[Federal Register Volume 81, Number 138 (Tuesday, July 19, 2016)]
[Notices]
[Pages 46958-46970]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-16925]


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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0141]


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 June 21, 2016, to July 1, 2016. The last 
biweekly notice was published on July 5, 2016 (81 FR 43646).

DATES: Comments must be filed by August 18, 2016. A request for a 
hearing must be filed by September 19, 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 http://www.regulations.gov and search for Docket ID: NRC-2016-0141. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: 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: Lynn Ronewicz, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-1927, email: [email protected].

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID: NRC-2016-0141 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID: NRC-2016-0141.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2016-0141, 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 http://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

[[Page 46959]]

submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment 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 http://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

[[Page 46960]]

would take place before the issuance of any amendment unless the 
Commission finds an imminent danger to the health or safety of the 
public, in which case it will issue an appropriate order or rule under 
10 CFR part 2.
    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 
September 19, 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 
September 19, 2016.

B. Electronic Submissions (E-Filing)

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

[[Page 46961]]

class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, 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 these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

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

    Date of amendment request: May 25, 2016. A publicly available 
version is in ADAMS under Accession No. ML16146A639.
    Description of amendment request: The amendment would replace the 
CR-3 Permanently Defueled Emergency Plan and its associated Emergency 
Action Level (EAL) Bases Manual with the Independent Spent Fuel Storage 
Installation (ISFSI)-Only Emergency Plan (IOEP) and its associated EAL 
Bases Manual. This IOEP will be used at CR-3 after all spent fuel has 
been transferred to the CR-3 ISFSI.
    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?
    The proposed amendment would modify the CR-3 facility operating 
license by revising the emergency plan and revising the EAL scheme. 
CR-3 has permanently ceased operation and is permanently defueled. 
The proposed amendment is conditioned on all spent nuclear fuel 
being removed from wet storage in the spent fuel pools and placed in 
dry storage within the ISFSI. Occurrence of postulated accidents 
associated with spent fuel stored in a spent fuel pool is no longer 
credible in a spent fuel pool devoid of such fuel. The proposed 
amendment has no effect on plant systems, structures, or components 
(SSC) and no effect on the capability of any plant SSC to perform 
its design function. The proposed amendment would not increase the 
likelihood of the malfunction of any plant SSC. The proposed 
amendment would have no effect on any of the previously evaluated 
accidents in the CR-3 Final Safety Analysis Report.
    Since CR-3 has permanently ceased operation, the generation of 
fission products has ceased and the remaining source term continues 
to decay. This continues to significantly reduce the consequences of 
previously evaluated postulated accidents. Therefore, the proposed 
amendment does not involve a significant increase in the 
consequences of a previously evaluated accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed amendment constitutes a revision of the emergency 
planning function commensurate with the ongoing and anticipated 
reduction in radiological source term at CR-3.
    The proposed amendment does not involve a physical alteration of 
the plant. No new or different types of equipment will be installed 
and there are no physical modifications to existing equipment as a 
result of the proposed amendment. Similarly, the proposed amendment 
would not physically change any SSC involved in the mitigation of 
any postulated accidents. Thus, no new initiators or precursors of a 
new or different kind of accident are created. Furthermore, the 
proposed amendment does not create the possibility of a new failure 
mode associated with any equipment or personnel failures. The 
credible events for the ISFSI remain unchanged.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Because the 10 CFR part 50 license for CR-3 no longer authorizes 
operation of the reactor or emplacement or retention of fuel into 
the reactor vessel, as specified in 10 CFR 50.82(a)(2), the 
occurrence of postulated accidents associated with reactor operation 
is no longer credible. With all spent nuclear fuel transferred out 
of wet storage from the spent fuel pools and placed in dry storage 
within the ISFSI, a fuel handling accident is no longer credible. 
There are no longer credible events that would result in 
radiological releases beyond the site boundary exceeding the EPA 
[Environmental Protection Agency] Protective Action Guide exposure 
levels, as detailed in the EPA's ``Protective Action Guide and 
Planning Guidance for Radiological Incidents,'' Draft for Interim 
Use and Public Comment dated March 2013 (PAG [Protective Action 
Guide] Manual).
    The proposed amendment does not involve a change in the plant's 
design, configuration, or operation. The proposed amendment does not 
affect either the way in which the plant structures, systems, and 
components perform their safety function or their design margins. 
Because there is no change to the physical design of the plant, 
there is no change to these margins.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, 550 South Tryon Street, 
Charlotte, NC 28202.
    NRC Branch Chief: Bruce A. Watson.

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

    Date of amendment request: May 10, 2016, as supplemented by letter 
dated May 18, 2016. Publicly-available versions are in ADAMS under 
Accession Nos. ML16131A891 and ML16139A161, respectively.
    Description of amendment request: The amendment would revise the 
safety function lift and lower setpoint tolerances of the safety/relief 
valves (SRVs) that are listed in Surveillance Requirements 3.4.3.1 and 
3.4.4.1 of the Technical Specifications (TSs).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.

