Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 13902-13920 [2015-05994]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 80, Number 51 (Tuesday, March 17, 2015)]
[Notices]
[Pages 13902-13920]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2015-05994]


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NUCLEAR REGULATORY COMMISSION

[NRC-2015-0055]


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 February 19, 2015 to March 4, 2015. The last

[[Page 13903]]

biweekly notice was published on March 3, 2015.

DATES: Comments must be filed by April 16, 2015. A request for a 
hearing must be filed by May 18, 2015.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2015-0055. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Beverly A. Clayton, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-3475, email: Beverly.Clayton@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2015-0055 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2015-0055.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2015-0055, facility name, unit 
number(s), application date, and subject in your comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at https://www.regulations.gov as well as entering the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Sec.  50.92 of title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.

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

    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Agency 
Rules of Practice and Procedure'' in 10 CFR part 2. Interested 
person(s) should consult a current copy of 10 CFR 2.309, which is 
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The 
NRC's regulations are accessible electronically from the NRC Library on 
the NRC's Web site at https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed by the above date, the Commission or a presiding officer 
designated by the Commission or by the Chief Administrative Judge of 
the Atomic Safety and Licensing Board Panel, will rule on the request 
and/or petition; and the Secretary or the Chief Administrative Judge of 
the Atomic Safety and Licensing Board will issue a notice of a hearing 
or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition

[[Page 13904]]

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 
ten 10 days prior to the filing deadline, the participant should 
contact the Office of the Secretary by email at hearing.docket@nrc.gov, 
or by telephone at 301-415-1677, to request (1) a digital 
identification (ID) certificate, which allows the participant (or its 
counsel or representative) to digitally sign documents and access the 
E-Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at https://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available

[[Page 13905]]

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 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Duke Energy Florida, Inc. (DEF), et al., Docket No. 50-302, Crystal 
River, Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida
    Date of amendment request: November 7, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14321A450.
    Description of amendment request: The amendment would reflect the 
transfer of ownership, held by eight minority co-owners, in CR-3 to 
DEF. The transfer of ownership will take place pursuant to the 
Settlement, Release and Acquisition Agreement, dated September 26, 
2014, wherein DEF will purchase the 6.52 percent combined ownership 
share in CR-3 held by these minority co-owners, leaving DEF and 
Seminole Electric Cooperative, Inc., as the remaining licensees for CR-
3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not involve a significant increase in 
the probability of any accident previously evaluated because no 
accident initiators or assumptions are affected. The proposed 
license transfers are administrative in nature and have no direct 
effect on any plant system, plant personnel qualifications, or the 
operation and maintenance of CR-3.
    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 create the possibility of a new or 
different kind of accident from any previously evaluated because no 
new accident initiators or assumptions are introduced by the 
proposed changes. The proposed license transfers are administrative 
in nature and have no direct effect on any plant system, plant 
personnel qualifications, or operation and maintenance of CR-3.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not involve a significant reduction in a 
margin of safety because the proposed changes do not involve changes 
to the initial conditions contributing to accident severity or 
consequences, or reduce response or mitigation capabilities. The 
proposed license transfers are administrative in nature and have no 
direct effect on any plant system, plant personnel qualifications, 
or operation and maintenance of CR-3.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, 550 South Tryon Street, 
Charlotte NC 28202.
    NRC Branch Chief: Douglas A. Broaddus.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating, Unit 2, Westchester County, New York
    Date of amendment request: December 9, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14353A015.
    Description of amendment request: The amendment would revise 
Technical Specification 5.5.14, ``Containment Leakage Rate Testing 
Program,'' to extend the frequency of the Containment Integrated Leak 
Rate Test or Type A Test from once every 10 years to once every 15 
years on a permanent basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment involves changes to the IP2 [Indian Point 
Unit No. 2] containment leakage rate testing program. The proposed 
amendment does not involve a physical change to the plant or a 
change in the manner in which the plant is operated or controlled. 
The primary containment function is to provide an essentially leak 
tight barrier against the uncontrolled release of radioactivity to 
the environment for

[[Page 13906]]

postulated accidents. As such, the containment itself and the 
testing requirements to periodically demonstrate the integrity of 
the containment exist to ensure the plant's ability to mitigate the 
consequences of an accident do not involve any accident precursors 
or initiators.
    Therefore, the probability of occurrence of an accident 
previously evaluated is not significantly increased by the proposed 
amendment.
    The proposed amendment adopts the NRC accepted guidelines of NEI 
94-01, Revision 2A, for development of the IP2 performance-based 
testing program for the Type A testing. Implementation of these 
guidelines continues to provide adequate assurance that during 
design basis accidents, the primary containment and its components 
would limit leakage rates to less than the values assumed in the 
plant safety analyses. The potential consequences of extending the 
ILRT [integrated leak rate test] interval to 15 years have been 
evaluated by analyzing the resulting changes in risk. The increase 
in risk in terms of person-rem per year within 50 miles resulting 
from design basis accidents was estimated to be acceptably small and 
determined to be within the guidelines published in RG 1.174. 
Additionally, the proposed change maintains defense-in-depth by 
preserving a reasonable balance among prevention of core damage, 
prevention of containment failure, and consequence mitigation. 
Entergy has determined that the increase in conditional containment 
failure probability due to the proposed change would be very small. 
Therefore, it is concluded that the proposed amendment does not 
significantly increase the consequences of an 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 2A, for the development of the IP2 performance-based 
leakage testing program, and establishes a 15-year interval for the 
performance of the containment ILRT. The containment and the testing 
requirements to periodically demonstrate the integrity of the 
containment exist to ensure the plant's ability to mitigate the 
consequences of an accident do not involve any accident precursors 
or initiators. The proposed change does not involve a physical 
change to the plant (i.e., no new or different type of equipment 
will be installed) or a change to the manner in which the plant is 
operated or controlled.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 2A, for the development of the IP2 performance-based 
leakage testing program, and establishes a 15-year interval for the 
performance of the containment ILRT. This amendment does not alter 
the manner in which safety limits, limiting safety system setpoints, 
or limiting conditions for operation are determined. The specific 
requirements and conditions of the containment leakage rate testing 
program, as defined in the TS [technical specifications], ensure 
that the degree of primary containment structural integrity and 
leak-tightness that is considered in the plant's safety analysis is 
maintained. The overall containment leakage rate limit specified by 
the TS is maintained, and the Type A containment leakage tests would 
be performed at the frequencies established in accordance with the 
NRC-accepted guidelines of NEI 94-01, Revision 2A with no change to 
the 60 month frequencies of Type B, and Type C tests.
    Containment inspections performed in accordance with other plant 
programs serve to provide a high degree of assurance that the 
containment would not degrade in a manner that is not detectable by 
an ILRT. A risk assessment using the current IP2 PSA [probabilistic 
safety assessment] model concluded that extending the ILRT test 
interval from ten years to 15 years results in a very small change 
to the risk profile.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Benjamin G. Beasley.
Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2 (NMP2), Oswego County, New York
    Date of amendment request: November 19, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14329A353.
    Description of amendment request: The proposed amendment would 
modify the Nine Mile Point (NMP) Nuclear Station, Unit 2 Technical 
Specifications (TS) by relocating specific surveillance frequencies to 
a licensee-controlled program with the adoption of Technical 
Specification Task Force (TSTF)-425, Revision 3, ``Relocate 
Surveillance Frequencies to Licensee Control--Risk Informed Technical 
Specification Task Force (RITSTF) Initiative 5b.'' The licensee's 
application dated November 19, 2014, Attachment 1, section 2.2, has 
identified some variations or deviations from the TSTF-425. 
Additionally, the change would add a new program, the Surveillance 
Frequency Control Program, to TS section 5, Administrative Controls. 
The NRC staff issued a notice of opportunity for comment in the Federal 
Register on December 5, 2008, 73 FR 74202, on possible amendments to 
revise the plant specific TS, to Relocate Surveillance Frequencies to 
Licensee Control--RITSTF Initiative 5b. The Notice included 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 July 6, 2009 (74 FR 31996). The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
November 19, 2014, which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes relocate the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program (SFCP). Surveillance 
frequencies are not an initiator to any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not significantly increased. The systems and components 
required by the technical specifications for which the surveillance 
frequencies are relocated are still required to be operable, meet 
the acceptance criteria for the surveillance requirements, and be 
capable of performing any mitigation function assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
changes. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of

