[Federal Register Volume 79, Number 189 (Tuesday, September 30, 2014)]
[Notices]
[Pages 58812-58831]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-23015]


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NUCLEAR REGULATORY COMMISSION

[NRC-2014-0207]


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 September 4, 2014 to September 17, 2014. The 
last biweekly notice was published on September 16, 2014.

DATES: Comments must be filed by October 30, 2014. A request for a 
hearing must be filed by December 1, 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 http://www.regulations.gov and search for Docket ID NRC-2014-0207. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected]. 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: Mable Henderson, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-3760, email: [email protected].

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2014-0207 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0207.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section.
     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-2014-0207 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into 
ADAMS. 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.

[[Page 58813]]

    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example, in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.

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

    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Agency 
Rules of Practice and Procedure'' in 10 CFR Part 2. Interested 
person(s) should consult a current copy of 10 CFR 2.309, which is 
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The 
NRC's regulations are accessible electronically from the NRC Library on 
the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed by the above date, the Commission or a presiding officer 
designated by the Commission or by the Chief Administrative Judge of 
the Atomic Safety and Licensing Board Panel, will rule on the request 
and/or petition; and the Secretary or the Chief Administrative Judge of 
the Atomic Safety and Licensing Board will issue a notice of a hearing 
or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment unless the 
Commission finds an imminent danger to the health or safety of the 
public, in which case it will issue an appropriate order or rule under 
10 CFR Part 2.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [email protected], or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://

[[Page 58814]]

www.nrc.gov/site-help/e-submittals/getting-started.html. System 
requirements for accessing the E-Submittal server are detailed in the 
NRC's ``Guidance for Electronic Submission,'' which is available on the 
agency's public Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may attempt to use other software not 
listed on the Web site, but should note that the NRC's E-Filing system 
does not support unlisted software, and the NRC Meta System Help Desk 
will not be able to offer assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with the NRC's guidance 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at http://www.nrc.gov/site-help/e-submittals.html, by email to 
[email protected], or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant (PNP), Van Buren County, Michigan
    Date of amendment request: July 29, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14211A520.
    Description of amendment request: The amendment would change the 
Operating License at PNP. Specifically, the amendment requests 
authorization to implement 10 CFR 50.61a, ``Alternate fracture 
toughness requirements for protection against pressurized thermal shock 
events,'' in lieu of 10 CFR 50.61, ``Fracture toughness requirements 
for protection against pressurized thermal shock events.'' PNP 
currently complies with 10 CFR 50.61. The 10 CFR 50.61 screening 
criteria define a limiting level of embrittlement beyond which plant 
operation cannot continue without further evaluation. As described in 
NUREG-1806, ``Technical Basis for Revision of the Pressurized Thermal 
Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61),'' August 
2007, the screening criteria in the PTS rule is overly conservative and 
the risk of through-wall cracking due to a PTS event is much lower than 
previously estimated. A publically-available version of NUREG-1806 is 
in ADAMS under Accession No. ML072830074.
    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

[[Page 58815]]

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.
    This amendment request would allow implementation of the 10 CFR 
50.61a alternate pressurized thermal shock (PTS) rule in lieu of the 
10 CFR 50.61 PTS rule, and would not involve a significant increase 
in the probability or consequences of an accident. Application of 10 
CFR 50.61a in lieu of 10 CFR 50.61 would not result in physical 
alteration of a plant structure, system or component, or 
installation of new or different types of equipment. Further, 
application of 10 CFR 50.61a would not significantly affect the 
probability of accidents previously evaluated in the Updated Final 
Safety Analysis Report (UFSAR) or cause a change to any of the dose 
analyses associated with the UFSAR accidents because accident 
mitigation functions would remain unchanged. Use of 10 CFR 50.61a 
would change how fracture toughness of the reactor vessel is 
assessed and does not affect reactor vessel neutron radiation 
fluence. As such, implementation of 10 CFR 50.61a in lieu of 10 CFR 
50.61 would not increase the likelihood of a malfunction.
    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 type of accident from any accident previously evaluated?
    Response: No.
    The amendment request would allow implementation of the 10 CFR 
50.61a alternate PTS rule in lieu of 10 CFR 50.61. No new accident 
scenarios, failure mechanisms, or limiting single failures are 
introduced as a result of the proposed change. No physical plant 
alterations are made as a result of the proposed change. The 
proposed change does not challenge the performance or integrity of 
any safety-related system. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The amendment request would authorize implementation of 10 CFR 
50.61a in lieu of 10 CFR 50.61. Regulation 10 CFR 50.61a would 
maintain the same functional requirements for the facility as 10 CFR 
50.61. It establishes screening criteria that limit levels of 
embrittlement beyond which operation cannot continue without further 
plant-specific evaluation or modifications. Sufficient safety 
margins are maintained to ensure that any potential increases in 
core damage frequency and large early release frequency resulting 
from implementation of 10 CFR 50.61a are negligible. As such, there 
would be no significant reduction in the margin of safety as a 
result of use of the alternate PTS rule. The margin of safety 
associated with the acceptance criteria of accidents previously 
evaluated in the UFSAR is unchanged. The proposed change would have 
no effect on the availability, operability, or performance of the 
safety-related systems and components.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White 
Plains, NY 10601.
    NRC Branch Chief: David L. Pelton.
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. 50-010, 50-237 and 50-249, 
Dresden Nuclear Power Station, Units 1, 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: August 11, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14224A245.
    Description of amendment request: The proposed changes would revise 
the description for the Emergency Response Organization (ERO) 
requalification training frequency for Exelon personnel defined in 
Exelon's governing Emergency Plans for the named stations from annually 
to ``once per calendar year not to exceed 18 months between training 
sessions.''
    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:

    Exelon has evaluated the proposed changes to the affected sites' 
Emergency Plans and determined that the changes do not involve a 
Significant Hazards Consideration. In support of this determination, 
an evaluation of each of the three (3) standards, set forth in 10 
CFR 50.92, ``Issuance of amendment,'' is provided below.
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not increase the probability or 
consequences of an accident. The proposed changes do not involve the 
modification of any plant equipment or affect plant operation. The 
proposed changes will have no impact on any safety-related 
Structures, Systems, or Components (SSC).
    The proposed changes would revise the ERO requalification 
frequency from an annual basis to once per calendar year not to 
exceed 18 months between training sessions defined in the Emergency 
Plan for the applicable Exelon facility. The proposed changes will 
align the Exelon legacy plants under one standard regarding the 
annual requalification training frequency for ERO personnel.
    Therefore, the proposed changes to the Emergency Plan 
requalification training frequency for the affected sites do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes have no impact on the design, function, or 
operation of any plant SSC. The proposed changes do not affect plant 
equipment or accident analyses. The proposed changes only affect the 
administrative aspects of the annual ERO requalification training 
frequency requirements.
    Therefore, the proposed changes to the Emergency Plan 
requalification training frequency for the affected sites do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to 
operate in the safety analyses. There is no change being made to 
safety analysis assumptions, safety limits, or limiting safety 
system settings that would adversely affect plant safety as a result 
of the proposed changes. Margins of safety are unaffected by the 
proposed changes to the frequency in the ERO requalification 
training requirements.
    Therefore, the proposed changes to the Emergency Plan 
requalification training frequency for the affected sites 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

[[Page 58816]]

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: Bradley Fewell, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and 
3, York and Lancaster Counties, Pennsylvania
    Date of application for amendments: July 25, 2014. A publicly-
available version is in ADAMS under Accession No. ML14211A017.
    Description of amendment request: The proposed amendment would 
change the definition in the PBAPS, Units 2 and 3, Technical 
Specifications (TS) for RECENTLY IRRADIATED FUEL. Specifically, the 
amendment would revise requirements pertaining to secondary containment 
hatches in order to facilitate activities performed during refueling 
outages.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, with NRC staff revisions provided in [brackets], which 
is presented below:

