Biweekly Notice, Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 36613-36626 [2016-13255]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 81, Number 109 (Tuesday, June 7, 2016)]
[Notices]
[Pages 36613-36626]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-13255]


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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0107]


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 May 10, 2016, to May 23, 2016. The last 
biweekly notice was published on May 24, 2016 (81 FR 32800).

DATES: Comments must be filed by July 7, 2016. A request for a hearing 
must be filed by August 8, 2016.

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-2016-0107. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individuals listed in the FOR FURTHER INFORMATION CONTACT section 
of this document.
     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: Kay Goldstein, Office of Nuclear 
Reactor Regulation, telephone: 301-415-1506, email: 
Kay.Goldstein@nrc.gov and Lynn Ronewicz, Office of Nuclear Reactor 
Regulation, telephone: 301-415-1927, email: Lynn.Ronewicz@nrc.gov. Both 
are staff of the U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2016-0107 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-2016-0107.
     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 of this document.
     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-2016-0107, 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 will post all comment submissions at https://www.regulations.gov as well as enter 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

[[Page 36614]]

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 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 if 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. If the Commission 
takes 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. If the Commission makes 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 within 60 days, 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 set forth 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 with respect to resolution of that person's admitted 
contentions, including the opportunity to present evidence and to 
submit a cross-examination plan for cross-examination of witnesses, 
consistent with NRC regulations, policies and procedures.
    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). If a hearing is requested, and the Commission 
has not made a final determination on the issue of no significant 
hazards consideration, 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

[[Page 36615]]

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.
    A State, local governmental body, federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission by 
August 8, 2016. The petition must be filed in accordance with the 
filing instructions in the ``Electronic Submissions (E-Filing)'' 
section of this document, and should meet the requirements for 
petitions for leave to intervene set forth in this section, except that 
under Sec.  2.309(h)(2) a State, local governmental body, or Federally-
recognized Indian Tribe, or agency thereof does not need to address the 
standing requirements in 10 CFR 2.309(d) if the facility is located 
within its boundaries. A State, local governmental body, Federally-
recognized Indian Tribe, or agency thereof may also have the 
opportunity to participate under 10 CFR 2.315(c).
    If a hearing is granted, any person who does not wish, or is not 
qualified, to become a party to the proceeding may, in the discretion 
of the presiding officer, be permitted to make a limited appearance 
pursuant to the provisions of 10 CFR 2.315(a). A person making a 
limited appearance may make an oral or written statement of position on 
the issues, but may not otherwise participate in the proceeding. A 
limited appearance may be made at any session of the hearing or at any 
prehearing conference, subject to the limits and conditions as may be 
imposed by the presiding officer. Persons desiring to make a limited 
appearance are requested to inform the Secretary of the Commission by 
August 8, 2016.
Electronic Submissions (E-Filing)
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the Internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten 10 days prior to the filing deadline, the participant should 
contact the Office of the Secretary by email at hearing.docket@nrc.gov, 
or by telephone at 301-415-1677, to request (1) a digital 
identification (ID) certificate, which allows the participant (or its 
counsel or representative) to digitally sign documents and access the 
E-Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at https://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or

[[Page 36616]]

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, in some instances, 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.
    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.

DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: March 22, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16082A309.
    Description of amendment request: The proposed amendment would 
allow for permanent extension of the Type A primary containment 
integrated leak rate test interval to 15 years and extension of the 
Type C test interval up to 75 months. The amendment also proposes two 
administrative changes to remove text that is no longer applicable. The 
first change revises technical specification (TS) 5.5.12 to remove a 
one-time extension of the Type A test frequency. The second change 
would revise the Fermi 2 Operating License, Section D, to remove a 
reference to an exemption regarding Appendix J testing of containment 
air locks.
    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 to the TS involves the extension of Fermi 
2 Type A containment test interval to 15 years and the extension of 
the Type C test interval to 75 months. The current Type A test 
interval of 10 years would be extended on a permanent basis to no 
longer than 15 years from the last Type A test. The current Type C 
test interval of 60 months for selected components would be extended 
on a performance basis to no longer than 75 months. Extensions of up 
to nine months (total maximum interval of 84 months for Type C 
tests) are permissible only for non-routine emergent conditions. The 
proposed amendment does not involve either a physical change to the 
plant or a change in the manner in which the plant is operated or 
controlled. The primary containment is designed to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. As such, 
the containment and the testing requirements invoked to periodically 
demonstrate the integrity of the containment exist to ensure the 
plant's ability to mitigate the consequences of an accident, and do 
not involve any accident precursors or initiators. RG [Regulatory 
Guide] 1.174 [sic] [ADAMS Accession No. ML023240437] provides 
guidance for determining the risk impact of plant-specific changes 
to the licensing basis. RG 1.174 defines very small changes in risk 
as resulting in increases of CDF [core damage frequency] below 1.0E-
06/yr and increases in LERF [large early release frequency] below 
1.0E-07/yr. Since the ILRT [integrated leak rate test] does not 
impact CDF, the relevant criterion is LERF. The increase in LERF 
resulting from a change in the Type A ILRT test interval from three 
in ten years to one in fifteen years is very conservatively 
estimated as 1.27E-08/yr using the EPRI [Electric Power Research 
Institute] guidance as written. As such, the estimated change in 
LERF is determined to be ``very small'' using the acceptance 
guidelines of RG 1.174.
    RG 1.174 also states that when the calculated increase in LERF 
is in the range of 1.0E-06 per reactor year to 1.0E-07 per reactor 
year, applications will be considered only if it can be reasonably 
shown that the total LERF is less than 1.0E-05 per reactor year. An 
additional assessment of the impact from external events was also 
made. In this case, the total LERF increase was conservatively 
estimated (with an external event multiplier of 15) as 1.90E-07 for 
Fermi 2 (the baseline total LERF for this case is 7.88E- 06/yr). 
This is well below the RG 1.174 acceptance criteria for total LERF 
of 1.0E-05.
    The change in Type A test frequency to once per 15 years, 
measured as an increase to the total integrated plant risk for those 
accident sequences influenced by Type A testing, is 1.14E-4 person-
rem/yr (a 0.00184% increase). EPRI Report No. 1009325, Revision 2-A, 
states that a very small population dose is defined as an increase 
of <=1.0 person-rem per year or <=1% of the total population dose, 
whichever is less restrictive for the risk impact assessment of the 
extended ILRT intervals. Moreover, the risk impact when compared to 
other severe accident risks is negligible.
    The increase in the CCFP [conditional containment failure 
probability] from the three in 10 year [sic] interval to one in 15 
year interval is 0.73%. EPRI Report No. 1009325, Revision 2-A, 
states that increases in CCFP of less than or equal to 1.5 
percentage points are very small. Therefore, this increase judged to 
be very small.
    The other two changes, to TS 5.5.12, item a, and Operating 
License, Provision D, are administrative in nature to remove old 
text that is no longer applicable.
    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 to the TS involves the extension of the 
Fermi 2 Type A containment test interval to 15 years and the 
extension of the Type C test interval to 75 months. The containment 
and the testing requirements to periodically demonstrate the 
integrity of the containment exist to ensure the plant's ability to 
mitigate the consequences of an accident and do not involve any 
accident precursors or initiators. The proposed change does not 
involve a physical change to the plant (e.g., no new or different 
type of equipment will be installed) or a change to the manner in 
which the plant is operated or controlled.
    The other two changes to TS 5.5.12, item a, and Operating 
License, Provision D, are administrative in nature to remove old 
text that is no longer needed. Therefore, these changes have no 
impact on the probability or consequences of an accident previously 
evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

