[Federal Register Volume 81, Number 206 (Tuesday, October 25, 2016)]
[Notices]
[Pages 73428-73447]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-25641]


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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0214]


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses and Combined Licenses Involving No Significant 
Hazards Considerations

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, 
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is 
publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued, and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from September 27, 2016, to October 7, 2016. The 
last biweekly notice was published on October 11, 2016.

DATES: Comments must be filed by November 25, 2016. A request for a 
hearing must be filed by December 27, 2016.

ADDRESSES: You may submit comments by any of the following methods.
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0214. 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-2016-0214, facility name, unit 
number(s), plant docket number, application date, and subject 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-2016-0214.
     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 this 
document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2016-0214, facility name, unit 
number(s), plant docket number, 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:/
/

[[Page 73429]]

www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that 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.

I. 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 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and a petition to intervene (petition) 
with respect to the action. Petitions shall be filed in accordance with 
the Commission's ``Agency Rules of Practice and Procedure'' in 10 CFR 
part 2. Interested persons 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 petition 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 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 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 
petitioner; (2) the nature of the petitioner's right under the Act to 
be made a party to the proceeding; (3) the nature and extent of the 
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 petitioner's interest. The petition 
must also set forth the specific contentions which the 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 
petitioner shall provide a brief explanation of the bases for the 
contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion to support 
its position on the issue. 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 proceeding. The contention must be one 
which, if proven, would entitle the petitioner to relief. A 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 consistent with the NRC's 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) through (iii).
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of any amendment unless the Commission finds an imminent 
danger to the health

[[Page 73430]]

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 December 27, 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 set forth in this section, except that under 10 CFR 
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. Details regarding the opportunity to 
make a limited appearance will be provided by the presiding officer if 
such sessions are scheduled.

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 (hereinafter 
``petition''), 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, as amended at 77 
FR 46562, August 3, 2012). The E-Filing process requires participants 
to submit and serve all adjudicatory documents over the internet, or in 
some cases to mail copies on electronic storage media. Participants may 
not submit paper copies of their filings unless they seek an exemption 
in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [email protected], or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a petition (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 available on the NRC's public Web site at 
http://www.nrc.gov/site-help/e-submittals/adjudicatory-sub.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 Electronic Filing Help Desk will not be 
able to offer assistance in using unlisted software.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a petition. 
Submissions should be in Portable Document Format (PDF). Additional 
guidance on PDF submissions is available on the NRC's public Web site 
at http://www.nrc.gov/site-help/electronic-sub-ref-mat.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 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 Electronic 
Filing 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 Electronic Filing Help Desk is available between 9 a.m. 
and 7 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 
stating why there is good cause for not filing electronically and 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal

[[Page 73431]]

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 petition 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.
    The Commission will issue a notice or order granting or denying a 
hearing request or intervention petition, designating the issues for 
any hearing that will be held and designating the Presiding Officer. A 
notice granting a hearing will be published in the Federal Register and 
served on the parties to the hearing.
    For further details with respect to these license amendment 
applications, see the application for amendment, which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

Dominion Energy Kewanee, Inc. (DEK), Docket No. 50-305, Kewanee Power 
Station (KPS), Carlton, Wisconsin

    Date of amendment request: September 14, 2015. A publicly available 
version is in ADAMS under Accession No. ML15261A238.
    Description of amendment request: The amendment would revise the 
KPS Permanently Defueled Emergency Plan (PDEP) and the Permanently 
Defueled Emergency Action Levels (EAL) Bases Document. DEK requests 
revisions of the PDEP and the EAL Bases Document that reflect DEK's 
plan to transfer all spent fuel to the independent spent fuel storage 
installation (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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would modify the KPS renewed facility 
operating license by revising the emergency plan and revising the 
EAL scheme. KPS 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 pool 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, and 
components (SSCs) 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 KPS Updated Safety Analysis Report 
(USAR).
    Since KPS 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 postulated accidents.
    Therefore, the proposed amendment does not involve a significant 
increase in the consequences of a previously evaluated accident.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment constitutes a revision of the emergency 
planning function commensurate with the ongoing and anticipated 
reduction in radiological source term at KPS.
    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 SSCs 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. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    Because the 10 CFR part 50 license for KPS 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 nuclear spent fuel pool transferred 
out of wet storage from the spent fuel pool 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 
any releases beyond the site boundary exceeding the EPA PAG 
[Environmental Protection Agency protective action guideline] 
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 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 any of these margins.
    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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
    NRC Branch Chief: Bruce A. Watson.

Dominion Energy Kewanee, Inc. (DEK), Docket No. 50-305, Kewanee Power 
Station (KPS), Carlton, Wisconsin

    Date of amendment request: July 28, 2016. A publicly available 
version is in ADAMS under Accession No. ML16216A187.
    Description of amendment request: The amendment would revise the 
KPS Updated Safety Analysis Report (USAR) Section 9.5.2.2.4, 
``Auxiliary Building Crane,'' to: (1) Add a description of a non-single 
failure proof intermediate lifting device that DEK intends to use 
during a specific spent fuel cask handling activity in the auxiliary 
building, and (2) incorporate a new load drop analysis applicable to 
the use of this intermediate lifting device. The amendment also 
includes (for information) a new Technical Requirements Manual section 
that governs the use of the non-single failure proof intermediate 
lifting device to ensure compliance with the required parameters in the 
load drop analysis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

[[Page 73432]]

    The probability of a heavy load drop onto fuel is unchanged by 
this amendment since the intermediate lift device is not used for 
handling loaded or unloaded spent fuel canisters in or around the 
spent fuel pool. Heavy load lifts in and around the spent fuel pool 
will continue to be performed per the current licensing basis.
    The proposed amendment has no effect on the capability of any 
plant systems, structures, and components (SSCs) to perform their 
design functions. The spent fuel pool is unaffected by the proposed 
amendment. The design function of the auxiliary building crane is 
not changed. Other lifting devices and interfacing lifting points 
associated with spent fuel cask handling are designed in accordance 
with applicable NRC guidance pertaining to single failure proof 
lifting systems. Therefore, 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 KPS USAR.
    Therefore, the proposed amendment does not involve a significant 
increase in the consequences of a previously evaluated accident.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment does not affect cask handling activities 
in or around the KPS spent fuel pool. Drops of heavy loads will 
continue to be very improbable events. Use of a different type of 
equipment to lift spent fuel canisters does not involve any new or 
different kind of accident.
    The proposed amendment does not involve a physical alteration of 
the plant. Similarly, the proposed amendment would not physically 
change any SSCs involved in the mitigation of any postulated 
accidents. The physical structure of the spent fuel canisters is not 
altered by this amendment.
    The possibility of a heavy load drop onto fuel remains non-
credible since the intermediate lift device is not used to handle 
spent fuel canisters in or around the spent fuel pool. Heavy load 
lifts in and around the spent fuel pool will continue to be 
performed per the current licensing basis. The proposed amendment 
does not impact safe shutdown equipment. The spent fuel pool, 
including its cooling and inventory makeup, is unaffected by the 
proposed amendment.
    The current licensing basis (USAR Section 14.2.1) includes 
evaluations of the consequences of a fuel handling accident 
involving failure of fuel cladding. Postulation of a canister load 
drop creates the possibility of a new initiator of this previously 
evaluated accident (failure of fuel cladding) caused by the 
postulated non-mechanistic single failure of the intermediate lift 
device. The analysis concludes that the postulated drop of a 
canister loaded with fuel assemblies would not result in failure of 
canister integrity (and therefore there would be no radiological 
release). The consequences of a canister drop are bounded by the 
current licensing scenario of a fuel handling accident.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    Heavy load handling will continue to be conducted in accordance 
with NRC approved methods. Analysis of a postulated load drop of a 
loaded spent fuel canister demonstrates satisfactory outcomes.
    The proposed amendment does not involve a change in the plant's 
design, configuration, or operation. The proposed amendment does not 
significantly 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 likewise no significant change to any of these 
margins.
    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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
    NRC Branch Chief: Bruce A. Watson.