    1. Will operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to revise the PBAPS, Units 2 and 3, TS 
definition for RECENTLY IRRADIATED FUEL do not introduce new 
equipment or new equipment operating modes, nor do the proposed 
changes alter existing system relationships. The proposed changes do 
not affect plant operation, [any] design function, or any analysis 
that verifies the capability of a Structure, System, or Component 
(SSC) to perform a design function. There are no changes or 
modifications to [any] plant SSC. The plant Engineered Safety 
Features (ESFs) will continue to function as designed in all modes 
of operation. There are no significant changes to procedures or 
training being introduced by the proposed changes to the TS 
definition.
    Based upon the results of the [fuel handling accident (FHA)] 
analysis, it has been demonstrated that, with the requested changes, 
the dose consequences remain within the regulatory guidance provided 
by the NRC as specified in 10 CFR 50.67 and associated Regulatory 
Guide (RG) 1.183 [ADAMS Accession No. ML003716792]. The calculations 
used to evaluate the consequences of the FHA accident in support of 
the proposed changes do not by themselves affect the plant response, 
but better represent the physical characteristics of the release, so 
that appropriate mitigation techniques may be applied. Therefore, 
the consequences of an accident previously evaluated are not 
significantly increased.
    There is no adverse impact on systems designed to mitigate the 
consequences of accidents. The proposed changes do not adversely 
affect system or component pressures, temperatures, or flowrates for 
systems designed to prevent accidents or mitigate the consequences 
of an accident. Since these conditions are not adversely affected, 
the likelihood of failure of [an] SSC is not increased.
    The proposed changes do not increase the likelihood of the 
malfunction of any SSC or impact any analyzed accident. 
Consequently, the probability or consequences of an accident 
previously evaluated are not affected.
    Based on the above, Exelon concludes that the proposed changes 
do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Will operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to revise the PBAPS, Units 2 and 3, TS 
definition for RECENTLY IRRADIATED FUEL do not alter the design 
function or operation of any SSC. There are no changes or 
modifications to [any] plant SSC. The plant ESFs will continue to 
function as designed. There is no new system component being 
installed, no new construction, and no performance of a new test or 
maintenance function. The proposed TS changes do not create the 
possibility of a new credible failure mechanism or malfunction. The 
proposed changes do not introduce new accident initiators or 
precursors of a new or different kind of accident. New equipment or 
personnel failure modes that might initiate a new type of accident 
are not created as a result of the proposed changes. [Secondary 
containment] integrity is not adversely impacted and radiological 
consequences from the analyzed FHA remain within specified 
regulatory limits. The proposed changes do not adversely impact 
system or component pressures, temperatures, or flowrates for 
systems designed to prevent accidents or mitigate the consequences 
of an accident. Since these conditions are not adversely impacted, 
the likelihood of failure of [an] SSC is not increased. 
Consequently, the proposed changes cannot create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    Based on the above, Exelon concludes that the proposed changes 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Will operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed changes to revise the PBAPS, Units 2 and 3, TS 
definition for RECENTLY IRRADIATED FUEL do not alter the design 
function or operation of any SSC. There are no changes or 
modifications to [any] plant SSC. The plant ESFs will continue to 
function as designed. The proposed changes do not increase system or 
component pressures, temperatures, or flowrates for systems designed 
to prevent accidents or mitigate the consequences of an accident.
    Safety margins and analytical conservatisms have been evaluated 
and have been found acceptable. The analyzed event has been 
evaluated and margin has been retained to ensure that the analysis 
adequately bounds the postulated FHA event. The dose consequences 
resulting from analyzing the FHA design basis accident comply with 
the requirements of 10 CFR 50.67 and the guidance of RG 1.183.
    The proposed changes continue to ensure that the doses at the 
Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) 
boundary, as well as the Main Control Room (MCR), remain within 
corresponding regulatory limits.
    Based on the above, Exelon concludes 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: J. Bradley Fewell, Esquire, Vice President 
and Deputy General Counsel, Exelon Generation Company, LLC, 200 Exelon 
Way, Kennett Square, PA 19348.
    Acting NRC Branch Chief: Robert G. Schaaf.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit 2, (BVPS-2) Beaver County, 
Pennsylvania
    Date of amendment request: June 2, 2014, as supplemented by letter 
dated August 8, 2014. Publicly-available versions are in ADAMS under 
Accession Nos. ML14153A388, and ML14223A540, respectively.
    Description of amendment request: The amendment would change the 
BVPS-2 technical specifications (TSs). Specifically, the proposed 
license amendment would revise TS 4.3.2, ``Drainage,'' to correct the 
minimum drain elevation for the spent fuel storage pool specified in 
the TS. In accordance with 10 CFR Part 50, Appendix B, Section XVI, 
``Corrective Action,'' the proposed amendment is required to resolve a 
TS discrepancy regarding an existing plant design feature.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 58817]]

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.
    Previously evaluated accidents including a fuel handling 
accident and spent fuel cask drop accident are not affected by the 
proposed amendment. Reducing the minimum water level above fuel 
stored in the spent fuel storage pool in the event of inadvertent 
draining as proposed would not involve a significant increase in the 
probability of a previously evaluated accident. Maloperation or 
passive piping failure causing inadvertent draining of the spent 
fuel storage pool is not postulated concurrent with the fuel 
handling or spent fuel cask drop accident. The proposed amendment 
would not result in any failure modes that could initiate an 
analyzed accident, and does not increase the likelihood of a 
malfunction of a system, structure or component; therefore, the 
probability of analyzed accidents is not affected.
    There are no changes to how the station will be operated, 
limiting conditions for operation, or limiting safety system 
settings. The proposed amendment does not affect the capability of a 
system, structure or component to perform a design function. Since 
design functions are not affected by the proposed amendment, the 
consequences of previously evaluated accidents 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.
    Reducing the minimum water level above fuel stored in the spent 
fuel storage pool in the event of inadvertent draining as proposed 
does not create any new failure mechanisms, malfunctions, or 
accident initiators and does not change design functions or system 
operation in a way that affects the ability of systems, structures, 
and components to perform design functions.
    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.
    General Design Criterion 61, ``Fuel storage and handling and 
radioactivity control,'' of 10 CFR 50, Appendix A, states in part 
that fuel storage and handling systems shall be designed with 
suitable shielding for radiation protection.
    The proposed change involves a reduction in the minimum 
elevation of piping and penetrations of the spent fuel storage pool 
specified in the Technical Specifications. In the event maloperation 
or passive piping failure causes inadvertent draining of the spent 
fuel storage pool, the remaining water level in the pool ensures the 
stored fuel remains covered, provides adequate shielding for 
personnel, and affords adequate assurance of safety when judged 
against the current regulatory standard of General Design Criterion 
61.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, 
Akron, OH 44308.
    Acting NRC Branch Chief: Robert G. Schaaf.
FirstEnergy Nuclear Operating Company (FENOC), Docket No. 50-440, Perry 
Nuclear Power Plant, Unit 1, Perry, OH
    Date of amendment request: June 23, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14174A633.
    Description of amendment request: The proposed amendment updates 
the technical specification (TS) pressure and temperature (P/T) figures 
using an NRC approved methodology to adjust the P/T limit curves for 
previously missing data, addresses the reactor coolant system (RCS) 
vacuum condition that can occur under certain conditions, and aligns 
the heatup/cooldown requirements of the TS with the limits in the 
associated P/T figures. Additionally editorial changes are proposed 
related to the P/T figures including clarifications and updates to the 
associated titles, labeling, and notes.
    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 any accident previously 
evaluated?
    Response: No.
    The P/T [pressure and temperature] limits define RCS [reactor 
coolant system] operational limits to avoid encountering pressure, 
temperature, and temperature rate of change conditions that reduce 
safety margins with respect to nonductile brittle failure of the 
reactor coolant pressure boundary (RCPB). The figures are not 
accident initiators or accident mitigating features, but preclude 
operation in an unanalyzed condition.
    This proposed amendment does not change the design function of 
the RCS or RCPB and does not change the way the plant is maintained 
or operated when using the P/T limit curves. This proposed amendment 
does not affect any plant systems that are accident initiators and 
does not affect any accident mitigating feature.
    The proposed amendment does not affect the operability 
requirements for the RCS, as verification of operating within the P/
T limits will continue to be performed, as required. Compliance with 
and continued verification of the P/T limits support the capability 
of the RCS to perform its required design functions, consistent with 
the plant safety analyses.
    Changing the figures will not change any of the dose analyses 
associated with the USAR [updated safety analysis report] Chapter 15 
accidents because they do not affect the source term, containment 
isolation or radiological release assumptions used in any accident 
previously evaluated. Plant accident mitigation functions and 
requirements remain unchanged.
    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 P/T limits define RCS operational parameters to protect the 
RCPB and are not accident initiators or accident mitigating 
features. The limits are conservatively calculated using an NRC 
approved methodology. This proposed amendment does not change the 
design function of the RCS or RCPB, and does not change the way the 
plant is operated or maintained. This proposed amendment does not 
affect any plant systems that are accident initiators, does not 
affect any accident mitigating feature, and does not create 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 change involve a significant reduction in 
the margin of safety?
    Response: No.
    The P/T limits define RCS operational parameters, which are 
established to protect the reactor vessel. The analysis supporting 
the curve changes utilize methods previously reviewed and approved 
by the NRC.
    Margin of safety is related to the ability of the fission 
product barriers (fuel cladding, reactor coolant system, and primary 
containment) to perform their design functions during and following 
postulated accidents. This proposed amendment does not directly 
involve or physically affect fuel cladding or the primary 
containment.
    The amendment request proposes to update the P/T limit figures 
using an NRC approved methodology. The curves maintain the margin of 
safety for RCPB materials that are exposed to neutron radiation.