[[Page 46962]]

    This proposed amendment has no influence on the probability or 
consequences of any accident previously evaluated. The lower safety 
setpoint tolerance change does not affect the operation of the SRVs 
and it does not affect the as-left setpoint tolerance band which is 
unchanged at 3% of the lift setpoint of the SRVs. The 
change only affects the lower tolerance for opening of the SRVs. The 
proposed amendment does not affect the upper tolerance for SRVs 
safety setpoints, which is the limit that protects from 
overpressurization.
    The proposed amendment does not involve any physical changes to 
the SRVs, nor does it change the safety function of the SRVs. The 
proposed TS revision involves no significant changes to the 
operation of any systems or components in normal or accident 
operating conditions as discussed in the technical evaluation for 
this [license amendment request]. Additionally, the proposed change 
does not involve any significant changes to existing structures, 
systems, or components.
    The proposed amendment does not change any other behavior or 
operation of the SRVs, and, therefore, has no significant impact on 
reactor operation. It also has no significant impact on response to 
any perturbation of reactor operation including transients and 
accidents previously analyzed in the [Final Safety Analysis Report 
(FSAR)].
    Therefore, the proposed amendment does not result in a 
significant increase in the probability or consequences of any 
previously evaluated accident.
    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 from -3% to -5% for the SRV safety setpoint 
lower tolerance only affects the criteria to determine when an as-
found SRV test is considered acceptable. The proposed change does 
not affect the criteria for the setpoint upper tolerance for the 
SRVs.
    The proposed change from -3% to -5% for the SRV safety setpoint 
lower tolerance does not adversely affect the operation of any 
safety-related components or equipment. Since the proposed amendment 
does not involve any hardware changes, significant changes to the 
operation of any systems or components, nor change to existing 
structures, systems, or components, there is no possibility that a 
new or different kind of accident is created.
    The proposed change from -3% to -5% for the SRV safety setpoint 
lower tolerance does not involve any physical changes to the SRVs, 
nor does it change the safety function of the SRVs. The proposed 
change does not require any physical change or alteration of any 
existing plant equipment. No new or different equipment is being 
installed. No installed equipment is being operated in a new or 
different manner. There is no alteration to the parameters within 
which the plant is normally operated. This change does not alter the 
manner in which equipment operation is initiated, nor will the 
functional demands on credited equipment be changed. No alterations 
in the procedures that ensure the plant remains within analyzed 
limits are being proposed. No changes are being made to the 
procedures relied upon to respond to off-normal events as described 
in the FSAR are being proposed by this change. The proposed change 
does not alter assumptions made in the safety analysis and licensing 
basis.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change from -3% to -5% for the SRV safety setpoint 
lower tolerance only affects the criteria to determine when an as-
found SRV test is considered acceptable. This change does not affect 
the criteria for the SRV safety setpoint upper tolerance. The TS 
setpoints for the SRVs are not changed. The as-left setpoint 
tolerances are not changed by the proposed amendment and remain at 
3%.
    The margin of safety is established through the design of the 
plant structures, systems, and components, the parameters within 
which the plant is operated, and the establishment of the setpoints 
for the actuation of equipment relied upon to respond to an event. 
The proposed change from -3% to -5% for the SRV safety setpoint 
lower tolerance does not significantly impact the condition or 
performance of structures, systems, and components relied upon for 
accident mitigation.
    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: William A. Horin, Esq., Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: Robert J. Pascarelli.

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: June 20, 2016. A publicly available 
version is in ADAMS under Accession No. ML16173A371.
    Description of amendment request: The amendments would revise the 
Technical Specification (TS) requirements associated with the storage 
inventory of lube oil for the emergency diesel generators (EDGs). 
Specifically, the TS volume requirements for stored EDG lube oil 
(currently specified in number of gallons) would be replaced with 
volume requirements based on EDG operating time (specified in number of 
days). The volume requirements, specified in number of gallons, along 
with the equivalent number of days of EDG operating time, would be 
included in the TS Bases. As such, the amendments would allow the 
licensee to make changes to the number of gallons using the provisions 
of 10 CFR 50.59, consistent with the TS Bases Control Program specified 
in TS 5.5.10. The proposed changes are based on Revision 1 to Technical 
Specification Task Force (TSTF) Improved Standard Technical 
Specifications Change Traveler TSTF-501, ``Relocate Stored Fuel Oil and 
Lube Oil Volume Values to Licensee Control.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change relocates the volume of diesel lube oil 
required to support 7-day operation of each onsite diesel generator, 
and the volume equivalent to a 6-day supply, to licensee control. 
The specific volume of lube oil equivalent to a 7-day and 6-day 
supply is based on the diesel generator manufacturer's consumption 
values for the run time of the diesel generator. Because the 
requirement to maintain a 7-day supply of diesel lube oil is not 
changed and is consistent with the assumptions in the accident 
analyses, and the actions taken when the volume of lube oil is less 
than a 6-day supply have not changed, neither the probability nor 
the consequences of any accident previously evaluated will be 
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 change does not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. The change 
does not alter assumptions made in the safety analysis but ensures 
that each diesel generator operates as assumed in the accident 
analysis. The proposed change is consistent with the safety analysis 
assumptions. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