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

    Date of amendment request: August 31, 2016. A publicly available 
version is in ADAMS under Accession No. ML16243A259.
    Description of amendment request: The amendment would revise the 
Operating License and associated Permanently Defueled Technical 
Specifications (PDTS) to reflect removal of all CR-3 spent nuclear fuel 
from the spent fuel pools and its transfer to dry cask storage within 
the onsite Independent Spent Fuel Storage Installation (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?
    Response: No.
    The proposed amendment would modify the CR-3 facility operating 
license and PDTS by deleting the portions of the license and PDTS 
that are no longer applicable to a facility with no spent nuclear 
fuel stored in the spent fuel pools, while modifying the remaining 
portions to correspond to all nuclear fuel stored within an ISFSI. 
This amendment will be implemented within 60 days following DEF's 
submittal of written notification to the NRC that all spent fuel 
assemblies have been transferred out of the spent fuel pools and 
placed in dry storage within the ISFSI.
    The definition of safety-related structures, systems, and 
components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are 
those relied on to remain functional during and following design 
basis events to assure:
    1. The integrity of the reactor coolant boundary;
    2. The capability to shutdown the reactor and maintain it in a 
safe shutdown condition; or
    3. The capability to prevent or mitigate the consequences of 
accidents which could result in potential offsite exposures 
comparable to the applicable guideline exposures set forth in 10 CFR 
50.43(a)(1) or 100.11.
    The first two criteria (integrity of the reactor coolant 
pressure boundary and safe shutdown of the reactor) are not 
applicable to a plant in a permanently defueled condition. The third 
criterion is related to preventing or mitigating the consequences of 
accidents that could result in potential offsite exposures exceeding 
limits. However, after all nuclear spent fuel assemblies have been 
transferred to dry cask storage within an ISFSI, none of the SSCs at 
CR-3 are required to be relied on for accident mitigation. 
Therefore, none of the SSCs at CR-3 meet the definition of a safety-
related SSC stated in 10 CFR 50.2. The proposed deletion of 
requirements in the PDTS does not affect systems credited in any 
accident analysis at CR-3.
    Section 14 of the CR-3 Final Safety Analysis Report (FSAR) 
described the design basis accidents (DBAs) related to the spent 
fuel pools. These postulated accidents are predicated on spent fuel 
being stored in the spent fuel pools. With the removal of the spent 
fuel from the spent fuel pools, there are no remaining spent fuel 
assemblies to be monitored and there are no credible accidents that 
require the actions of a Certified Fuel Handler, Shift Manager, or a 
Non-certified Operator to prevent occurrence or mitigate the 
consequences of an accident.
    The proposed changes do not have an adverse impact on the 
remaining decommissioning activities or any of their postulated 
consequences.
    The proposed changes related to the relocation of certain 
administrative requirements do not affect operating procedures or 
administrative controls that have the function of preventing or 
mitigating any accidents applicable to the safe management of 
irradiated fuel or decommissioning of the facility.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

[[Page 73433]]

    Response: No.
    The proposed changes eliminate the operational requirements and 
certain design requirements associated with the storage of the spent 
fuel in the spent fuel pools, and relocate certain administrative 
controls to the Quality Assurance Program Description or other 
licensee controlled document.
    After the removal of the spent fuel from the spent fuel pools 
and transfer to the ISFSI, there are no spent fuel assemblies that 
remain in the spent fuel pools. Coupled with a prohibition against 
storage of fuel in the spent fuel pools, the potential for fuel 
related accidents is removed. The proposed changes do not introduce 
any new failure modes.
    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 removal of all spent nuclear fuel from the spent fuel pools 
into storage in casks within an ISFSI, coupled with a prohibition 
against future storage of fuel within the spent fuel pools, removes 
the potential for fuel related accidents.
    The design basis and accident assumptions within the CR-3 FSAR 
and the PDTS relating to safe management and safety of spent fuel in 
the spent fuel pools are no longer applicable. The proposed changes 
do not affect remaining plant operations, systems, or components 
supporting decommissioning activities.
    The requirements for systems, structures, and components (SSCs) 
that have been removed from the CR-3 PDTS are not credited in the 
existing accident analysis for any applicable postulated accident; 
and as such, do not contribute to the margin of safety associated 
with the accident analysis.
    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.

Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear 
Power Plant (HNP), Unit 1, Wake and Chatham Counties, North Carolina

    Date of amendment request: June 29, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16182A387.
    Description of amendment request: The amendment would revise HNP 
Technical Specifications (TSs) to (1) delete the Gaseous Radwaste 
Treatment System definition from TSs, (2) relocate Explosive Gas 
Mixture TS requirements and Liquid Holdup Tanks TS requirements to a 
licensee-controlled program in the Procedures and Programs TSs section, 
and (3) modify the Gas Storage Tank Radioactivity Monitoring Program 
TSs into an Explosive Gas and Storage Tank Radioactivity Monitoring 
Program to include controls for potentially explosive gas mixtures and 
the quantity of radioactivity contained in unprotected outdoor liquid 
storage tanks.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes are administrative in nature and alter only the 
format and location of programmatic controls and procedural details 
relative to explosive gas monitoring and liquid holdup tanks. 
Existing TS containing procedural details are being relocated to 
licensee control. Compliance with applicable regulatory requirements 
will continue to be maintained. In addition, the proposed changes do 
not alter the conditions or assumptions in any of the previous 
accident analyses. Because the previous accident analyses remain 
bounding, the radiological consequences previously evaluated are not 
adversely affected by the proposed changes.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed changes do not involve any change to the configuration 
or method of operation of any plant equipment. Accordingly, no new 
failure modes have been defined for any plant system or component 
important to safety nor has any new limiting single failure been 
identified as a result of the proposed changes. Also, there will be 
no change in types or increase in the amounts of any effluents 
released offsite.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not involve a significant reduction in a 
margin of safety and are considered administrative in nature. The 
proposed changes do not involve any actual change in the methodology 
used in the monitoring of explosive gas mixtures contained in the 
Gaseous Waste Processing System. HNP does not currently utilize 
unprotected outdoor liquid storage tanks; therefore, there are no 
associated methodology changes with this request. These changes 
provide for the relocation of procedural details outside of the 
technical specifications with the addition of appropriate 
administrative controls to provide continued assurance of compliance 
to applicable regulatory requirements. 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: Lara Nichols, Deputy General Counsel, Duke 
Energy Corporation, 550 South Tryon St., M/C DEC45A, Charlotte, NC 
28202.
    NRC Acting Branch Chief: Jeanne A. Dion.

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

    Date of amendment request: July 12, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16194A515.
    Description of amendment request: The amendment would reduce the 
minimum reactor dome pressure associated with the critical power 
correlation from 785 pounds per square inch gauge (psig) to 685 psig in 
Technical Specification (TS) 2.1.1, ``Reactor Core SLs [Safety 
Limits],'' and associated bases.
    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 change does not involve a modification of any plant 
hardware; the probability and consequence of the Pressure Regulator 
Failure Open (PRFO) transient are essentially unchanged. The 
reduction in the reactor dome pressure safety limit (SL) from 785 
psig to 685 psig provides greater margin to accommodate the pressure 
reduction

[[Page 73434]]

during the transient within the revised TS limit.
    The proposed change will continue to support the validity range 
for the correlations and the calculation of Minimum Core Power Ratio 
(MCPR) as approved. The proposed TS revision involves no significant 
changes to the operation of any systems or components in normal, 
accident or transient operating conditions.
    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 previously evaluated?
    Response: No.
    The proposed reduction in the reactor dome pressure SL from 785 
psig to 685 psig is a change based upon previously approved 
documents and does not involve changes to the plant hardware or its 
operating characteristics. As a result, no new failure modes are 
being introduced.
    Therefore, the change does not introduce a new or different kind 
of accident from those previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through the design of the 
plant structures, systems, and components, and through the 
parameters for safe operation and setpoints for the actuation of 
equipment relied upon to respond to transients and design basis 
accidents. The proposed change in reactor dome pressure enhances the 
safety margin, which protects the fuel cladding integrity during a 
depressurization transient, but does not change the requirements 
governing operation or availability of safety equipment assumed to 
operate to preserve the margin of safety. The change does not alter 
the behavior of plant equipment, which remains unchanged. The 
available pressure range is expanded by the change, thus offering 
greater margin for pressure reduction during the transient.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: January 15, 2016, as supplemented by 
letters dated June 3, 2016, and September 19, 2016. Publicly available 
versions are available in ADAMS under Accession Nos. ML16015A456, 
ML16155A326, and ML16263A237, respectively.
    Description of amendment request: The licensee has provided a 
formal notification to the NRC of the intention to permanently cease 
power operations of JAF at the end of the current operating cycle. Once 
certifications for permanent cessation of operation and permanent 
removal of fuel from the reactor are submitted to the NRC, certain 
staffing and training Technical Specifications (TSs) administrative 
controls will no longer be applicable or appropriate for the 
permanently defueled condition. Therefore, ENO is requesting approval 
of changes to the staffing and training requirements in Section 5.0, 
``Administrative Controls,'' of the JAF TSs. Specifically, the 
amendment would revise and remove certain requirements in TS Sections 
5.1, ``Responsibility''; 5.2, ``Organization''; and 5.3, ``Plant Staff 
Qualifications,'' and add additional definitions to TS Section 1.1, 
``Definitions.'' The proposed amendment would not be effective until 
the certification of permanent cessation of operation and certification 
of permanent removal of fuel from the reactor vessel are submitted to 
the NRC.
    The license amendment request was originally noticed in the Federal 
Register on March 1, 2016 (81 FR 10678). The notice is being reissued 
in its entirety to include the revised scope and description of the 
amendment request.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, with NRC staff revisions provided in [brackets], which 
is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would not take effect until JAF has 
permanently ceased operation and entered a permanently defueled 
condition. The proposed amendment would modify the JAF TS by 
deleting the portions of the TS that are no longer applicable to a 
permanently defueled facility, while modifying the other sections to 
correspond to the permanently defueled condition.
    The deletion and modification of provisions of the 
administrative controls do not directly affect the design of 
structures, systems, and components (SSCs) necessary for safe 
storage of irradiated fuel or the methods used for handling and 
storage of such fuel in the fuel pool. The changes to the 
administrative controls are administrative in nature and do not 
affect any accidents applicable to the safe management of irradiated 
fuel or the permanently shutdown and defueled condition of the 
reactor.
    In a permanently defueled condition, the only credible accident 
is the fuel handling accident.
    The probability of occurrence of previously evaluated accidents 
is not increased, since extended operation in a 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 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 have no impact on facility SSCs affecting 
the safe storage of irradiated fuel, or on the methods of operation 
of such SSCs, or on the handling and storage of irradiated fuel 
itself. The administrative removal of or modifications of the TS 
that are related only to administration of facility cannot result in 
different or more adverse failure modes or accidents than previously 
evaluated because the reactor will be permanently shutdown and 
defueled and JAF will no longer be authorized to operate the 
reactor.
    The proposed deletion of requirements of the JAF TS do not 
affect systems credited in the accident analysis for the fuel 
handling accident at JAF. The proposed TS will continue to require 
proper control and monitoring of safety significant parameters and 
activities.
    The proposed amendment does not result in any new mechanisms 
that could initiate damage to the remaining relevant safety barriers 
for defueled plants (fuel cladding and spent fuel 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.
    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.
    Because the 10 CFR part 50 license for JAF will no longer 
authorize operation of the reactor or emplacement or retention of 
fuel into the reactor vessel once the certifications required by 10 
CFR 50.82(a)(1) are submitted, as specified in 10 CFR 50.82(a)(2), 
the occurrence of postulated accidents associated with reactor 
operation is no longer credible. The only remaining credible 
accident is a fuel handling accident (FHA). The proposed