3. Does the proposed change involve a significant reduction in a margin 
of safety?

    Response: No.
    The proposed amendment to TS 5.5.12 involves the extension of 
the Fermi 2 Type A containment test interval to 15 years and the 
extension of the Type C test interval to 75 months for selected 
components. This amendment does not alter the manner in which safety 
limits, limiting safety system set points, or limiting conditions 
for operation are determined. The specific requirements and 
conditions of the TS Containment Leak Rate Testing Program exist to 
ensure that the

[[Page 36617]]

degree of containment structural integrity and leak-tightness that 
is considered in the plant safety analysis is maintained. The 
overall containment leak rate limit specified by TS is maintained.
    The proposed surveillance interval extension is bounded by the 
15 year ILRT interval and the 75 month Type C test interval 
currently authorized within NEI 94-01, Revision 3-A. Industry 
experience supports the conclusion that Type B and Type C testing 
detects a large percentage of containment leakage paths and the 
percentage of containment leakage paths that are detected only by 
Type A testing is small. The containment inspections preformed in 
accordance with ASME [American Society of Mechanical Engineers] 
Section XI, Maintenance Rule, and TS serve to provide a high degree 
of assurance that the containment would not degrade in a manner that 
is detectable only by Type A testing. The combination of these 
factors ensures that the margin of safety in the plant safety 
analysis is maintained. The design, operation, testing methods, and 
acceptance criteria for Type A, Type B, and Type C containment 
leakage tests specified in applicable codes and standards would 
continue to be met with the acceptance of this proposed change since 
these are not affected by the changes to the Type A and Type C test 
intervals.
    The other two changes to TS 5.5.12, item a, and Operating 
License, Provision D, are administrative in nature to remove old 
text that is no longer needed. Therefore, these changes have no 
impact on the probability or consequences of an accident previously 
evaluated.
    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: Jon P. Christinidis, DTE Energy, Expert 
Attorney--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-
1279.
    NRC Branch Chief: David J. Wrona.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: March 24, 2016. A publicly available 
version is in ADAMS under Accession No. ML16089A228.
    Description of amendment request: The amendments would modify 
Technical Specification 3.6.13, ``Ice Condenser Doors,'' to revise 
Condition B for an ice condenser lower inlet door invalid open alarm to 
preclude plant shutdown caused by an invalid ``OPEN'' alarm from the 
``Inlet Door Position Monitoring System.''
    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 proposed change will not increase the probability of 
accident previously evaluated. The Ice Condenser performs an 
entirely mitigative function. The proposed change does not result in 
any physical change to the plant which would affect any accident 
initiators. No structures, systems, or components (SSCs) involved in 
the initiation of postulated accidents will be operated in any 
different manner. The probability of occurrence of a previously 
evaluated accident will not be significantly increased. The proposed 
change involves use of an alternate method of verifying that the 
lower inlet doors to the ice condenser are closed. This proposed 
change has no effect on the ability of the ice condenser to perform 
its function.
    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 alter the design function or 
operation of any SSC that may be involved in the initiation of an 
accident. The Ice Condenser will not become the source of a new type 
of accident. No new accident causal mechanisms will be created. The 
proposed change does not create new failure mechanisms, 
malfunctions, or accident initiators.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

3. Does the proposed change involve a significant reduction in the 
margin of safety?

    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their intended functions. 
These barriers include the fuel cladding, the reactor coolant system 
pressure boundary, and the containment barriers. The proposed change 
involves use of a method to verify the lower inlet doors to the ice 
condenser are closed when an invalid alarm is providing indication 
of an open door. This proposed change has no effect on the ability 
of the ice condenser to perform its function. Hence, the proposed 
change will not affect containment barriers. Nor does the proposed 
change have any effect on fuel cladding or the reactor coolant 
pressure boundary.
    Therefore, existing safety margins will be preserved, and the 
proposed change does not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.

Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick 
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

    Date of amendment request: April 13, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16111B203.
    Description of amendment request: The amendments would revise the 
Allowable Values (AVs) of Surveillance Requirements (SRs) contained in 
Technical Specification 3.3.8.2, ``RPS Electric Power Monitoring,'' by 
amending the Reactor Protection System electric power monitoring 
assembly AVs for overvoltage and undervoltage contained within SRs 
3.3.8.2.2 and 3.3.8.2.3.
    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 change to the Allowable Values of Surveillance 
Requirements contained in Technical Specifications 3.3.8.2 does not 
impact the physical function of plant structures, systems, or 
components (SSC) or the manner in which SCCs [sic] perform their 
design function. The proposed change does not authorize the addition 
of any new plant equipment or systems, nor does it alter the 
assumptions of any accident analyses. The Electrical Protection 
Assemblies are not accident initiators. They operate in response to 
off-normal voltage conditions on Class 1E buses to protect the 
connected loads. The proposed change does not adversely affect 
accident initiators or precursors, nor does it alter the design 
assumptions, conditions, and configuration or the manner in which 
the plant is operated and maintained.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 36618]]

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 change to the Allowable Values of Surveillance 
Requirements contained in Technical Specifications 3.3.8.2 does not 
require any modification to the plant (i.e., other than the setpoint 
changes) or change equipment operation or testing. The proposed 
change will not introduce failure modes that could result in a new 
accident, and the change does not alter assumptions made in the 
safety analysis. The proposed change will not alter the design 
configuration, or method of operation of plant equipment beyond its 
normal functional capabilities. The proposed change does not create 
any new credible failure mechanisms, malfunctions, or accident 
initiators.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from those that have been 
previously evaluated.