[[Page 46963]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change relocates the volume of diesel lube oil 
required to support 7-day operation of each onsite diesel generator, 
and the volume equivalent to a 6-day supply, to licensee control. As 
the bases for the existing limits on diesel lube oil are not 
changed, no change is made to the accident analysis assumptions and 
no margin of safety is reduced as part of this change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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. 50-219, Oyster Creek Nuclear 
Generating Station (OCNGS), Ocean County, New Jersey

    Date of amendment request: May 17, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16138A129.
    Description of amendment request: The proposed amendment would 
revise OCNGS's Technical Specification (TS) Section 6.0, 
``Administrative Controls.''
    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 amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes would not take effect until OCNGS has 
permanently ceased operation and entered a permanently defueled 
condition. The proposed changes would revise the OCNGS TS by 
deleting or modifying certain portions of the TS administrative 
controls described in Section 6.0 of the TS that are no longer 
applicable to a permanently shutdown and defueled facility.
    The proposed changes do not involve any physical changes to 
plant Structures, Systems, and Components (SSCs) or the manner in 
which SSCs are operated, maintained, modified, tested, or inspected. 
The proposed changes do not involve a change to any safety limits, 
limiting safety system settings, limiting control settings, limiting 
conditions for operation, surveillance requirements, or design 
features.
    The deletion and modification of provisions of the 
administrative controls do not directly affect the design of SSCs 
necessary for safe storage of spent irradiated fuel or the methods 
used for handling and storage of such fuel in the Spent Fuel Pool 
(SFP). The proposed changes are administrative in nature and do not 
affect any accidents applicable to the safe management of spent 
irradiated fuel or the permanently shutdown and defueled condition 
of the reactor.
    In a permanently defueled condition, the only credible accidents 
are the Fuel Handling Accident (FHA), Radioactive Liquid Waste 
System Leak, and Postulated Radioactive Releases Due to Liquid Tank 
Failures. Other accidents such as Loss of Coolant Accident, Loss of 
Feedwater, and Reactivity and Power Distribution Anomalies will no 
longer be applicable to a permanently defueled reactor plant.
    The probability of occurrence of previously evaluated accidents 
is not increased, since extended operation in a permanently defueled 
condition will be the only operation allowed, and therefore, bounded 
by the existing analyses. Additionally, the occurrence of postulated 
accidents associated with reactor operation is no longer credible in 
a permanently defueled reactor. This significantly reduces the scope 
of applicable accidents.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to delete and/or modify certain TS 
administrative controls have no impact on facility SSCs affecting 
the safe storage of spent irradiated fuel, or on the methods of 
operation of such SSCs, or on the handling and storage of spent 
irradiated fuel itself. The proposed changes do not result in 
different or more adverse failure modes or accidents than previously 
evaluated because the reactor will be permanently shut down and 
defueled and OCNGS will no longer be authorized to operate the 
reactor.
    The proposed changes do not affect systems credited in the 
accident analysis for the FHA, Radioactive Liquid Waste System Leak, 
and Postulated Radioactive Releases Due to Liquid Tank Failures at 
OCNGS. The proposed changes will continue to require proper control 
and monitoring of safety significant parameters and activities. The 
proposed changes do not result in any new mechanisms that could 
initiate damage to the remaining relevant safety barriers in support 
of maintaining the plant in a permanently shutdown and defueled 
condition (e.g., fuel cladding and SFP cooling). Since extended 
operation in a defueled condition will be the only operation 
allowed, and therefore bounded by the existing analyses, such a 
condition does not create the possibility of a new or different kind 
of accident.
    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. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes involve deleting and/or modifying certain 
TS administrative controls once the OCNGS facility has been 
permanently shutdown and defueled. As specified in 10 CFR 
50.82(a)(2), the 10 CFR 50 license for OCNGS will no longer 
authorize operation of the reactor or emplacement or retention of 
fuel into the reactor vessel following submittal of the 
certifications required by 10 CFR 50.82(a)(1). As a result, the 
occurrence of certain design basis postulated accidents are no 
longer considered credible when the reactor is permanently defueled.
    The only remaining credible accident is a fuel handling accident 
(FHA). The proposed changes do not adversely affect the inputs or 
assumptions of any of the design basis analyses that impact the FHA.
    The proposed changes are limited to those portions of the TS 
administrative controls that are related to the safe storage and 
maintenance of spent irradiated fuel. The requirements that are 
proposed to be revised and/or deleted from the OCNGS TS are not 
credited in the existing accident analysis for the remaining 
applicable postulated accident (i.e., FHA); therefore, they do not 
contribute to the margin of safety associated with the accident 
analysis. Certain postulated DBAs [design-basis accidents] involving 
the reactor are no longer possible because the reactor will be 
permanently shut down and defueled and OCNGS will no longer be 
authorized to operate the reactor.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Acting Branch Chief: Shaun M. Anderson.