[[Page 73435]]

amendment does 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 OL 
[operating license] and TS that are not related to the safe storage 
of irradiated fuel. The requirements that are proposed to be revised 
or deleted from the JAF OL and TS are not credited in the existing 
accident analysis for the remaining applicable as such, do not 
contribute to the margin of safety associated with the accident 
analysis. Postulated DBAs [design-basis accidents] involving the 
reactor are no longer possible because the reactor will be 
permanently shutdown and defueled and JAF will no longer be 
authorized to operate the reactor.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Travis L. Tate.

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

    Date of amendment request: July 26, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16210A227.
    Description of amendment request: The amendments would revise 
technical specification (TS) requirements relating to: (1) The 
inservice inspection (ISI) program required by the American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Code (Code), and (2) 
the inservice testing (IST) program required by the ASME Code for 
Operation and Maintenance of Nuclear Power Plants (OM Code). The 
proposed changes are based, in part, on Technical Specifications Task 
Force (TSTF) Traveler TSTF-545, Revision 3, ``TS Inservice Testing 
Program Removal & Clarify SR Usage Rule Application to Section 5.5 
Testing.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff edits in square 
brackets:

    1. 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 4.0.5, Surveillance Requirements 
for inservice inspection and testing of ASME Code Class 1, 2 & 3 
components, by revising the Inservice Testing Program and Inservice 
Inspection Program specification.
    Most requirements in the IST Program are removed, as they are 
duplicative of requirements in the ASME OM Code, as clarified by 
Code Case OMN-20, ``Inservice Test Frequency.'' The remaining 
requirements in the TS Section 4.0.5, IST Program are eliminated 
because the NRC has determined their inclusion in the TS is contrary 
to regulations. A new defined term, ``Inservice Testing Program,'' 
is added to the TS, which references the requirements of 10 CFR 
50.55a(f).
    Similarly, the requirements in the ISI Program are revised, as 
they are [ ] duplicative of requirements in Section XI of the ASME 
Boiler and Pressure Vessel Code and applicable Addenda.
    Performance of inservice testing or inservice inspection 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 two years may be extended by up to six months to 
facilitate test scheduling and consideration of plant operating 
conditions that may not be suitable for performance of the required 
testing. The testing frequency extension will not affect the ability 
of the components to mitigate any accident previously evaluated as 
the components are required to be operable during the testing period 
extension. Performance of inservice tests utilizing the allowances 
in OMN-20 will not significantly affect the reliability of the 
tested components. As a result, the availability of the affected 
components, as well as their ability to mitigate the consequences of 
accidents previously evaluated, is not affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed 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 or inservice inspection performed. In most cases, 
the frequency of inservice testing and inservice inspection is 
unchanged. However, the frequency of testing or inspection 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 provisions from the TS in 
lieu of provisions in the ASME Code, as modified by use of Code Case 
OMN-20 (IST) or ASME Boiler and Pressure Vessel Code (ISI). 
Compliance with the ASME Code is required by 10 CFR 50.55a. The 
proposed change also allows inservice tests with frequencies greater 
than two years to be extended by six months to facilitate test 
scheduling and consideration of plant operating conditions that may 
not be suitable for performance of the required testing. The testing 
frequency extension will not affect the ability of the components to 
respond to an accident as the components are required to be operable 
during the testing period extension. The proposed change will 
eliminate the existing TS SR 4.0.2 allowance to perform a specified 
surveillance time interval with a maximum allowable extension not to 
exceed 25% of the surveillance interval, unless there is a specific 
SR referencing usage of the INSERVICE TESTING PROGRAM and TS SR 
4.0.3 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. 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: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: Douglas A. Broaddus.

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

    Date of amendment request: July 15, 2016. A publicly-available 
version is in ADAMS under Package Accession No. ML16201A306.
    Description of amendment request: The amendment would revise the

[[Page 73436]]

Radiological Emergency Plan Annex for TMI-1. The proposed changes would 
decrease the radiation protection technician staffing from three to two 
technicians, remove two maintenance technicians currently assigned to 
the repair and corrective action function, and eliminate the on-shift 
Operations Support Center director position.
    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 to the TMI Emergency Plan do not increase 
the probability or consequences of an accident. The proposed changes 
do not impact the function of plant Structures, Systems, or 
Components (SSCs). The proposed changes do not affect accident 
initiators or accident precursors, nor do the changes alter design 
assumptions. The proposed changes do not alter or prevent the 
ability of the onsite ERO [emergency response organization] to 
perform their intended functions to mitigate the consequences of an 
accident or event. The proposed changes remove onsite ERO positions 
no longer credited or considered necessary in support of Emergency 
Plan implementation.
    Therefore, the proposed changes to the Emergency Plan do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    The proposed changes have no impact on the design, function, or 
operation of any plant SSCs. The proposed changes do not affect 
plant equipment or accident analyses. The proposed changes do not 
involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed), a change in the 
method of plant operation, or new operator actions. The proposed 
changes do not introduce failure modes that could result in a new 
accident, and the proposed changes do not alter assumptions made in 
the safety analysis. The proposed changes remove onsite ERO 
positions no longer credited or considered necessary in support of 
Emergency Plan implementation.
    Therefore, the proposed changes to the Emergency Plan 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.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public.
    The proposed changes do not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to 
operate in the safety analyses. There are no changes being made to 
safety analysis assumptions, safety limits, or limiting safety 
system settings that would adversely affect plant safety as a result 
of the proposed changes. Margins of safety are unaffected by the 
proposed changes to the ERO minimum on-shift staffing.
    The proposed changes are associated with the Emergency Plan 
staffing and do not impact operation of the plant or its response to 
transients or accidents. The proposed changes do not affect the 
Technical Specifications. The proposed changes do not involve a 
change in the method of plant operation, and no accident analyses 
will be affected by the proposed changes. Safety analysis acceptance 
criteria are not affected by these proposed changes. The proposed 
changes to the Emergency Plan will continue to provide the necessary 
onsite ERO response staff.
    Therefore, the proposed changes to the Emergency Plan do not 
involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: August 31, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16244A493.
    Brief description of amendment request: The amendments would revise 
the Required Actions and associated Completion Times to Technical 
Specification (TS) 3.8.7, ``Inverters--Operating.'' Specifically, 
Condition B would be deleted and current Condition C would be re-
lettered to Condition B. Additionally, the Required Actions and 
associated Completion Times for Condition A would be modified to 
require restoration of one inoperable inverter to operability within 24 
hours. These changes conform to Improved Standard Technical 
Specification TS 3.8.7.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the TS for the purpose of 
eliminating a non-conservative Required Action. The proposed TS 
change does not introduce new equipment or new equipment operating 
modes, nor does the proposed change alter existing system 
relationships. The proposed change does not affect normal plant 
operation. Further, the proposed change does not increase the 
likelihood of the malfunction of any SSC [structure, system and 
component] or impact any analyzed accident. Consequently, the 
probability of an accident previously evaluated is not affected and 
there is no significant increase in the consequences of any accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the TS for the purpose of 
eliminating a non-conservative Required Action. 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 operations. The proposed change does 
not alter assumptions made in the safety analysis. Further, the 
proposed change does not introduce new accident initiators.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction a 
margin of safety?
    Response: No.
    The proposed change revises the TS for the purpose of 
eliminating a non-conservative Required Action. The proposed change 
does not alter the manner in which safety limits, limiting safety 
system settings, or limiting conditions for operation are 
determined. The safety analysis assumptions and acceptance criteria 
are not affected by this change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

[[Page 73437]]

Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: David J. Wrona.