3. Does the proposed amendment involve a significant reduction in a 
margin of safety?

    Response: No
    The proposed change to the Allowable Values of Surveillance 
Requirements contained in Technical Specifications 3.3.8.2 does not 
alter or exceed a design basis or safety limit. There is no change 
being made to safety analysis assumptions or the safety limits that 
would adversely affect plant safety as a result of the proposed 
change. Margins of safety are unaffected by the proposed change and 
the applicable requirements of 10 CFR 50.36(c)(2)(ii) and 10 CFR 50, 
Appendix A will continue to be met.
    Therefore, the proposed change does not involve any reduction in 
a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, 
550 South Tryon Street, M/C DEC45A, Charlotte NC 28202.
    NRC Branch Chief: Benjamin G. Beasley.
Entergy Operations, Inc. (Entergy), Docket No. 50-368, Arkansas Nuclear 
One, Unit No. 2 (ANO-2), Pope County, Arkansas
    Date of amendment request: March 25, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16088A186.
    Description of amendment request: The amendment will revise the 
Technical Specifications (TSs) to eliminate TS 6.5.8, ``Inservice 
Testing Program.'' A new defined term, ``Inservice Testing [IST] 
Program,'' will be added to TS 1.0, ``Definitions,'' section. The 
licensee has noted that while the request is consistent with TS Task 
Force (TSTF)-545, Revision 3, ``TS Inservice Testing Program Removal & 
Clarify SR [Surveillance Requirement] Usage Rule Application to Section 
5.5 Testing,'' there are various deviations from the TSTF-545, Revision 
3. ANO-2 TSs are of an older standard version and have not been 
converted to the improved standard TSs (ISTSs) based on NUREG 1432, 
``Standard Technical Specifications--Combustion Engineering Plants,'' 
Revision 4. As such, Entergy stated there are several administrative-
type variations (TS numbering, wording, etc.) but these variations do 
not result in any technical conflict with the intent of TSTF-545, 
Revision 3 or the associated model safety evaluation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, with NRC edits in [brackets], which is presented below:

1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    Response: No.
    The proposed change revises TS Chapter 6, ``Administrative 
Controls,'' Section 6.5, ``Programs and Manuals,'' by eliminating 
the ``Inservice Testing Program'' specification. Most requirements 
in the IST Program are removed, as they are duplicative of 
requirements in the ASME [American Society of Mechanical Engineers] 
OM Code [ASME Code for Operation and Maintenance of Nuclear Power 
Plants], as clarified by Code Case OMN-20, ``Inservice Test 
Frequency.'' The remaining requirements in the Section 6.5 IST 
Program are eliminated because the NRC has determined their 
inclusion in the TS is contrary to regulations. A new defined term, 
``Inservice Testing Program,'' is added to the TS, which references 
the requirements of 10 CFR 50.55a(f).
    Performance of inservice testing is not an initiator to any 
accident previously evaluated. As a result, the probability of 
occurrence of an accident is not significantly affected by the 
proposed change. Inservice test frequencies under Code Case OMN-20 
are equivalent to the current testing period allowed by the TS with 
the exception that testing frequencies greater than 2 years may be 
extended by up to 6 months to facilitate test scheduling and 
consideration of plant operating conditions that may not be suitable 
for performance of the required testing. The testing frequency 
extension will not affect the ability of the components to mitigate 
any accident previously evaluated as the components are required to 
be operable during the testing period extension. Performance of 
inservice tests utilizing the allowances in OMN-20 will not 
significantly affect the reliability of the tested components. As a 
result, the availability of the affected components, as well as 
their ability to mitigate the consequences of accidents previously 
evaluated, is not affected.
    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 alter the design or configuration 
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. The proposed change does not alter the types of 
inservice testing performed. In most cases, the frequency of 
inservice testing is unchanged. However, the frequency of testing 
would not result in a new or different kind of accident from any 
previously evaluated since the testing methods are not altered.
    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 eliminates some requirements from the TS in 
lieu of requirements in the ASME Code, as modified by use of Code 
Case OMN-20. Compliance with the ASME Code is required by 10 CFR 
50.55a. The proposed change also allows inservice tests with 
frequencies greater than 2 years to be extended by 6 months to 
facilitate test scheduling and consideration of plant operating 
conditions that may not be suitable for performance of the required 
testing. The testing frequency extension will not affect the ability 
of the components to respond to an accident as the components are 
required to be operable during the testing period extension. The 
proposed change will eliminate the existing TS Surveillance 
Requirement (SR) 4.0.3 (referenced as SR 3.0.3 in the ISTS) 
allowance to defer performance of missed inservice tests up to the 
duration of the specified testing frequency, and instead will 
require an assessment of the missed test on equipment operability. 
This assessment will consider the effect on a margin of safety 
(equipment operability). Should the component be inoperable, the 
Technical Specifications provide actions to ensure that the margin 
of safety is protected. The proposed change also eliminates a 
statement that nothing in the ASME Code should be construed to 
supersede the requirements of any TS. The NRC has determined that 
statement to be incorrect. However, elimination of the statement 
will have no effect on plant operation or safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.