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

    Date of amendment request: April 27, 2016. A publicly available 
version is in

[[Page 46964]]

ADAMS under Accession No. ML16120A432.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TSs) by eliminating Section 5.5.8, 
``Inservice Testing Program,'' and adding a new defined term, 
``Inservice Testing Program,'' to the TS Definitions section. The 
proposed amendments are consistent with Technical Specification Task 
Force (TSTF) Traveler TSTF-545, Revision 3, ``TS Inservice Testing 
Program Removal & Clarify SR [Surveillance Requirement] Usage Rule 
Application to Section 5.5 Testing,'' dated October 21, 2015 (ADAMS 
Accession No. ML15294A555).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS Chapter 5, ``Administrative 
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating 
the ``Inservice Testing Program'' specification. Most requirements 
in the Inservice Testing Program are removed, as they are 
duplicative of requirements in the [American Society of Mechanical 
Engineers (ASME) Operations and Maintenance (OM) Code], as clarified 
by Code Case OMN-20, ``Inservice Test Frequency.'' The remaining 
requirements in the Section 5.5.8 [Inservice Testing (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 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 amendment 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 amendment 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 existing TS 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: Timothy P. Matthews, Esq., Morgan, Lewis and 
Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Branch Chief: Robert J. Pascarelli.

NextEra Energy Seabrook LLC, Docket No. 50-443, Seabrook Station, Unit 
No. 1, Rockingham County, New Hampshire

    Date of amendment request: March 31, 2016, as supplemented by 
letter dated May 31, 2016. Publicly-available versions are in ADAMS 
under Accession Nos. ML16095A278 and ML16159A194, respectively.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 6.15, ``Containment Leakage Rate Testing 
Program,'' to require a program that is in accordance with Nuclear 
Energy Institute (NEI) Topical Report NEI 94-01, Revision 3-A, 
``Industry Guideline for Implementing Performance-Based Option of 10 
CFR part 50, Appendix J'' (ADAMS Accession No. ML12221A202). The 
proposed change would allow extension of the Type A test interval up to 
one test in 15 years, and extension of the Type C test interval up to 
75 months, based on acceptable performance history as defined in NEI 
94-01, Revision 3-A.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, ``Industry Guideline for Implementing 
Performance-Based Option of 10 CFR part 50, Appendix J,'' for 
development of the Seabrook performance-based containment testing 
program. NEI 94-01 allows, based on risk and performance, an 
extension of Type A and Type C containment leak test intervals. 
Implementation of these guidelines continues to provide adequate 
assurance that during design basis accidents, the primary 
containment and its components will limit leakage rates to less than 
the values assumed in the plant safety analyses.
    The findings of the Seabrook risk assessment confirm the general 
findings of previous studies that the risk impact with extending the 
containment leak rate is small. Per the guidance provided in 
Regulatory Guide 1.174, an extension of the leak test interval in 
accordance with NEI 94-01, Revision 3-A results in an estimated 
change within the small change region.
    Since the change is implementing a performance-based containment 
testing

[[Page 46965]]

