Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 64219-64232 [2014-25357]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 79, Number 208 (Tuesday, October 28, 2014)]
[Notices]
[Pages 64219-64232]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2014-25357]


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NUCLEAR REGULATORY COMMISSION

[NRC-2014-0239]


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 2, 2014 to October 15, 2014. The 
last biweekly notice was published on October 14, 2014.

DATES: Comments must be filed by November 28, 2014. A request for a 
hearing must be filed by December 29, 2014.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2014-0239. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: 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: Shirley Rohrer, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-5411, email: Shirley.Rohrer@nrc.gov

SUPPLEMENTARY INFORMATION: 

[[Page 64220]]

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2014-0239 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2014-0239.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section.
     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-2014-0239 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 https://www.regulations.gov as well as entering the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

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

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

[[Page 64221]]

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 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 hearing.docket@nrc.gov, 
or by telephone at 301-415-1677, to request (1) a digital 
identification (ID) certificate, which allows the participant (or its 
counsel or representative) to digitally sign documents and access the 
E-Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at https://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having

[[Page 64222]]

granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

Duke Energy Florida, Inc. et al. (DEF), Docket No. 50-302, Crystal 
River, Unit 3, Nuclear Generating Plant (CR-3), Citrus County, Florida

    Date of amendment request: October 29, 2013, as supplemented by 
letters dated May 7, 2014 and June 17, 2014. Publicly-available 
versions are in ADAMS under Accession Nos. ML13316C083, ML14139A006, 
and ML14178B284.
    Description of amendment request: The amendment would revise the 
CR-3 Facility Operating License (FOL) to remove and revise certain 
License Conditions. This amendment also proposes to extensively revise 
the CR-3 Improved Technical Specifications (ITS) in order to create the 
CR-3 Permanently Defueled Technical Specifications (PDTS).
    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.
    CR-3 has permanently ceased operation. The proposed amendment 
would modify the CR-3 FOL and ITS by proposing to delete certain 
License Conditions (LCs) and ITS that are no longer applicable to a 
permanently defueled facility, while modifying the remaining 
portions to correspond to the permanently shutdown condition. 
Changes proposed to LCs will make them consistent with the non-
operating status of CR-3. Other proposed LCs changes will eliminate 
LCs that were designed for one time implementation and have been 
satisfied, or are no longer required due to changes to Part 50 or 
Part 73 regulations that accomplish the same result or eliminate the 
requirement for the LC. The proposed changes to the ITS are 
consistent with the criteria set forth in 10 CFR 50.36 for the 
contents of ITS.
    Chapter 14 of the CR-3 Final Safety Analysis Report (FSAR) 
described the design basis accident (DBA) and transient scenarios 
applicable to CR-3 during power operations. With the reactor in a 
permanently defueled condition, the spent fuel pool and its cooling 
systems are dedicated only to spent fuel storage. In this condition, 
the spectrum of credible accidents is much smaller than for an 
operational plant. As a result of the certifications submitted by 
CR-3 in accordance with 10 CFR 50.82(a)(1), and the consequent 
removal of authorization to operate the reactor or to place or 
retain fuel in the reactor vessel in accordance with 10 CFR 
50.82(a)(2), the majority of the accident scenarios originally 
postulated in the FSAR are no longer possible and have been removed 
from the FSAR under 10 CFR 50.59.
    The definition of safety-related structures, systems, and 
components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are 
those relied on to remain functional during and following design 
basis events to assure:
    1. The integrity of the reactor coolant boundary;
    2. The capability to shutdown the reactor and maintain it in a 
safe shutdown condition; or
    3. The capability to prevent or mitigate the consequences of 
accidents which could result in potential offsite exposures 
comparable to the applicable guideline exposures set forth in 10 CFR 
50.34(a)(1) or 100.11.
    The first two criteria, integrity of the reactor coolant 
pressure boundary and safe shutdown of the reactor, are not 
applicable to a plant in a permanently defueled condition. The third 
criterion is related to preventing or mitigating the consequences of 
accidents that could result in potential offsite exposures exceeding 
limits. However, after the termination of reactor operations at CR-3 
and the permanent removal of the fuel from the reactor vessel 
(following 4 years of decay time after shutdown) and purging of the 
contents of the waste gas decay tanks, none of the SSCs at CR-3 are 
required to be relied on for accident mitigation. Therefore, none of 
the SSCs at CR-3 meet the definition of a safety-related SSC stated 
in 10 CFR 50.2 (with the exception of the passive spent fuel pool 
structure).
    The deletion of ITS definitions and rules of usage and 
application, that are currently not applicable in a defueled 
condition, has no impact on facility SSCs or the methods of 
operation of such SSCs. The deletion of design features and safety 
limits not applicable to the permanently shutdown and defueled 
status of CR-3 has no impact on the remaining DBA [design basis 
accidents] (the Fuel Handling Accident in the Auxiliary Building) or 
the proposed Radioactive Waste Handling Accident. The removal of 
LCOs [Limiting Conditions for Operation] or SRs [Surveillance 
Requirements] that are related only to the operation of the nuclear 
reactor or accidents do not affect mitigation of the applicable DBAs 
previously evaluated since these DBAs are no longer applicable in 
the defueled mode. The safety functions involving core reactivity 
control, reactor heat removal, reactor coolant system inventory 
control, and containment integrity are no longer applicable at CR-3 
as a permanently defueled plant. The analyzed accidents involving 
damage to the reactor coolant system, main steam lines, reactor 
core, and the subsequent release of radioactive material are no 
longer possible at CR-3
    Since CR-3 has permanently ceased operation, the generation of 
fission products has ceased and the remaining source term will 
decay. The radioactive decay of the irradiated fuel since shutdown 
of the reactor have reduced the consequences of the Fuel Handling 
Accident (FHA) to levels well below those previously analyzed. The 
relevant parameter (water level) associated with the fuel pool 
provides an initial condition for the FHA analysis and is included 
in the PDTS.
    The spent fuel pool water level, spent fuel pool boron 
concentration, and spent fuel pool storage LCOs are retained to 
preserve the current requirements for safe storage of irradiated 
fuel.
    Fuel pool cooling and makeup related equipment and support 
equipment (e.g., electrical power systems) are not required to be 
continuously available since there is sufficient time to effect 
repairs, establish alternate sources of makeup flow, or establish 
alternate sources of cooling in the event of a loss of cooling and 
makeup flow to the spent fuel pool.
    The deletion and modification of provisions of the 
Administrative Controls do not directly affect the design of SSCs 
necessary for the safe storage of irradiated fuel or the methods 
used for handling and