[[Page 36619]]


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

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: March 25, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16088A181.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) to eliminate TS Section 5.5.8, 
``Inservice Testing [IST] Program.'' A new defined term, ``Inservice 
Testing Program,'' will be added to TS 1.1, ``Definitions.'' This 
amendment request is consistent with TS Task Force (TSTF)-545, Revision 
3, ``TS Inservice Testing Program Removal & Clarify SR [Surveillance 
Requirement] Usage Rule Application to Section 5.5 Testing,'' under the 
consolidated line item improvement process.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, with NRC edits in [brackets], which is presented below:

1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    Response: No.
    The proposed change revises TS Chapter 5, ``Administrative 
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating 
the ``Inservice Testing Program'' specification. Most requirements 
in the IST Program are removed, as they are duplicative of 
requirements in the ASME [American Society of Mechanical Engineers] 
OM Code [ASME Code for Operation and Maintenance of Nuclear Power 
Plants], as clarified by Code Case OMN-20, ``Inservice Test 
Frequency.'' The remaining requirements in the Section 5.5 IST 
Program are eliminated because the NRC has determined their 
inclusion in the TS is contrary to regulations. A new defined term, 
``Inservice Testing Program,'' is added to the TS, which references 
the requirements of 10 CFR 50.55a(f).
    Performance of inservice testing is not an initiator to any 
accident previously evaluated. As a result, the probability of 
occurrence of an accident is not significantly affected by the 
proposed change. Inservice test frequencies under Code Case OMN-20 
are equivalent to the current testing period allowed by the TS with 
the exception that testing frequencies greater than 2 years may be 
extended by up to 6 months to facilitate test scheduling and 
consideration of plant operating conditions that may not be suitable 
for performance of the required testing. The testing frequency 
extension will not affect the ability of the components to mitigate 
any accident previously evaluated as the components are required to 
be operable during the testing period extension. Performance of 
inservice tests utilizing the allowances in OMN-20 will not 
significantly affect the reliability of the tested components. As a 
result, the availability of the affected components, as well as 
their ability to mitigate the consequences of accidents previously 
evaluated, is not affected.
    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 alter the design or configuration 
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. The proposed change does not alter the types of 
inservice testing performed. In most cases, the frequency of 
inservice testing is unchanged. However, the frequency of testing 
would not result in a new or different kind of accident from any 
previously evaluated since the testing methods are not altered.
    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 eliminates some requirements from the TS in 
lieu of requirements in the ASME Code, as modified by use of Code 
Case OMN-20. Compliance with the ASME Code is required by 10 CFR 
50.55a. The proposed change also allows inservice tests with 
frequencies greater than 2 years to be extended by 6 months to 
facilitate test scheduling and consideration of plant operating 
conditions that may not be suitable for performance of the required 
testing. The testing frequency extension will not affect the ability 
of the components to respond to an accident as the components are 
required to be operable during the testing period extension. The 
proposed change will eliminate the existing TS Surveillance 
Requirement (SR) 3.0.3 allowance to defer performance of missed 
inservice tests up to the duration of the specified testing 
frequency, and instead will require an assessment of the missed test 
on equipment operability. This assessment will consider the effect 
on a margin of safety (equipment operability). Should the component 
be inoperable, the Technical Specifications provide actions to 
ensure that the margin of safety is protected. The proposed change 
also eliminates a statement that nothing in the ASME Code should be 
construed to supersede the requirements of any TS. The NRC has 
determined that statement to be incorrect. However, elimination of 
the statement will have no effect on plant operation or safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 
and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and 
Lancaster Counties, Pennsylvania

    Date of amendment request: March 24, 2016, as supplemented by 
letter dated May 11, 2016. A publicly-available version is in ADAMS 
under Accession Nos. ML16084A567 and ML16132A440.
    Description of amendment request: The amendments would revise the 
frequency for cycling of the recirculation pump discharge valves as 
specified in Technical Specification (TS) Surveillance Requirement (SR) 
3.5.1.5. Specifically, SR 3.5.1.5 requires verification that each 
recirculation pump discharge valve cycles through one complete cycle of 
full travel or is de-energized in the closed position. Currently, this 
SR needs to be performed once each plant startup prior to exceeding 23 
percent rated thermal power (RTP), if the SR had not been performed 
within the previous 31 days. The amendments would change the frequency 
for the SR such that it is performed in accordance with the Inservice 
Testing Program.
    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 36620]]



1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    Response: No.
    The proposed change revises the frequency for cycling the 
recirculation pump discharge valves from ``Once each startup prior 
to exceeding 23% RTP,'' as modified by a Note stating, ``Not 
required to be performed if performed within the previous 31 days'' 
to ``In accordance with the Inservice Testing Program''. Testing of 
the recirculation pump discharge valves is not an initiator of any 
accident previously evaluated. As the recirculation pump discharge 
valves are still required to be Operable, the ability to mitigate 
any accident previously evaluated is not affected. The proposed 
change does not adversely affect the design assumptions, conditions, 
or configuration of the facility. The proposed change does not alter 
or prevent the ability of structures, systems, and components (SSCs) 
from performing their intended function.
    Therefore, this change does not significantly increase 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 the frequency for cycling the 
recirculation pump discharge valves from ``Once each startup prior 
to exceeding 23% RTP,'' as modified by a Note stating, ``Not 
required to be performed if performed within the previous 31 days'' 
to ``In accordance with the Inservice Testing Program''. This 
revision will not impact the accident analysis. The change will not 
alter the methods of operation of the recirculation pump discharge 
valves. No new or different accidents result. The change does not 
involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The change 
does not alter assumptions made in the safety analysis.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not created.

3. Does the proposed change involve a significant reduction in a margin 
of safety?

    Response: No.
    The proposed change revises the frequency for cycling the 
recirculation pump discharge valves from ``Once each startup prior 
to exceeding 23% RTP,'' as modified by a Note stating, ``Not 
required to be performed if performed within the previous 31 days'' 
to ``In accordance with the Inservice Testing Program.'' The 
proposed change does not alter the manner in which safety limits, 
limiting safety system settings or limiting conditions for operation 
are determined. The safety analysis acceptance criteria are not 
affected by this change. The proposed change will not result in 
plant operation in a configuration outside the design basis. The 
frequency of testing the recirculation pump discharge valves will be 
consistent with the frequency of testing other valves in the 
Emergency Core Cooling System.
    Therefore, this 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: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL 
60555.
    NRC Branch Chief: Douglas A. Broaddus.