program, 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 requirement for containment leakage rate 
acceptance will not be changed by this amendment. Therefore, the 
containment will continue to perform its design function as a 
barrier to fission product releases.
    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 change to implement a performance-based containment 
testing program, associated with integrated leakage rate test 
frequency, does not change the design or operation of structures, 
systems, or components of the plant.
    The proposed changes would continue to ensure containment 
integrity and would ensure operation within the bounds of existing 
accident analyses. There are no accident initiators created or 
affected by these changes. Therefore, the proposed changes will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    Margin of safety is related to confidence in the ability of the 
fission product barriers (fuel cladding, reactor coolant system, and 
primary containment) to perform their design functions during and 
following postulated accidents. The proposed change to implement a 
performance-based containment testing program, associated with 
integrated leakage rate test frequency, does not affect plant 
operations, design functions, or any analysis that verifies the 
capability of a structure, system, or component of the plant to 
perform a design function. In addition, this change does not affect 
safety limits, limiting safety system setpoints, or limiting 
conditions for operation.
    The specific requirements and conditions of the TS Containment 
Leakage Rate Testing Program exist to ensure that the degree of 
containment structural integrity and leak-tightness that is 
considered in the plant safety analysis is maintained. The overall 
containment leak rate limit specified by TS is maintained. This 
ensures that the margin of safety in the plant safety analysis is 
maintained. The design, operation, testing methods and acceptance 
criteria for Type A, B, and C containment leakage tests specified in 
applicable codes and standards would continue to be met, with the 
acceptance of this proposed change, since these are not affected by 
implementation of a performance-based containment testing program.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William Blair, Managing Attorney--Nuclear, 
Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-
0420.
    NRC Branch Chief: Douglas A. Broaddus.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: May 11, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16132A374.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) requirements by deleting TS Action 
Statement 3.4.2.1.b concerning stuck open safety/relief valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed TS change deletes Action Statement 3.4.2.1.b 
concerning safety/relief valves. The two (2) minute action 
represents detailed methods of responding to an event, and 
therefore, if eliminated, would not result in increasing the 
probability of the event, nor act as an initiator of an event. 
Limiting condition for operation 3.6.2.1, ``Depressurization 
Systems--Suppression Chamber,'' and plant procedures provide 
operators with appropriate direction for response to a suppression 
pool high temperature (which could be caused by a stuck open relief 
valve). Providing specific direction to close the valve within two 
(2) minutes does not provide additional plant protection beyond what 
is provided for in plant procedures and TS 3.6.2.1.
    Therefore, this action can be eliminated, and will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed TS change deletes Action Statement 3.4.2.1.b 
concerning safety/relief valves. This change does not change the 
design or configuration of the plant. No new operation or failure 
modes are created, nor is a system-level failure mode created that 
is different than those that already exist.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not involve a significant reduction in 
a margin of safety, nor does it affect any analytical limits. There 
are no changes to accident or transient core thermal hydraulic 
conditions, or fuel or reactor coolant boundary design limits, as a 
result of the proposed change. The proposed change will not alter 
the assumptions or results of the analysis contained in the Updated 
Final Safety Analysis Report (UFSAR).
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC-N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Douglas A. Broaddus.

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

    Date of amendment request: May 17, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16138A431.
    Description of amendment request: The amendment request proposes 
changes to the Updated Final Safety Analysis Report (UFSAR) in the form 
of departures from the incorporated plant-specific Design Control 
Document (DCD) Tier 2 information and involves changes to related Tier 
1 information, with corresponding changes to the associated Combined 
License (COL) Appendix C information. Pursuant to the provisions of 10 
CFR 52.63(b)(1), an exemption from elements of the design as certified 
in the 10 CFR part 52, Appendix D, ``Design Certification Rule for the 
AP1000 Design,'' is also requested for the plant-specific DCD Tier 1 
material departures. Specifically, the requested amendment proposes 
changes to the concrete wall thickness tolerance for the column line N 
wall, from column lines 2 to 4 from elevation 100'-0'' to 135'-3'', 
from plus or minus 1 inch to plus 4 inches.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 46966]]

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.
    As indicated in the UFSAR Subsection 3.8.4.1.2, the auxiliary 
building contains structural modules in the south side of the 
building that include the spent fuel pool, fuel transfer canal, and 
cask loading and washdown pits. The increase in tolerance associated 
with the concrete thickness of the concrete wall for the column line 
N from column line 2 to 4 and the deviation from ACI 349-01 does not 
involve any accident initiating components or events, thus leaving 
the probabilities of an accident unaltered. The increased tolerance 
does not adversely affect any safety-related structures or equipment 
nor does the increased tolerance reduce the effectiveness of a 
radioactive material barrier. Thus, the proposed changes would not 
affect any safety-related accident mitigating function served by the 
containment internal structures.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed tolerance increase and code deviation from ACI 349-
01 does not change the performance of the affected radiologically 
controlled portion of the auxiliary building. As demonstrated by the 
continued conformance to the other applicable codes and standards 
governing the design of the structures, and in conjunction with the 
analysis of a special system of construction in accordance with ACI 
349-01 Section 1.4, the wall with an increased concrete thickness 
tolerance continues to withstand the same effects as previously 
evaluated. There is no change to the design function of the affected 
module and wall, and no new failure mechanisms are identified as the 
same types of accidents are presented to the wall before and after 
the change.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to increase the concrete thickness tolerance 
for the column line N wall from column line 2 to 4 identified in COL 
Appendix C Table 3.3-1 does not alter any design function, design 
analysis, or safety analysis input or result, and sufficient margin 
exists to justify departure from the ACI 349-01 requirements for the 
wall. As such, because the system continues to respond to design 
basis accidents in the same manner as before without any changes to 
the expected response of the structure, no safety analysis or design 
basis acceptance limit/criterion is challenged or exceeded by the 
proposed changes. Accordingly, no safety margin is reduced by the 
increase of the wall concrete thickness tolerance.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Acting Branch Chief: Jennifer Dixon-Herrity.