[[Page 58818]]

    The proposed amendment does not involve a physical change to the 
plant, does not change methods of plant operation within prescribed 
limits, and does not change methods of maintenance on equipment 
important to safety. Therefore, the proposed amendment does not 
involve a significant reduction in a margin of safety.
    Based on the responses to the three questions above, FENOC 
[FirstEnergy Nuclear Operating Company] concludes that the proposed 
amendment does not involve a significant hazards consideration under 
the standards set forth in 10 CFR 50.92(c), and, accordingly, a 
finding of ``no significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop. A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Travis L. Tate.
Florida Power and Light Company, et al. (FPL), Docket Nos. 50-335 and 
50-389, St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida
    Date of amendment request: June 9, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14175A121.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 6.2, Organization, specifically TS 
6.2.2.e. to allow the station technical assistant (STA) position to be 
manned by a single STA, a shift supervisor who meets the qualifications 
for the STA, or an individual with a senior reactor operator's license 
who meets the qualifications for the STA on each unit in MODES 1, 2, 3, 
or 4. This criterion was omitted from FPL's license amendment request 
dated July 26, 2013 (ADAMS Accession No. ML13219A840), that addressed 
shift staffing requirements. As a result, it was omitted from the 
corresponding license amendments dated February 7, 2014 (ADAMS 
Accession No. ML14016A248). This criterion was previously approved by 
the NRC and incorporated into the St. Lucie Units 1 and 2 TSs by 
Amendment Nos. 173 and 113, respectively.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes define the method for manning the shift 
technical advisor (STA) position and do not reduce the unit staffing 
requirements. In addition, the changes correct a typographical 
error. The changes do not affect the minimum shift compliment in any 
mode of operation nor decrease the effectiveness of shift personnel. 
The STA position will continue to be manned by qualified personnel. 
The proposed changes are administrative and editorial in nature and 
will not result in any significant increase in the probability of 
consequences of an accident as previously evaluated. Further, the 
proposed changes do not introduce additional risk or greater 
potential for consequences of an accident that has not previously 
been evaluated. Therefore, the proposed changes do 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 define the method for manning the shift 
technical advisor position and do not reduce the unit staffing 
requirements. In addition, the changes correct a typographical 
error. The proposed changes are administrative and editorial in 
nature. No new or different type of equipment will be installed. The 
proposed changes will not introduce new failure modes/effects that 
could lead to an accident for which consequences exceed that of 
accidents previously analyzed. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes define the method for manning the STA 
position and do not reduce the unit staffing requirements. In 
addition, the changes correct a typographical error. The changes do 
not affect the minimum shift compliment in any mode of operation nor 
decrease the effectiveness of shift personnel. The STA position will 
continue to be manned by qualified personnel. The proposed changes 
will not involve a significant reduction in a margin of safety in 
that the changes are administrative and editorial in nature. No 
plant equipment or accident analyses will be affected. Additionally, 
the proposed changes will not relax any criteria used to establish 
safety limits, safety system settings, or the bases for any limiting 
conditions for operation. Safety analysis acceptance criteria are 
not affected. Plant operation will continue within the design basis.
    The proposed changes do not adversely affect systems that 
respond to safely shutdown the plant, and maintain the plant in a 
safe shutdown condition. Consequently, 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 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.
    Acting NRC Branch Chief: Lisa M. Regner.
Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant, 
Unit 2, St. Lucie County, Florida
    Date of amendment request: January 30, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14049A284.
    Description of amendment request: The amendment would revise the 
Technical Specification (TS) surveillance requirements (SRs) for 
snubbers to conform to revisions to the Snubber Testing Program 
allowing a year extension to the existing interval for the snubber 
program transition. This revision would meet the requirements of the 
Operation and Maintenance (OM) Code and Subsection ISTD, ``Preservice 
and Inservice Examination and Testing of Dynamic Restraints (Snubbers) 
in Light Water Reactor Nuclear Power Plants,'' of the American Society 
of Mechanical Engineers OM Code, 2004 Edition with 2005 and 2006 
Addenda.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes would revise SR 4.7.9 to conform the TS to the 
revised surveillance program for snubbers. Snubber examination, testing 
and service life monitoring will continue to meet the requirements of 
10 CFR 50.55a(g). Snubber examination, testing and service life 
monitoring is not an initiator of any accident previously evaluated. 
Therefore, the probability of an accident previously evaluated is not 
significantly increased. Snubbers will continue to be demonstrated 
OPERABLE by performance of a

[[Page 58819]]

program for examination, testing and service life monitoring in 
compliance with 10 CFR 50.55a or authorized alternatives. The proposed 
change to TS ACTION 3.7.9 for inoperable snubbers is administrative in 
nature and is required for consistency with the proposed change to SR 
4.7.9. The proposed change does not adversely affect plant operations, 
design functions or analyses that verify the capability of systems, 
structures, and components to perform their design functions therefore, 
the consequences of accidents previously evaluated are not 
significantly increased. Therefore, it is concluded that this 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 changes do not involve any physical alteration of 
plant equipment. The proposed changes do not alter the method by which 
any safety-related system performs its function. As such, no new or 
different types of equipment will be installed, and the basic operation 
of installed equipment is unchanged. The methods governing plant 
operation and testing remain consistent with current safety analysis 
assumptions. Therefore, it is concluded that this 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 changes ensure snubber examination, testing and 
service life monitoring will continue to meet the requirements of 10 
CFR 50.55a(g). Snubbers will continue to be demonstrated OPERABLE by 
performance of a program for examination, testing and service life 
monitoring in compliance with 10 CFR 50.55a or authorized alternatives.
    The proposed change to TS ACTION 3.7.9 for inoperable snubbers is 
administrative in nature and is required for consistency with the 
proposed change to SR 4.7.9. Therefore, it is concluded that 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 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.
    Acting NRC Branch Chief: Lisa M. Regner.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station (CNS), Nemaha County, Nebraska
    Date of amendment request: July 14, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14202A205.
    Description of amendment request: The proposed amendment would 
delete Technical Specification 5.5.3, ``Post Accident Sampling,'' 
thereby eliminating the program requirements to have and maintain the 
post-accident sampling system. The changes are consistent with NRC-
approved Industry/Technical Specification Task Force (TSTF) Standard 
Technical Specification Change Traveler, TSTF-413, ``Elimination of 
Requirements for a Post Accident Sampling System (PASS).'' The 
availability of this technical specification improvement was announced 
in the Federal Register on March 20, 2002, as part of the consolidated 
line item improvement process. CNS will continue to have the ability to 
obtain samples, utilizing PASS, following an accident.
    Basis for proposed no significant hazards consideration 
determination: The licensee stated in its application that it reviewed 
the proposed no significant hazards consideration determination 
published on December 27, 2001 (66 FR 66949), as part of the 
consolidated line item improvement process. The licensee stated that it 
concluded that the proposed determination presented in the notice is 
applicable to CNS and the determination is incorporated by reference to 
satisfy the requirements of 10 CFR 50.91(a). As required by 10 CFR 
50.91(a), an analysis of the issue of no significant hazards 
consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 [Three Mile Island, Unit 2] accident. The specific 
intent of the PASS was to provide a system that has the capability 
to obtain and analyze samples of plant fluids containing potentially 
high levels of radioactivity, without exceeding plant personnel 
radiation exposure limits. Analytical results of these samples would 
be used largely for verification purposes in aiding the plant staff 
in assessing the extent of core damage and subsequent offsite 
radiological dose projections. The system was not intended to and 
does not serve a function for preventing accidents and its 
elimination would not affect the probability of accidents previously 
evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident from any Previously 
Evaluated.
    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.