[[Page 13907]]

equipment will be installed) or a change in the methods governing 
normal plant operation. In addition, the changes do not impose any 
new or different requirements. The changes do not alter assumptions 
made in the safety analysis. The proposed changes are consistent 
with the safety analysis assumptions and current plant operating 
practice.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, Exelon 
will perform a probabilistic risk evaluation using the guidance 
contained in NRC approved NEI 04-10, Rev. 1 in accordance with the 
TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable 
acceptance guidelines and methods for evaluating the risk increase 
of proposed changes to surveillance frequencies consistent with 
Regulatory Guide 1.177.
    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: J. Bradley Fewell, Senior Vice President, 
Regulatory Affairs, Nuclear, and General Counsel, Exelon Generation 
Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Benjamin G. Beasley.
Exelon Generation Company LLC (), Docket Nos. STN 50-456 and STN 50-
457, Braidwood Station, Units 1 and 2, Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 and 2, 
Ogle County, Illinois
    Date of amendment request: December 18, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14352A204.
    Description of amendment request: The proposed amendment would 
increase the voltage limit for the diesel generator (DG) full load 
rejection test specified by technical specification (TS) Surveillance 
Requirement (SR) 3.8.1.10. Additionally, the proposed amendment would 
add Note 3 to TS SR 3.8.1.10 for alignment with the Standard Technical 
Specifications documented in NUREG-1431, April 2012 (ADAMS Accession 
No. ML12100A222).
    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:

    EGC [Exelon Generation Company] has evaluated the proposed 
change for Braidwood Station and Byron Station, using the criteria 
in 10 CFR 50.92, and has determined that the proposed change does 
not involve a significant hazards consideration. The following 
information is provided to support a finding of no significant 
hazards consideration.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The DGs design function is to mitigate an accident and there are 
no analyzed scenarios where the DGs are initiators of any previously 
evaluated accident. Since DGs do not initiate accidents, this change 
does not increase the probability of occurrence of a previously 
evaluated accident. The proposed change to the testing approach of 
the DGs is consistent with the original design of the DGs. The 
proposed change is in accordance with RG [Regulatory Guide] 1.9 
Revision 3, and this change to the testing approach does not impact 
the DGs ability to mitigate accidents. The DGs will continue to 
operate within the parameters and conditions assumed within the 
accident analysis. This change does not result in an increase in the 
likelihood of malfunction of the DGs or their supported equipment. 
Since the DGs will continue to perform its required function, there 
is no increase in the consequences of previously evaluated 
accidents.
    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 amendment does not change the DGs operation or 
ability to perform its design function. The proposed change to TS SR 
3.8.1.10 at increased voltage will ensure the DGs ability to perform 
at rated power factor while meeting its requirements. The change to 
TS SR 3.8.1.10 does not result in DG operation that would create a 
new failure mode of the DGs that could create a new initiator of an 
accident. This is because the DGs ability to perform its design 
function is maintained in the same manner as originally designed. 
The proposed change does not change the single failure capabilities 
of the electrical power system or create a potential for loss of 
power since the design operation of the DGs is maintained.
    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 margin of safety is established through the design of the 
plant structures, systems, and components, the parameters within 
which the plant is operated, and the setpoints for the actuation of 
equipment relied upon to respond to an event. The proposed change 
does not modify the safety limits or setpoints at which protective 
actions are initiated. The proposed change increases the voltage 
limit for the DG full load rejection test which results in new test 
acceptance criterion that is more restrictive than the existing 
acceptance criteria. The proposed change ensures the availability 
and operability of safety-related DGs.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above evaluation, EGC concludes that the proposed 
amendment presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92, paragraph (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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Travis L. Tate.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant (PNPP), Unit 1, Perry, Ohio
    Date of amendment request: November 24, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14328A665.
    Description of amendment request: The proposed amendment is 
intended to revise the battery capacity testing surveillance 
requirements in the technical specifications to reflect test 
requirements when the battery is near end of life.
    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:


[[Page 13908]]


    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not change the design function of 
the Class 1 E divisional battery systems and does not change the way 
the plant is maintained or operated when performing battery 
surveillance testing. The proposed amendment does not affect any 
accident mitigating feature or increase the likelihood of 
malfunction for plant structures, systems and components.
    The proposed amendment does not affect the operability 
requirements of the Class 1 E divisional battery systems. 
Verification of operating the plant within prescribed limits will 
continue to be performed, as currently required. Compliance with and 
continued verification of the prescribed limits support the 
capability of the Class 1 E divisional battery systems to perform 
their required design functions during all plant operating, 
accident, and station blackout conditions, consistent with the plant 
safety analyses.
    The proposed amendment will not change any of the analyses 
associated with the PNPP Updated Safety Analysis Report Chapter 15 
accidents because plant operation, plant structures, systems, 
components, accident initiators, and accident mitigation functions 
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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not change the design function of 
the Class 1 E divisional battery systems, and does not change the 
way the plant is operated or maintained. The proposed amendment does 
not create a credible failure mechanism, malfunction or accident 
initiator not already considered in the design and licensing basis.
    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.
    Safety margins are applied to design and licensing basis 
functions and to the controlling values of parameters to account for 
various uncertainties and to avoid exceeding regulatory or licensing 
limits. The proposed amendment does not involve a physical change to 
the plant, does not change methods of plant operation within 
prescribed limits, or affect design and licensing basis functions or 
controlling values of parameters for plant systems, structures, and 
components.
    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, 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., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida
    Date of amendment request: August 7, 2014 (ADAMS Accession No. 
ML14225A630).
    Description of amendment request: The amendment would revise the 
Technical Specifications to add a short Allowed Outage Time to restore 
an inoperable system for conditions under which the existing 
specifications require a plant shutdown. The proposed amendment is 
consistent with an NRC-approved change identified as Technical 
Specifications Task Force (TSTF) Traveler TSTF-426, Revision 5, 
``Revise or Add Actions to Preclude Entry into LCO [Limiting Condition 
for Operation] 3.0.3--RITSTF [Risk-Informed TSTF] Initiatives 6b & 6c'' 
(see 78 FR 32476, May 30, 2013). The Allowed Outage Time would be added 
to specifications governing the boron injection flow paths of the 
reactivity control systems, pressurizer heaters, containment spray 
trains, shield building ventilation systems, and control room emergency 
air cleanup systems.
    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 reproduced 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 provides a short Allowed Outage Time to 
restore an inoperable system for conditions under which the existing 
Technical Specifications require a plant shutdown to begin within 
one hour in accordance with Limiting Condition for Operation (LCO) 
3.0.3. Entering into Technical Specification Actions is not an 
initiator of any accident previously evaluated. As a result, the 
probability of an accident previously evaluated is not significantly 
increased. The consequences of any accident previously evaluated 
that may occur during the proposed Allowed Outage Times are no 
different from the consequences of the same accident during the 
existing one-hour allowance. As a result, 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.
    No new or different accidents [would] result from utilizing the 
proposed change. The changes [to the TSs] do 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 changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
[any] safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change increases the time the plant may operate 
without the ability to perform an assumed safety function. The 
analyses in [the NRC-approved topical report] WCAP-16125-NP-A, 
``Justification for Risk-Informed Modifications to Selected 
Technical Specifications for Conditions Leading to Exigent Plant 
Shutdown,'' Revision 2, August 2010, demonstrated that there is an 
acceptably small increase in risk due to a limited period of 
continued operation in these conditions and that this risk is 
balanced by avoiding the risks associated with a plant shutdown. As 
a result, the change to the margin of safety provided by requiring a 
plant shutdown within one hour is not significant.
    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 determines 
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, 
the NRC staff proposes to determine that the proposed amendment 
involves no significant hazards consideration.
    Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB, 
Juno Beach, FL 33408-0420.
    NRC Branch Chief: Shana R. Helton.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida
    Date of amendment request: December 5, 2014 (ADAMS Accession No. 
ML14353A016).
    Description of amendment request: The proposed amendment will 
modify the Technical Specification (TS)