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

    Date of amendment request: July 20, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16203A006.
    Description of amendment request: The amendment would revise the 
Hope Creek Generating Station (Hope Creek) Technical Specifications 
(TS), Section 6.8.4.i, ``Inservice Testing Program,'' to remove 
requirements duplicated in the American Society of Mechanical Engineers 
(ASME) Code for Operations and Maintenance of Nuclear Power Plants Case 
OMN-20, ``Inservice Test Frequency.'' A new defined term, ``Inservice 
Testing Program,'' will be added to the TS 1.0, ``Definitions,'' 
section. The licensee stated that the proposed change to the TS is 
consistent with Technical Specifications Task Force (TSTF) Traveler 
TSTF-545, Revision 3, ``TS Inservice Testing Program Removal & Clarity 
SR Usage Rule Application to Section 5.5 Testing'' (ADAMS Accession No. 
ML15294A555), with no proposed variations or deviations. However, the 
Hope Creek TS uses different numbering for surveillance requirements 
than the Standard Technical Specifications on which TSTF-545 was based, 
so the licensee changed the TSTF-545 numbering to be consistent with 
the Hope Creek TS numbering.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff edits in square 
brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS Chapter 6, ``Administrative 
Controls,'' Section 6.8, ``Procedures and Programs,'' by eliminating 
the ``Inservice Testing Program'' specification. Most requirements 
in the Inservice Testing Program are removed, as they are 
duplicative of requirements in the ASME OM Code, as clarified by 
Code Case OMN-20, ``Inservice Test Frequency.'' The remaining 
requirements in the Section 6.8 IST Program are eliminated [. . .]. 
A new defined term, ``Inservice Testing Program,'' is added to the 
TS, which references the requirements of 10 CFR 50.55a(f).
    Performance of inservice testing is not an initiator to any 
accident previously evaluated. As a result, the probability of 
occurrence of an accident is not significantly affected by the 
proposed change. Inservice test frequencies under Code Case OMN-20 
are equivalent to the current testing period allowed by the TS with 
the exception that testing frequencies greater than 2 years may be 
extended by up to 6 months to facilitate test scheduling and 
consideration of plant operating conditions that may not be suitable 
for performance of the required testing. The testing frequency 
extension will not affect the ability of the components to mitigate 
any accident previously evaluated as the components are required to 
be operable during the testing period extension. Performance of 
inservice tests utilizing the allowances in OMN-20 will not 
significantly affect the reliability of the tested components. As a 
result, the availability of the affected components, as well as 
their ability to mitigate the consequences of accidents previously 
evaluated, is not affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not alter the design or configuration 
of the plant. The proposed change does not involve a physical 
alteration of the plant; no new or different kind of equipment will 
be installed. The proposed change does not alter the types of 
inservice testing performed. In most cases, the frequency of 
inservice testing is unchanged. However, the frequency of testing 
would not result in a new or different kind of accident from any 
previously evaluated since the testing methods are not altered.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change eliminates some requirements from the TS in 
lieu of requirements in the ASME Code, as modified by use of Code 
Case OMN-20. Compliance with the ASME Code is required by 10 CFR 
50.55a. The proposed change also allows inservice tests with 
frequencies greater than 2 years to be extended by 6 months to 
facilitate test scheduling and consideration of plant operating 
conditions that may not be suitable for performance of the required 
testing. The testing frequency extension will not affect the ability 
of the components to respond to an accident as the components are 
required to be operable during the testing period extension. The 
proposed change will eliminate the existing TS 4.0.3 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 TS 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. [. 
. .] 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: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Douglas A. Broaddus.

South Carolina Electric & Gas Company and South Carolina Public Service 
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear 
Station, Units 2 and 3, Fairfield, South Carolina

    Date of amendment request: September 15, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16259A310.
    Description of amendment request: The amendments would revise 
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer 
Nuclear Station, Units 2 and 3. The amendments propose changes to the 
Updated Final Safety Analysis Report (UFSAR) in the form of departures 
from the incorporated plant-specific Design Control Document Tier 2* 
information. Specifically, the proposed changes would revise the 
Combined Licenses to clarify information in WCAP-17179, 
``AP1000[supreg] Component Interface Module Technical Report,'' which 
demonstrates design compliance with licensing bases requirements. WCAP-
17179 is incorporated by reference into the UFSAR to provide additional 
details regarding the component interface module (CIM) system design. 
The requested amendments also propose a change to the CIM internal 
power supply that will enable proper functioning of the field 
programmable gate arrays.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

[[Page 73438]]

    Response: No.
    The proposed change to the CIM internal power supply enables the 
field programmable gate array (FPGA) to function properly. The 
proposed change to the FPGA core power has no adverse effect on the 
operation of the output actuation relays. The function of the 
internal power supply has no input to plant safety analysis. The 
change to the CIM internal power supply has a negligible effect on 
the 24 Vdc [volts direct current] supplies and ultimately the plant 
electrical system load and has no adverse effect on the CIM 
functionality.
    The proposed changes to clarify how licensing basis design 
documentation reflects compliance with license basis requirements, 
and the proposed change to the ownership of safety remote node 
controller (SRNC) and CIM intellectual property, are not technical 
changes. The proposed changes do not affect any accident initiator 
in the UFSAR, or affect the radioactive material releases in the 
UFSAR accident analyses. The proposed change does not alter the 
ability of the facility to prevent and mitigate abnormal events, 
e.g., accidents, anticipated operational occurrences, earthquakes, 
floods and turbine missiles, or their safety or design analyses. No 
safety-related structure, system, or component (SSC) or function is 
adversely affected. The change does not involve or interface with 
any SSC accident initiator or initiating sequence of events, and 
thus, the probabilities of the accidents evaluated in the UFSAR are 
not affected. This activity does not involve a new fission product 
release path, nor a new fission product barrier failure mode, nor 
create a new sequence of events that would result in significant 
fuel cladding failures. Because the proposed changes do not change 
any safety-related SSC or function credited in the mitigation of an 
accident, the consequences of the accidents evaluated in the UFSAR 
are not affected.
    Therefore, the requested amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change to the CIM internal power supply enables the 
FPGA to function properly and does not involve accident initiators. 
The change to the CIM internal power supply has a negligible effect 
on the 24 Vdc supplies and ultimately the plant electrical system 
load and has no adverse effect on CIM functionality.
    The proposed clarified descriptions and the proposed change to 
the ownership of SRNC and CIM intellectual property are not 
technical changes. The proposed changes do not affect other plant 
equipment or adversely affect the design of the CIM. Therefore, the 
proposed changes do not affect any safety-related equipment itself, 
nor do they affect equipment whose failure could initiate an 
accident or a failure of a fission product barrier. No analysis is 
adversely affected by the proposed changes. No system or design 
function or equipment qualification would be adversely affected by 
the proposed changes. Furthermore, the proposed changes do not 
result in a new failure mode, malfunction or sequence of events that 
could affect safety or safety-related equipment.
    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 the CIM internal power supply enables the 
FPGA to function properly. The function of the internal power supply 
has no input to plant safety analysis. The change to the CIM 
internal power supplies has a negligible effect on the 24 Vdc 
supplies and ultimately the plant electrical system load and has no 
adverse effect on the CIM functionality.
    The proposed clarified descriptions and the proposed change to 
the ownership of SRNC and CIM intellectual property are not 
technical changes. The proposed changes do not adversely affect the 
design, construction, or operation of any plant SSCs, including any 
equipment whose failure could initiate an accident or a failure of a 
fission product barrier. No analysis is adversely affected by the 
proposed changes. Furthermore, no system function, design function, 
or equipment qualification will be adversely affected by the 
changes. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the proposed changes, thus no 
margin of safety is reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius, LLC, 1111 Pennsylvania Ave. NW., Washington, DC 20004-2514.
    NRC Branch Chief: Michael T. Markley.