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

    Date of amendment request: April 4, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16095A285.
    Description of amendment request: The proposed changes would revise 
technical specification (TS) limiting condition for operation (LCO) 
3.10.1, and the associated Bases, to expand its scope to include 
provisions for temperature excursions greater than 200 degrees 
Fahrenheit as a consequence of in-service leak and hydrostatic testing, 
and as a consequence of scram time testing initiated in conjunction 
with an in-service leak or hydrostatic test, while considering 
operational conditions to be in Mode 4.
    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.
    Technical Specifications currently allow for operation at 
greater than 200 degrees F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact the probability or 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 change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    Response: No.
    Technical Specifications currently allow for operation at 
greater than 200 degrees F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. No new operational conditions beyond those currently allowed by 
LCO 3.10.1 are introduced. The changes 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 or eliminate any existing 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 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.
    Technical Specifications currently allow for operation at 
greater than 200 degrees F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact any margin of safety. Allowing completion of 
inspections and testing and supporting completion of scram time 
testing initiated in conjunction with an in-service leak or 
hydrostatic test prior to power operation results in enhanced safe 
operations by eliminating unnecessary maneuvers to control reactor 
temperature and pressure. 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: Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: G. Ed Miller (Acting)

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station (LGS), Units 1 and 2, Montgomery County, 
Pennsylvania

    Date of amendment request: April 4, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16095A275.
    Description of amendment request: The amendments would revise the 
high pressure coolant injection (HPCI) and reactor core isolation 
cooling (RCIC)

[[Page 36621]]

system actuation instrumentation Technical Specification (TS) 
requirements.
    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 involve the addition of clarifying 
footnotes to the HPCI and RCIC actuation instrumentation TS to 
reflect the as-built plant design and operability requirements of 
HPCI and RCIC instrumentation as described in the LGS Updated Final 
Safety Analysis Report (UFSAR).
    HPCI and RCIC are not an initiator of any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not increased. In addition, the automatic start of HPCI 
on high drywell pressure, and the manual initiation of HPCI and 
RCIC, are not credited to mitigate the consequences of design basis 
accidents, transients or special events within the current LGS 
design and licensing basis.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

2. Do the proposed changes create the possibility of a new or different 
kind of accident from any accident previously evaluated?

    Response: No.
    The proposed changes do not alter the protection system design, 
create new failure modes, or change any modes of operation. The 
proposed changes do not involve a physical alteration of the plant, 
and no new or different kind of equipment will be installed. 
Consequently, there are no new initiators that could result in a new 
or different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.

3. Do the proposed changes involve a significant reduction in a margin 
of safety?

    Response: No.
    The proposed changes have no adverse effect on plant operation. 
The plant response to the design basis accidents does not change. 
The proposed changes do not adversely affect existing plant safety 
margins or the reliability of the equipment assumed to operate in 
the safety analyses.
    There is no change being made to safety analysis assumptions, 
safety limits or limiting safety system settings that would 
adversely affect plant safety as a result of the proposed changes.
    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: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Acting Branch Chief: Andrew Hon.

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

    Date of amendment request: April 29, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16125A253.
    Description of amendment request: The amendments would revise 
Appendix B (Environmental Protection Plan (EPP)) of the Unit 1 and Unit 
2 Operating Licenses to incorporate the revised Section 8.4, ``Terms 
and Conditions'' of the currently applicable Biological Opinion issued 
by the National Marine Fisheries Service (NMFS) on March 24, 2016. In 
addition, the amendments would clarify in the EPP that the licensee 
must adhere to the currently applicable Biological Opinion. This 
clarification would preclude the need for a new license amendment in 
the event that NMFS issues a new Biological Opinion.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. Operation of the Facility in Accordance With the Proposed Amendments 
Would Not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated

    The changes are administrative in nature and would in no way 
affect the initial conditions, assumptions, or conclusions of the 
St. Lucie Unit 1 or Unit 2 accident analyses. In addition, the 
proposed changes would not affect the operation or performance of 
any equipment assumed in the accident analyses. Based on the above 
information, we conclude that the proposed changes would not 
significantly increase the probability or consequences of an 
accident previously evaluated.

2. Use of the Modified Specification Would Not Create the Possibility 
of a New or Different Kind of Accident From any Previously Evaluated

    The changes are administrative in nature and would in no way 
impact or alter the configuration or operation of the facilities and 
would create no new modes of operation. We conclude that the 
proposed changes would not create the possibility of a new or 
different kind of accident.

3. Use of the Modified Specification Would Not Involve a Significant 
Reduction in a Margin of Safety

    The changes are administrative in nature and would in no way 
affect plant or equipment operation or the accident analysis. We 
conclude that the proposed changes would not result in a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: April 4, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16099A097.
    Description of amendment request: The proposed amendment would 
revise technical specification (TS) 3.8.4, ``DC Sources--Operating,'' 
Surveillance Requirement (SR) 3.8.4.2 to increase the required 125 Volt 
(V) Direct Current (DC) subsystems battery charger output current and 
to remove the second method specified to perform the surveillance. The 
first proposed change is to increase the required 125 Volt VDC battery 
charger output current specified as the first option under SR 3.8.4.2 
to resolve a non-conservative TS condition. The second proposed change 
is to remove from SR 3.8.4.2 an alternative option for meeting the 
surveillance requirement. This alternative requires verifying each 
battery charger can recharge the battery to the fully charged state 
within the required time period, 24 hours for the 250 VDC and 8 hours 
for the 125 VDC subsystems, respectively, while supplying the largest 
combined continuous steady state loads, after a battery discharge to 
the bounding design basis event discharge state.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 36622]]