[[Page 58820]]

    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the analysis and, based on this review, 
it appears that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    Acting NRC Branch Chief: Eric R. Oesterle.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska
    Date of amendment request: July 17, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14203A045.
    Description of amendment request: The proposed amendment would move 
the Linear Heat Generation Rate (LHGR) and Single Loop Operation LHGR 
Limit from the Technical Requirements Manual (TRM) to the Technical 
Specifications (TS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    LHGR limits have been defined to provide sufficient margin 
between the steady-state operating condition and any fuel damage 
condition to accommodate uncertainties and to assure that no fuel 
damage results even during the worst anticipated transient condition 
at any time. The proposed change to move the LHGR limits from the 
TRM to TS, including the change to TS 3.4.1, Recirculation Loops 
Operating, and TS 3.7.7, Main Turbine Bypass System, does not modify 
the limits, change assumptions for the accident analysis, or change 
operation of the station.
    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 modify the limits, change 
assumptions for the accident analysis, or change operation of the 
station.
    The proposed change does move LHGR limits that have been defined 
to provide sufficient margin between the steady-state operating 
condition and any fuel damage condition to accommodate uncertainties 
and to assure that no fuel damage results even during the worst 
anticipated transient condition at any time from the TRM to TS.
    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 to move the LHGR limits from the TRM to TS, 
including the change to TS 3.4.1 and TS 3.7.7, does not modify the 
limits, change assumptions for the accident analysis, or change 
operation of the station.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    Acting NRC Branch Chief: Eric R. Oesterle.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center, Linn County, Iowa
    Date of amendment request: June 23, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14175B387.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) requirements to address NRC Generic 
Letter (GL) 2008-01, ``Managing Gas Accumulation in Emergency Core 
Cooling, Decay Heat Removal, and Containment Spray Systems,'' as 
described in Technical Specification Task Force (TSTF) Change Traveler 
TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas 
Accumulation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the Proposed Change Involve a Significant Increase in 
the Probability or Consequences of an Accident Previously Evaluated?
    Response: No.
    The proposed change revises or adds SRs [Surveillance 
Requirements] that require verification that the Emergency Core 
Cooling Systems (ECCS), Residual Heat Removal (RHR) System, and the 
Reactor Core Isolation Cooling (RCIC) System are not rendered 
inoperable due to accumulated gas and to provide allowances which 
permit performance of the revised verification. Gas accumulation in 
the subject systems is not an initiator of any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not significantly increased. The proposed SRs ensure 
that the subject systems continue to be capable to perform their 
assumed safety function and are not rendered inoperable due to gas 
accumulation. Thus, the consequences of any accident previously 
evaluated are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the Proposed Change Create the Possibility of a New or 
Different Kind of Accident from any Accident Previously Evaluated?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, RHR System, and RCIC 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, and RCIC System are not 
rendered inoperable due to accumulated gas and to

[[Page 58821]]

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.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. James Petro, P. O. Box 14000 Juno Beach, 
FL 33408-0420.
    NRC Branch Chief: David L. Pelton.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
1, (Seabrook) Rockingham County, New Hampshire
    Date of amendment request: July 24, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14209A918.
    Description of amendment request: The proposed amendment would 
revise Seabrook Technical Specifications (TSs) by increasing the 
voltage limit for a full load rejection test of the emergency diesel 
generator specified in surveillance requirement 4.8.1.1.2.f.3 of TS 
3.8.1.1, ``A.C. [alternating current] Sources--Operating.'' The 
proposed amendment also revises the TS definition of the terms 
``Operable--Operability.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, with NRC staff revisions provided in [brackets], which 
is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to increase in the [emergency diesel 
generator] EDG full load rejection overvoltage limit from 4784 
[volts] V to 4992V is not an accident initiator. The overvoltage 
transient is an expected response to a full load rejection. The 
magnitude and duration of the proposed overvoltage limit have been 
considered and determined to have no detrimental effects on the 
connected equipment that is exposed to the voltage transient. The 
proposed change does not affect the EDG design function or how the 
EDG is operated. Since the EDG is not impacted, the EDG remains 
capable of performing its intended design function of supplying 
power to emergency safeguards equipment. The proposed change to the 
definition of operable--operability is administrative in nature and 
does not alter the meaning of the defined terms.
    Therefore, the proposed changes do 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 revise the definition of the terms 
operable--operability and to increase the EDG full load rejection 
overvoltage limit from 4784V to 4992V are not accident initiators. 
The overvoltage transient is an expected response to a full load 
rejection. The magnitude and duration of the proposed overvoltage 
limit have been considered and determined to have no detrimental 
effects on the connected equipment that is exposed to the voltage 
transient. The proposed changes do not introduce any new failure 
modes.
    The changes do not involve a physical alteration to the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods for operating the plant. The proposed changes 
do not affect the EDG design function or how the EDG is operated. 
Since the EDG is not impacted, the EDG remains capable of performing 
its intended design function of supplying power to emergency 
safeguards equipment. The change to the definition of operable--
operability makes grammatical corrections and adds clarity but makes 
no change to the meaning of the terms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to increase in the EDG full load rejection 
overvoltage limit from 4784V to 4992V has been evaluated with 
consideration of the effect on the EDG and connected equipment that 
would be exposed to the higher voltage transient. Based on review of 
equipment specifications, test data, and manufacturer's input, it 
was concluded that there would be no detrimental effects to the EDG 
or connected equipment that is exposed to the higher voltage 
transient. The EDG remains capable of performing its intended design 
function of supplying power to emergency safeguards equipment.
    The proposed change to the definition of operable--operability 
is administrative in nature and does not alter any criterion used to 
establish operability of plant structure, systems, or components.
    The proposed amendment does not involve changes to any safety 
analyses assumptions, safety limits, or limiting safety system 
settings. The changes do not adversely impact plant operating 
margins or the reliability of equipment credited in the safety 
analyses.
    Therefore, the proposed changes do 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: William Blair, Managing Attorney, Florida 
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    Acting NRC Branch Chief: Robert G. Schaaf.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
1, (Seabrook) Rockingham County, New Hampshire
    Date of amendment request: July 24, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14209A919.
    Description of amendment request: The proposed amendment would 
revise the Seabrook Technical Specifications (TS). The proposed change 
modifies TS 3.3.3.1, ``Radiation Monitoring for Plant Operations,'' to 
eliminate duplicate requirements, resolve an inconsistency, and correct 
a deficiency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, with NRC staff revisions provided in [brackets], which 
is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The instruments involved with the proposed changes to the 
technical specifications (TS) are not initiators of any accidents 
previously evaluated, and the probability and consequences of 
accidents previously evaluated are unaffected by the proposed 
changes. There is no change to any equipment response or accident 
scenario, and the changes impose no additional challenges to fission 
product barrier integrity. The proposed changes do not alter the 
design, function, operation, or configuration of any plant 
structure, system, or component (SSC). As a result, the outcomes of 
accidents previously evaluated are unaffected. The

[[Page 58822]]

proposed changes modify the TS to eliminate duplicate requirements, 
resolve an inconsistency, and correct a deficiency.
    Therefore, the proposed changes do not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
The changes do not challenge the integrity or performance of any 
safety-related systems. No plant equipment is installed or removed, 
and the changes do not alter the design, physical configuration, or 
method of operation of any plant SSC. No physical changes are made 
to the plant, so no new causal mechanisms are introduced.
    Therefore, the proposed changes to the TS do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The ability of any operable SSC to perform its designated safety 
function is unaffected by the proposed changes. The proposed changes 
do not alter any safety analyses assumptions, safety limits, 
limiting safety system settings, or method of operating the plant. 
The changes do not adversely impact plant operating margins or the 
reliability of equipment credited in the safety analyses.
    Therefore, the proposed change 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: William Blair, Managing Attorney, Florida 
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    Acting NRC Branch Chief: Robert G. Schaaf.
NextEra Energy Seabrook LLC, Docket No. 50-443, Seabrook Station, Unit 
1, Rockingham County, New Hampshire
    Date of amendment request: July 24, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14216A404.
    Description of amendment request: The proposed amendment would 
incorporate revised reactor coolant system (RCS) pressure-temperature 
limits in the Technical Specification (TS) applicable to 55 effective 
full-power years. The change will also provide new overpressure 
protection setpoints and lower the RCS temperature at which the TS is 
applicable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the Technical Specifications (TS) do not 
impact the physical function of plant structures, systems, or 
components (SSCs) or the manner in which SSCs perform their design 
function. Operation in accordance with the proposed TS will ensure 
that all analyzed accidents will continue to be mitigated by the 
SSCs as previously analyzed. The proposed changes do not alter or 
prevent the ability of operable SSCs to perform their intended 
function to mitigate the consequences of an initiating event within 
assumed acceptance limits. The proposed changes neither adversely 
affect accident initiators or precursors, nor alter design 
assumptions.
    Therefore, the proposed changes do 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 previously evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed), create new failure modes for existing equipment, or 
create any new limiting single failures. The changes to the 
pressure--temperature limits, power operated relief valve setpoints, 
and the over pressure protection system effective temperature will 
continue to ensure that appropriate fracture toughness margins are 
maintained to protect against reactor vessel failure, during both 
normal and low temperature operation. The proposed changes are 
consistent with the applicable NRC approved methodologies (i.e., 
WCAP-14040, Rev. 4 and ASME Code Case N-641). Plant operation will 
not be altered, and all safety functions will continue to perform as 
previously assumed in accident analyses.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed changes will not 
adversely affect the operation of plant equipment or the function of 
any equipment assumed in the accident analysis. The proposed changes 
were developed using NRC approved methodologies and will continue to 
ensure an acceptable margin of safety is maintained. The safety 
analysis acceptance criteria are not affected by this change. The 
proposed changes will not result in plant operation in a 
configuration outside the design basis. The proposed changes do not 
adversely affect systems that respond to safely shutdown the plant 
and to maintain the plant in a safe shutdown condition.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William Blair, Managing Attorney, Florida 
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    Acting NRC Branch Chief: Robert G. Schaaf.
Northern States Power Company--Minnesota (NSPM), Docket No. 50-263, 
Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota
    Date of amendment request: June 17, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14168A486.
    Description of amendment request: The NSPM proposes to revise MNGP 
Technical Specification (TS) 3.5.1, ``ECCS [Emergency Core Cooling 
System]--Operating,'' to correct the requirements for the Alternate 
Nitrogen System pressure. TS Surveillance Requirement (SR) 3.5.1.3 
requires verification of limits for automatic depressurization system 
(ADS) pneumatic pressure for both ADS pneumatic supplies. The proposed 
change would revise the TS SR 3.5.1.3.b pressure limit for determining 
operability of the Alternate Nitrogen System from greater than or equal 
to (>=) 410 pounds per square inch gauge (psig) to a corrected value of 
>= 700 psig.
    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 provided 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 the TS SR for the purpose of 
restoring a value to be consistent with the licensing basis. The 
proposed TS change does not introduce new