[[Page 13909]]

requirements related to Completion Times for Required Actions to 
provide the option to calculate longer, risk-informed Completion Times. 
The proposed amendment will also add a new program, the Risk Informed 
Completion Time Program, to TS section 6.0, ``Administrative 
Controls.'' The methodology for using the Risk Informed Completion Time 
Program is described in Nuclear Energy Institute topical report NEI 06-
09, ``Risk-Informed Technical Specifications Initiative 4b, Risk-
Managed Technical Specifications (RMTS) Guidelines,'' Revision 0-A, 
which was approved by the NRC on May 17, 2007. The proposed amendment 
is consistent with the NRC-approved industry-proposed Technical 
Specification Task Force-505, Revision 1, ``Provide Risk-Informed 
Extended Completion Times--RITSTF Initiative 4b.''
    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 reproduced below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change permits the extension of Completion Times 
provided the associated risk is assessed and managed in accordance 
with the NRC[-]approved Risk Informed Completion Time Program. The 
proposed change does not involve a significant increase in the 
probability of an accident previously evaluated because the change 
involves no change to the plant or its modes of operation. The 
proposed change does not increase the consequences of an accident 
because the design-basis mitigation function of the affected systems 
is not changed and the consequences of an accident [occurring] 
during the extended Completion Time are no different from those 
[occurring] during the existing Completion Time.
    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 does not change the design, configuration, 
or method of operation of the plant. The proposed change does not 
involve a physical alteration of the plant (no new or different kind 
of equipment will be installed).
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change permits the extension of Completion Times 
provided risk is assessed and managed in accordance with the NRC[-
]approved Risk Informed Completion Time Program. The proposed change 
implements a risk-informed configuration management program to 
assure that adequate margins of safety are maintained. Application 
of these new specifications and the configuration management program 
considers cumulative effects of multiple systems or components being 
out of service and does so more effectively than the current TS.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and determines 
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, 
the NRC staff proposes to determine that the proposed amendment 
involves no significant hazards consideration.
    Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB, 
Juno Beach, FL 33408-0420.
    NRC Branch Chief: Shana R. Helton.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
    Date of amendment request: December 17, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14356A022.
    Description of amendment request: The proposed amendment would 
amend the Appendix A technical specifications to Facility Operating 
Licenses DPR-58 and DPR-74, to modify the notes to TS 3.8.1, ``AC 
Sources--Operating,'' to allow surveillance testing of the onsite 
standby emergency diesel generators (DGs) during modes in which it is 
currently prohibited. Specifically, the license amendment request 
proposes removing the mode restrictions for the following Surveillance 
Requirements (SRs): 3.8.1.10 (DG single largest load rejection test), 
3.8.1.11 (DG full load rejection test), and 3.8.1.15 (DG endurance 
run).
    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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The design of plant equipment is not being modified by the 
proposed changes. In addition, the DGs and their associated 
emergency loads are accident mitigating features. As such, testing 
of the DGs themselves is not associated with any potential accident-
initiating mechanism.
    Therefore, there will be no significant impact on any accident 
probabilities by the approval of the requested changes.
    The changes include an increase in the time that a DG under test 
will be paralleled to the grid while the unit is in Modes 1 or 2. As 
such, the ability of the tested DG to respond to a DBA [design-basis 
accident] could be minimally adversely impacted by the proposed 
changes. However, the impacts are not considered significant based, 
in part, on the ability of the remaining DG to mitigate a DBA or 
provide safe shutdown. Experience shows that testing for these SRs 
typically does not perturb the electrical distribution system. In 
addition, operating experience supports the conclusion that the 
proposed changes do not involve any significant increases in the 
likelihood of a safety-related bus blackout or damage to plant 
loads.
    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 capability to synchronize a DG to the offsite source (via 
the associated plant bus) and test the DG in such a configuration is 
a design feature of the DGs, including the test mode override in 
response to a safety injection signal. Paralleling the DG for longer 
periods of time during plant operation may slightly increase the 
probability of incurring an adverse effect from the offsite source, 
but this increase in probability is judged to be still quite small 
and such a possibility is not a new or previously unrecognized 
consideration.
    The proposed change does not introduce a new mode of plant 
operation and does not involve physical modification to the plant. 
The change does not introduce new accident initiators or impact 
assumptions made in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not exceed or alter a design basis or 
safety limit, so there is no significant reduction in the margin of 
safety. The margin of safety is related to the confidence in the 
ability of the fission product barriers to perform their design 
functions during and following an accident situation. These barriers 
include the fuel cladding, the reactor coolant system, and the 
containment system. The proposed changes do not directly affect 
these barriers, nor do they involve any significantly adverse impact 
on the DGs which serve to support these barriers in the event of an 
accident concurrent with a LOOP [loss of offsight power]. The 
proposed changes to the testing requirements for the plant DGs do 
not affect the OPERABILITY requirements for the DGs,

[[Page 13910]]