[[Page 64223]]

storage of such fuel in the fuel pool. Deletion of Programs are 
administrative in nature and do not affect any accidents applicable 
to the safe management of irradiated fuel or the permanently 
shutdown and defueled condition of the reactor.
    The proposed LC revisions reflect the CR-3 functions that are 
still authorized in the permanently defueled condition, and remove 
authorizations that suggest the reactor can be placed in operation. 
LCs that are being removed due to their one time applicability being 
previously satisfied have no bearing on future functions at CR-3. 
Other LCs are being removed that are not required by regulation for 
a permanently defueled and decommissioning plant. These changes 
cannot increase the probability or consequences of any accident that 
remains credible. The probability of occurrence of previously 
evaluated accidents is not increased, since extended operation in a 
defueled condition is the only operation currently allowed, and is 
therefore bounded by the existing analyses. Additionally, the 
occurrence of postulated accidents associated with reactor operation 
is no longer credible in a permanently defueled reactor. This 
significantly reduces the scope of applicable accidents.
    Therefore, the proposed amendment 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 changes have no impact on facility SSCs affecting 
the safe storage of irradiated fuel, or on the methods of operation 
of such SSCs, or on the handling and storage of irradiated fuel 
itself. The removal of ITS that are related only to the operation of 
the nuclear reactor or only to the prevention, diagnosis, or 
mitigation of reactor-related transients or accidents cannot result 
in different or more adverse failure modes or accidents than 
previously evaluated because the reactor is permanently shutdown and 
defueled, and CR-3 is no longer authorized to operate the reactor.
    The proposed deletion of requirements of the CR-3 ITS do not 
affect safe storage of nuclear fuel. The proposed PDTS continue to 
require proper control and monitoring of safety significant 
parameters. The proposed restriction on the fuel pool level is 
fulfilled by normal operating conditions and preserves initial 
conditions assumed in the analyses of the postulated DBA. The spent 
fuel pool water level, spent fuel pool boron concentration, and 
spent fuel pool storage LCOs are retained to preserve the current 
requirements for safe storage of irradiated fuel.
    The proposed amendment does not result in any new mechanisms 
that could initiate damage to the remaining relevant safety barriers 
for defueled plants (i.e., fuel cladding and spent fuel cooling). 
Since extended operation in a defueled condition is the only 
operation currently allowed, and therefore bounded by the existing 
analyses, such a condition does not create the possibility of a new 
or different kind of accident.
    Therefore, the proposed amendment 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.
    Because the 10 CFR Part 50 license for CR-3 no longer authorizes 
operation of the reactor or emplacement or retention of fuel into 
the reactor vessel, as specified in 10 CFR 50.82(a)(2), the 
occurrence of postulated accidents associated with reactor operation 
are no longer credible. The only remaining credible accident is a 
FHA. The proposed amendment does not adversely affect the inputs or 
assumptions of any of the design basis analyses that impact a FHA.
    The proposed changes are limited to those portions of the LCs 
and ITS that are not related to the safe storage of irradiated fuel. 
The requirements for SSCs that have been deleted from the CR-3 ITS 
are not credited in the existing accident analysis for the remaining 
applicable postulated accident; and as such, do not contribute to 
the margin of safety associated with the accident analysis. 
Postulated DBAs involving the reactor are no longer possible because 
the reactor is permanently shutdown and defueled and CR-3 is no 
longer authorized to operate the reactor.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety because the current design limits 
continue to be met for the accident of concern.

    The NRC staff has 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: Lara S. Nichols, 550 South Tryon Street, 
Charlotte NC 28202.
    NRC Branch Chief: Douglas A. Broaddus.

Entergy Nuclear Indian Point 3, LLC, Docket No. 50-286, Indian Point, 
Unit 3, Westchester County, NY