[[Page 58823]]

equipment or new equipment operating modes, nor does the proposed 
change alter existing system relationships. The proposed change does 
not affect plant operation, design function or any analysis that 
verifies the capability of a system, structure or component (SSC) to 
perform a design function. Further, the proposed change does not 
increase the likelihood of the malfunction of any SSC or impact any 
analyzed accident. Consequently, the probability of an accident 
previously evaluated is not affected and there is not significant 
increase in the consequences of any accident previously evaluated.
    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.
    There proposed change revises the TS SR for the purpose of 
restoring a value to be consistent with the licensing basis. The 
change does not involve a physical alteration to the plant (i.e., no 
new or different type of equipment will be installed) or a change in 
the methods governing normal plant operations. The proposed change 
does not alter assumptions made in the safety analysis for the 
components supplied by the Alternate Nitrogen System. Further, the 
proposed change does not introduce new accident initiators.
    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 the TS SR for the purpose of 
restoring a value to be consistent with the licensing basis. The 
proposed change does not alter the manner in which safety limits, 
limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis assumptions and 
acceptance criteria are not affected by this change.
    Therefore, the proposed change 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: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: David L. Pelton.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, 
California
    Date of amendment request: July 28, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14209B074.
    Description of amendment request: The proposed amendments would 
modify the technical specifications (TS) to risk-inform requirements 
regarding selected Required Action End States. The proposed changes to 
the Required Action End States are described in Table 1 of the 
Enclosure to the licensee's letter dated July 28, 2014. The changes are 
consistent with Technical Specification Task Force (TSTF) Traveler 
TSTF-432, Revision 1, ``Change in Technical Specifications End States 
(WCAP-16294)'' (ADAMS Accession No. ML103430249).
    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 modifies the end state (e.g., mode or other 
specified condition) which the Required Actions specify must be 
entered if compliance with the Limiting Conditions for Operation 
(LCO) is not restored. The requested Technical Specifications (TS) 
permit an end state of Mode 4 rather than an end state of Mode 5 
contained in the current TS. In some cases, other Conditions and 
Required Actions are revised to implement the proposed change. 
Required Actions are not an initiator of any accident previously 
evaluated. Therefore, the proposed change does not affect the 
probability of any accident previously evaluated. The affected 
systems continued to be required to be operable by the TS and the 
Completion Times specified in the TS to restore equipment to 
operable status or take other remedial Actions remain unchanged. 
WCAP-16294-NP-A, Revision 1, ``Risk-Informed Evaluation of Changes 
to [Technical Specification] Required Action Endstates for 
Westinghouse NSSS [Nuclear Steam Supply System] PWRs [Pressurized 
Water Reactors],'' demonstrates that the proposed change does not 
significantly increase the consequences of any accident previously 
evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different accident from any accident previously evaluated?
    Response: No.
    The proposed change modifies the end state (e.g., mode or other 
specified condition) which the Required Actions specify must be 
entered if compliance with the LCO is not restored. In some cases, 
other Conditions and Required Actions are revised to implement the 
proposed change. The 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 change does not impose any new 
requirements. The change does not alter assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different 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 modifies the end state (e.g., mode or other 
specified condition) which the Required Actions specify must be 
entered if compliance with the LCO is not restored. In some cases, 
other Conditions and Required Actions are revised to implement the 
proposed change. Remaining within the Applicability of the LCO is 
acceptable because WCAP-16294-NP-A demonstrates that the plant risk 
in MODE 4 is similar to, or lower than, MODE 5. As a result, no 
margin of safety is significantly affected.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    Acting NRC Branch Chief: Eric R. Oesterle.
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 12, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14164A098.
    Description of amendment request: The proposed license amendment 
request (LAR) proposes to revise Plant Specific Tier 2* material within 
the Updated Final Safety Analysis Report (UFSAR) by making editorial 
and consistency corrections.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 58824]]

