[Federal Register Volume 84, Number 168 (Thursday, August 29, 2019)]
[Notices]
[Pages 45537-45552]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2019-18617]


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NUCLEAR REGULATORY COMMISSION

[NRC-2019-0168]


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 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 July 30, 2019 to August 12, 2019. The last 
biweekly notice was published on August 13, 2019.

DATES: Comments must be filed by September 30, 2019. A request for a 
hearing must be filed by October 28, 2019.

ADDRESSES: You may submit comments by any of the following methods:
     Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2019-0168. Address 
questions about NRC docket IDs in Regulations.gov to Jennifer Borges; 
telephone: 301-287-9127; email: [email protected]. For technical 
questions, contact the individual listed in the FOR FURTHER INFORMATION 
CONTACT section of this document.
     Mail comments to: Office of Administration, Mail Stop: 
TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, ATTN: Program Management, Announcements and Editing Staff.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Kay Goldstein, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-1506, email: [email protected].

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2019-0168, 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 Website: Go to https://www.regulations.gov/ and search for Docket ID NRC-2019-0168.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``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-2019-0168, 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 
https://www.regulations.gov/ as well as enter

[[Page 45538]]

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.

II. Background

    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.

III. 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 section 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 petition for leave 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. The NRC's regulations are accessible 
electronically from the NRC Library on the NRC's website at https://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of 
the regulations is available at the NRC's Public Document Room, located 
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. If a petition is filed, the 
Commission or a presiding officer will rule on the petition and, if 
appropriate, a notice of a hearing will be issued.
    As required by 10 CFR 2.309(d) the petition should specifically 
explain the reasons why intervention should be permitted with 
particular reference to the following general requirements for 
standing: (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.
    In accordance with 10 CFR 2.309(f), the petition must also set 
forth the specific contentions which the petitioner seeks to have 
litigated in 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 must 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 the specific sources and 
documents on which the petitioner intends to rely to support its 
position on the issue. The petition must include sufficient information 
to show that a genuine dispute exists with the applicant or licensee on 
a material issue of law or fact. Contentions must 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 the requirements at 10 CFR 2.309(f) 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. 
Parties have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that party's admitted 
contentions, including the opportunity to present evidence, consistent 
with the NRC's regulations, policies, and procedures.
    Petitions must be filed no later than 60 days from the date of 
publication of this notice. Petitions and motions for leave to file new 
or amended contentions that are filed after the 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). The petition must be filed in 
accordance with the filing instructions in the ``Electronic Submissions 
(E-Filing)'' section of this document.
    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 establish when the hearing is held. If the final determination is 
that the amendment request involves no significant hazards 
consideration, the

[[Page 45539]]

Commission may issue the amendment and make it immediately effective, 
notwithstanding the request for a hearing. Any hearing 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 the amendment 
unless the Commission finds an imminent danger to the health or safety 
of the public, in which case it will issue an appropriate order or rule 
under 10 CFR part 2.
    A State, local governmental body, Federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission no later 
than 60 days from the date of publication of this notice. 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. Alternatively, a State, local 
governmental body, Federally-recognized Indian Tribe, or agency thereof 
may participate as a non-party under 10 CFR 2.315(c).
    If a hearing is granted, any person who is not a party to the 
proceeding and is not affiliated with or represented by a party may, at 
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 
his or her 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 and petition for leave to intervene (petition), any 
motion or other document filed in the proceeding prior to the 
submission of a request for hearing or petition to intervene, and 
documents filed by interested governmental entities that request to 
participate 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. Detailed 
guidance on making electronic submissions may be found in the Guidance 
for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/site-help/e-submittals.html. 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 (1) request a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign submissions and access the E-Filing 
system for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a petition or 
other adjudicatory document (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 website at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a 
digital ID certificate and a docket has been created, the participant 
can then submit adjudicatory documents. Submissions must be in Portable 
Document Format (PDF). Additional guidance on PDF submissions is 
available on the NRC's public website at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the 
time the document is 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 
document on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before adjudicatory documents are 
filed so that they can obtain access to the documents via the E-Filing 
system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic 
Filing Help Desk through the ``Contact Us'' link located on the NRC's 
public website at https://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 6 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, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff. Participants filing adjudicatory documents 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.