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

1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    Response: No.
    The proposed TS changes revise the battery charger surveillance 
requirements in SR 3.8.4.2. The DC electrical power system, 
including associated battery chargers, is not an initiator of any 
accident sequence analyzed in the Updated Safety Analysis Report 
(USAR). Rather, the DC electrical power system supports operation of 
equipment used to mitigate accidents. Operation in accordance with 
the proposed TS continues to ensure that the DC electrical power 
system is capable of performing its specified safety functions as 
described in the USAR. Therefore, the mitigating functions supported 
by the DC electrical power system will continue to provide the 
protection assumed by the analysis.
    Accidents are initiated by the malfunction of plant equipment, 
or the catastrophic failure of plant structures, systems, or 
components (SSCs). Performance of battery testing is not a precursor 
to any accident previously evaluated, nor does it change the manner 
in which the batteries and battery chargers are operated. The 
proposed testing requirements will not contribute to the failure of 
the batteries nor any plant SSC. NSPM has determined that the 
proposed TS changes provide an equivalent level of assurance that 
the batteries and battery chargers are capable of performing their 
intended safety functions. Thus, the proposed changes do not affect 
the probability 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 change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    Response: No.
    The DC electrical power system, including the associated battery 
chargers, is not an initiator of any accident sequence analyzed in 
the USAR. The proposed TS changes do not involve operation of the DC 
electrical power system in a manner or configuration different from 
those previously evaluated. Performance of battery testing is not a 
precursor to any accident previously evaluated. NSPM has determined 
that the proposed TS changes provide an equivalent level of 
assurance that the batteries and battery chargers are capable of 
performing their intended safety functions. Therefore, the 
mitigating functions supported by the DC electrical power system 
will continue to provide the protection assumed in the safety 
analyses.
    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 equipment 
design, the operating parameters, and the setpoints at which 
automatic actions are initiated. The equipment margins will be 
maintained in accordance with the plant-specific design bases as a 
result of the proposed changes. The proposed changes do not 
adversely affect operation of plant equipment. The proposed TS 
changes do not result in a change to the setpoints at which 
protective actions are initiated. Sufficient DC capacity to support 
operation of mitigation equipment continues to be ensured. The 
equipment fed by the DC electrical sources will continue to provide 
adequate power to safety-related loads in accordance with safety 
analysis assumptions.
    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: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: David J. Wrona.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1, 
Fairfield County, South Carolina

    Date of amendment request: April 7, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16104A027.
    Description of amendment request: The amendment would revise the 
Emergency Feedwater System pump performance testing requirements in 
Technical Specification (TS) 3/4.7.1.2, ``Emergency Feedwater System,'' 
Surveillance Requirements 4.7.1.2.a.1 and 4.7.1.2.a.2.
    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 with NRC staff edits in square 
brackets:

1. Do the proposed changes [sic] involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    Response: No.
    The proposed change deletes an allowed outage time that is no 
longer applicable and revises the Surveillance Requirements (SRs) 
that confirm the Emergency Feedwater (EFW) pump performance to be 
more consistent with the STS [Standard Technical Specifications--
Westinghouse Plants]. The change has been determined not to 
adversely affect the safe operation of the plant. The affected TS 
requirements are not initiating conditions for any accident 
previously evaluated. In addition, changes that are consistent with 
the STS have been previously evaluated by plants adopting the STS 
and found not to adversely affect the safe operation of Westinghouse 
NSSS [Nuclear Steam Supply System] plants. Based on the conclusions 
of the plant specific evaluation associated with the change and the 
evaluations performed in developing the STS, the proposed change 
does not result in operating conditions that will significantly 
increase the probability of initiating an analyzed event. The 
proposed change was also evaluated to assure that it does not alter 
the safety analysis assumptions relative to mitigation of an 
accident or transient event and that the resulting TS requirements 
continue to ensure the necessary equipment is operable consistent 
with the safety analyses or that the plant is placed in an operating 
Mode where the system is no longer required operable. As such the 
proposed change also does not result in operating conditions that 
will significantly increase the consequences of an analyzed event.
    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 includes the deletion of an expired allowed 
outage time extension and the revision of the SRs that confirm the 
EFW pump performance to be more consistent with the corresponding 
STS SR. Consistent with the STS SR, the proposed change would remove 
the specific pump head and flow values from the current SRs and 
require that the SR be performed in accordance with the Inservice 
Testing Program. The removal of the specific pump head and flow 
values from the SR is necessary to support the implementation of a 
plant modification that would change the current EFW pump head and 
flow values in the SR. The plant modification is being performed 
under the provisions of 10CFR50.59. The proposed TS change does not 
involve a change in the methods governing normal plant operation. 
The proposed change also does not change any system functions nor 
does the proposed TS change affect any safety analysis or design 
basis requirements. The proposed TS change will continue to ensure 
the EFW System is operable in a similar manner as before. As such, 
the proposed change does not create new failure modes or mechanisms 
that are not identifiable during testing, and no new accident 
precursors are generated.
    Therefore, the proposed changes do [sic] not create the 
possibility of a new or different kind of accident from any 
previously evaluated.

[[Page 36623]]

3. Does this [proposed] change involve a significant reduction in a 
margin of safety?

    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed change does not physically alter safety-
related systems, nor does it affect the way in which safety related 
systems perform their functions. The setpoints at which protective 
actions are initiated are not altered by the proposed change. 
Therefore, in a similar manner as before, sufficient equipment 
remains available to actuate upon demand for the purpose of 
mitigating an analyzed event. The proposed change results in TS 
requirements that are consistent with the plant safety analyses. As 
such, the change does not result in operating conditions that 
significantly reduce any margin of safety.
    Therefore, the proposed changes do [sic] 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. Hagood Hamilton, Jr., South Carolina 
Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 
29218.
    NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Inc.; Georgia Power Company; 
Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia

    Date of amendment request: August 11, 2015, as supplemented by 
letters dated March 16, 2014, and April 4, 2016. Publicly-available 
versions are in ADAMS under Accession Nos. ML15226A276, ML16076A453, 
and ML16095A373, respectively.
    Description of amendment request: The amendments would revise the 
technical specification (TS) requirements related to direct current 
(DC) electrical systems in TS Limiting Condition for Operation (LCO) 
3.8.4, ``DC Sources--Operating''; LCO 3.8.5, ``DC Sources--Shutdown''; 
and LCO 3.8.6, ``Battery Cell Parameters.'' A new battery monitoring 
and maintenance program is being proposed for Section 5.5, 
``Administrative Controls--Programs and Manuals.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. Does the proposed change involve a significant increase in the 
probability or consequences of any accident previously evaluated?