    Date of amendment request: April 1, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14099A333.
    Description of amendment request: The amendment would revise 
Technical Specifications (TS) Figure 3.4.3-1, Heatup Limitations for 
Reactor Coolant System, Figure 3.4.3-2, Cooldown Limitations for 
Reactor Coolant System, and Figure 3.4.3-3, Hydrostatic and Inservice 
Leak Testing Limitations for Reactor Coolant System, to indicate that 
the curves are applicable for vacuum fill.
    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. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated.
    The proposed TS changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
There are no physical changes to the plant being introduced by the 
proposed changes to the heatup, cooldown and hydrostatic inservice 
leak testing limitation curves. The proposed changes do not modify 
the RCS pressure boundary. That is, there are no changes in 
operating pressure, materials, or seismic loading. The proposed 
changes do not adversely affect the integrity of the RCS pressure 
boundary such that its function in the control of radiological 
consequences is affected. The heatup, cooldown and hydrostatic 
inservice leak testing limitation curves were established in 
compliance with the methodology used to calculate and predict 
effects of radiation on embrittlement of RPV beltline materials and 
remain valid during vacuum fill.
    Consequently, the proposed changes do not involve a significant 
increase in the probability or the consequences of an accident 
previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. No new modes of operation are introduced by the proposed 
changes. The proposed changes will not create any failure mode not 
bounded by previously evaluated accidents.
    Consequently, the proposed changes do not create the possibility 
of a new or different kind of accident, from any accident previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety.
    The proposed TS changes do not involve a significant reduction 
in the margin of safety. The changes clarify that the heatup, 
cooldown and hydrostatic inservice leak testing limitation curves 
remain valid during vacuum fill (to 0 psia) in accordance with 
current regulations. Because operation will be within these limits, 
the RCS materials will continue to behave in a non-brittle manner 
consistent with the original design bases.
    Therefore, Entergy has concluded that 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

[[Page 64224]]

proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Benjamin Beasley.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Units 1 and 2, Ogle County, Illinois

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

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

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

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: July 14, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14195A172.
    Description of amendment request: The proposed amendment would 
revise and add technical specification (TS) surveillance requirements 
to address the concerns discussed in NRC Generic Letter 2008-01, 
``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat 
Removal, and Containment Spray Systems,'' dated January 11, 2008 (ADAMS 
Accession No. ML072910759). The proposed TS changes are based on NRC-
approved TS Task Force (TSTF) Traveler TSTF-523, Revision 2, ``Generic 
Letter 2008-01, Managing Gas Accumulation,'' dated February 21, 2013 
(ADAMS Accession No. ML13053A075). The NRC staff issued a Notice of 
Availability for TSTF-523, Revision 2, for plant-specific adoption 
using the Consolidated Line Item Improvement Process, in the Federal 
Register on January 15, 2014 (79 FR 2700).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises or adds Surveillance Requirements 
(SRs) that require verification that the Emergency Core Cooling 
System (ECCS), the Decay Heat Removal (DHR)/Residual Heat Removal 
(RHR)/Shutdown Cooling (SDC) System, the Containment Spray (CS) 
System, and the Reactor Core Isolation Cooling (RCIC) System, as 
applicable, are not rendered inoperable due to accumulated gas and 
to provide allowances which permit performance of the revised 
verification. Gas accumulation in the subject systems is not an 
initiator of any accident previously evaluated. As a result, the 
probability of any accident previously evaluated is not 
significantly increased. The proposed SRs ensure that the subject 
systems continue to be capable to perform their assumed safety 
function and are not rendered inoperable due to gas accumulation. 
Thus, the consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, the DHR/RHR/SDC System, the CS System, 
and the RCIC System, as applicable, are not rendered inoperable due 
to accumulated gas and to provide allowances which permit 
performance of the revised verification. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation. In addition, the proposed 
change does not impose any new or different requirements that could 
initiate an accident. The proposed change does not alter assumptions 
made in the safety analysis and is consistent with the safety 
analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    2. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, the DHR/RHR/SDC System, the CS System, 
and the RCIC System, as applicable, are not rendered inoperable due 
to accumulated gas and to provide allowances which permit 
performance of the revised verification. The proposed change adds 
new requirements to manage gas accumulation in order to ensure the 
subject systems are capable of performing their assumed safety 
functions. The proposed SRs are more comprehensive than the current 
SRs and will ensure that the assumptions of the safety analysis are 
protected. The proposed change does not adversely affect any current 
plant safety margins or the reliability of the equipment assumed in 
the safety analysis. Therefore, there are no changes being made to 
any safety analysis assumptions, safety limits or limiting safety 
system settings that would adversely affect plant safety as a result 
of the proposed change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Esquire, Vice President 
and Deputy General Counsel, Exelon Generation Company, LLC, 4300 
Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Travis L. Tate.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station (BVPS), Units 1 and 2, Beaver 
County, Pennsylvania

    Date of amendment request: September 4, 2014. A publicly-available 
version is available in ADAMS under Accession No. ML14247A512.
    Description of amendment request: The amendment proposes changes to 
align the BVPS Emergency Planning Zone (EPZ) boundary with the boundary 
that is currently in use by the emergency management agencies of the 
three counties that implement public protective actions around BVPS.
    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.
    This amendment request would alter portions of the outer EPZ 
boundary defined in the BVPS EPP [Emergency Preparedness Plan] to 
align with the EPZ boundaries implemented by the Columbiana County, 
Hancock County, and Beaver County emergency management agencies. The 
proposed amendment does not involve any modifications or physical 
changes to plant systems, structures, or components. The proposed 
amendment does not change plant operations or maintenance of plant 
systems, structures, or components. Nor does the proposed amendment 
alter any BVPS EPP