as verification of such OPERABILITY will continue to be performed as 
required (except during different allowed modes). The changes have 
an insignificant impact on DG availability, as the DGs remain 
available to perform their required function of providing emergency 
power to plant equipment that supports or constitutes the fission 
product barriers. Only one DG is to be tested at a time, so that the 
remaining DG will be available to safety shut down the plant if 
required. Consequently, performance of the fission product barriers 
will not be impacted by implementation of the proposed amendment.
    In addition, the proposed changes involve no changes to 
setpoints or limits established or assumed by the accident analysis.
    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: Robert B. Haemer, Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: David L. Pelton.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska
    Date of amendment request: January 15, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15021A127.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to add a limiting condition 
for operation, applicability, required actions, completion times, and 
surveillance requirements for the residual heat removal (RHR) 
containment spray system consistent with the guidance in NUREG-1433, 
Revision 4, ``Standard Technical Specifications General Electric BWR 
[Boiling Water Reactor]/4 Plants,'' dated April 2012 (ADAMS Accession 
No. ML12104A192). New TS section 3.6.1.9, ``Residual Heat Removal (RHR) 
Containment Spray,'' would be added to reflect the reliance on 
containment spray to maintain the drywell within design temperature 
limits during a small steam line break. In addition, the ``Drywell 
Pressure--High'' function that serves as an interlock permissive to 
allow RHR containment spray mode alignment would be relocated from the 
Technical Requirements Manual (TRM) to TS 3.3.5.1, ``Emergency Core 
Cooling System (ECCS) Instrumentation.''
    The requirements for the RHR containment spray function and 
``Drywell Pressure--High'' function are currently contained in TRM 
sections T3.6.1, ``RHR Containment Spray,'' and T3.3.2, ``ECCS and 
Reactor Core Isolation Cooling Instrumentation,'' respectively. These 
TRM sections established specific guidance and criteria related to the 
applicability, operation, and testing for the RHR containment spray 
system. The TRM requirements for the RHR containment spray system would 
be removed once the TS requirements are approved.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to establish the RHR Containment Spray 
requirement in TS does not introduce new equipment or new equipment 
operating modes, nor do the proposed changes alter existing system 
relationships. The proposed change does not affect plant operation, 
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 the RHR system. The RHR 
system will continue to function as designed in all modes of 
operation, including the Containment Spray function. There are no 
significant changes to procedures or training related to the 
operation of the Containment Spray function. Primary containment 
integrity is not adversely impacted and radiological consequences 
from the accidents analyzed in the Updated Safety Analysis Report 
(USAR) are not increased. Containment parameters are not increased 
beyond those previously evaluated and the potential for failure of 
the containment is not increased.
    There is no adverse impact on systems designed to mitigate the 
consequences of accidents. The proposed change does not increase 
system or component pressures, temperatures, and flowrates for 
systems designed to prevent accidents or mitigate the consequences 
of an accident. Since these conditions do not change, the likelihood 
of failure of SSC is not increased.
    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 to establish the RHR Containment Spray 
requirement in TS does not alter the design function or operation of 
any SSC. The Containment system will continue to function as 
designed in all modes of operation, including RHR Containment Spray 
function. There is no new system component being installed, no new 
construction, and no performance of a new test or maintenance 
function. The proposed TS change does not create the possibility of 
a new credible failure mechanism or malfunction. The proposed change 
does not modify the design function or operation of any SSC. The 
proposed change does not introduce new accident initiators. Primary 
containment integrity is not adversely impacted and radiological 
consequences from the accident analyzed in the USAR are not 
increased. Containment parameters are not increased beyond those 
previously evaluated and the potential for failure of the 
containment is not increased. The proposed change does not increase 
system or component pressures, temperatures, and flowrates for 
systems designed to prevent accidents or mitigate the consequences 
of an accident. Since these conditions do not change, the likelihood 
of failure of an SSC is not increased.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not increase system or component 
pressures, temperatures, and flowrates for systems designed to 
prevent accidents or mitigate the consequences of an accident. 
Containment parameters are not increased beyond those previously 
evaluated and the potential for failure of the containment is not 
increased.
    The proposed change to establish the RHR Containment Spray 
requirement in TS is needed in order to reflect the current safety 
function of Containment Spray related to the small steam line break 
accident. The proposed change does not exceed or alter a design 
basis or a safety limit parameter that is described in the USAR.
    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.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota
    Date of amendment request: February 20, 2013, as supplemented by 
letters dated June 25, 2013; September 15,

[[Page 13911]]

2014; and February 26, 2015. Publicly-available versions are in ADAMS 
under Accession Nos. ML13053A199, ML13178A024, ML14258A089, and 
ML15057A480, respectively.
    Brief description of amendment request: The proposed amendments 
would remove the technical specification (TS) 3.5.3 ``ECCS [Emergency 
Core Cooling System]-Shutdown,'' Limiting Condition for Operation (LCO) 
Note 1 to eliminate information to the plant operators that could cause 
non-conservative operation, and would revise the LCO Applicability 
statement to apply to all of Mode 4.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration, 
which the Commission previously issued in the Federal Register on 
August 20, 2013 (78 FR 51229). The licensee revised its analysis of the 
issue of no significant hazards consideration, which is presented 
below, to consider expansion of the scope of the amendments by revising 
the LCO Applicability statement to include all of Mode 4.

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment request proposes to revise the Technical 
Specification for ECCS operability requirements in Mode 4 by 
removing the LCO Note which allows the RHR [residual heat removal] 
subsystem to be considered operable for ECCS when aligned for 
shutdown cooling and revising the Applicability statement to include 
all of Mode 4. These changes will require one train of RHR to be 
aligned for ECCS operation throughout Mode 4.
    The proposed changes do not affect the ECCS and RHR subsystem 
design, the interfaces between the RHR subsystem and other plant 
systems' operating functions, or the reliability of the RHR 
subsystem. The proposed changes do not change or impact the 
initiators and assumptions of the analyzed accidents. Therefore, the 
ECCS and RHR subsystems will be capable of performing their accident 
mitigation functions, and the proposed TS changes do not involve an 
increase in the probability of an accident.
    The proposed TS changes will require that one train of RHR is 
aligned for ECCS operation during Mode 4 which assures that one 
train of ECCS is operable to mitigate the consequences of a loss of 
coolant accident. Thus the proposed TS changes do not involve a 
significant increase in the consequences of an accident.
    Therefore, the proposed Technical Specification changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This license amendment request proposes to revise the Technical 
Specification for ECCS operability requirements in Mode 4 by 
removing the LCO Note which allows the RHR subsystem to be 
considered operable for ECCS when aligned for shutdown cooling and 
revising the Applicability statement to include all of Mode 4. These 
changes will require one train of RHR to be aligned for ECCS 
operation throughout Mode 4.
    The proposed Technical Specification changes involve changes to 
when system trains are operated, but they do not change any system 
functions or maintenance activities. The changes do not involve 
physical alteration of the plant, that is, no new or different type 
of equipment will be installed. The changes do not alter assumptions 
made in the safety analyses but ensure that one train of ECCS is 
operable to mitigate the consequences of a loss of coolant accident. 
These changes do not create new failure modes or mechanisms which 
are not identifiable during testing and no new accident precursors 
are generated.
    Therefore, the proposed Technical Specification changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This license amendment request proposes to revise the Technical 
Specification [TS] for ECCS operability requirements in Mode 4 by 
removing the LCO Note which allows the RHR subsystem to be 
considered operable for ECCS when aligned for shutdown cooling and 
revising the Applicability statement to include all of Mode 4. These 
changes will require one train of RHR to be aligned for ECCS 
operation throughout Mode 4.
    This license amendment proposes Technical Specification changes 
which assure that the ECCS--Shutdown TS LCO requirements are met if 
a Mode 4 LOCA were to occur. With these changes, other TS 
requirements for shutdown cooling in Mode 4 will continue to be met. 
Based on review of plant operating experience, there is no 
discernable change in cooldown rates when utilizing a single train 
of RHR for shutdown cooling. Thus, no margin of safety is reduced as 
part of this change.
    Therefore, the proposed Technical Specification 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 requests involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: David L. Pelton.
Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-321 and 
50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, GA
    Date of amendment request: January 13, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15014A411.
    Description of amendment request: The licensee proposes to adopt 
Technical Specification Task Force (TSTF) change number 523, revision 
2, ``Generic Letter 2008-01, Managing Gas Accumulation,'' for the Hatch 
Nuclear Plant, Unit 1 and 2, technical specifications (TS). The 
proposed change would revise or add Surveillance Requirements to verify 
that the system locations susceptible to gas accumulation are 
sufficiently filled with water and to provide allowances which permit 
performance of the verification.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises or adds Surveillance Requirement(s) 
(SRs) that require verification that the Emergency Core Cooling 
System (ECCS), the Residual Heat Removal (RHR) System, the RHR 
Shutdown Cooling (SDC) System, the Containment Spray (CS) 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, the RHR, the RHR SDC System, the CS 
System, and the RCIC System are not rendered inoperable due to 
accumulated gas and to