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 editorial and consistency update does not involve a 
technical change, i.e., there is no design parameter or requirement, 
calculation, analysis, function, or qualification change. No 
structure, system, component (SSC), design, or function would be 
adversely affected. No design or safety analysis would be adversely 
affected. The proposed changes do not adversely affect any accident 
initiating event or component failure, thus the probabilities of the 
accidents previously evaluated are not adversely affected. No 
function used to mitigate a radioactive material release and no 
radioactive material release source term is involved, thus the 
radiological releases in the accident analyses are not adversely 
affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed editorial and consistency update would not affect 
the design or function of any SSC, but will instead provide 
consistency between the SSC designs and functions and the 
discussions currently presented in the UFSAR via Tier 2* 
information. The proposed nontechnical changes would not introduce a 
new failure mode, fault, or sequence of events that could result in 
a radioactive material release.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed editorial and consistency update is nontechnical 
and thus would not affect any design parameter, function, or 
analysis. There would be no change to an existing design basis, 
design function, regulatory criterion, or analyses. No safety 
analysis or design basis acceptance limit/criterion is involved.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW, Washington, DC, 20004-2514.
    NRC Branch Chief: Lawrence J. Burkhart.
South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028, 
Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, 
South Carolina
    Date of amendment request: August 28, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14245A601.
    Description of amendment request: The proposed license amendment 
request (LAR) proposes to revise Tier 2* and Tier 2 information related 
to the design details of connections in several locations between the 
steel plate composite construction (SC) used for the shield building 
and the standard reinforced concrete (RC) walls, floors, and roofs of 
the auxiliary building and the lowers of the shield building.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The design function of the nuclear island structures is to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the nuclear island. 
The nuclear island structures are structurally designed to meet 
seismic Category I requirements as defined in Regulatory Guide 1.29.
    The changes to the detail design of connections between the RC 
and SC structures do not have an adverse impact on the response of 
the nuclear island structures to safe shutdown earthquake ground 
motions or loads due to anticipated transients or postulated 
accident conditions. The changes to the detail design do not impact 
the support, design, or operation of mechanical and fluid systems. 
There is no change to plant systems or the response of systems to 
postulated accident conditions. There is no change to the predicted 
radioactive releases due to postulated accident conditions. The 
plant response to previously evaluated accidents or external events 
is not adversely affected, nor do the changes described create any 
new accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes are to the detail design of connections 
between the RC and SC structures. The changes to the detail design 
of connections do not change the criteria and requirements for the 
design and analysis of the nuclear island structures. The changes to 
the detail design of connections do not change the design function, 
support, design, or operation of mechanical and fluid systems. The 
changes to the detail design of connections do not change the 
methods used to connect the RC to SC. The changes of the detail 
design of connections do not result in a new failure mechanism for 
the nuclear island structures or new accident precursors. As a 
result, the design functions of the nuclear island structures are 
not adversely affected by the proposed changes.
    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.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes; and thus, no margin 
of safety is reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW, Washington, DC, 20004-2514.
    NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company Docket Nos.: 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: July 29, 2014. A publicly-available 
version is available in ADAMS under Accession No. ML14210A646.
    Description of amendment request: The proposed license amendment 
request would revise the Combined Licenses (COLs) with regard to Tier 1 
material and promote consistency with the Updated Final Safety Analysis 
Report Tier 2.
    Nuclear Operating Company has also requested an exemption from the 
provisions of 10 CFR part 52, appendix D, section III.B, ``Design 
Certification Rule for the AP1000 Design, Scope and Contents,'' to 
allow a departure from the elements of the certification information in 
Tier 1 of the generic Design Control Document.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 58825]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. Does the requested amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No
    The proposed editorial and consistency COL Appendix C and 
corresponding plant-specific Tier 1 update does not involve a 
technical change, e.g., there is no design parameter or requirement, 
calculation, analysis, function or qualification change. No 
structure, system, or component (SSC) design or function would be 
affected. No design or safety analysis would be affected. The 
proposed changes do not affect any accident initiating event or 
component failure, thus the probabilities of the accidents 
previously evaluated are not affected. No function used to mitigate 
a radioactive material release and no radioactive material release 
source term is involved, thus the radiological releases in the 
accident analyses are not affected.
    Therefore, the requested amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the requested amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No
    The proposed editorial and consistency COL Appendix C and 
corresponding plant-specific Tier 1 update would not affect the 
design or function of any SSC, but will instead provide consistency 
between the SSC designs and functions currently presented in the 
UFSAR, COL Appendix C, and the Tier 1 information. The proposed 
changes would not introduce a new failure mode, 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.
    Therefore, the requested amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No
    The proposed editorial and consistency COL Appendix C and 
corresponding plant-specific Tier 1 update would not affect the 
design or function of any SSC, but will instead provide consistency 
between the SSC designs and functions currently presented in the 
UFSAR, COL Appendix C, and the Tier 1 information. The proposed 
changes would not introduce a new failure mode, fault or sequence of 
events that could result in a radioactive material release. 
Therefore, the requested amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. M. Stanford Blanton, Blach & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence Burkhart.
Southern Nuclear Operating Company Docket Nos.: 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: July 30, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14211A666.
    Description of amendment request: The proposed license amendment 
request would revise the combined licenses (COLs) with regard to Tier 
2* material within the Updated Final Safety Analysis Report (UFSAR) to 
resolve inconsistencies with other Tier 2* information elsewhere in the 
UFSAR.
    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 requested amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No.
    The proposed editorial and consistency update does not involve a 
technical change, i.e., there is no design parameter or requirement, 
calculation, analysis, function, or qualification change. No 
structure, system, or component, design, or function would be 
adversely affected. No design or safety analysis would be adversely 
affected. The proposed changes do not adversely affect any accident 
initiating event or component failure, thus the probabilities of the 
accidents previously evaluated are not adversely affected. No 
function used to mitigate a radioactive material release and no 
radioactive material release source term is involved, thus the 
radiological releases in the accident analyses are not adversely 
affected.
    Therefore, the requested amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the requested amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed editorial and consistency update would not affect 
the design or function of any structure, system, or component, but 
will instead provide consistency between the structure, system, and 
component designs and functions and the discussions currently 
presented in the UFSAR via Tier 2* information. The proposed non-
technical changes would not introduce a new failure mode, fault, or 
sequence of events that could result in a radioactive material 
release.
    Therefore, the requested amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed editorial and consistency update is non-technical 
and thus would not affect any design parameter, function, or 
analysis. There would be no change to an existing design basis, 
design function, regulatory criterion, or analyses. No safety 
analysis or design basis acceptance limit/criterion is involved.
    Therefore, the requested amendment does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. M. Stanford Blanton, Blach & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence Burkhart.
Southern Nuclear Operating Company, Inc. Docket Nos.: 52-025 and 52-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, 
Georgia
    Date of amendment request: April 11, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14101A459. The amendment 
request was supplemented by letter dated April 18, 2014. A publicly-
available version is in ADAMS under Accession No. ML14108A093. The 
amendment request was further supplemented by two letters dated August 
28, 2014. Publicly-available versions of the two letters are in ADAMS 
under Accession Nos. ML14241A250 and ML14241A264.
    Description of amendment request: The license amendment request was 
originally noticed in the Federal Register on June 6, 2014 (79 FR 
32771). This notice is being reissued in its entirety to include the 
revised analysis of the issue of no significant hazards consideration 
submitted by the licensee in its August 28, 2014, submission (ADAMS 
Accession No. ML14241A250). The proposed license amendment request 
would depart from the plant-specific Design Control Document Tier 1 and 
Tier 2 material to describe modifications to increase the efficiency of 
the return of condensate utilized by the passive core cooling system to 
the

[[Page 58826]]

in-containment refueling water storage tank to support the capability 
for long term cooling.
    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 containment condensate flow path changes provide 
sufficient condensate return flow to maintain In-containment 
Refueling Water Storage Tank (IRWST) level above the top of the 
Passive Residual Heat Removal Heat Exchanger (PRHR HX) tubes long 
enough to prevent PRHR HX performance degradation from that 
considered in the UFSAR [Updated Final Safety Analysis Report] 
Chapter 15 safety analyses. The added components are seismically 
qualified and constructed of only those materials appropriately 
suited for exposure to the reactor coolant environment as described 
in UFSAR Section 6.1. No aluminum is permitted to be used in the 
construction of these components so that they do not contribute to 
hydrogen production in containment.
    The proposed changes clarify the design basis for the PRHR HX, 
which removes decay heat from the Reactor Coolant System (RCS) 
during a non-loss of coolant accident (non-LOCA). With operator 
action to avoid unnecessary Automatic Depressurization System (ADS) 
actuation based on RCS conditions, PRHR HX operation can be extended 
longer than would be maintained automatically by the protection 
system. Though analysis shows significantly greater capacity, the 
extent of the capability of the PRHR HX would be changed from 
operating indefinitely to operating for at least 72 hours. If PRHR 
HX capability were exhausted after 72 hours, the ADS would be 
actuated, which could result in significant containment floodup. 
However, probabilistic analysis shows the probability of design 
basis containment floodup after PRHR HX operation during a non-LOCA 
event is significantly lower than the probability of a small break 
LOCA, for which comparable containment floodup is anticipated.
    Therefore, the probability of significant containment floodup is 
not increased.
    The proposed changes do not affect any components whose failure 
could initiate a previously evaluated event, thus the probabilities 
of the accidents previously evaluated are not affected. The affected 
equipment does not adversely affect or interact with safety-related 
equipment or another radioactive material barrier. The proposed 
changes clarify the post-accident performance requirements for the 
PRHR HX. However, the proposed changes do not prevent the engineered 
safety features from performing their safety-related accident 
mitigating functions. The radioactive material source terms and 
release paths used in the safety analyses are unchanged, thus the 
radiological releases in the UFSAR accident analyses are not 
affected.
    Therefore, the proposed amendment does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The long-term safe shutdown analysis results show that the PRHR 
HX continues to meet its acceptance criterion, i.e., to cool the 
Reactor Coolant System (RCS) to below 420[ordm]F in 36 hours. The 
added equipment does not adversely interface with any component 
whose failure could initiate an accident, or any component that 
contains radioactive material. The modified components do not 
incorporate any active features relied upon to support normal 
operation. The downspout and gutter return components are 
seismically qualified to remain in place and functional during 
seismic and dynamic events. The containment condensate flow path 
changes do not create a new fault or sequence of events that could 
result in a radioactive material release.
    The proposed change quantifies the duration that the PRHR HX is 
capable of maintaining adequate core cooling, and specifies that if 
PRHR HX cooling capability is exhausted, the ADS would be actuated. 
This involves the possibility of opening the ADS valves after the 
IRWST water level has decreased below the spargers, which promote 
steam condensation in the IRWST. During this condition, the loads on 
the IRWST, spargers and any internal structures or components in the 
IRWST would still be less than their limiting loads, and these SSCs 
would not be adversely affected or cause a different mode of 
operation. Therefore, no new type of accident could be created by 
this condition.
    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 proposed changes do not reduce the redundancy or diversity 
of any safety-related function. The added components are classified 
as safety-related, seismically qualified, and are designed to comply 
with applicable design codes. The proposed containment condensate 
flow path changes provide sufficient condensate return flow to 
maintain adequate IRWST water level for those events using the PRHR 
HX cooling function. The long-term Shutdown Temperature Evaluation 
results in UFSAR Chapter 19E show the PRHR HX continues to meet its 
acceptance criterion. The UFSAR Chapters 6 and 15 analyses results 
are not affected, thus margins to their regulatory acceptance 
criteria are unchanged. The former design basis, which stated the 
PRHR HX could bring the plant to 420[deg]F within 36 hours, is 
changed to state the heat exchanger can establish safe, stable 
conditions in the reactor coolant system after a design basis event. 
Such safe stable conditions may not coincide with an RCS temperature 
of 420[deg]F. However, the PRHR HX is able to bring the RCS to a 
sufficiently low temperature such that RCS conditions would be 
comparable to those achieved at 420[deg]F--peak cladding 
temperatures, departure from nucleate boiling, and pressurizer level 
would be maintained within acceptable limits of the evaluation 
criteria. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the proposed changes, thus no 
margin of safety is significantly reduced.
    Therefore, the proposed amendment does not reduce 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: Mr. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company, Inc. Docket Nos.: 52-025 and 52-
026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke 
County, Georgia
    Date of amendment request: July 14, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14195A296.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-91 and NPF-92 for the VEGP, Units 3 and 4 to 
modify the fire area fire barriers of the turbine building switchgear 
rooms on Elevations 141'-3'' and 158'-7'' of the turbine building to 
accommodate the revised layout of the low and medium voltage switchgear 
and associated equipment. The proposed changes also provide an 
editorial change to a fire area number. The requested amendment 
requires changes to Updated Final Safety Analysis Report (UFSAR) 
information, which include changes to plant-specific Tier 2* 
information and changes to Tier 2 information that involve changes to 
this 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.