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

    Date of amendment request: May 5, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16126A276.
    Description of amendment request: The proposed changes would revise 
the Combined Licenses (COLs) concerning the design details of the 
safety-related passive core cooling system (PXS), the nonsafety-related 
normal residual heat removal system (RNS), and the nonsafety-related 
containment air filtration system (VFS). The amendment request proposes 
changes to the Updated Final Safety Analysis Report (UFSAR) in the form 
of departures from the plant-specific Design Control Document (DCD) 
Tier 2 information and involves changes to related plant-specific DCD 
Tier 1 information, with corresponding changes to the associated COL 
Appendix C information. Because this proposed change would require a 
departure from Tier 1 information in the Westinghouse Advanced Passive 
1000 DCD, the licensee also requests an exemption from the requirements 
of the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that initiate an analyzed accident or alter any 
structures, systems, and components (SSCs) accident initiator or 
initiating sequence of events. The proposed changes result from 
identifying PSX, RNS, and VFS piping lines required to be described 
in the licensing basis as ASME [American Society of Mechanical 
Engineers] Code Section III, evaluated to meet the LBB [leak-before-
break] design criteria, or designed to withstand combined normal and 
seismic design basis loads without a loss of functional capability. 
Neither planned or inadvertent operation nor failure of the PXS, 
RNS, or VFS is an accident initiator or part of an initiating 
sequence of events for an accident previously evaluated. Therefore, 
the probabilities of the accidents evaluated in the UFSAR are not 
affected.
    The proposed changes do not have an adverse impact on the 
ability of the PXS, RNS, or VFS to perform their design functions. 
The design of the PXS, RNS, and VFS continues to meet the same 
regulatory acceptance criteria, codes, and standards as required by 
the UFSAR. In addition, the changes ensure that the capabilities of 
the PXS, RNS, and VFS to mitigate the consequences of an accident 
meet the applicable regulatory acceptance criteria, and there is no 
adverse effect on any safety-related SSC or function used to 
mitigate an accident. The changes do not affect the prevention and 
mitigation of other abnormal events, e.g., anticipated operational 
occurrences, earthquakes, floods and turbine missiles, or their 
safety or design analyses. Therefore, the consequences of the 
accidents evaluated in the UFSAR are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that may initiate a new or different kind of accident, 
or alter any SSC such that a new accident initiator or initiating 
sequence of events is created. The proposed changes result from 
identifying PXS, RNS, and VFS piping lines required to be described 
in the licensing basis as ASME Code Section III, evaluated to meet 
the LBB design criteria, or designed to withstand combined normal 
and seismic design basis loads without a loss of functional 
capability. These proposed changes do not adversely affect any other 
PXS, RNS, VFS, or SSC design functions or methods of operation in a 
manner that results in a new failure mode, malfunction, or sequence 
of events that affect safety-related or nonsafety-related equipment. 
Therefore, this activity does not allow for a new fission product 
release path, result in a new fission product barrier failure mode, 
or create a new sequence of events that results in significant fuel 
cladding failures.
    Therefore, the requested amendment does not create the 
possibility of a new or different

[[Page 46967]]

kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes maintain existing safety margins. The 
proposed changes ensure that PXS, RNS, and VFS design requirements 
and design functions are met. The proposed changes maintain existing 
safety margin through continued application of the existing 
requirements of the UFSAR, while adding additional design features 
to ensure the PXS, RNS, and VFS perform the design functions 
required to meet the existing safety margins. Therefore, the 
proposed changes satisfy the same design functions in accordance 
with the same codes and standards as stated in the UFSAR. These 
changes do not adversely affect any design code, function, design 
analysis, safety analysis input or result, or design/safety margin. 
Because no safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the proposed changes, no 
margin of safety is reduced.
    Therefore, the requested amendment does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Acting Branch Chief: Jennifer Dixon-Herrity.

STP Nuclear Operating Company (STPNOC), Docket No. 50-498, South Texas 
Project (STP), Unit 1, Matagorda County, Texas