[[Page 13912]]

provide allowances which permit performance of the revised 
verification. The proposed change does not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. In addition, the proposed change does not impose any new 
or different requirements that could initiate an accident. The 
proposed change does not alter assumptions made in the safety 
analysis and is consistent with the safety analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, the RHR, RHR SDC System, the CS System, 
and the RCIC System are not rendered inoperable due to accumulated 
gas and to provide allowances which permit performance of the 
revised verification. The proposed change adds new requirements to 
manage gas accumulation in order to ensure the subject systems are 
capable of performing their assumed safety functions. The proposed 
SRs are more comprehensive than the current SRs and will ensure that 
the assumptions of the safety analysis are protected. The proposed 
change does not adversely affect any current plant safety margins or 
the reliability of the equipment assumed in the safety analysis. 
Therefore, there are no changes being made to any safety analysis 
assumptions, safety limits or limiting safety system settings that 
would adversely affect plant safety as a result of the proposed 
change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed change 
presents no 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: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, 40 Inverness Center 
Parkway, Birmingham, AL 35201.
    NRC Branch Chief: Robert J. Pascarelli.
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: January 27, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15028A537.
    Description of amendment request: The proposed change, if approved, 
would revise, in part, the description and scope of human factors 
engineering (HFE) operational sequence analysis (OSA) task and delete a 
reference to document WCAP-15847, which are both identified as Tier 2* 
information in the Updated Final Safety Analysis Report (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 proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed deletion of WCAP-15847 removes obsolete and 
superseded procedures from the licensing basis. The amendment of the 
operational sequence analysis (OSA) task alters the automatic 
depressurization system (ADS) testing from Mode 1 to Mode 5. The 
proposed changes to the procedures do not involve any accident 
initiating component/system failure or event, and the change to the 
ADS testing mode helps prevent accidents that would occur if the 
tests were performed in Mode 1. Thus, the probabilities of the 
accidents previously evaluated are not affected. The affected 
procedures and requirements do not adversely affect or interact with 
safety-related equipment or a radioactive material barrier, and this 
activity does not involve the containment of radioactive material. 
Thus, the proposed changes would not affect any safety-related 
accident mitigating function. The radioactive material source terms 
and release paths used in the safety analyses are unchanged, thus 
the radiological releases in the Updated Final Safety Analysis 
Report 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.
    Removing WCAP-15847 from the UFSAR and amending the OSA task 
regarding ADS valve testing does not adversely affect the design or 
operation of safety-related equipment or equipment whose failure 
could initiate an accident other than what is already described in 
the licensing basis. These changes do not adversely affect safety-
related equipment or fission product barriers. No safety analysis or 
design basis acceptance limit/criterion is challenged or exceeded by 
the requested change.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to remove WCAP-15847 from the UFSAR and 
amend the OSA task do not adversely affect any safety-related 
equipment, design code compliance, design function, design analysis, 
safety analysis input or result, or design/safety margin because 
NQA-1 requirements are maintained in other Westinghouse procedures 
and testing of the ADS valves is still performed. No safety analysis 
or design basis acceptance limit/criterion is challenged or exceeded 
by the proposed changes, thus no margin of safety is reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW, Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, 
Georgia
    Date of amendment request: January 30, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15030A505.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-91 and NPF-92 for the Vogtle Electric 
Generating Plant (VEGP) Units 3 and 4. The requested amendment proposes 
changes to Tier 2* information contained within the Human Factors 
Engineering Design Verification, Task Support Verification and 
Integrated System Validation (ISV) plans. These documents are 
incorporated by reference into the VEGP Units 3 and 4 Updated Final 
Safety Analysis Report, and will additionally require changes to be 
made to affected 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?

[[Page 13913]]

    Response: No.
    The proposed amendment includes changes to Integrated System 
Validation (ISV) activities, which are performed on the AP1000 plant 
simulator to validate the adequacy of the AP1000 human system 
interface design and confirm that it meets human factors engineering 
principles. The proposed changes involve administrative details 
related to performance of the ISV, and no plant hardware or 
equipment is affected whose failure could initiate an accident, or 
that interfaces with a component that could initiate an accident, or 
that contains radioactive material. Therefore, these changes have no 
effect on any accident initiator in the Updated Final Safety 
Analysis Report (UFSAR), nor do they affect the radioactive material 
releases in the UFSAR accident analysis.
    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 proposed amendment includes changes to ISV activities, which 
are performed on the AP1000 plant simulator to validate the adequacy 
of the AP1000 human system interface design and confirm that it 
meets human factors engineering principles. The proposed changes 
involve administrative details related to performance of the ISV, 
and no plant hardware or equipment is affected whose failure could 
initiate an accident, or that interfaces with a component that could 
initiate an accident, or that contains radioactive material. 
Although the ISV may identify a need to initiate changes to add, 
modify, or remove plant structures, systems, or components, these 
changes will not be made directly as part of the ISV.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment includes changes to ISV activities, which 
are performed on the AP1000 plant simulator to validate the adequacy 
of the AP1000 human system interface design and confirm that it 
meets human factors engineering principles. The proposed changes 
involve administrative details related to performance of the ISV, 
and do not affect any safety-related equipment, design code 
compliance, design function, design analysis, safety analysis input 
or result, or design/safety margin. No safety analysis or design 
basis acceptance limit/criterion is challenged or exceeded by the 
proposed changes, thus no margin of safety is reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia
    Date of amendment request: September 13, 2012, as supplemented 
August 2, 2013, July 3, July 17, November 11, and December 12, 2014. 
Publicly-available versions are in ADAMS under Accession Nos. 
ML12258A055, ML13217A072, ML14189A554, ML14198A574, ML14315A051 and 
ML14346A643, respectively.
    Description of amendment request: The proposed amendment would 
modify certain Technical Specification (TS) requirements related to 
Completion Times for Required Actions to provide the option to 
calculate a longer, risk-informed Completion Time. The allowance will 
be described in a new program, ``Risk Informed Completion Time Program 
(RICT),'' to be approved by NRC and to be added to Chapter 5, 
``Administrative Controls,'' of the Technical Specifications. The 
methodology for using the RICT Program is described in an industry 
document NEI 06-09, ``Risk-Informed Technical Specifications Initiative 
4b, Risk-Managed Technical Specifications (RMTS) Guidelines,'' which 
was approved by the Nuclear Regulatory Commission (NRC) on May 17, 
2007. Adherence to NEI 06-09 is required by the proposed RICT Program. 
The proposed amendment is also consistent with the methodologies 
presented in an industry initiative identified as TSTF-505, Revision 1, 
``Provide Risk-Informed Extended Completion Times--RITSTF Initiative 
4b.'' Although the proposed amendment is consistent with TSTF-505, the 
licensee is not proposing adoption of TSTF-505 with this proposed 
amendment; the proposed amendment is a site-specific action.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change permits the extension of Completion Times 
provided risk is assessed and managed within the Risk Informed 
Completion Time Program. The proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated because the changes involve no change to the plant or its 
modes of operation. This proposed change does not increase the 
consequences of an accident because the design-basis mitigation 
function of the affected systems is not changed and the consequences 
of an accident during the extended Completion Time are no different 
from those during the existing Completion Time.
    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 does not change the design, configuration, 
or method of operation of the plant. The proposed change does not 
involve a physical alteration of the plant (no new or different kind 
of equipment will be installed).
    Therefore, the proposed 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 change permits the extension of Completion Times 
provided risk is assessed and managed within the Risk Informed 
Completion Time Program. The proposed change implements a risk-
informed configuration management program to assure that adequate 
margins of safety are maintained. Application of these new 
specifications and the configuration management program considers 
cumulative effects of multiple systems or components being out of 
service and does so more effectively than the current TS.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendment involve no significant hazards consideration.
    Attorney for licensee: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, 40 Inverness Center 
Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Robert J. Pascarelli.