[[Page 58827]]

    The proposed reconfiguration of the turbine building switchgear 
rooms, the control system cabinet room, the new electrical equipment 
room, and the associated heating, ventilation and air conditioning 
(HVAC) room and the proposed editorial change would not adversely 
affect any safety-related equipment or function. The modified 
configuration will maintain the fire protection function (i.e., 
barrier) as evaluated in Updated Final Safety Analysis Report 
(UFSAR) Appendix 9A, thus, the probability of a spread of a fire 
from these areas 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 affect 
nonsafety-related electrical switchgear and do not involve any 
accident, initiating event, or component failure; thus, the 
probabilities of the accidents previously evaluated are not 
affected. The proposed changes do not interface with or affect any 
system containing radioactivity or affect any radiological material 
release source terms; 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 proposed changes to the fire zones in the turbine building 
related to the turbine building switchgear rooms, the control system 
cabinet room, the new electrical equipment room, the associated HVAC 
room, and stairway will maintain the fire barrier fire protection 
function as evaluated in the UFSAR Appendix 9A. The changes to the 
fire areas and fire zones do not affect the function of any safety-
related structure, system, or component, and thus, do not introduce 
a new failure mode. The affected turbine building areas and 
equipment do not interface with any safety-related equipment or any 
equipment associated with radioactive material and, thus, do not 
create a new fault or sequence of events that could result in 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 proposed reconfiguration of the fire zones associated with 
the turbine building switchgear rooms, the electrical equipment 
room, and the associated HVAC room and the proposed editorial change 
will maintain the fire barrier fire protection function as evaluated 
in the UFSAR Appendix 9A. The fire barriers and equipment in the 
turbine building do not interface with any safety-related equipment 
or affect any safety-related function. The changes to the area 
barriers associated with the turbine building switchgear and 
associated HVAC continue to comply with the existing design codes 
and regulatory criteria, and do not affect any safety analysis.
    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: Mr. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence J. Burkhart.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee
    Date of amendment request: July 24, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14210A051.
    Description of amendment request: The amendment would revise the 
reactor coolant pump (RCP) flywheel inspection surveillance 
requirements to extend the allowable inspection interval to 20 years, 
consistent with Technical Specification Task Force (TSTF) Standard 
Technical Specification Change Traveler, TSTF-421, ``Revision to RCP 
Flywheel Inspection Program (WCAP-15666).'' The Nuclear Regulatory 
Commission (NRC) staff believes that this amendment made use of both 
TSTF-421 and TSTF-237, Revision 1, ``Relaxation of Reactor Coolant Pump 
Flywheel Examinations.''
    The NRC staff published a notice of opportunity for comment in the 
Federal Register on June 24, 2003 (68 FR 37590), on possible amendments 
adopting TSTF-421, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line-item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on October 22, 
2003 (68 FR 60422). The licensee affirmed the applicability of the 
model NSHC determination in its application dated July 24, 2014.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration determination is presented 
below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change to the RCP flywheel examination frequency 
does not change the response of the plant to any accidents. The RCP 
will remain highly reliable and the proposed change will not result 
in a significant increase in the risk of plant operation. Given the 
extremely low failure probabilities for the RCP motor flywheel 
during normal and accident conditions, the extremely low probability 
of a loss-of-coolant accident (LOCA) with loss of offsite power 
(LOOP), and assuming a conditional core damage probability (CCDP) of 
1.0 (complete failure of safety systems), the core damage frequency 
(CDF) and change in risk would still not exceed the NRC's acceptance 
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per 
year). Moreover, considering the uncertainties involved in this 
evaluation, the risk associated with the postulated failure of an 
RCP motor flywheel is significantly low. Even if all four RCP motor 
flywheels are considered in the bounding plant configuration case, 
the risk is still acceptably low.
    The proposed change does not adversely affect accident 
initiators or precursors, nor alter the design assumptions, 
conditions, or configuration of the facility, or the manner in which 
the plant is operated and maintained; alter or prevent the ability 
of structures, systems, components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits or affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the type or amount of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiations exposure. The proposed change is 
consistent with the safety analysis assumptions and resultant 
consequences.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident from Any Accident Previously 
Evaluated

    The proposed change in flywheel inspection frequency does not 
involve any change in the design or operation of the RCP. Nor does 
the change to examination frequency affect any existing accident 
scenarios, or create any new or different accident scenarios. 
Further, the change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or alter the methods governing normal plant operation. In 
addition, the change does not impose any new or different 
requirements or eliminate any existing requirements, and does not 
alter any assumptions made in the safety analysis. The proposed 
change is consistent with the

[[Page 58828]]

safety analysis assumptions and current plant operating practice.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by this change. The proposed change will not result 
in plant operation in a configuration outside of the design basis. 
The calculated impact on risk is insignificant and meets the 
acceptance criteria contained in RG 1.174. There are no significant 
mechanisms for inservice degradation of the RCP flywheel.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Jessie F. Quichocho.
Virginia Electric and Power Company, Docket Nos.: 50-280 and 50-281, 
Surry Power Station, Units 1 and 2, Surry County, Virginia
    Date of amendment request: June 3, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14160A607.
    Description of amendment request: The amendments would revise the 
Surry Power Station (Surry) Units 1 and 2, Technical Specifications 
(TS). Specifically, TS Figures 3.1-1 and 3.1-2, Surry, Units 1 and 2, 
Reactor Coolant System Heatup Limitations and Surry, Units 1 and 2, 
Reactor Coolant System Cooldown Limitations, respectively, are being 
revised for clarification and to be fully representative of the 
allowable operating conditions during Reactor Coolant System startup 
and cooldown evolutions.
    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 clarification of TS Figures 3.1-1 and 3.1-2 does 
not involve a physical change to the plant and does not change the 
manner in which plant systems or components are operated or 
controlled. The proposed change does not alter or prevent the 
ability of structures, system, and components (SSCs) to perform 
their intended function to mitigate the consequences of an 
initiating event within the assumed acceptance limits. The P/T 
limits curves on TS Figures 3.1-1 and 3.1-2 are not being modified 
and remain valid.
    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 clarification of TS Figures 3.1-1 and 3.1-2 does 
not involve any physical alteration of plant equipment; 
consequently, no new or different types of equipment will be 
installed. The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The P/T limits curves 
on TS Figures 3.1-1 and 3.1-2 are not being modified, and the basic 
operation of installed plant systems and components is unchanged.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The existing RCS P/T limits curves on TS Figures 3.1-1 and 3.1-2 
are not being modified. The proposed clarification of TS Figures 
3.1-1 and 3.1-2 does not alter any plant equipment, does not change 
the manner in which the plant is operated or controlled, and has no 
impact on any safety analysis assumptions. The proposed change does 
not alter the manner in which safety limits, limiting safety system 
settings, or limiting conditions for operation are determined. The 
proposed change does not result in plant operation in a 
configuration outside the analyses or design basis and does not 
adversely affect systems that respond to safely shut down the plant 
and to maintain the plant in a safe shutdown condition.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Arizona Public Service Company, et al., Docket Nos.: STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 
2, and 3, Maricopa County, Arizona
    Date of application for amendment: September 27, 2013, as 
supplemented by letter dated December 12, 2013.
    Brief description of amendment: The amendments revised Technical 
Specification (TS) 3.3.3, ``Control Element Assembly Calculators 
(CEACs),'' to reinstate an inadvertently omitted 4-hour completion time 
for