    Response: No.
    The proposed changes restructure the Technical Specifications 
(TS) for the direct current (DC) electrical power system and are 
consistent with TSTF-500, Revision 2. The proposed changes modify TS 
Actions relating to battery and battery charger inoperability. The 
DC electrical power system, including associated battery chargers, 
is not an initiator of any accident sequence analyzed in the Final 
Safety Analysis Report (FSAR). Rather, the DC electrical power 
system supports equipment used to mitigate accidents. The proposed 
changes to restructure TS and change surveillances for batteries and 
chargers to incorporate the updates included in TSTF-500, Revision 
2, will maintain the same level of equipment performance required 
for mitigating accidents assumed in the FSAR. Operation in 
accordance with the proposed TS would ensure that the DC electrical 
power system is capable of performing its specified safety function 
as described in the FSAR. Therefore, the mitigating functions 
supported by the DC electrical power system will continue to provide 
the protection assumed by the analysis.
    The relocation of preventive maintenance surveillances, and 
certain operating limits and actions, to a licensee-controlled 
Battery Monitoring and Maintenance Program will not challenge the 
ability of the DC electrical power system to perform its design 
function. Appropriate monitoring and maintenance that are consistent 
with industry standards will continue to be performed. In addition, 
the DC electrical power system is within the scope of 10 CFR 50.65, 
``Requirements for monitoring the effectiveness of maintenance at 
nuclear power plants,'' which will ensure the control of maintenance 
activities associated with the DC electrical power system.
    The integrity of fission product barriers, plant configuration, 
and operating procedures as described in the FSAR will not be 
affected by the proposed changes. Therefore, the consequences of 
previously analyzed accidents will not increase by implementing 
these changes.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?

    Response: No.
    The proposed changes involve restructuring the TS for the DC 
electrical power system. The DC electrical power system, including 
associated battery chargers, is not an initiator to any accident 
sequence analyzed in the FSAR. Rather, the DC electrical power 
system supports equipment used to mitigate accidents. The proposed 
changes to restructure the TS and change surveillances for batteries 
and chargers to incorporate the updates included in TSTF-500, 
Revision 2, will maintain the same level of equipment performance 
required for mitigating accidents assumed in the FSAR. 
Administrative and mechanical controls are in place to ensure the 
design and operation of the DC systems continues to meet the plant 
design basis described in the FSAR.
    Therefore, operation of the facility in accordance with this 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in the 
margin of safety?

    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The equipment margins will be maintained in 
accordance with the plant-specific design bases as a result of the 
proposed changes. The proposed changes will not adversely affect 
operation of plant equipment. These changes will not result in a 
change to the setpoints at which protective actions are initiated. 
Sufficient DC capacity to support operation of mitigation equipment 
is ensured. The changes associated with the new Battery Monitoring 
and Maintenance Program will ensure that the station batteries are 
maintained in a highly reliable manner. The equipment fed by the DC 
electrical sources will continue to provide adequate power to 
safety-related loads in accordance with analysis assumptions.
    TS changes made in accordance with TSTF-500, Revision 2, 
maintain the same level of equipment performance stated in the FSAR 
and the current TSs. Therefore, the proposed changes do not involve 
a significant reduction 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: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, Inc., 40 Iverness Center 
Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, Houston 
County, Alabama

    Date of amendment request: April 25, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16120A294.
    Description of amendment request: The license proposed three 
changes to

[[Page 36624]]

modifications specified in the March 10, 2015, NFPA [National 
Environmental Policy Act]-805 amendment, Attachment S, Table S-2, 
``Plant Modifications Committed.'' The three proposed modifications 
are: (1) Delete Fire Area 1-041 information from Table S-2, (2) add 
information on item 11, Pyro Panel modification, and, (3) change cable 
2VCHAL07P to cable 2VCFARK2P.
    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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The licensee's analysis 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 updates Attachments M, S, and W of the 
previously approved NFPA-805 LAR [license amendment request] 
submittal for FNP. The attachment revisions are based on the three 
changes to Table S-2 proposed in this LAR. One of the changes is 
justified based on negligible risk impact to Core Damage Frequency 
or Large Early Release Frequency associated with not performing the 
committed modification. The other two changes have no impact on 
accident analysis as they are clarifying or administrative in 
nature.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed changes do not 
adversely affect the ability of structures, systems and components 
(SSCs) to perform their intended safety function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change does not increase the probability or 
consequence of an accident as verified by the risk analysis 
performed.
    Therefore, this proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously identified.

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 updates Attachments M, S, and W of the 
previously approved NFPA-805 LAR submittal for FNP. The attachment 
revisions are based on the three changes to Table S-2 proposed in 
this LAR. One of the changes is justified based on negligible risk 
impact to Core Damage Frequency or Large Early Release Frequency 
associated with not performing the committed modification. The other 
two changes have no impact on accident analysis as they are 
clarifying or administrative in nature. The proposed change relates 
to the availability of fire PRA [probabilistic risk analysis] 
credited component in given fire scenarios.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

3. Does the proposed amendment involve a significant reduction in a 
margin of safety?

    Response: No.
    The proposed amendment updates Attachments M, S, and W of the 
previously approved NFPA-805 LAR submittal for FNP. The attachment 
revisions are based on the three changes to Table S-2 proposed in 
this LAR. One of the changes is justified based on negligible risk 
impact to Core Damage Frequency or Large Early Release Frequency 
associated with not performing the committed modification. The other 
two changes have no impact on accident analysis as they are 
clarifying or administrative in nature.
    The proposed change does not increase the probability or 
consequence of an accident and does not reduce the margin of safety 
as verified by the risk analysis performed.
    Therefore, this 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: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, 40 Iverness Center 
Parkway, Birmingham, AL 35201.
    NRC Branch Chief: Michael T. Markley.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: June 29, 2015.
    Brief description of amendment: The amendment approved a change to 
the Waterford Steam Electric Station, Unit 3, Cyber Security Plan 
Implementation Schedule Milestone 8 full implementation date and a 
related change to the existing operating license physical protection 
license condition.
    Date of issuance: May 10, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 247. A publicly-available version is in ADAMS under 
Accession No. ML16077A270; 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.
    Date of initial notice in Federal Register: September 1, 2015 (80 
FR 52805).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 10, 2016.
    No significant hazards consideration comments received: No.