South Carolina Electric & Gas Company, Inc., Docket Nos. 52-027 and 52-
028, Virgil C. Summer Nuclear Station Units 2 and 3, Fairfield, South 
Carolina

    Date of amendment request: September 28, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16272A373.
    Description of amendment request: The amendment request proposes 
changes to revise plant-specific Tier 1, plant-specific Tier 2, and 
Combined License (COL) Appendix C information concerning the details of 
the Class 1E direct current and uninterruptible power supply system 
(IDS), specifically adding seven Class 1E fuse panels to the IDS 
design. These proposed changes provide electrical isolation between the 
non-Class 1E IDS battery monitors and their respective Class 1E battery 
banks. Because, this proposed change requires a departure from Tier 1 
information in the Westinghouse Electric Company's AP1000 Design 
Control Document (DCD), the licensee also requested 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 with NRC staff edits in square 
brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to revise plant-specific Tier 1, COL 
Appendix C, and [Updated Final Safety Analysis Report (UFSAR)] 
information concerning details of the IDS, specifically the addition 
of seven Class 1E fuse isolation panels at the interconnection of 
the non-Class 1E IDS battery monitors and Class 1E IDS circuits, are 
necessary to conform to Regulatory Guide 1.75 Rev. 2 (consistent 
with UFSAR Appendix 1A exceptions) and IEEE 384-1981 to prevent a 
fault on non-Class 1E circuits or equipment from degrading the 
operation of Class 1E IDS circuits and equipment below an acceptable 
level. The proposed changes do not adversely affect the design 
functions of the IDS, including the Class 1E battery banks and the 
battery monitors.
    These proposed changes to revise plant-specific Tier 1, COL 
Appendix C, and UFSAR information concerning details of the IDS, 
specifically the addition of seven Class 1E fuse isolation panels at 
the interconnection of the non-Class 1E IDS battery monitors and 
Class 1E IDS circuits as described in the current licensing basis do 
not have an adverse effect on any of the design functions of any 
plant systems. The proposed changes do not adversely affect any 
plant electrical system and do not affect the support, design, or 
operation of mechanical and fluid systems required to mitigate the 
consequences of an accident. There is no change to plant systems or 
the response of systems to postulated accident conditions. There is 
no change to the predicted radioactive releases due to postulated 
accident conditions. The plant response to previously evaluated 
accidents or external events is not adversely affected, nor do the 
proposed changes create any new accident precursors.
    Therefore, the requested 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.

[[Page 73439]]

    The proposed changes to revise plant-specific Tier 1, COL 
Appendix C, and UFSAR information concerning details of the IDS, 
specifically the addition of seven Class 1E fuse isolation panels at 
the interconnection of the non-Class 1E IDS battery monitors and 
Class 1E IDS circuits, are necessary to conform to Regulatory Guide 
1.75 Rev. 2 (consistent with UFSAR Appendix 1A exceptions) and IEEE 
384-1981 to prevent a fault on non-Class 1E circuits or equipment 
from degrading the operation of Class 1E IDS circuits and equipment 
below an acceptable level. The proposed changes do not adversely 
affect any plant electrical system and do not adversely affect the 
design function, support, design, or operation of mechanical and 
fluid systems. The proposed changes do not result in a new failure 
mechanism or introduce any new accident precursors. No design 
function described in the UFSAR is adversely affected by the 
proposed changes.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    There is no safety-related [structure, system, and component 
(SSC)] or function adversely affected by the proposed change to add 
IDS fuse isolation panels to non-Class 1E IDS battery monitors and 
Class 1E IDS circuits. No safety analysis or design basis acceptance 
limit/criterion is challenged or exceeded by the proposed changes 
and no margin or safety is reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius, LLC, 1111 Pennsylvania Ave. NW., Washington, DC 20004-2514.
    NRC 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: December 22, 2015, as supplemented by 
letter dated July 27, 2016. Publicly-available versions are in ADAMS 
under Accession Nos. ML15356A655 and ML16209A477, respectively.
    Description of amendment request: The proposed changes would revise 
the Combined License (COL) Appendix C and corresponding plant-specific 
Tier 1 information to add two turbine building sump pumps to 
accommodate the increased flow that will be experienced during 
condensate polishing system rinsing operations, for each unit, 
respectively. The proposed changes include information in the combined 
license, Appendix C. An exemption request relating to the proposed 
changes to the AP1000 DCD Tier 1 is included with the request.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff edits in square 
brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to identify that there is more than one 
turbine building sump and to add two turbine building sump pumps 
(WWS-MP-07A and WWS-MP-07B) to COL Appendix C Subsection 2.3.29 and 
corresponding Table 2.3.29-1 will provide consistency within the 
current licensing basis. The main turbine building sumps and sump 
pumps are not safety-related components and do not interface with 
any systems, structures, or components (SSCs) accident initiator or 
initiating sequence of events; thus, the probability of accidents 
evaluated within the [Updated Final Safety Analysis Report (UFSAR)] 
are not affected. The proposed changes do not involve a change to 
the predicted radiological releases due to accident conditions, thus 
the consequences of 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 to identify that there is more than one 
turbine building sump and to add two turbine building sump pumps to 
the non-safety waste water system (WWS) do not affect any safety-
related equipment, nor do they add any new interface to safety-
related SSCs. No system or design function or equipment 
qualification is affected by these changes. The changes do not 
introduce a new failure mode, malfunction, or sequence of events 
that could affect safety or safety-related equipment.
    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 WWS is a non-safety-related system that does not interface 
with any safety-related equipment. The proposed changes to identify 
that there is more than one turbine building sump and to add two 
turbine building sump pumps do not affect any design code, function, 
design analysis, safety analysis input or result, or design/safety 
margin. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the proposed change.
    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 Branch Chief: Jennifer Dixon-Herrity.

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

    Date of amendment request: August 23, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16236A266.
    Description of amendment request: The proposed changes would amend 
Combined License Nos. NPF-91 and NPF-92 for the Vogtle Electric 
Generating Plant, Units 3 and 4. The amendments propose changes to the 
Updated Final Safety Analysis Report (UFSAR) in the form of departures 
from the incorporated plant-specific Design Control Document Tier 2 
information and involve related changes to the Combined Operating 
License Appendix C (and corresponding plant-specific design control 
document Tier 1) information. Specifically, the proposed departures 
consist of changes to the design reliability assurance program (D-RAP) 
to identify the covers for the in-containment refueling water storage 
tank vents and overflow weirs as the risk-significant components 
included in the D-RAP and to differentiate between the rod drive motor-
generator (MG) sets field control relays and the rod drive power supply 
control cabinets in which the relays are located.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

[[Page 73440]]