[[Page 58829]]

Required Action B.2.2. Additionally, the amendments revised a test 
frequency note in Surveillance Requirement (SR) 3.3.6.2 under TS 3.3.6, 
``Engineered Safety Features Actuation System (ESFAS) Logic and Manual 
Trip,'' which should have been included in the license amendment 
request for Technical Specifications Task Force (TSTF) change traveler 
TSTF-425, Revision 3, ``Relocate Surveillance Frequencies to Licensee 
Control--RITSTF [Risk-Informed TSTF] Initiative 5b.''
    Date of issuance: September 9, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: Unit 1--194; Unit 2--194; Unit 3--194. A publicly-
available version is in ADAMS under Accession No. ML14202A378; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: 
The amendment revised the Operating Licenses and Technical 
Specifications.
    Date of initial notice in Federal Register: February 4, 2014 (79 FR 
6640). The supplemental letter dated December 12, 2013, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 9, 2014.
    No significant hazards consideration comments received: No.
Dairyland Power Cooperative, Docket Nos.: 50-409 and 72-046, La Crosse 
Boiling Water Reactor (LACBWR), La Crosse County, Wisconsin
    Date of application for amendment: August 6, 2013, supplemented by 
letters dated January 16, 2014, and April 14, 2014.
    Brief description of amendment: The amendment approves changes to 
the Emergency Plan, including removal of the various emergency actions 
related to the former spent fuel pool, the transfer of responsibility 
for implementing the Emergency Plan to the Security Shift Supervisors 
at the ISFSI, a revised emergency plan organization, removal of the 
fire brigade, and abandonment of the LACBWR Control Room consistent 
with the current state of decommissioning, in that all of the spent 
nuclear fuel has now been transferred from the spent fuel pool to an 
independent spent fuel storage installation.
    Date of issuance: September 8, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 73.
    Possession Only License No. DPR-45.
    Date of initial notice in Federal Register: October 29, 2013 (78 FR 
64543). The supplements dated January 16, 2014, and April 14, 2014, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register on 
October 29, 2013 (78 FR 64543).
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated September 8, 2014.
    No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos.: 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
    Date of application for amendments: October 28, 2013, as 
supplemented by letter dated June 3, 2014.
    Brief description of amendments: The amendments modify Technical 
Specification (TS) 3.8.4. Specifically, the change allows a one-time 
extension of the completion time for Required Action A.2.2 to support 
replacement of the existing shared 125 VDC vital batteries.
    Date of issuance: September 10, 2014.
    Effective date: This license amendment is effective as of its date 
of issuance and shall be implemented within 30 days of issuance.
    Amendment Nos.: 274 and 254. A publicly-available version is in 
ADAMS under Accession No. ML14231A634; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the licenses and technical specifications.
    Date of initial notice in Federal Register: January 21, 2014 (79 FR 
3415). The supplemental letter dated June 3, 2014, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 10, 2014.
    No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit 2, Westchester County, New York
    Date of amendment request: January 16, 2014, as supplemented by 
letters dated April 2, and April 15, 2014.
    Brief description of amendment: The amendment revises Indian Point 
Nuclear Generating Unit 2 Technical Specification (TS) 5.5.7, ``Steam 
Generator (SG) Program,'' to exclude portions of the SG tubes below the 
top of the SG tubesheet from periodic inspections and plugging by 
implementing the alternate repair criteria ``H*.'' In addition, TS 
3.4.13, ``RCS [reactor coolant system] Operational Leakage,'' is being 
revised to reduce the allowable primary to secondary leakage through 
any one SG from 150 to 85 gallons per day and TS 5.6.7, ``Steam 
Generator Tube Inspection Program,'' is being revised to include 
additional reporting requirements.
    Date of issuance: September 5, 2014.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 277. A publicly-available version is in ADAMS under 
Accession No. ML14198A161; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. DPR-26: The amendment revised the 
Facility Operating License and the TSs.
    Date of initial notice in Federal Register: March 18, 2014 (79 FR 
15147). The supplemental letters provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 5, 2014.
    No significant hazards consideration comments received: No.

[[Page 58830]]

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 application for amendment: May 21, 2013, as supplemented by 
letter dated October 4, 2013.
    Brief description of amendment: The amendments revised the 
Technical Specifications (TSs) moderator temperature coefficient 
surveillance requirements associated with the implementation of Topical 
Report WCAP-16011-P-A, ``Startup Test Activity Reduction (STAR) 
Program,'' which describes the methods to be used for the 
implementation of reduction in the startup testing requirements. The 
changes are consistent with the NRC-approved Technical Specification 
Task Force (TSTF) Standard Technical Specifications change TSTF-486, 
Revision 2 as included in NUREG-1432, Revision 4.0, Standard Technical 
Specifications--Combustion Engineering Plants.
    The NRC staff published a notice of opportunity for comment in the 
Federal Register on July 27, 2007 (72 FR 41360), on possible amendments 
adopting TSTF-486 using the NRC's consolidated line-item improvement 
process for amending licensees' TSs, which included a model safety 
evaluation (SE) and model no significant hazards consideration (NSHC) 
determination. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on September 6, 2007 (72 FR 
51259), which included the resolution of public comments on the model 
SE and model NSHC determination. The licensee affirmed in its 
application dated May 21, 2013, that the proposed changes to the TSs 
satisfy the intent of TSTF-486.
    Date of issuance: September 16, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 219 and 168 (ADAMS Accession No. ML14218A180). 
Documents related to these amendments are provided in an SE enclosed 
with the amendments.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the License and TSs.
    Date of initial notice in Federal Register: July 23, 2013 (78 FR 
44173). The supplement dated October 4, 2013, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in an SE dated September 16, 2014.
    No significant hazards consideration comments received: No.
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: November 21, 2013, as supplemented by 
letters dated February 4 and April 1, 2014.
    Brief description of amendment: The amendments revised the date of 
cyber security plan (CSP) full implementation schedule (Milestone 8) 
and the existing license condition 2.H in the facility operating 
licenses NPF-87 and NPF-89 for CPNPP, Units 1 and 2, respectively. The 
CSP and the implementation schedule for CPNPP, Units 1 and 2, were 
previously approved by the NRC staff by letter dated July 26, 2011 
(ADAMS Accession No. ML111780745).
    Date of issuance: September 8, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: Unit 1--163; Unit 2--163. A publicly-available 
version is in ADAMS under Accession No. ML14183A342; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: April 8, 2014 (79 FR 
19399). The supplements dated February 4 and April 1, 2014, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 8, 2014.
    No significant hazards consideration comments received: No.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
1, Rockingham County, New Hampshire
    Date of amendment request: January 30, 2012, as supplemented by 
letters dated May 10, 2012, September 20, 2012, March 27, 2013, 
December 20, 2013, and January 29, 2014.
    Description of amendment request: The original application proposed 
revisions to the technical specifications (TSs) for new and spent fuel 
storage as a result of the new criticality analyses for the new fuel 
vault (NFV) and spent fuel pool (SFP). By letter dated December 20, 
2013 (ADAMS Accession No. ML13360A045), NextEra requested that the SFP 
and NFV be separated into two separate license amendment requests. This 
amendment revised the TSs related to spent fuel storage as a result of 
new criticality analyses for the SFP. The license amendment request for 
the NFV will be processed under TAC No. MF3283.
    Date of issuance: September 3, 2014.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 142. A publicly-available version is in ADAMS under 
Accession No. ML14184A795; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-86: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 14, 2012 (77 FR 
48559). The supplemental letters dated May 10, 2012, September 20, 
2012, March 27, 2013, December 20, 2013, and January 29, 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.
    Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated September 3, 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 4, 2014, as supplemented by the 
letter dated May 27, 2014.
    Brief description of amendment: The amendment revises Tier 2* 
information, incorporated into the VEGP Units 3 and 4 Updated Final 
Safety Analysis Report (UFSAR). Specifically, the amendment revises the 
details regarding the structural floor of the Auxiliary Building and 
its constructability. Notes are added to drawings in subsection 3H.5 of 
the UFSAR in order to clarify variations in detail design such as size

[[Page 58831]]

and spacing or reinforcement and spans of the noncritical sections of 
floors.
    Date of issuance: July 3, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 21. A publicly-available version is in ADAMS under 
Accession No. ML14150A133; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: April 29, 2014 (79 FR 
24025).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 3, 2014.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 22nd day of September 2014.

    For the Nuclear Regulatory Commission.
George A. Wilson,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2014-23015 Filed 9-29-14; 8:45 am]
BILLING CODE 7590-01-P