[[Page 13914]]

Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328, 
Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
    Date of amendment request: December 2, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14339A539.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 6.8.4.h, ``Containment Leakage Rate 
Testing Program,'' by adopting Nuclear Energy Institute (NEI) 94-01, 
Revision 3-A, ``Industry Guideline for Implementing Performance-Based 
Option of 10 CFR part 50, Appendix J,'' as the implementation document 
for the performance-based Option B of 10 CFR part 50, Appendix J. The 
proposed changes would permanently extend the Type A containment 
integrated leak rate testing (ILRT) interval from 10 years to 15 years, 
and the Type C local leakage rate testing (LLRT) intervals from 60 
months to 75 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    1. Does the proposed amendment involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    Response: No.
    The proposed revision to TS 6.8.4.h changes the testing period 
to a permanent 15-year interval for Type A testing (10 CFR part 50, 
Appendix J, Option B, ILRT) and a 75-month interval for Type C 
testing (10 CFR part 50, Appendix J, Option B, LLRT). The current 
type A test interval of 10 years would be extended to 15 years from 
the last Type A test. The proposed extension to Type A testing does 
not involve a significant increase in the consequences of an 
accident because research documented in NUREG-1493, ``Performance-
Based Containment System Leakage Testing Requirements [sic] 
[Performance-Based Containment Leak-Test Program],'' September 1995, 
has found that, generically, very few potential containment leakage 
paths are not identified by Type B and C tests. NUREG-1493 concluded 
that reducing the Type A testing frequency to one per twenty years 
was found to lead to an imperceptible increase in risk. A high 
degree of assurance is provided through testing and inspection that 
the containment will not degrade in a manner detectable only by Type 
A testing. The last Type A test (performed October 27, 2007 for SQN, 
Unit 1 and December 30, 2006 for SQN, Unit 2) shows leakage to be 
below acceptance criteria, indicating a very leak tight containment. 
Inspections required by the ASME [American Society of Mechanical 
Engineers] Code section Xl (subsections IWE and IWL) and Maintenance 
Rule monitoring (10 CFR 50.65, ``Requirements for Monitoring the 
Effectiveness of Maintenance at Nuclear Power Plants''), are 
performed in order to identify indications of containment 
degradation that could affect that leak tightness. Types B and C 
testing required by TSs will identify any containment opening such 
as valves that would otherwise be detected by the Type A tests. 
These factors show that a Type A test interval extension will not 
represent a significant increase in the consequences of an accident.
    The proposed amendment involves changes to the SQN, Units 1 and 
2, 10 CFR 50 Appendix J Testing Program Plan. The proposed amendment 
does not involve a physical change to the plant or a change in the 
manner in which the units are operated or controlled. The primary 
containment function is to provide an essentially leak tight barrier 
against the uncontrolled release of radioactivity to the environment 
for postulated accidents. As such, the containment itself and the 
testing requirements to periodically demonstrate the integrity of 
the containment exist to ensure the plant's ability to mitigate the 
consequences of an accident, and do not involve any accident 
precursors or initiators.
    Therefore, the probability of occurrence of an accident 
previously evaluated is not significantly increased by the proposed 
amendment.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, for development of the SQN, Units 1 and 2, 
performance-based leakage testing program. Implementation of these 
guidelines continues to provide adequate assurance that during 
design basis accidents, the primary containment and its components 
will limit leakage rates to less than the values assumed in the 
plant safety analyses. The potential consequences of extending the 
ILRT interval from 10 years to 15 years have been evaluated by 
analyzing the resulting changes in risk. The increase in risk in 
terms of person-rem per year resulting from design basis accidents 
was estimated to be very small, and the increase in the LERF [large 
early release frequency] resulting from the proposed change was 
determined to be within the guidelines published in NRC RG 
[Regulatory Guide] 1.174. Additionally, the proposed change 
maintains defense-in-depth by preserving a reasonable balance among 
prevention of core damage, prevention of containment failure, and 
consequence mitigation. TVA has determined that the increase in CCFP 
[conditional containment failure probability] due to the proposed 
change would be very small.
    Based on the above discussions, the proposed changes do 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 proposed revision to TS 6.8.4.h changes the testing period 
to a permanent 15-year interval for Type A testing (10 CFR part 50, 
Appendix J, Option B, ILRT) and a 75-month interval for Type C 
testing (10 CFR part 50, Appendix J, Option B, LLRT). The current 
test interval of 10 years, based on past performance, would be 
extended to 15 years from the last Type A test (performed October 
27, 2007 for SQN, Unit 1 and December 30, 2006 for SQN, Unit 2). The 
proposed extension to Type A and Type C test intervals does not 
create the possibility of a new or different type of accident 
because there are no physical changes being made to the plant and 
there are no changes to the operation of the plant that could 
introduce a new failure mode creating an accident or affecting the 
mitigation of an accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed revision to TS 6.8.4.h changes the testing period 
to a permanent 15-year interval for Type A testing (10 CFR part 50, 
Appendix J, Option B, ILRT) and a 75-month interval for Type C 
testing (10 CFR part 50, Appendix J, Option B, LLRT). The current 
test interval of 10 years, based on past performance, would be 
extended to 15 years from the last Type A test (performed October 
27, 2007 for SQN, Unit 1 and December 30, 2006 for SQN, Unit 2). The 
proposed extension to Type A testing will not significantly reduce 
the margin of safety. NUREG-1493, ``Performance-Based Containment 
System Leakage Testing Requirements [sic] [Performance-Based 
Containment Leak-Test Program],'' September 1995, generic study of 
the effects of extending containment leakage testing, found that a 
20-year extension to Type A leakage testing resulted in an 
imperceptible increase in risk to the public. NUREG-1493 found that, 
generically, the design containment leakage rate contributes about 
0.1% to the individual risk and that the decrease in Type A testing 
frequency would have a minimal effect on this risk since 95% of the 
potential leakage paths are detected by Type C testing. Regular 
inspections required by the ASME Code section Xl (subsections IWE 
and IWL) and maintenance rule monitoring (10 CFR 50.65, 
``Requirements for Monitoring the Effectiveness of Maintenance at 
Nuclear Power Plants'') will further reduce the risk of a 
containment leakage path going undetected.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, for development of the SQN, Units 1 and 2, 
performance-based leakage testing program, and establishes a 15-year 
interval for the performance of the primary containment ILRT and a 
75-month interval for Type C testing. The amendment does not alter 
the manner in which safety limits, limiting safety system setpoints, 
or limiting conditions for operation are determined. The specific 
requirements and conditions of the 10 CFR part 50, Appendix J 
Testing Program Plan, as defined in the TS, ensure that the degree 
of primary containment structural integrity and leak-tightness that 
is considered in the plant safety analyses is maintained. The 
overall containment leakage rate limit specified by the TS is 
maintained, and the

[[Page 13915]]

Type A, B, and C containment leakage tests will continue to be 
performed at the frequencies established in accordance with the NRC-
accepted guidelines of NEI 94-01, Revision 3-A.
    Containment inspections performed in accordance with other plant 
programs serve to provide a high degree of assurance that the 
containment will not degrade in a manner that is detectable only by 
an ILRT. This ensures that evidence of containment structural 
degradation is identified in a timely manner. Furthermore, a risk 
assessment using the current SQN, Units 1 and 2, PRA model concluded 
that extending the ILRT test interval from 10 years to 15 years 
results in a very small change to the SQN, Units 1 and 2, risk 
profile.
    Accordingly, 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Branch Chief: Shana R. Helton.