[[Page 64225]]

facility or equipment. Changing the EPZ boundaries cannot increase 
the probability of an accident since emergency plan functions would 
be implemented after a postulated accident occurs. The proposed 
amendment does not alter or prevent the ability of the BVPS 
emergency response organization to perform intended emergency plan 
functions to mitigate the consequences of and to respond adequately 
to radiological emergencies.
    Therefore, the proposed amendment 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.
    This amendment request alters the EPZ boundary described in the 
BVPS EPP. The proposed amendment does not involve any design 
modifications or physical changes to the plant, does not change 
plant operation or maintenance of equipment, and does not alter BVPS 
EPP facilities or equipment. The proposed amendment to the BVPS EPP 
does not alter any BVPS emergency actions that would be implemented 
in response to postulated accident events. The proposed amendment 
does not create any credible new failure mechanisms, malfunctions, 
or accident initiators not previously considered.
    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.
    This amendment request would alter portions of the EPZ boundary 
defined in the BVPS EPP. The proposed amendment does not involve any 
design or licensing basis functions of the plant, no physical 
changes to the plant are made, does not impact plant operation or 
maintenance of equipment, and does not alter BVPS EPP facilities or 
equipment. This change does not alter any BVPS emergency actions 
that would be implemented in response to postulated accident events. 
The BVPS EPP continues to meet 10 CFR 50.47 and 10 CFR 50, Appendix 
E requirements for emergency response.
    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, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, 
Akron, OH 44308.
    NRC Branch Chief: Meena K. Khanna.

Florida Power and Light Company, et al. (the licensee), Docket Nos. 50-
335 and 50-389, St. Lucie Plant, Units 1 and 2, St. Lucie County, 
Florida

    Date of amendment request: July 14, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14198A074.
    Description of amendment request: The amendments would revise or 
add surveillance requirements (SRs) to verify that the system locations 
susceptible to gas accumulation are sufficiently filled with water and 
to provide allowances that permit performance of the verification. The 
licensee proposed the changes to address NRC Generic Letter 2008-01, 
``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat 
Removal, and Containment Spray Systems'' (ADAMS Accession No. 
ML072910759), as described in Revision 2 of Technical Specification 
Task Force No. 523, ``Generic Letter 2008-01, Managing Gas 
Accumulation'' (ADAMS Accession No. ML13053A075).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the Proposed Change Involve a Significant Increase in 
the Probability or Consequences of an Accident Previously Evaluated?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the Emergency Core Cooling Systems (ECCS), 
Residual Heat Removal (RHR) System, Shutdown Cooling (SDC) System, 
and Containment Spray (CS) System are not rendered inoperable due to 
accumulated gas and to provide allowances which permit performance 
of the revised verification. Gas accumulation in the subject systems 
is not an initiator of any accident previously evaluated. As a 
result, the probability of any accident previously evaluated is not 
significantly increased. The proposed SRs ensure that the subject 
systems continue to be capable of performing their assumed safety 
function and are not rendered inoperable due to gas accumulation. 
Thus, the consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the Proposed Change Create the Possibility of a New or 
Different Kind of Accident from any Accident Previously Evaluated?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, RHR System, SDC System, and CS System 
are not rendered inoperable due to accumulated gas and to provide 
allowances which permit performance of the revised verification. The 
proposed change does not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. In addition, 
the proposed change does not impose any new or different 
requirements that could initiate an accident. The proposed change 
does not alter assumptions made in the safety analysis and is 
consistent with the safety analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the Proposed Change Involve a Significant Reduction in a 
Margin of Safety?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, RHR System, SDC System, and CS System 
are not rendered inoperable due to accumulated gas and to provide 
allowances which permit performance of the revised verification. The 
proposed change adds new requirements to manage gas accumulation in 
order to ensure that the subject systems are capable of performing 
their assumed safety functions. The proposed SRs are more 
comprehensive than the current SRs and will ensure that the 
assumptions of the safety analysis are protected. The proposed 
change does not adversely affect any current plant safety margins or 
the reliability of the equipment assumed in the safety analysis. 
Therefore, there are no changes being made to any safety analysis 
assumptions, safety limits, or limiting safety system settings that 
would adversely affect plant safety as a result of the proposed 
change.

    The NRC staff has 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, 700 Universe Blvd., MS LAW/JB, Juno 
Beach, Florida 33408-0420.
    NRC Acting Branch Chief: Lisa M. Regner.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida

    Date of amendment request: August 8, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14225A654.
    Description of amendment request: The amendments would modify the 
Technical Specifications 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.

[[Page 64226]]

The licensee requested that these requirements be relocated to the 
Updated Final Safety Analysis Report (UFSAR) and 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. 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, relocating the current TS 
requirements to the UFSAR 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, 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. The proposed change 
does not introduce new modes of plant operation and 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. 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.
    NRC Acting Branch Chief: Lisa M. Regner.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Nuclear Power Plant (CPNPP), Units 1 and 2, Somervell 
County, Texas