    Response: No.
    The in-containment refueling water storage tank (IRWST) provides 
flooding of the refueling cavity for normal refueling. The tank also 
serves as a heat sink during Passive Residual Heat Removal (PRHR) 
Heat Exchanger (HX) operation and in the event of a loss-of-coolant-
accident (LOCA) provides injection in support of long-term RCS 
[reactor coolant system] cooling. This activity adds normally closed 
covers to the IRWST vents and overflow weirs to prevent debris from 
entering the tank, prevent over-pressurization and accommodate 
volume and mass increases in the tank. The vent and overflow weir 
covers open upon differential pressures between the IRWST and 
containment.
    The rod drive MG sets provide the power to the control rod drive 
mechanisms through the reactor trip switchgear. This activity 
revises the equipment description and equipment tag associated with 
the risk-significant control relays which open to de-energize the 
rod drive MG sets and permit rods to drop.
    The proposed changes to add the IRWST vent and overflow weir 
covers and to change the description of the equipment and equipment 
tag related to the rod drive MG sets does not inhibit the SSCs from 
performing their safety-related function. The design bases of the 
IRWST vents and overflow weirs are not modified as a result of the 
addition of the covers to the vents and overflow weirs and the 
change to the control cabinet relay description and equipment tag. 
This proposed amendment does not have an adverse impact on the 
response to anticipated transients or postulated accident conditions 
because the functions of the SSCs are not changed. Required IRWST 
venting is not affected for any accident conditions. Required DAS 
functions are not affected for any accident conditions. Safety-
related structure, system, component (SSC) or function is not 
adversely affected by this change. The changes to include the IRWST 
covers and to change the control cabinet relay description and tag 
number do not involve an interface with any SSC accident initiator 
or initiating sequence of events, and thus, the probabilities of the 
accidents evaluated in the UFSAR are not affected. The proposed 
changes do not involve a change to the predicted radiological 
releases due to postulated accident conditions, thus, the 
consequences of the accidents evaluated in the UFSAR are not 
affected. Probabilistic Risk Assessment (PRA) modeling and analyses 
associated with the SSCs are not impacted by this change.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the design of the IRWST vent and 
overflow weir covers do not adversely affect any safety-related 
equipment, and do not add any new interfaces to safety-related SSCs. 
No system or design function or equipment qualification is affected 
by these changes. The changes do not introduce a new failure mode, 
malfunction or sequence of events that could affect plant safety or 
safety-related equipment as the simplistic design of the cover 
louvers and hinged flappers are not considered unique designs. No 
new credible failure modes are introduced by the addition of the 
covers.
    The proposed changes to the description and equipment tag 
associated with the risk-significant control relays for the rod 
drive MG sets do not adversely affect any safety-related equipment, 
and do not add any new interfaces to safety-related SSCs. No system 
or design function or equipment qualification is affected by these 
changes. The changes do not introduce a new failure mode, 
malfunction or sequence of events that could affect plant safety or 
safety-related equipment because the design function of the control 
relays, control cabinets, or rod drive MG sets is not changed.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes maintain compliance with the applicable 
Codes and Standards, thereby maintaining the margin of safety 
associated with these SSCs. The proposed changes do not alter any 
applicable design codes, code compliance, design function, or safety 
analysis. Consequently, no safety analysis or design basis 
acceptance limit/criterion is challenged or exceeded by the proposed 
change, thus the margin of safety is not reduced. Because no safety 
analysis or design basis acceptance limit/criterion is challenged or 
exceeded by these changes, no margin of safety is reduced.
    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: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: August 31, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16244A836.
    Description of amendment request: The amendments propose changes to 
the Updated Final Safety Analysis Report (UFSAR) in the form of 
departures from the incorporated plant-specific Design Control Document 
Tier 2* information. Specifically, the proposed changes would revise 
the Combined Licenses for the Vogtle Electric Generating Plant, Units 3 
and 4, to clarify information in WCAP-17179, ``AP1000[supreg] Component 
Interface Module Technical Report,'' which demonstrates design 
compliance with licensing bases requirements. WCAP-17179 is 
incorporated by reference into the UFSAR to provide additional details 
regarding the component interface module (CIM) system design. The 
requested amendments also propose a change to the CIM internal power 
supply that will enable proper functioning of the field programmable 
gate arrays.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the CIM internal power supply enables the 
field programmable gate array (FPGA) to function properly. The 
proposed change to the FPGA core power has no adverse effect on the 
operation of the output actuation relays. The function of the 
internal power supply has no input to plant safety analysis. The 
change to the CIM internal power supply has a negligible effect on 
the 24 Vdc [volts direct current] supplies and ultimately the plant 
electrical system load and has no adverse effect on the CIM 
functionality.
    The proposed changes to clarify how licensing basis design 
documentation reflects compliance with license basis requirements, 
and the proposed change to the ownership of safety remote node 
controller (SRNC) and CIM intellectual property, are not technical 
changes. The proposed changes do not affect any accident initiator 
in the UFSAR, or affect the radioactive material releases in the 
UFSAR accident analyses. The proposed change does not alter the 
ability of the facility to prevent and mitigate abnormal events, 
e.g., accidents, anticipated operational occurrences, earthquakes, 
floods and turbine missiles, or their safety or design analyses. No 
safety-related structure, system, or component (SSC) or function is 
adversely affected. The change does not involve or interface with 
any SSC accident initiator or initiating sequence of events, and 
thus, the probabilities of the accidents evaluated in the UFSAR are 
not affected. This activity does not involve a new fission product 
release path, nor a new fission product barrier failure mode, nor 
create a new sequence of events that would result in significant 
fuel cladding failures. Because the proposed changes do

[[Page 73441]]

not change any safety-related SSC or function credited in the 
mitigation of an accident, the consequences of the accidents 
evaluated in the UFSAR are not affected.
    Therefore, the requested amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change to the CIM internal power supply enables the 
FPGA to function properly and does not involve accident initiators. 
The change to the CIM internal power supply has a negligible effect 
on the 24 Vdc supplies and ultimately the plant electrical system 
load and has no adverse effect on CIM functionality.
    The proposed clarified descriptions and the proposed change to 
the ownership of SRNC and CIM intellectual property are not 
technical changes. The proposed changes do not affect other plant 
equipment or adversely affect the design of the CIM. Therefore, the 
proposed changes do not affect any safety-related equipment itself, 
nor do they affect equipment whose failure could initiate an 
accident or a failure of a fission product barrier. No analysis is 
adversely affected by the proposed changes. No system or design 
function or equipment qualification would be adversely affected by 
the proposed changes. Furthermore, the proposed changes do not 
result in a new failure mode, malfunction or sequence of events that 
could affect safety or safety-related equipment.
    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 the CIM internal power supply enables the 
FPGA to function properly. The function of the internal power supply 
has no input to plant safety analysis. The change to the CIM 
internal power supplies has a negligible effect on the 24 Vdc 
supplies and ultimately the plant electrical system load and has no 
adverse effect on the CIM functionality.
    The proposed clarified descriptions and the proposed change to 
the ownership of SRNC and CIM intellectual property are not 
technical changes. The proposed changes do not adversely affect the 
design, construction, or operation of any plant SSCs, including any 
equipment whose failure could initiate an accident or a failure of a 
fission product barrier. No analysis is adversely affected by the 
proposed changes. Furthermore, no system function, design function, 
or equipment qualification will be adversely affected by the 
changes. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the proposed changes, thus no 
margin of safety is reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Docket No. 50-364, Joseph M. Farley 
Nuclear Plant, Unit 2, Houston County, Alabama

    Date of amendment request: September 8, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16256A135.
    Description of amendment request: The amendment would correct an 
error in the Joseph M. Farley Nuclear Plant, Unit 2, Renewed Facility 
Operating License No. NPF-8, for Condition 2.C.(23). Specifically, the 
Unit 2 referenced date prior to the period of extended operation was 
incorrectly entered as June 25, 2017. This date corresponds to the Unit 
1 period of extended operation. The Unit 2 correct date for this 
license condition is March 31, 2021.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment contains no technical changes; all 
proposed changes are administrative. These changes are consistent 
with the intent of what has already been approved by the Nuclear 
Regulatory Commission (NRC). There are no accidents affected by this 
change, and therefore no increase in the probability or consequences 
of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment contains no technical changes; all 
proposed changes are administrative. These changes are consistent 
with the intent of what has already been approved by the Nuclear 
Regulatory Commission (NRC). There are no accidents affected by this 
change, and therefore no possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment contains no technical changes; all 
proposed changes are administrative. These changes are consistent 
with the intent of what has already been approved by the Nuclear 
Regulatory Commission (NRC). There are no accidents affected by this 
change, and therefore no reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, Inc., 40 Inverness Center 
Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: July 27, 2016, as supplemented by letter 
dated September 13, 2016. Publicly-available versions are in ADAMS 
under Accession Nos. ML16210A001 and ML16257A598, respectively.
    Description of amendment request: The amendments would revise 
Technical Specification 3.6.4.1, ``Secondary Containment,'' 
Surveillance Requirement (SR) 3.6.4.1.3 to provide an allowance for 
brief, inadvertent, simultaneous opening of redundant secondary 
containment access doors during normal entry and exit conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, along with NRC edits in square 
brackets:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of any accident previously evaluated?
    Response: No.
    The proposed change does not involve any physical change to 
structures, systems, or components (SSCs) and do not alter the 
method of operation of any SSCs. The proposed change addresses a 
temporary condition during which Secondary Containment SRs are not 
met. The Secondary Containment is not an initiator of any accident 
previously evaluated. As a result, the probability of any accident 
previously evaluated is not increased. [Two accidents credit the 
Secondary Containment from a dose consequence perspective. They are 
the

[[Page 73442]]

loss-of-coolant accident (LOCA) and fuel/equipment handling 
accident. Each accident requires time to drawdown the secondary 
containment to less than atmospheric pressure. The brief, 
inadvertent, simultaneous opening of both an inner and outer 
personnel access door during normal entry and exit conditions 
followed by prompt closure does not challenge the design basis 
drawdown time and does not result in an increase in any on-site or 
offsite dose for the LOCA dose analysis. All dose consequences are 
within the regulatory limits established for the fuel handling 
accident and bound the case in which airlock doors are briefly, 
inadvertently opened.] As a result, the consequences of any accident 
previously evaluated is not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
any plant equipment. No new equipment is being introduced, and 
installed equipment is not being operated in a new or different 
manner. There are not setpoints, at which protective or mitigative 
actions are initiated, affected by the proposed change. The proposed 
change does not alter the manner in which equipment operation is 
initiated, nor will the function of credited equipment be changed. 
No alterations in the procedures that ensure the plant remains 
within analyzed limits are being proposed, and no changes are being 
made to the procedures relied upon to respond to an off-normal event 
described in the FSAR [Final Safety Analysis Report]. As such, no 
new failure modes are being introduced. The change does not alter 
the assumptions made in the safety analysis and licensing basis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed change addresses temporary conditions 
during which the Secondary Containment SR is not met. The allowance 
for both an inner and outer Secondary Containment access door to be 
open simultaneously for entry and exit does not affect the safety 
function of the reactor enclosure and refuel area Secondary 
Containments as the doors are promptly closed after entry of exit, 
thereby restoring the Secondary Containment boundary. In addition, 
brief, inadvertent simultaneous opening and closing of redundant 
Secondary Containment personnel access doors during normal entry and 
exit conditions does not affect the ability of the SGTS to establish 
the required Secondary Containment vacuum. Therefore, the safety 
function of the Secondary Containment is not affected.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Damon D. Obie, Esquire, Associate General 
Counsel, Talen Energy Supply, LLC, 835 Hamilton St., Suite 150, 
Allentown, PA 18101.
    NRC Branch Chief: Douglas A. Broaddus.

Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant 
(WBN), Unit 2, Rhea County, Tennessee

    Date of amendment request: September 30, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16277A477.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) to allow a one-time extension of the 
frequency for performing TS Surveillance Requirements (SRs) related to 
verifying the operability of the containment ice bed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with NRC edits in brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The requested action is a one-time extension to the performance 
interval for TS SRs [3.6.11.2] and 3.6.11.3. The performance of 
these surveillances, or the extension of these surveillances, is not 
a precursor to an accident. Performing these surveillances or 
failing to perform these surveillances does not affect the 
probability of an accident.
    Therefore, the proposed delay in performance of the SRs in this 
amendment request does not increase the probability of an accident 
previously evaluated.
    A delay in performing these surveillances does not result in a 
system being unable to perform its required function. In the case of 
this one-time extension request, the short period of additional time 
that the systems and components will be in service before the next 
performance of the surveillance will not affect the ability of those 
systems to operate as designed. Therefore, the systems required to 
mitigate accidents will remain capable of performing their required 
function. No new failure modes have been introduced because of this 
action and the consequences remain consistent with previously 
evaluated accidents. On this basis, the proposed delay in 
performance of the SRs in this amendment request does not involve a 
significant increase in the consequences of an 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not involve a physical alteration of 
any system, structure, or component (SSC) or a change in the way any 
SSC is operated. The proposed amendment does not involve operation 
of any SSCs in a manner or configuration different from those 
previously recognized or evaluated. No new failure mechanisms will 
be introduced by the one-time SR extensions being requested.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment is a one-time extension of the 
performance interval of TS SRs [3.6.11.2] and 3.6.11.3. Extending 
these surveillance requirements does not involve a modification of 
any TS limiting conditions for operation. Extending these SRs does 
not involve a change to any limit on accident consequences specified 
in the license or regulations. Extending these SRs does not involve 
a change in how accidents are mitigated or a significant increase in 
the consequences of an accident. Extending these SRs does not 
involve a change in a methodology used to evaluate consequences of 
an accident. Extending these SRs does not involve a change in any 
operating procedure or process.
    Based on the limited additional period of time that the systems 
and components will be in service before the surveillances are next 
performed, as well as the operating experience that these 
surveillances are typically successful when performed, it is 
reasonable to conclude that the margins of safety associated with 
these SRs will not be affected by the requested extension.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Sherry A. Quirk, Executive Vice 
President and General Counsel, Tennessee Valley Authority, 400 West 
Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Acting Branch Chief: Jeanne A. Dion.

[[Page 73443]]

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: May 18, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16146A540.
    Description of amendment request: The amendments would revise the 
Surry Power Station, Unit Nos. 1 and 2, Technical Specification (TS) 
3.14, ``Circulating and Service Water Systems,'' to extend the allowed 
outage time (AOT) for only one operable service water (SW) flow path to 
the charging pump service water (CPSW) subsystem and to the main 
control room/emergency switchgear room (MCR/ESGR) air conditioning (AC) 
subsystem. TS 3.14.A.5 and TS 3.14.A.7 require two SW flow paths to the 
CPSW subsystem and to the MCR/ESGR AC subsystem, respectively, to be 
operable. Currently, the TS 3.14.C AOT for only one operable CPSW or 
MCR/ESGR AC flow path is 24 hours. The proposed revision will extend 
the AOT for only one operable CPSW or MCR/ESGR AC flow path from 24 
hours to 72 hours.
    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 extends the AOT for only one operable CPSW 
or MCR/ESGR AC flow path from 24 hours to 72 hours. The CPSW 
subsystem is a support system for the Charging/High Head Safety 
Injection (HHSI) pumps; the proposed CPSW AOT extension aligns the 
CPSW support system AOT with the AOT for the supported components 
(i.e., the Charging/HHSI pumps). The proposed MCR/ESGR AC AOT 
extension revises the AOT to be the same as the CPSW AOT since both 
subsystems share common piping. The design function of the CPSW 
system, which is to provide cooling to the charging pump 
intermediate seal coolers and the charging pump lubricating oil 
coolers, is not impacted by the proposed revision, nor is the design 
function of the Charging/HHSI pumps impacted. Furthermore, the 
design functions of the MCR/ESGR AC subsystem and the MCR/ESGR 
chillers are not impacted by the proposed revision. In addition, the 
proposed change deletes the now expired and no longer necessary 
requirements for the temporary SW jumper to the CCHXs [component 
cooling heat exchangers]. The deletion of these temporary 
requirements is administrative in nature. As a result, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change extends the AOT for only one operable CPSW 
or MCR/ESGRAC flow path from 24 hours to 72 hours. In addition, the 
proposed change deletes the now expired and no longer necessary 
requirements for the temporary SW jumper to the CCHXs. The proposed 
change does not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) and does not 
impact plant operation. Furthermore, the proposed change does not 
impose any new or different requirements that could initiate an 
accident. The proposed change does not alter assumptions made in the 
safety analysis and is consistent with the safety analysis 
assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change extends the AOT for only one operable CPSW 
or MCR/ESGR AC flow path from 24 hours to 72 hours. The proposed 
change does not adversely affect any current plant safety margins or 
the reliability of the equipment assumed in the safety analysis. 
There are no changes being made to any safety analysis assumptions, 
safety limits, or limiting safety system settings that would 
adversely affect plant safety as a result of the proposed change. 
Furthermore, as noted above, a supporting PRA [probabilistic risk 
assessment] was performed for the proposed AOT changes. The PRA 
concluded that the increase in risk associated with the proposed 
changes is consistent with the RG [Regulatory Guide] 1.174 and RG 
1.177 acceptance guidelines for a permanent TS AOT change. This PRA 
evaluation demonstrates that defense-in-depth will not be 
significantly impacted by changing the AOTs for only one operable SW 
flow path to the CPSW subsystem and to the MCR/ESGR AC subsystem 
from 24 to 72 hours. In addition, the proposed change deletes the 
now expired and no longer necessary requirements for the temporary 
SW jumper to the CCHXs. The deletion of these temporary requirements 
is administrative in nature. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
    NRC Branch Chief: Michael T. Markley.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: July 14, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16202A068.
    Description of amendment request: The amendments would revise the 
Surry Power Station, Unit Nos. 1 and 2, Technical Specification (TS) 
3.14, ``Circulating and Service Water Systems,'' to extend the allowed 
outage time (AOT) for emergency service water (ESW) pump inoperability. 
The proposed revision would extend the TS 3.14.B AOT for one inoperable 
ESW pump from 7 days to 14 days to provide operational flexibility for 
ESW pump maintenance and repairs.
    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 design function of the ESW pumps is to ensure that water can 
be provided to the intake canal (i.e., the ultimate heat sink) when 
power is not available to the Circulating Water (CW) pumps. The 
proposed extension of the AOT for one inoperable ESW pump from 7 to 
14 days does not impact the design function of the ESW pumps. In 
addition, the number of ESW pumps required to be operable for the 
specified plant operating conditions is not impacted by the proposed 
AOT extension. As a result, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant (i.e., no new or different type of equipment will be 
installed) and does not impact plant operation. Furthermore, the 
proposed change does not impose any new or different requirements 
that could initiate an accident. The proposed change does not alter 
assumptions made in the safety analysis and is consistent with the 
safety analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different

[[Page 73444]]

kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not adversely affect any current plant 
safety margins or the reliability of the equipment assumed in the 
safety analysis. There are no changes being made to any safety 
analysis assumptions, safety limits, or limiting safety system 
settings that would adversely affect plant safety as a result of the 
proposed change. Furthermore, as noted above, a supporting PRA 
[probabilistic risk assessment] was performed for the proposed AOT 
change. The PRA concluded that the increase in risk associated with 
the proposed change is consistent with the RG [Regulatory Guide] 
1.174 and RG 1.177 acceptance guidelines for a permanent TS AOT 
change. This PRA evaluation demonstrates that defense-in-depth will 
not be significantly impacted by changing the AOT for one inoperable 
ESW pump from 7 to 14 days.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
    NRC Branch Chief: Michael T. Markley.

