaunee, Inc. (DEK), Docket No. 50-305, Kewaunee Power 
Station (KPS), Kewaunee County, Wisconsin

    Date of amendment request: January 16, 2014, as supplemented by 
letters dated June 19, 2014, and September 9, 2014.
    Brief description of amendment: The amendment authorizes revision 
to the Dominion Energy Kewaunee, Inc., emergency plan and emergency 
action level scheme to conform to the permanent shut down and defueled 
status of Kewaunee Power Station (KPS). The review considered the 
storage of spent nuclear fuel in the spent fuel pool and the 
independent spent fuel storage installation, and the low likelihood of 
any credible accident resulting in radiological releases requiring 
offsite protective measures. The U.S. Nuclear Regulatory Commission 
staff has concluded that the changes to the KPS emergency plan and 
emergency action level scheme would provide: (1) An adequate basis for 
an acceptable state of emergency preparedness, and (2) reasonable 
assurance that adequate protective measures can and will be taken in 
the event of a radiological emergency based on the permanently shut 
down and defueled status of the KPS facility.
    Date of issuance: October 31, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 214. A publicly-available version is in ADAMS under 
Accession No. ML14279A482; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-43: The amendment 
authorizes revision to the Dominion Energy Kewaunee, Inc., Renewed 
Facility Operating License emergency plan and emergency action level 
scheme.
    Date of initial notice in Federal Register: August 5, 2014 (79 FR 
45472). The supplemental letters dated June 19, 2014, and September 9, 
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 October 31, 2014.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: November 26, 2013, as supplemented by 
letters dated July 11, September 11, October 3, and October 16, 2014.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 4.3.4, ``Heavy Loads,'' by modifying the limit 
imposed on the maximum weight that could travel over the irradiated 
fuel in the spent fuel pool. The amendment also revised TS 4.3.4 to 
reflect the removal of the energy absorbing pad from the spent fuel 
pool and installation of a leveling platform.
    Date of issuance: October 31, 2014.
    Effective date: As of the date of issuance, and shall be 
implemented prior to the start of the dry cask storage operations.
    Amendment No.: 240. A publicly-available version is in ADAMS under 
Accession No. ML13346A026; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-35: Amendment revised 
the License and TSs.
    Date of notices in Federal Register: July 22, 2014 (79 FR 42545), 
as supplemented on September 22, 2014 (79 FR 56608). The supplement 
dated July 11, 2014, 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 (NSHC) determination as published in 
the Federal Register on July 22, 2014 (79 FR 42545).
    The supplement dated September 11, 2014, expanded the scope of the 
application as originally noticed and, therefore, the September 11, 
2014, supplement was published in the Federal Register on September 22, 
2014 (79 FR 56608). The supplements dated October 3 and October 16, 
2014, provided additional information that clarified the September 11, 
2014, supplement, did not expand the scope as noticed, and did not 
change the NRC staff's proposed NSHC determination as published in the 
Federal Register on September 22, 2014 (79 FR 56608).
    No significant hazards consideration comments received: Yes.
    The notice provided an opportunity to submit comments on the 
Commission's proposed NSHC determination. Several comments were 
received and evaluated.
    The Commission's related evaluation of the amendment and final NSHC 
determination, including the comments received, are contained in a 
safety evaluation dated October 31, 2014.

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: December 19, 2013, as supplemented by 
letter dated June 25, 2014.
    Description of amendment request: The amendment would change the 
Vermont Yankee Cyber Security Plan Implementation Schedule Milestone 8 
full implementation date from December 15, 2015, to June 30, 2016. The 
amendment would also revise the existing operating license Security 
Plan license condition.
    Date of Issuance: November 7, 2014.
    Effective date: As of the date of issuance, and shall be 
implemented by December 15, 2014.
    Amendment No.: 259. A publicly-available version is in ADAMS under 
Accession No. ML14206A710; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. DPR-28: Amendment revised the 
License.
    Date of initial notice in Federal Register: February 27, 2014 (79 
FR 11149). The supplement letter dated

[[Page 70219]]

June 25, 2014, 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 this amendment is contained 
in a Safety Evaluation dated November 7, 2014.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine 
Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York