    Date of amendment request: July 1, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14192A338.
    Description of amendment request. The amendments would revise 
Technical Specification (TS) 3.8.1, ``AC [Alternating Current] 
Sources--Operating,'' to extend, on a one-time basis, the Completion 
Time (CT) of Required Action A.3 from 72 hours to 14 days. By letter 
dated September 18, 2013 (ADAMS Accession No. ML13232A143), the NRC 
staff issued Amendment No. 160 to Facility Operating License No. NPF-87 
and Amendment No. 160 to Facility Operating License No. NPF-89 for 
CPNPP, Units 1 and 2, respectively. The amendments revised TS 3.8.1 to 
extend the CT for Required Action A.3 on a one-time basis from 72 hours 
to 14 days. The CT extension from 72 hours to 14 days was to be used 
twice while completing the plant modification to install alternate 
startup transformer (ST) XST1A and was to expire on March 31, 2014.
    The first 14-day CT was successfully completed, on October 14, 
2013. However, the licensee inadvertently cut the wrong offsite power 
cable during the second 14-day CT resulting in a total loss-of-offsite 
power (LOOP) to both units and the modification had to be abandoned. 
Due to the cut-cable event and the subsequent efforts to determine the 
causes and corrective actions, the modification could not be completed 
by March 31, 2014. The licensee has requested an extension of the CT 
from 72 hours to 14 days on one-time basis to complete the plant 
modification. Installation of the alternate ST XST1A will result in 
improved offsite power system reliability.
    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 will revise the CT for the loss of one 
offsite source from 72 hours to 14 days to allow a one-time, 14-day 
CT. The proposed one-time extension of the CT for the loss of one 
offsite power circuit does not significantly increase the 
probability of an accident previously evaluated. The TS will 
continue to require equipment that will power safety related 
equipment necessary to perform any required safety function. The 
one-time extension of the CT to 14 days does not affect the design 
of the STs, the interface of the STs with other plant systems, the 
operating characteristic of the STs, or the reliability of the STs.
    The consequence of a LOOP event has been evaluated in the CPNPP 
Final Safety Analysis Report (Reference 8.1 [of the licensee's 
letter dated July 1, 2014]) and the Station Blackout evaluation. 
Increasing the CT for one offsite power source on a one-time basis 
from 72 hours to 14 days does not increase the consequences of a 
LOOP event nor change the evaluation of LOOP events.
    Therefore, the proposed change does 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 change does not result in a change in the manner in 
which the electrical distribution subsystems provide plant 
protection. The proposed change will only affect the time allowed to 
restore the operability of the offsite power source through a ST. 
The proposed change does not affect the configuration, or operation 
of the

[[Page 64227]]

plant. The proposed change to the CT will facilitate installation of 
a plant modification which will improve plant design and will 
eliminate the necessity to shut down both Units if XST1 fails or 
requires maintenance that goes beyond the current TS CT of 72 hours. 
This change will improve the long-term reliability of the 138 
[kiloVolt (kV)] offsite circuit ST which is common to both CPNPP 
Units.
    There are no changes to the STs or the supporting systems 
operating characteristics or conditions. The change to the CT does 
not change any existing accident scenarios, nor create any new or 
different accident scenarios. In addition, the change does not 
impose any new or different requirements or eliminate any existing 
requirements. The change does not alter any of the assumptions made 
in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not affect the acceptance criteria for 
any analyzed event nor is there a change to any safety limit. The 
proposed change does not alter the manner in which safety limits, 
limiting safety system settings, or limiting conditions for 
operation are determined. Neither the safety analyses nor the safety 
analysis acceptance criteria are affected by this change. The 
proposed change will not result in plant operation in a 
configuration outside the current design basis. The proposed 
activity only increases a one-time pre-planned occurrence, the 
period when the plant may operate with one offsite power source. The 
margin of safety is maintained by maintaining the ability to safely 
shut down the plant and remove residual heat.
    Therefore, the proposed change does not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Timothy P. Matthews, Esq., Morgan, Lewis and 
Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Branch Chief: Michael T. Markley.

Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, 
Goodhue County, Minnesota

    Date of amendment request: August 21, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14233A431.
    Description of amendment request: The proposed amendments would 
revise the PINGP, Units 1 and 2, licensing basis analysis for waste gas 
decay tank rupture.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment request proposes to revise the licensing 
basis waste gas decay tank rupture analysis. The proposed analysis 
was updated to include the current fuel type, current fuel cycle 
lengths and plant operation to sixty years.
    The proposed waste gas decay tank rupture analysis changes are 
not accident initiators, and therefore the proposed changes do not 
involve an increase in the probability of an accident.
    The original waste gas decay tank rupture analysis demonstrated 
that the doses were a small fraction of the regulatory guidelines 
and that the waste gas system design prevents release of undue 
amounts of radioactivity. The revised waste gas decay tank rupture 
analysis demonstrates that the doses are well within the regulatory 
guidelines and that the waste gas system design continues to prevent 
release of undue amounts of radioactivity, and thus the proposed 
changes do not involve a significant increase in the consequences of 
an accident.
    Therefore, the proposed licensing basis 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.
    This license amendment request proposes to revise the licensing 
basis waste gas decay tank rupture analysis. The proposed analysis 
was updated to include the current fuel type, current fuel cycle 
lengths and plant operation to sixty years.
    The proposed waste gas decay tank rupture analysis includes 
plant changes that have previously been evaluated. This analysis 
applies the same methodology as the previous analysis. The proposed 
revision to the waste gas decay tank rupture analysis does not 
change any system operations or maintenance activities. The changes 
do not involve physical alteration of the plant; that is, no new or 
different type of equipment will be installed. These changes do not 
create new failure modes or mechanisms which are not identifiable 
during testing and no new accident precursors are generated.
    Therefore, the proposed licensing basis 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.
    This license amendment request proposes to revise the licensing 
basis waste gas decay tank rupture analysis. The proposed analysis 
was updated to include the current fuel type, current fuel cycle 
lengths and plant operation to sixty years.
    This revised analysis applies the same methodology as the 
original waste gas decay tank rupture analysis. The original waste 
gas decay tank rupture analysis demonstrated that the doses were a 
small fraction of the regulatory guidelines and that the waste gas 
system design prevents release of undue amounts of radioactivity. 
The revised waste gas decay tank rupture analysis demonstrates that 
the doses are well within the regulatory guidelines and that the 
waste gas system design continues to prevent release of undue 
amounts of radioactivity.
    Therefore, the proposed licensing basis change does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401
    NRC Branch Chief: David L. Pelton.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station (SNGS), Units 1 and 2, Salem County, New Jersey