    Date of amendment request: April 7, 2016, as supplemented by letter 
dated May 25, 2016. Publicly-available versions are in ADAMS under 
Accession Nos. ML16110A297 and ML16162A196, respectively.
    Description of amendment request: The amendment would revise 
Technical Specification 5.3.2 for STP, Unit 1, to allow permanent 
operation with 56 full-length control rods with no control rod assembly 
in core location D-6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    STPNOC has performed a multi-cycle assessment on previous Unit 1 
reactor cores and evaluated the consequences associated with removal 
of Control Rod D-6. The assessment indicates that removal of Control 
Rod D-6 does impact reactivity parameters (e.g., shutdown margin and 
trip reactivity); however, sufficient margin exists to ensure the 
Updated Final Safety Analysis Report (UFSAR) accident analysis 
limits continue to be met. The physical changes associated with the 
removal of Control Rod D-6 do not impact the probability of 
occurrence of a previously evaluated accident. 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.
    Operation of STP Unit 1 with Control Rod D-6 removed will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. To preserve the reactor coolant 
system flow characteristics in the reactor core, a flow restrictor 
will be installed at the top of the D-6 guide tube housing. 
Installation of this component will not prevent the remaining 56 
control rods from performing the required design function of 
providing adequate shutdown margin. No new operator actions are 
created as a result of the proposed change. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Operation of STP Unit 1 with Control Rod D-6 removed will not 
involve a significant reduction in a margin of safety. The margin of 
safety is established by setting safety limits and operating within 
those limits. The proposed change does not alter a UFSAR design 
basis or safety limit and does not change any setpoint at which 
automatic actuations are initiated. STPNOC will continue to confirm 
all safety analysis limits remain bounding on a cycle-specific basis 
using an NRC-approved Westinghouse core reload evaluation 
methodology. 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendment involves no significant hazards consideration.
    Attorney for licensee: Kym Harshaw, General Counsel, STP Nuclear 
Operating Company, P.O. Box 289, Wadsworth, TX 77483.
    NRC Branch Chief: Robert J. Pascarelli.

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

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

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2 (MPS2), New London County, Connecticut

    Date of amendment request: December 17, 2012, as supplemented by 
letters dated February 25, 2013; May 28, 2013; July 21, 2015; December 
18, 2015; and June 1, 2016.
    Brief description of amendment: The amendment revised the MPS2 
Technical Specifications (TSs) to reflect the results and constraints 
of a new criticality safety analysis for fuel assembly storage in the 
MPS2 fuel storage racks.
    Date of issuance: June 23, 2016.

[[Page 46968]]

    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 327. A publicly-available version is in ADAMS under 
Accession No. ML16003A008; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-65: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: June 11, 2013 (78 FR 
35060). The supplemental letters dated May 28, 2013; July 21, 2015; 
December 18, 2015; and June 1, 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 June 23, 2016.
    No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-369, 50-370, 50-413, and 50-
414, McGuire Nuclear Station (McGuire), Units 1 and 2, Mecklenburg 
County, North Carolina, and Catawba Nuclear Station (Catawba), Units 1 
and 2, York County, South Carolina

    Date of amendment request: August 20, 2015.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) to allow the use of Optimized Zirlo\TM\. 
Specifically, the proposed changes modify TS 4.2.1 to add Optimized 
Zirlo\TM\ as an allowable cladding and TS 5.6.5.b to add associated 
methodologies for determining the core operating limits report.
    Date of issuance: June 21, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: McGuire--288 (Unit 1) and 267 (Unit 2); Catawba--
284 (Unit 1) and 280 (Unit 2). A publicly-available version is in ADAMS 
under Accession No. ML16105A326; documents related to these amendments 
are listed in the Safety Evaluation enclosed with the amendments.
    Facility Operating License Nos. NPF-9, NPF-17, NPF-35, and NPF-52: 
Amendments revised the Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: November 24, 2015 (80 
FR 73236).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 21, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 9, 2015, as supplemented by letter 
dated January 7, 2016.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' to resolve an operable but degraded non-conforming 
issue associated with the reactor coolant pump under-frequency trip 
setpoint allowable value for the McGuire Nuclear Station, Units 1 and 
2.
    Date of issuance: June 21, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 287 (Unit 1) and 266 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16109A084; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: October 13, 2015 (80 FR 
61479). The supplemental letter dated January 7, 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 June 21, 2016.
    No significant hazards consideration comments received: No.

Entergy Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1 (RBS), West Feliciana Parish, Louisiana

    Date of amendment request: June 29, 2015, as supplemented by letter 
dated December 3, 2015.
    Brief description of amendment: The amendment revised the full 
implementation date (Milestone 8) of the RBS Cyber Security Plan and 
revised the associated license condition for the Facility Operating 
License. The license was also revised, in part, to include 
administrative and editorial corrections.
    Date of issuance: June 21, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 190. A publicly-available version is in ADAMS under 
Accession No. ML16124A688; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: April 5, 2016 (81 FR 
19647).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 21, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 17, 2011, as supplemented by 
letters dated January 26, September 27 and October 16, 2012; May 16, 
June 26, and December 18, 2013; June 11, 2014; March 12, April 10, May 
14, August 27, September 8, September 24, and October 13, 2015; and 
January 18, 2016.
    Brief description of amendment: The amendment permits the licensee 
to adopt a new risk-informed, performance-based fire protection 
licensing basis for Waterford 3, in accordance with the requirements in 
10 CFR 50.48(a) and (c) and the guidance in NRC Regulatory Guide 1.205, 
``Risk-Informed, Performance-Based Fire Protection for Existing Light-
Water Nuclear Power Plants,'' December 2009; National Fire Protection 
Association (NFPA) 805, ``Performance-Based Standard for Fire 
Protection for Light Water Reactor Electric Generating Plants'' (2001 
Edition); and Nuclear Energy Institute 04-02, ``Guidance for 
Implementing a Risk-Informed, Performance-Based Fire Protection Program 
under 10 CFR 50.48(c),'' Revision 2.
    Date of issuance: June 27, 2016.
    Effective date: As of the date of issuance and shall be implemented 
as described in the transition license conditions.
    Amendment No.: 248. A publicly-available version is in ADAMS under 
Accession No. ML16126A033; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.