    Date of application for amendment: October 7, 2013, as supplemented 
by letter dated June 18, 2014.
    Brief description of amendment: The amendment modified the Nine 
Mile Point Units 1 and 2 Technical Specifications (TS) definition of 
``Shutdown Margin'' (SDM) to require calculation of the SDM at a 
reactor moderator temperature of 68 [deg]F or a higher temperature that 
represents the most reactive state throughout the operating cycle. This 
change is needed to address new Boiling Water Reactor (BWR) fuel 
designs which may be more reactive at shutdown temperatures above 68 
[deg]F.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on November 19, 2012 (77 FR 69507), on possible 
amendments to revise the plant-specific TS, to modify the TS definition 
of ``Shutdown Margin'' (SDM) to require calculation of the SDM at a 
reactor moderator temperature of 68 [deg]F or a higher temperature that 
represents the most reactive state throughout the operating cycle, 
including a model safety evaluation and model no significant hazards 
consideration determination, using the consolidated line-item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on February 26, 2013, (78 FR 
13100).
    Date of issuance: October 30, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 1: 216, Unit 2: 146. A publicly-available 
version is in ADAMS under Accession No. ML14248A084; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License No. DPR-63 and NPF-69: Amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: November 12, 2013 (78 
FR 67411). The supplemental letter dated June 18, 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 October 30, 2014.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: October 4, 2013, as supplemented 
on April 29, 2014.
    Brief description of amendment: The amendment changes the Site 
Emergency Plan (SEP) for the Monticello Nuclear Generating Plant (MNGP) 
by eliminating the Radwaste Operator position as one of the 60-minute 
responders.
    Date of issuance: October 31, 2014.
    Effective date: This amendment is effective as of its date of 
issuance and shall be implemented within 60 days from the date of 
issuance.
    Amendment No.: 183. A publicly-available version is in the ADAMS 
under Accession No. ML14196A328; documents related to this amendment 
are listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. DPR-22: This amendment revised the 
MNGP SEP.
    Date of initial notice in Federal Register: July 8, 2014 (79 FR 
38591).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 31, 2014.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 4, 2014.
    Brief description of amendment: The amendment revised Technical 
Specification 3.5.1, ``ECCS [Emergency Core Cooling System]--
Operating,'' by removing Condition F. Condition F provides for a 72-
hour completion time to restore one core spray subsystem to an operable 
status when both core spray subsystems are inoperable. NSPM requested 
approval to remove the option of having a limiting condition of 
operation with both core spray subsystems inoperable based on an 
evaluation that at least one core spray subsystem is necessary to 
maintain adequate post-accident long-term core cooling.
    Date of issuance: November 3, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 184. A publicly-available version is in ADAMS under 
Accession No. ML14246A449; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-22: This amendment 
revised the Renewed Facility Operating License and the Technical 
Specifications.
    Date of initial notice in Federal Register: August 19, 2014 (79 FR 
49107).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 3, 2014.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 10, 2014, as supplemented by 
letter dated June 9, 2014.
    Brief description of amendment: The amendment revised Technical 
Specification Surveillance Requirement 3.2, ``Equipment and Sampling 
Tests,'' Table 3-5, ``Minimum Frequencies for Equipment Tests,'' Item 3 
for the pressurizer safety valves from a refueling frequency (i.e., 18 
months <plus-minus> 25 percent) to be consistent with the Inservice 
Testing Program, and made editorial changes to Table 3-5.
    Date of issuance: November 6, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 277. A publicly-available version is in ADAMS under 
Accession No. ML14279A275; 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 Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2014 (79 FR 
38592).

[[Page 70220]]

    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated November 6, 2014.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas, 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: March 13, September 25 (two letters for 
Licensing Amendment Request (LAR) 13-16 and LAR 13-17) and October 3, 
2013 (two letters for LAR 13-18 and LAR 13-19), as supplemented by 
letters dated October 3, 2013, and February 10 and June 6, 2014.
    Brief description of amendment: The proposed amendment involves 
changes to the five Human Factors Engineering (HFE) Reports 
(Westinghouse Electric Company and the U.S. Nuclear Regulatory 
Commission reviewed these reports as part of the AP1000 Design 
Certification Rule) that are incorporated by reference in the VCSNS 
Updated Final Safety Analysis Report. These are:

<bullet> HFE Design Verification Plan (APP-OCS-GEH-120) (LAR 13-16)
<bullet> HFE Task Support Verification Plan (APP-OCS-GEH-220) (LAR 13-
17)
<bullet> HFE Integrated System Validation (APP-OCS-GEH-320) (LAR 13-10)
<bullet> Human Engineering Discrepancy Resolution Process (APP-OCS-GEH-
420) (LAR 13-18)
<bullet> Plant Startup HFE Design Verification Plan (APP-OCS-GEH-520) 
(LAR 13-19)

    Date of issuance: July 31, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 16. A publicly-available version is in ADAMS under 
Accession No. ML14177A486; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: May 28, 2013 (78 FR 
31984 for LAR 13-10), November 12, 2013 (78 FR 67412 for LAR 13-16, 78 
FR 67411 for LAR 13-17, 78 FR 67413 for LAR 13-18, and 78 FR 67413 for 
LAR 13-19).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 31, 2014.
    No significant hazards consideration comments received: No.

South Carolina Electric and Gas, Inc. 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: July 17, 2013, as supplemented by 
letters dated July 8, and July 11, 2014.
    Brief description of amendment: The amendment revises Tier 2* and 
associated Tier 2 information, incorporated into the VCSNS Units 2 and 
3 Updated Final Safety Analysis Report (UFSAR). Specifically, the 
amendment revises the following information related to fire area 
boundaries: (1) Various Annex Building and Turbine Building layout 
changes, (2) Turbine Building Stairwell S08 changes to support egress 
functions, and (3) an Annex Building Heating, Ventilation and Air 
Conditioning shaft UFSAR figure clarification.
    Date of issuance: September 9, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 17. A publicly-available version is in ADAMS under 
Accession No. ML14218A687; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: October 1, 2013 (78 FR 
60321). The supplements dated July 8 and July 11, 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 September 9, 2014.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 29, 2014, as supplemented by 
letter dated May 27, 2014.
    Brief description of amendment: The amendments revised the Cyber 
Security Plan Implementation Milestone No. 8 completion date and the 
physical protection license condition.
    Date of issuance: September 29, 2014.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: Unit 1-286, Unit 2-312, and Unit 3-217. A publicly-
available version is in ADAMS under Accession No. ML14247A536; 
documents related to these amendments are listed in the Safety 
Evaluation (SE) enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68. 
The amendments revised the Operating Licenses.
    Date of initial notice in Federal Register: July 8, 2014 (79 FR 
38582). The supplemental letter dated May 27, 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 the SE dated September 29, 2014.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 17th day of November 2014.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2014-27630 Filed 11-24-14; 8:45 am]
BILLING CODE 7590-01-P


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