    Date of amendment request: July 28, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14210A484.
    Description of amendment request: The proposed amendment would 
modify technical specification (TS) requirements regarding steam 
generator tube inspections and reporting as described in Technical 
Specification Task Force (TSTF) traveler TSTF-510, Revision 2, 
``Revision to Steam Generator Program Inspection Frequencies and Tube 
Sample Selection.'' In addition, the proposed amendment would revise 
the SNGS, Unit 2, TSs 6.8.4.i, ``Steam Generator (SG) Program,'' TS 
6.9.1.10, ``Steam Generator Tube Inspection Report,'' and the bases 
section of 3/4.4.6, ``Steam Generator (SG) Tube Integrity,'' to remove 
unnecessary information related to the original Salem Unit 2 
Westinghouse steam generators.
    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

[[Page 64228]]

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 change revises the Steam Generator (SG) Program to 
modify the frequency of verification of SG tube integrity and SG 
tube sample selection. A steam generator tube rupture (SGTR) event 
is one of the design basis accidents that are analyzed as part of a 
plant's licensing basis. The proposed SG tube inspection frequency 
and sample selection criteria will continue to ensure that the SG 
tubes are inspected such that the probability of a SGTR is not 
increased. The consequences of a SGTR are bounded by the 
conservative assumptions in the design basis accident analysis. The 
proposed change will not cause the consequences of a SGTR to exceed 
those assumptions.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes to the Salem Unit 2 Technical 
Specifications (TS) that are not associated with TSTF-510, removing 
unnecessary information related to W* [pronounced ``W star,'' which 
refers to the length of the steam generator tube required to be 
inspected within the hot-leg tube sheet] that is only applicable to 
Westinghouse steam generators, is an administrative change that 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 changes to the Steam Generator Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
proposed change does not affect the design of the SGs or their 
method of operation. In addition, the proposed change does not 
impact any other plant system or component.
    The proposed changes to the Salem Unit 2 Technical 
Specifications (TS) that are not associated with TSTF-510, removing 
unnecessary information related to W* that is only applicable to 
Westinghouse steam generators, is an administrative change that does 
not affect the design of the SGs or their method of operation.
    Therefore, it is concluded that these changes do not create the 
possibility of a new of different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change will continue to require monitoring of the physical 
condition of the SG tubes such that there will not be a reduction in 
the margin of safety compared to the current requirements.
    The proposed changes to the Salem Unit 2 Technical 
Specifications (TS) that are not associated with TSTF-510, removing 
unnecessary information related to W* that is only applicable to 
Westinghouse steam generators, is an administrative change that does 
not involve a significant reduction in a margin of safety.
    Therefore, it is concluded that 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: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Meena K. Khanna.

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: June 20, 2014, as supplemented by letter 
dated August 6, 2014. Publicly-available versions are in ADAMS under 
Accession Nos. ML14174B176 and ML14218A809.
    Description of amendment request: The proposed license amendment 
request (LAR) would revise the Updated Final Safety Analysis Report 
(UFSAR) in regard to Tier 2* information related to fire area 
boundaries. These changes add three new fire zones in the middle 
annulus to provide enclosures for the Class 1E electrical containment 
penetrations in accordance with UFSAR Appendix 9A, Subsection 
9A.3.1.1.15. The addition of the three new fire zones extended the fire 
area boundaries for three existing fire areas and, therefore, 
constitutes a change to Tier 2* information. Additionally, the licensee 
proposed changes that require revisions to UFSAR Tier 2 information 
involving changes to plant-specific Tier 2* information.
    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 middle annulus fire barrier reconfiguration for the 
electrical penetrations would not adversely affect any safety-
related equipment or function. The modified configuration for the 
Class 1E electrical containment penetration enclosures will maintain 
the fire protection function (i.e., barrier) as evaluated in Updated 
Final Safety Analysis Report (UFSAR), thus, the probability of a 
Class 1E electrical containment penetration failure is not 
significantly increased. The safe shutdown fire analysis is not 
affected, and the fire protection analysis results are not adversely 
affected. The proposed changes do not involve any accident, 
initiating event or component failure; thus, the probabilities of 
previously evaluated accidents 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 addition of enclosures constructed of three-hour rated fire 
barriers to separate the fire zones in the middle annulus for the 
Class 1E electrical penetration assemblies will maintain the fire 
protection function as evaluated in the UFSAR. The addition of the 
fire barriers does not affect the function of the Class 1E 
electrical containment penetrations or electrical penetration 
assemblies, and thus, does not introduce a new failure mode. The 
addition of the fire barriers 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.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The use of enclosures constructed of three-hour rated fire 
barriers to separate the fire zones in the middle annulus for the 
Class 1E electrical penetration assemblies will maintain the fire 
protection function as evaluated in the UFSAR. The use of the fire 
barriers does not affect the ability of the Class

[[Page 64229]]

1E electrical containment penetrations, electrical penetration 
assemblies, or the containment to perform their design function. The 
Class 1E electrical containment penetrations and electrical 
penetration assemblies within the enclosures continue to comply with 
the existing design codes and regulatory criteria, and do not affect 
any safety limit. The use of fire barriers and enclosures to 
separate the Class 1E electrical penetration assemblies does not 
adversely affect any margin of safety.
    Therefore, the proposed amendment does not involve a significant 
reduction in the 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.