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

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

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

    Date of amendment request: September 1, 2015, as supplemented by 
letter dated March 24, 2016.
    Brief description of amendment: The amendment revised the MPS2 
Technical Specifications (TSs) to add the evaluation model EMF-
2328(P)(A), Supplement 1, ``PWR [pressurized water reactor] Small Break 
LOCA [loss-of coolant accident] Evaluation Model S-RELAP5 Based,'' and 
EMF-92-116(P)(A), Supplement 1, ``Generic Mechanical Design Criteria 
for PWR Fuel Designs,'' to the TS Section 6.9.1.8.b list of analytical 
methods used to determine core operating limits as a result of 
reanalyzing the small break LOCA.
    Date of issuance: September 30, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 329. A publicly-available version is in ADAMS under 
Accession No. ML16249A001; 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 Operating License and TSs.
    Date of initial notice in Federal Register: December 8, 2015 (80 FR 
76318). The supplemental letter dated March 24, 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 September 30, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 29, 2014, as supplemented by 
letters dated May 19, June 16, July 21, August 12, September 22, 
November 4, and November 17, 2015; and January 15, March 25, April 7, 
May 19, and August 29, 2016.
    Brief description of amendment: The amendment authorized the 
transition of the ANO-1 fire protection program to a risk-informed, 
performance-based program based on National Fire Protection Association 
(NFPA) 805, in accordance with 10 CFR 50.48(c). NFPA 805 allows the use 
of performance-based methods such as fire modeling and risk-informed 
methods such as fire probabilistic risk assessment to demonstrate 
compliance with the nuclear safety performance criteria.
    Date of issuance: October 7, 2016.
    Effective date: As of the date of issuance and shall be implemented 
as described in the transition license conditions.
    Amendment No.: 256. A publicly-available version is in ADAMS under 
Accession No. ML16223A481; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-51: The amendment 
revised the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2014 (79 FR 
38589). The supplemental letters dated May 19, June 16, July 21, August 
12, September 22, November 4, and November 17, 2015; and January 15, 
March 25, April 7, May 19, and August 29, 2016, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 7, 2016.
    No significant hazards consideration comments received: No.

[[Page 73445]]

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

    Date of amendment request: October 15, 2015, as supplemented by a 
letter dated May 6, 2016.
    Description of amendment request: The amendments revised the St. 
Lucie Plant, Unit Nos. 1 and 2, Technical Specifications (TSs) and 
licensing bases to reflect the use of the commercially available 
computer code ``Generation of Thermal-Hydraulic Information for 
Containments (GOTHIC Version 7.2b(QA))'' to model the containment 
response following the inadvertent actuation of the containment spray 
system during normal plant operation (referred to as the vacuum 
analysis). The amendments also updated the licensing bases to credit 
the design basis ability of the containment vessel to withstand a 
higher external pressure differential of 1.04 pounds per square inch 
(psi) for Unit No. 1 and 1.05 psi for Unit No. 2, and updated TS 
3.6.1.4 for each unit to revise the allowable containment operating 
pressure range.
    Date of issuance: October 5, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 236 (Unit No. 1) and 186 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML16166A424; 
documents related to these amendments are listed in the Safety 
Evaluation (SE) enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: February 16, 2016 (81 
FR 7839). The supplemental letter dated May 6, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in an SE dated October 5, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 12, 2015, as supplemented by 
letters dated December 9, 2015, and March 14, March 29, April 7, April 
20, August 16, September 16, September 21, and September 29, 2016.
    Brief description of amendments: By order dated May 6, 2016, as 
published in the Federal Register on May 23, 2016 (81 FR 32350), the 
NRC approved the transfer of Facility Operating License (FOL) Nos. NPF-
87 and NPF-89 for Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2, 
and the general license for the independent spent fuel storage 
installation facility from the current holder, Luminant Generation 
Company LLC, to Comanche Peak Power Company LLC, as owner, and TEX 
Operations Company LLC, as operator. The conforming amendments revised 
the FOLs to reflect the direct transfer of ownership and the indirect 
transfer of control of the licenses.
    Date of issuance: October 3, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 7 days from the date of issuance.
    Amendment Nos.: 167 (Unit No. 1) and 167 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML16266A005; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Facility Operating License Nos. NPF-87 and NPF-89: Amendments 
revised the FOLs.
    Date of initial notice in Federal Register: February 8, 2016 (81 FR 
6545). The supplemental letters dated March 14, March 29, April 7, 
April 20, August 16, September 16, September 21, and September 29, 
2016, provided additional information that clarified the application 
and did not expand the scope of the application as originally noticed.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 6, 2016.

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

    Date amendment request: October 12, 2015.
    Brief description of amendments: The amendments revised Salem 
Nuclear Generating Station, Unit Nos. 1 and 2, Technical Specifications 
(TSs) by adding MODE 4 to the applicability of TS 3.6.2.3, 
``Containment Cooling System.'' The amendments also revised TS 3.7.1.1, 
``Safety Valves,'' to correct discrepancies between the applicable 
modes and the action statements.
    Date of issuance: September 29, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 315 (Unit No. 1) and 296 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML16229A519; 
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 and TSs.
    Date of initial notice in Federal Register: January 5, 2016 (81 FR 
264).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 29, 2016.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company and South Carolina Public Service 
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear 
Station, Units 2 and 3, Fairfield County, South Carolina

    Date of amendment request: March 14, 2016, as supplemented by 
letters dated May 12, 2016, and July 12, 2016.
    Description of amendment: The amendments incorporated changes that 
are consistent with those generically approved in WCAP-17524-P-A, 
Revision 1, ``AP1000 Core Reference Report,'' dated February 19, 2015. 
The amendments also approved changes to the Updated Final Safety 
Analysis Report (UFSAR) in the form of departures from the incorporated 
plant-specific Design Control Document Tier 2 licensing basis 
information, involves changes to the UFSAR information that has been 
designated as Tier 2* information, and involves changes to the plant-
specific Technical Specifications.
    Date of issuance: September 20, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 50 days of issuance.
    Amendment Nos.: 52 (Unit 2) and 52 (Unit 3). A publicly-available 
version is in ADAMS under Package Accession No. ML16144A591; documents 
related to these amendments are listed in the Safety Evaluation 
enclosed with the amendments.
    Facility Combined Licenses Nos. NPF-93 and NPF-94: Amendments 
revised the Facility Combined Licenses and Technical Specifications.
    Date of initial notice in Federal Register: May 10, 2016 (81 FR 
28900). The supplemental letters dated May 12, 2016, and July 12, 2016, 
provided additional information that clarified the

[[Page 73446]]

application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 20, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 10, 2014, as supplemented by 
letters dated May 4, 2015; October 15, 2015; and August 26, 2016.
    Brief description of amendments: The amendments revised Technical 
Specifications (TSs) by adopting 18 previously NRC-approved Technical 
Specifications Task Force (TSTF) Travelers, two TSTF T-Travelers, and 
one feature of the Improved Standard Technical Specifications not 
associated with a Traveler.
    Date of issuance: September 29, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 279 (Unit 1) and 223 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16231A041; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: March 31, 2015 (80 FR 
17095). The supplemental letters dated May 4, 2015; October 15, 2015; 
and August 26, 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 September 29, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 15, 2015, as supplemented by 
letters dated March 16, May 9, and May 16, 2016.
    Brief description of amendments: The amendments revised the 
Technical Specifications Surveillance Requirement 3.6.4.1.3 to increase 
the allowable time from 2 minutes to 10 minutes for the standby gas 
treatment system to draw down the secondary containment to negative 
pressure.
    Date of issuance: September 30, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 280 (Unit No. 1) and 224 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML16235A287; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 2015 (80 
FR 73240). The supplemental letters dated March 16, May 9, and May 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 amendments is contained 
in a Safety Evaluation dated September 30, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 16, 2016.
    Brief description of amendments: The amendments consisted of change 
to the Completion Date of Cyber Security Plan (CSP) Implementation 
Milestone 8--full implementation of the CSP from October 31, 2016 to 
December 31, 2017.
    Date of issuance: October 3, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 337 (Unit 1) and 330 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16228A096; documents related 
to these amendments are listed in the Safety Evaluation (SE) enclosed 
with the amendments.
    Facility Operating License Nos. DPR-77 and DPR-79. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2016 (81 FR 
44665).
    The Commission's related evaluation of the amendments is contained 
in an SE dated October 3, 2016.
    No significant hazards consideration comments received: No.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.

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

    Date of amendment request: March 11, 2016, as supplemented by 
letters dated May 31, 2016, and July 22, 2016.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) by adding a new Condition A to TS 3.7.8, 
``Essential Raw Cooling Water (ERCW) System,'' to extend the allowed 
completion time to restore ERCW System train to OPERABLE status from 72 
hours to 7 days for planned maintenance when the opposite unit is 
defueled or in Mode 6 following defueled under certain restrictions.
    Date of issuance: September 29, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 336 (Unit 1) and 329 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16225A276; documents related 
to these amendments are listed in the Safety Evaluation (SE) enclosed 
with the amendments.
    Facility Operating License Nos. DPR-77 and DPR-79. Amendments 
revised the TSs.
    Date of initial notice in Federal Register: April 12, 2016 (81 FR 
21603). The supplemental letters dated May 31, 2016, and July 22, 2016, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in an SE dated September 29, 2016.

[[Page 73447]]

    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 13th day of October 2016.

    For the Nuclear Regulatory Commission.
George A. Wilson, Jr.,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2016-25641 Filed 10-24-16; 8:45 am]
 BILLING CODE 7590-01-P