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit 2, Westchester County, New York
    Date of amendment request: February 12, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15044A471.
    Brief description of amendment request: The proposed amendment 
would allow a revision to the acceptance criteria for the Surveillance 
Requirement 3.1.4.2 for Control Rod G-3. During the last two 
performances of this Surveillance on September 18, 2014, and December 
11, 2014, Control Rod G-3 misalignment occurred with Shutdown Bank B 
group movement as displayed by Individual Rod Position Indication and 
Plant Instrument Computer System. The proposed change is to defer 
subsequent testing of the Control Rod G-3 until repaired during the 
next refuel outage (March 2016) or forced outage long enough to repair 
the Control Rod.
    Date of publication of individual notice in Federal Register: March 
2, 2015 (80 FR 11236).
    Expiration date of individual notice: April 1, 2015 (public 
comments); May 1, 2015 (hearing requests).
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323 for 
Diablo Canyon Nuclear Power Plant (DCPP), Units 1 and 2, Docket No. 72-
26 for Diablo Canyon Independent Spent Fuel Storage Installation 
(ISFSI), San Luis Obispo County, California
    Date of amendment request: September 24, 2013, as supplemented by 
letters dated December 18, 2013 (security-related), and May 15, 2014. 
Publicly-available versions of the letters dated September 24, 2013, 
and May 15, 2014, are in ADAMS under Accession Nos. ML13268A398 and 
ML14135A379, respectively.
    Brief description of amendment request: The proposed amendments 
would modify the licenses to reflect a grant of section 161A of the 
Atomic Energy Act, to authorize the licensee the authority to possess 
and use certain firearms, ammunition, and other devices such as large-
capacity ammunition feeding devices, to implement the NRC-approved 
security plan for DCPP, Unit Nos. 1 and 2, and the Diablo Canyon ISFSI.
    Date of publication of individual notice in Federal Register: 
February 18, 2015 (80 FR 8706).
    Expiration date of individual notice: March 20, 2015 (public 
comments); April 19, 2015 (hearing requests).
Southern California Edison Company, et al., Docket Nos. 50-361, 50-362, 
and 72-41, San Onofre Nuclear Generating Station, Units 2 and 3, and 
Independent Spent Fuel Storage Installation, San Diego County, 
California
    Date of amendment request: August 28, 2013, as supplemented by 
letters dated December 31, 2013, May 15, 2014, and February 10, 2015. 
Publicly-available versions are in ADAMS under Accession Nos. 
ML13242A277, ML14007A496, ML14139A424, and ML15044A047, respectively.
    Brief description of amendment request: The licensee is requesting 
that the Commission grant it preemption authority consistent with the 
Commission's authority under section 161A of the Atomic Energy Act of 
1954, as amended, to authorize the security personnel of designated 
classes of licensees to possess, use, and access covered weapons for 
the physical security of SONGS, Units 2 and 3, and the Independent 
Spent Fuel Storage Installation, notwithstanding Federal, State, or 
local laws prohibiting such possession or use. If the amendment request 
is granted, the licenses would be modified to reflect the Commission's 
granting of section 161A preemption authority.
    Date of publication of individual notice in Federal Register: 
February 18, 2015 (80 FR 8701).
    Expiration date of individual notice: March 20, 2015 (public 
comments); April 20, 2015 (hearing requests).

IV. Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined Licenses and Final Determination of No Significant Hazards 
Consideration and Opportunity for a Hearing (Exigent Public 
Announcement or Emergency Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual notice of 
consideration of issuance of amendment, proposed no significant hazards 
consideration determination, and opportunity for a hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the

[[Page 13916]]

Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 51.12(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 
application for amendment, (2) the amendment to Facility Operating 
License or Combined License, as applicable, 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.

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

    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC's Web 
site, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR's Reference staff 
at 1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. 
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 petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the 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 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. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.

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

[[Page 13917]]

submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten 10 days prior to the filing deadline, the participant should 
contact the Office of the Secretary by email at hearing.docket@nrc.gov, 
or by telephone at 301-415-1677, to request (1) a digital 
identification (ID) certificate, which allows the participant (or its 
counsel or representative) to digitally sign documents and access the 
E-Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at https://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) first class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating 
Station, Unit 2, Montgomery County, Pennsylvania
    Date of amendment request: February 12, 2015.
    Description of amendment request: The amendment extends the 
implementation period for Amendment No. 174, ``Leak Detection System 
Setpoint and Allowable Value Changes,'' which was issued on December 
29, 2014. Amendment No. 174 was effective as of the date of issuance 
(i.e., on December 29, 2014) and was required to be implemented within 
60 days (i.e., by February 27, 2015). Amendment No. 177 extends the 
implementation period for Amendment No. 174 from 60 days to prior to 
startup from the spring 2015 refueling outage.
    Date of issuance: February 25, 2015.
    Effective date: As of its date of issuance and shall be implemented 
prior to startup from the Spring 2015 Unit 2 Refueling Outage.
    Amendment No.: 177. A publicly-available version is in ADAMS under 
Accession No. ML15049A084; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.

[[Page 13918]]