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

    Date of amendment request: May 15, 2014, as supplemented by letter 
dated July 10, 2014. Publicly-available versions are in ADAMS under 
Accession Nos. ML14164A341 and ML14282A185, respectively.
    Description of amendment request: The amendment would update the 
Emergency Action Levels (EALs) used at STP, Units 1 and 2 from the 
current scheme based on Nuclear Management and Resources Council, Inc. 
(NUMARC)/Nuclear Environmental Studies Project (NESP) report NUMARC/
NESP-007, Revision 2, ``Methodology for Development of Emergency Action 
Levels,'' dated January 1992 (ADAMS Accession No. ML041120174), to the 
NRC-endorsed scheme contained in Nuclear Energy Institute (NEI) 99-01, 
Revision 6, ``Development of Emergency Action Levels for Non-Passive 
Reactors,'' dated November 2012 (ADAMS Accession No. ML12326A805). The 
EAL scheme in NEI 99-01, Revision 6 includes an EAL for Independent 
Spent Fuel Storage Installations (ISFSI), which is needed in order to 
implement dry cask storage operations at STP Units 1 and 2. 
Additionally, there are three EALs that require Spent Fuel Pool level 
instrument values which are designed to address lessons learned from 
the Fukushima Dai-ichi accident.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    Response: No.
    The change revises the STPNOC [STP Nuclear Operating Company] 
Emergency Action Levels to be consistent with the NRC endorsed EAL 
scheme contained in NEI 99-01, Revision 6, ``Methodology for 
Development of Emergency Action Levels,'' but does not alter any of 
the requirements of the Operating License or the Technical 
Specifications. In addition to replacing the current STP EALs, the 
new EAL scheme includes an EAL related to the planned STP 
Independent Spent Fuel Storage Installation, and EALs related to 
planned changes to the Spent Fuel Pool level instrumentation that 
will address lessons learned from Fukushima Daiichi. The proposed 
change does not modify any plant equipment and does not impact any 
failure modes that could lead to an accident. Additionally, the 
proposed change has no effect on the consequences of any analyzed 
accident since the change does not affect any equipment related to 
accident mitigation. Based on this discussion, the proposed 
amendment does not increase the probability or consequence of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The change revises the STPNOC Emergency Action Levels to be 
consistent with the NRC endorsed EAL scheme contained in NEI 99-01, 
Revision 6, ``Methodology for Development of Emergency Action 
Levels,'' but does not alter any of the requirements of the 
Operating License or the Technical Specifications. In addition to 
replacing the current STP EALs, the new EAL scheme includes an EAL 
related to the planned STP Independent Spent Fuel Storage 
Installation, and EALs related to planned changes to the Spent Fuel 
Pool level instrumentation that will address lessons learned from 
Fukushima Daiichi. The proposed change does not modify any plant 
equipment and there is no impact on the capability of the existing 
equipment to perform their intended functions. No system setpoints 
are being modified. No new failure modes are introduced by the 
proposed change. The proposed amendment does not introduce any 
accident initiators or malfunctions that would cause a new or 
different kind of accident. 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 change revises the STPNOC Emergency Action Levels to be 
consistent with the NRC endorsed EAL scheme contained in NEI 99-01, 
Revision 6, ``Methodology for Development of Emergency Action 
Levels,'' but does not alter any of the requirements of the 
Operating License or the Technical Specifications. In addition to 
replacing the current STP EALs, the new EAL scheme includes an EAL 
related to the planned STP Independent Spent Fuel Storage 
Installation, and EALs related to planned changes to the Spent Fuel 
Pool level instrumentation that will address lessons learned from 
Fukushima Daiichi. The proposed change does not affect any of the 
assumptions used in the accident analysis, nor does it affect any 
operability requirements for equipment important to plant safety. 
Therefore, the proposed change will not result in a significant 
reduction in the margin of safety in operation of the facility as 
discussed in this license amendment request.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Branch Chief: Michael T. Markley.

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

[[Page 64230]]

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.

Duke Energy Progress, Inc., Docket No. 50-261, H. B. Robinson Steam 
Electric Plant (HBRSEP), Unit 2, Darlington County, South Carolina

    Date of application for amendment: September 30, 2013, as 
supplemented by letter dated August 6, 2014.
    Brief description of amendment: The amendment implements the NRC-
approved Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-491, ``Removal of Main Steam and 
Main Feedwater Valve Isolation Times from Technical Specifications,'' 
via the Consolidated Line Item Improvement Process. This amendment 
modifies the current Technical Specifications (TSs) 3.7.2, Main Steam 
Isolation Valves and 3.7.3, Main Feedwater Isolation Valves, Main 
Feedwater Regulation Valves and Bypass Valves by relocating the 
specific isolation time for the isolation valves from the associated 
Surveillance Requirements (SRs). The isolation time in the TS SRs is 
replaced with the requirement to verify the valve isolation time is 
``within limits.'' The specific isolation times will be maintained in 
the HBRSEP, Unit 2, Technical Requirements Manual.
    Date of issuance: October 10, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 237. A publicly-available version is in ADAMS under 
Accession No. ML14252A221; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-23. Amendment revised 
the Facility Operating License and TSs.
    Date of initial notice in Federal Register: December 10, 2013 (78 
FR 74180). The supplemental letter dated August 6, 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 10, 2014.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: May 1, 2014, as supplemented by letter 
dated August 21, 2014.
    Brief description of amendment: The amendment revises Technical 
Specification 2.0, ``Safety Limits (SLs),'' by changing the safety 
limit minimum critical power ratio for both single and dual 
recirculation loop operation.
    Date of issuance: September 30, 2014.
    Effective date: As of the date of issuance, and shall be 
implemented within [30] days.
    Amendment No.: 307. A publicly-available version is in ADAMS under 
Accession No. ML14258B201; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-59: The amendment 
revised the License and the Technical Specifications.
    Date of initial notice in Federal Register: August 5, 2014 (79 FR 
45487). The supplemental letter dated August 21, 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.
    The Commission's related evaluation of the amendment and final NSHC 
determination is contained in a Safety Evaluation dated September 30, 
2014.
    No significant hazards consideration comments received. No.