[[Page 36625]]

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

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

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

    Date of application for amendments: August 18, 2015, as 
supplemented by letter dated April 14, 2016.
    Brief description of amendments: The amendments revised the reactor 
steam dome pressure specified in the technical specification safety 
limits.
    Date of issuance: May 11, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 209, 250, 243, 262, and 257. A publicly-available 
versions is in ADAMS under Accession No. ML16111A104. Documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. : NPF-62, DPR-19, DPR-25, DPR-29, 
and DPR-30. Amendments revised the Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: October 27, 2015 (80 FR 
65812). The supplemental letter dated April 14, 2016, 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 May 11, 2016.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant (DCPP), Units 1 and 2, San Luis Obispo 
County, California

    Date of application for amendments: September 16, 2015.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.4.1, ``RCS [Reactor Coolant System] Pressure, 
Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,'' 
to delete current Tables 3.4.1-1, ``Reduction in Percent RATED THERMAL 
POWER for Reduced RCS Flow Rate, Unit 1,'' and 3.4.1-2, ``Reduction in 
Percent RATED THERMAL POWER for Reduced RCS Flow Rate, Unit 2,'' and 
add RCS thermal design flow (TDF) values to the requirements of TS 
3.4.1. The change also relocates the RCS minimum measured flow (MMF) 
values to the DCPP, Units 1 and 2, core operating limits reports (COLR) 
with a reference to the MMF values in TS 3.4.1 and Surveillance 
Requirements 3.4.1.3 and 3.4.1.4. Figure 2.1.1-1, ``Reactor Core Safety 
Limit,'' has been revised to delete a footnote with references to 
Tables 3.4.1-1 and 3.4.1-2. The change is consistent with NUREG-1431, 
Volume 1, Revision 4.0, ``Standard Technical Specifications, 
Westinghouse Plants,'' April 2012; NRC-approved Technical Specification 
Task Force (TSTF) Change Traveler 339-A, Revision 2, ``Relocate TS 
Parameters to COLR,'' dated June 13, 2000; and NRC-approved WCAP-14483-
A, ``Generic Methodology for Expanded Core Operating Limits Report,'' 
January 1999.
    The change is necessary to correct a non-conservative TS 3.4.1 
total RCS flow rate value for DCPP, Unit 1. The change also ensures 
that the TS stays conservative, if the cycle-specific minimum RCS flow 
is higher than the minimum TDF.
    Date of issuance: May 19, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: Unit 1--226; Unit 2--228. A publicly-available 
version is in ADAMS under Accession No. ML16117A252; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: November 10, 2015 (80 
FR 69714).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 19, 2016.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of amendment request: August 27, 2014, as supplemented by 
letters dated October 31, 2014; February 12, May 12, September 10, and 
November 5, 2015; and January 14 and March 4, 2016.
    Brief description of amendment: The amendment approved a change to 
the Virgil C. Summer Nuclear Station licensing basis to incorporate a 
supplemental analysis for the steam generator tube rupture accident.
    Date of issuance: May 16, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within120 days of issuance.
    Amendment No.: 205. A publicly-available version is in ADAMS under 
Accession No. ML15231A605; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-12: Amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: October 14, 2014 (79 FR 
61661). The supplemental letters dated October 31, 2014; February 12, 
May 12, September 10, and November 5, 2015; and January 14 and March 4, 
2016, 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 May 16, 2016.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

    Date of amendment request: August 31, 2015, as supplemented by 
letters dated January 28, 2016, and March 11, 2016.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.4.14, ``RCS Pressure Isolation Valve (PIV) 
Leakage,'' to eliminate the requirements for the residual heat removal 
system suction valve auto closure interlock function.
    Date of issuance: May 17, 2016.
    Effective date: As of the date of issuance and shall be implemented 
as follows: Unit 1--prior to the first entry into Mode 4, following the 
end-of-cycle refueling outage 27 (scheduled for fall 2016), and Unit 
2--prior to the first entry into Mode 4, following the end-of-cycle 
refueling outage 25 (scheduled for fall 2017).
    Amendment Nos.: 201 (Unit 1) and 197 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16083A265; documents related

[[Page 36626]]

to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-2 and NPF-8: The amendments 
revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: October 27, 2015 (80 FR 
65815). The supplemental letters dated January 28, 2016, and March 11, 
2016, 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 May 17, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 27, 2014, as supplemented by 
letters dated July 2, 2015; September 21, 2015; November 11, 2015; and 
January 29, 2016.
    Brief description of amendments: The amendments modified the SSES 
technical specifications (TSs). Specifically, the amendments modified 
the TSs by relocating specific surveillance frequencies to a licensee-
controlled program, the Surveillance Frequency Control Program, with 
implementation of Nuclear Energy Institute (NEI) 04-10, Revision 1, 
``Risk-Informed Technical Specifications Initiative 5b, Risk-Informed 
Method for Control of Surveillance Frequencies.'' The changes are 
consistent with NRC-approved Technical Specification Task Force 
Improved Standard Technical Specifications Change Traveler (TSTF)-425, 
Revision 3, ``Relocate Surveillance Frequencies to Licensee Control--
RITSTF Initiative 5b.'' The Federal Register notice published on July 
6, 2009 (74 FR 31996), announced the availability of this TSTF 
improvement and included a model no significant hazards consideration 
and safety evaluation (SE).
    This license amendment request was submitted by PPL Susquehanna, 
LLC; however, on June 1, 2015, the NRC staff issued an amendment 
changing the name on the SSES license from PPL Susquehanna, LLC to 
Susquehanna Nuclear, LLC (ADAMS Accession No. ML15054A066). These 
amendments were issued subsequent to an order issued on April 10, 2015, 
to SSES, approving an indirect license transfer of the SSES license to 
Talen Energy Corporation (ADAMS Accession No. ML15058A073).
    Date of issuance: May 20, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment Nos.: 266 (Unit 1) and 247 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16005A234; documents related 
to these amendments are listed in the SE enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-14 and NPF-22: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: March 3, 2015 (80 FR 
11479). The supplemental letters dated July 2, 2015; September 21, 
2015; November 11, 2015; and January 29, 2016, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in an SE dated May 20, 2016.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 27th day of May, 2016.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2016-13255 Filed 6-6-16; 8:45 am]
 BILLING CODE 7590-01-P