[[Page 46969]]

    Date of initial notice in Federal Register: April 10, 2012 (77 FR 
21597). The supplements dated September 27 and October 16, 2012; May 
16, June 26, and December 18, 2013; June 11, 2014; March 12, April 10, 
May 14, August 27, September 8, September 24, and October 13, 2015; and 
January 18, 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 June 27, 2016.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-244, R. E. Ginna Nuclear 
Power Plant, Wayne County, New York

    Date of amendment request: June 4, 2015, as supplemented by letters 
dated February 3, 2016; March 29, 2016; and June 16, 2016.
    Brief description of amendment: The amendment relocated specific 
technical specification surveillance frequencies to a licensee-
controlled program with the adoption of Technical Specification Task 
Force (TSTF) Traveler TSTF-425, Revision 3, ``Relocate Surveillance 
Frequencies to Licensee Control--Risk Informed Technical Specification 
Task Force Initiative 5b''. Additionally, the change added a new 
program, the Surveillance Frequency Control Program, to Technical 
Specification Section 5, Administrative Controls.
    Date of issuance: June 28, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 122. A publicly-available version is in ADAMS under 
Accession No. ML16125A485; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 13, 2015 (80 FR 
61482). The supplemental letters dated February 3, 2016; March 29, 
2016; and June 16, 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 June 28, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 6, 2015, as supplemented by 
letter dated March 25, 2016.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) related to moderator temperature 
coefficient requirements.
    Date of issuance: June 20, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos: 271 (Unit No. 3) and 266 (Unit No. 4). A publicly-
available version is in ADAMS under Accession No. ML16120A473; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: March 8, 2016 (81 FR 
12141). The supplemental letter dated March 25, 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 June 20, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 15, 2015.
    Brief description of amendment: The amendment adopts the NRC-
approved Technical Specifications Task Force (TSTF) Standard Technical 
Specifications Change Traveler TSTF-523, Revision 2, ``Generic Letter 
2008-01, Managing Gas Accumulation.''
    Date of issuance: June 21, 2016.
    Effective date: As of the date of issuance and shall be implemented 
prior to the startup from the 2017 refueling outage.
    Amendment No.: 189. A publicly-available version is in ADAMS under 
Accession No. ML16125A165; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-22: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 13, 2015 (80 FR 
61484).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 21, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 10, 2016.
    Brief description of amendments: The amendments extend the 
implementation period for the Salem, Unit No. 1, License Amendment No. 
311, and the Salem, Unit No. 2, License Amendment No. 292, which were 
effective as of the date of issuance (i.e., March 7, 2016). 
Specifically, the implementation period for the above amendments has 
been extended from July 5, 2016 (i.e., 120 days from the date of 
issuance), to prior to entry into Mode 6 for the Salem, Unit No. 1, 
Fall 2017 refueling outage (1R25), and prior to entry into Mode 6 for 
the Salem, Unit No. 2, Spring 2017 refueling outage (2R22), to align 
with the outages for which the replacement of the source range and 
intermediate range detectors is scheduled.
    Date of issuance: June 29, 2016.
    Effective date: As of the date of issuance and shall be implemented 
by July 5, 2016.
    Amendment Nos.: 314 (Unit No. 1) and 295 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML16137A579; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-70 and DPR-75: The 
amendments revised the Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: May 23, 2016 (81 FR 
32351).
    The Commission's related evaluation of the amendments and final no 
significant hazards consideration determination are contained in a 
Safety Evaluation dated June 29, 2016.
    No significant hazards consideration comments received: No.

[[Page 46970]]

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

    Date of amendment request: March 4, 2016.
    Brief description of amendment: The amendment revised the date of 
the Cyber Security Plan implementation schedule Milestone 8 and 
paragraph 2.E in the Facility Operating License.
    Date of issuance: June 23, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 14 days of issuance.
    Amendment No.: 106. A publicly-available version is in ADAMS under 
Accession No. ML16146A745; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-90: Amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: April 19, 2016 (81 FR 
23011).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 23, 2016.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 8th day of July 2016.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2016-16925 Filed 7-18-16; 8:45 am]
 BILLING CODE 7590-01-P