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

    Date of amendment request: September 20, 2013, as supplemented by 
letter dated June 30, 2014.
    Brief description of amendment: The amendment revised the LSCS, 
Units 1 and 2, allowable values for the loss of voltage relay voltage 
setpoints in Technical Specification Table 3.3.8.1-1, ``Loss of Power 
Instrumentation.''
    Date of issuance: September 29, 2014.
    Effective date: As of the date of issuance. For LSCS Unit 1, the 
amendment shall be implemented prior to entering MODE 4 following the 
spring 2016 refueling outage (L1R16). For LSCS, Unit 2, the amendment 
shall be implemented prior to entering MODE 4 following the spring 2015 
refueling outage (L2R15).
    Amendment No.: 209 and 196. A publicly-available version is in 
ADAMS under Accession No. ML14252A913; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendment.
    Facility Operating License Nos. NPF-11 and NPF-18: Amendment 
revised the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2013 (78 
FR 74182). The supplemental letter dated June 30, 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 29, 2014.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit (TMI-1), Dauphin County, Pennsylvania

    Date of application for amendment: May 7, 2014.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Surveillance Requirements (SRs) 4.12.1, ``Emergency 
Control Room Air Treatment System,'' and 4.12.4, ``Fuel Handling 
Building [Engineered Safety Feature] ESF Air Treatment System.'' The 
amendment revised the TSs to replace the existing SRs to operate 
ventilation systems with charcoal filters for a 10-hour period at a 
frequency controlled in accordance with the Surveillance Frequency 
Control Program (SFCP) with a requirement to operate the systems for 
greater than or equal to 15 continuous minutes at a frequency 
controlled in accordance with the SFCP. These changes are consistent 
with Technical Specification Task Force (TSTF) Traveler TSTF-522, 
Revision 0, ``Revise Ventilation System Surveillance Requirements to 
Operate for 10 hours per Month.''
    Date of issuance: October 14, 2014.

[[Page 64231]]

    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 282. A publicly-available version is in ADAMS under 
Accession No. ML14240A348; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-50: Amendment revised 
the license and the technical specifications.
    Date of initial notice in Federal Register: August 5, 2014 (79 FR 
45476).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 14, 2014.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant (CNP), Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: April 9, 2014, as supplemented by 
letter dated August 15, 2014.
    Brief description of amendments: The amendments revised the CNP 
Technical Specifications (TSs) 3.4.3, ``[Reactor Coolant System] RCS 
Pressure and Temperature Limits.'' The changes to TSs clarify that 
pressure limits are considered to be met for pressures that are below 0 
psig (i.e., up to and including full vacuum conditions). Vacuum fill 
operations for the RCS can result in system pressures below 0 psig.
    Date of issuance: October 1, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 323 (Unit 1) and 306 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML14259A549; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revise the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2014 (79 FR 
38591). The supplemental letter dated 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 October 1, 2014.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 18, 2014, as supplemented by the 
letter dated July 30, 2014.
    Brief description of amendment: The license amendment revises the 
Updated Final Safety Analysis Report (UFSAR) in regard to Tier 2* 
information related to fire area boundaries. These changes add three 
new fire zones in the middle annulus to provide enclosures for the 
Class 1E electrical containment penetrations in accordance with UFSAR 
Appendix 9A, Subsection 9A.3.1.1.15. Additionally, the license 
amendment revises UFSAR Tier 2 information involving changes to plant-
specific Tier 2* information.
    Date of issuance: October 8, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 25. A publicly-available version is in ADAMS under 
Accession No. ML14248A243; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: May 27, 2014 (79 FR 
30189).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 8, 2014.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 27, 2014, as supplemented by the 
letter dated July 23, 2014.
    Brief description of amendment: The amendment revises the VEGP 
Units 3 and 4 Emergency Plan and changes the combined licenses (COL), 
Appendix C, plant-specific emergency planning inspections, tests, 
analyses, and acceptance criteria (ITAAC) to reflect the relocation of 
the Operations Support Centers and changes the description of the plant 
monitoring system.
    Date of issuance: October 7, 2014
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 24. A publicly-available version is in ADAMS under 
Accession No. ML14245A075; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: May 13, 2014 (79 FR 
27345).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 7, 2014.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: April 29, 2014. A redacted version was 
provided 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 60 days.
    Amendment Nos.: Unit 1-333 and Unit 2-326. A publicly-available 
version is in ADAMS under Accession No. ML14245A179; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. DPR-77 and DPR-79. The amendments 
revised the Operating License.
    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 a Safety Evaluation dated September 29, 2014.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: April 29, 2014. A redacted version was 
provided 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.

[[Page 64232]]

    Date of issuance: September 29, 2014.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: Unit 1-333 and Unit 2-326. A publicly-available 
version is in ADAMS under Accession No. ML14245A179; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. DPR-77 and DPR-79. The amendments 
revised the Operating License.
    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 a Safety Evaluation dated September 29, 2014.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 29, 2014, as supplemented by 
letter dated May 27, 2014.
    Brief description of amendment: The amendment 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 30 days of issuance.
    Amendment No.: 97. A publicly-available version is in ADAMS under 
Accession No. ML14255A152; documents related to this amendment are 
listed in the Safety Evaluation (SE) enclosed with the amendment.
    Facility Operating License No. NPF-90. Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: July 8, 2014 (79 FR 
38581). 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 16th day of October 2014.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
 Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2014-25357 Filed 10-27-14; 8:45 am]
BILLING CODE 7590-01-P