    Renewed Facility Operating License Nos. NPF-85: Amendment revised 
the Renewed Facility Operating License to extend the implementation 
date of Amendment No. 174, issued on December 29, 2014, to prior to 
startup from the Spring 2015 Unit 2 Refueling Outage.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. Public notice of the proposed amendment was 
published in The Pottstown Mercury, located in in Pottstown, 
Pennsylvania, on February 15, and February 16, 2015. The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. Comments were received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, public comments, and final 
NSHC determination are contained in a safety evaluation dated February 
25, 2015.
    Attorney for licensee: J. Bradley Fewell, Esquire, Vice President 
and Deputy General Counsel, Exelon Generation Company, LLC, 200 Exelon 
Way, Kennett Square, PA 19348.
    NRC Branch Chief: Douglas A. Broaddus.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Respectively, 
Limestone County, Alabama
    Date of amendment request: February 12, 2015.
    Brief description of amendment request: The amendments revised 
Technical Specification (TS) 5.6.5, ``Core Operating Limits Report 
(COLR),'' to add the date of a previously issued NRC safety evaluation 
(SE) that stated it was acceptable for the licensee to use new 
analytical methods supporting the use of ATRIUM 10XM (10XM) fuel. In 
its letter dated February 12, 2015, the licensee stated BFN, Unit 2, is 
entering an outage on March 14, 2015, and is scheduled to commence 
loading 10XM fuel on March 17, 2015. Because the TSs do not reference 
the aforementioned NRC evaluation, the licensee would not be able to 
issue a COLR for the Unit 2 transition cycle unless the notation to the 
latest NRC SE is added. Therefore, the licensee requested that NRC 
process the license amendment request under exigent circumstances in 
accordance with 10 CFR 50.91(a)(6). The NRC staff determined that the 
provisions of 10 CFR 50.91(a)(6) were applicable for processing the 
licensee's request under exigent circumstances.
    Date of issuance: February 26, 2015.
    Effective date: As of the date of issuance and shall be implemented 
during the refueling outages in fall of 2016 for Unit 1, in spring of 
2015 for Unit 2, and in spring of 2016 for Unit 3.
    Amendment Nos.: 288, 313, and 272, which are available in ADAMS 
under Accession No. ML15051A337. Documents related to these amendments 
are listed in the SE enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the TSs.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): The public notice was published in ``The 
Huntsville Times,'' located in Huntsville, Alabama, on February 18 and 
20, 2015. The notice provided an opportunity to submit comments on the 
Commission's proposed NSHC determination. No comments have been 
received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a safety evaluation dated February 26, 2015.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, TN 37902.
    NRC Branch Chief: Shana R. Helton.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan
    Date of application for amendment: December 12, 2012, as 
supplemented by letters dated February 21, September 30, October 24, 
and December 2, 2013; April 2, May 7, June 17, August 14, November 4, 
and December 18, 2014.
    Brief description of amendment: The amendment authorizes the 
transition of the Palisades Nuclear Plant fire protection program to a 
risk-informed, performance-based program based on National Fire 
Protection Association (NFPA) 805, in accordance with 10 CFR 50.48(c). 
NFPA 805 allows the use of performance-based methods such as fire 
modeling and risk-informed methods such as fire probabilistic risk 
assessment to demonstrate compliance with the nuclear safety 
performance criteria.
    Date of issuance: February 27, 2015.
    Effective date: As of its date of issuance and shall be implemented 
by six months from the date of issuance.
    Amendment No.: 254. A publicly-available version is in ADAMS under 
Accession No. ML15007A191; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-20: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: February 27, 2014 (79 
FR 11148). The supplements dated April 2, May 7, June 17, August 14, 
November 4, and December 18, 2014, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.

[[Page 13919]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 27, 2015.
    No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
1, Pope County, Arkansas
    Date of amendment request: March 26, 2013, as supplemented by 
letters dated November 14, 2013, and August 18, October 22, and 
December 5, 2014.
    Brief description of amendment: The amendment revised the Technical 
Specification (TS) requirements for end states associated with the 
implementation of the NRC-approved Topical Report BAW-2441-A, Revision 
2, ``Risk-Informed Justification for LCO End-State Changes,'' as well 
as Required Actions revised by a specific Note in TS Task Force (TSTF) 
change traveler TSTF-431, Revision 3, ``Change in Technical 
Specifications End States (BAW-2441).''
    Date of issuance: March 3, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 253. A publicly-available version is in ADAMS under 
Accession No. ML15023A147; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the TSs/license.
    Date of initial notice in Federal Register: July 23, 2013 (78 FR 
44170). The supplemental letters dated November 14, 2013, and August 
19, October 22, and December 5, 2014, provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 3, 2015.
    No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana
    Date of amendment request: December 9, 2013, as supplemented by 
letters dated October 1, 2014, and December 17, 2014.
    Brief description of amendment: The amendment revised the Technical 
Specifications for the Waterford Steam Electric Station, Unit 3 to 
improve clarity, correct administrative and typographical errors, or 
establish consistency with NUREG-1432, ``Standard Technical 
Specifications--Combustion Engineering Plants,'' Revision 4.0.
    Date of issuance: February 23, 2015.
    Effective date: As of the date of issuance and shall be implemented 
90 days from the date of issuance.
    Amendment No.: 242. A publicly-available version is in ADAMS under 
Accession No. ML15005A126; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 5, 2014 (79 FR 
45475). The supplements dated October 1, 2014, and December 17, 2014, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 23, 2015.
    No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida
    Date of amendment request: February 26, 2014, as supplemented by 
letters dated May 29 and July 25, 2014.
    Brief description of amendment: The amendments revised the 
Technical Specifications (TSs), modifying requirements for mode change 
limitations in Limiting Condition for Operation 3.0.4 and Surveillance 
Requirement (SR) 4.0.4 to adopt the provisions of Industry/TS Task 
Force (TSTF)-359, Rev. 9, ``Increase Flexibility in MODE Restraints.'' 
The language of SR 4.0.1 is revised to conform to the language of 
NUREG-1432, ``Standard Technical Specifications for Combustion 
Engineering Plants,'' to resolve language incongruences and ensure 
conservative implementation of the TSTF-359, Rev. 9, changes.
    Date of issuance: February 27, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 220 and 170. A publicly-available version is in 
ADAMS under Accession No. ML14343A918; documents related to these 
amendments are listed in the Safety Evaluation (SE) enclosed with the 
amendments.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: May 27, 2014 (79 FR 
30187). The supplements dated May 29 and July 25, 2014, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the 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 SE dated February 27, 2015.
    No significant hazards consideration comments received: No.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County, 
Texas
    Date of amendment request: July 1, 2014.
    Brief description of amendment: The amendments revised Technical 
Specification 3.8.1, ``AC [Alternating Current] Sources--Operating,'' 
to extend on a one-time basis the Completion Time (CT) of Required 
Action A.3, ``Restore required offsite circuit to OPERABLE status,'' 
from 72 hours to 14 days. The CT extension from 72 hours to 14 days 
will be used while completing the plant modification to install 
alternate startup transformer XST1A and will expire on March 31, 2017.
    Date of issuance: February 24, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within [licensee requested number] days from the date of issuance.
    Amendment Nos.: Unit 1--164; Unit 2--164. A publicly-available 
version is in ADAMS under Accession No. ML15008A133; 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: October 28, 2014 (79 FR 
64226).
    The Commission's related evaluation of the amendments is contained 
in a

[[Page 13920]]

Safety Evaluation dated February 24, 2015.
    No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit 1, Washington County, Nebraska
    Date of amendment request: April 30, 2014, as supplemented by 
letter dated January 27, 2015.
    Brief description of amendment: The amendment revised Technical 
Specification section 3.2, Table 3-5, for Fort Calhoun Station, Unit 
No. 1, to add a new surveillance requirement to verify the correct 
position of the valves required to restrict flow in the high pressure 
safety injection system.
    Date of issuance: February 20, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment No.: 280. A publicly-available version is in ADAMS under 
Accession No. ML15015A413; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the license and Technical Specifications.
    Date of initial notice in Federal Register: August 19, 2014 (79 FR 
49108). The supplemental letter dated January 27, 2015, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated February 20, 2015.
    No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas
    Date of amendment request: August 14, 2014, as supplemented by 
letter dated December 18, 2014.
    Brief description of amendments: The amendments revised 
Administrative Controls Technical Specification (TS) 6.9.1.6, ``Core 
Operating Limits Report (COLR),'' with respect to the analytical 
methods used to determine the core operating limits.
    Date of issuance: February 27, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: Unit 1--204; Unit 2--192. A publicly-available 
version is in ADAMS under Accession No. ML15049A129; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: December 2, 2014 (79 FR 
71455). The supplemental letter dated December 18, 2014, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 27, 2015.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 9th day of March 2015.

For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2015-05994 Filed 3-16-15; 8:45 am]
 BILLING CODE 7590-01-P