[[Page 45540]]

    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the 
Commission or the presiding officer. If you do not have an NRC-issued 
digital ID certificate as described above, click ``Cancel'' when the 
link requests certificates and you will be automatically directed to 
the NRC's electronic hearing dockets where you will be able to access 
any publicly available documents in a particular hearing docket. 
Participants are requested not to include personal privacy information, 
such as social security numbers, home addresses, or personal phone 
numbers in their filings, unless an NRC regulation or other law 
requires submission of such information. For example, in some 
instances, individuals provide home addresses in order to demonstrate 
proximity to a facility or site. With respect to copyrighted works, 
except for limited excerpts that serve the purpose of the adjudicatory 
filings and would constitute a Fair Use application, participants are 
requested not to include copyrighted materials in their submission.
    For further details with respect to these license amendment 
application(s), 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.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

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

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

Exelon Generation Company, LLC and Exelon FitzPatrick, LLC, Docket No. 
50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New 
York

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

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

Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine 
Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York

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

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

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

    Date of amendment request: June 25, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19176A498.
    Description of amendment request: The amendments would revise 
instrument testing and calibration definitions in the technical 
specifications (TS) for each facility to incorporate the surveillance 
frequency control program. The proposed amendments are based on 
Technical Specification Task Force (TSTF) traveler TSTF-563, Revision 
0, ``Revise Instrument Testing Definitions to Incorporate the 
Surveillance Frequency Control Program'' (ADAMS Accession No. 
ML17130A819).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the TS definitions of Channel 
Calibration, Channel Functional Test, Channel Operational Test, and 
Trip Actuating Device Operational Test to allow the frequency for 
testing the components or devices in each step to be determined in 
accordance with the TS Surveillance Frequency Control Program, as 
applicable. All components in the channel continue to be calibrated. 
The frequency at which a channel calibration is performed is not an 
initiator of any accident previously evaluated, so the probability 
of an accident is not affected by the proposed change. The channels 
surveilled in accordance with the affected definitions continue to 
be required to be operable and the acceptance criteria of the 
surveillances are unchanged. As a result, any mitigating functions 
assumed in the accident analysis will continue to be performed.
    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 previously evaluated?
    Response: No.
    The proposed change revises the TS definitions of Channel 
Calibration, Channel Functional Test, Channel Operational Test, and 
Trip Actuating Device Operational Test to allow the frequency for 
testing the components or devices in each step to be determined in 
accordance with the TS Surveillance Frequency Control Program, as 
applicable. The design function or operation of the components 
involved are not affected and there is no physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed). No credible new failure mechanisms, malfunctions, or 
accident initiators not considered in the design and licensing bases 
are introduced. The changes do not alter assumptions made in the 
safety analysis. The proposed changes are 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 previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises the TS definitions of Channel 
Calibration, Channel Functional Test, Channel Operational Test, and 
Trip Actuating Device Operational Test to allow the frequency for 
testing the components or devices in each step to be determined in 
accordance with the TS Surveillance Frequency Control Program, as 
applicable. The Surveillance Frequency Control Program assures 
sufficient safety margins are maintained, and that design, 
operation, surveillance methods, and acceptance criteria specified 
in applicable codes and standards (or alternatives approved for use 
by the NRC) will continue to be met as described in the plants' 
licensing basis. The proposed change does not adversely affect 
existing plant safety margins or the reliability of the equipment 
assumed to operate in the safety analysis. As such, 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 change. Margins of safety are 
unaffected by method of determining surveillance test intervals 
under an NRC-approved licensee-controlled program.
    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

[[Page 45541]]

review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Acting Branch Chief: Lisa M. Regner.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

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

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

Exelon Generation Company, LLC and Exelon FitzPatrick, LLC, Docket No. 
50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New 
York

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois


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

Exelon Generation Company, LLC, Docket No. 50-220, Nine Mile Point 
Nuclear Station Unit No. 1, Oswego County, New York

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

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

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

    Date of amendment request: June 27, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19178A291.
    Description of amendment request: The amendments would revise the 
requirements in the technical specifications for each facility related 
to the unavailability of barriers. The proposed amendments are based on 
Technical Specification Task Force (TSTF) traveler TSTF-427, Revision 
2, ``Allowance for Non Technical Specification Barrier Degradation on 
Supported System OPERABILITY'' (ADAMS Accession No. ML061240055).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided (via incorporation by reference) its analysis of the issue of 
no significant hazards consideration, which is presented below:

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

    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an unavailable hazard barrier if risk is assessed and 
managed. The postulated initiating events which may require a 
functional barrier are limited to those with low frequencies of 
occurrence, and the overall TS system safety function would still be 
available for the majority of anticipated challenges. Therefore, the 
probability of an accident previously evaluated is not significantly 
increased, if at all. The consequences of an accident while relying 
on the allowance provided by proposed [Limiting Condition for 
Operation] LCO 3.0.9 are no different than the consequences of an 
accident while relying on the TS required actions in effect without 
the allowance provided by proposed LCO 3.0.9. Therefore, the 
consequences of an accident previously evaluated are not 
significantly affected by this change. The addition of a requirement 
to assess and manage the risk introduced by this change will further 
minimize possible concerns. Therefore, this change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to an unavailable hazard barrier, if 
risk is assessed and managed, will not introduce new failure modes 
or effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.

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

    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an unavailable 
hazard barrier, if risk is assessed and managed. The postulated 
initiating events which may require a functional barrier are limited 
to those with low frequencies of occurrence, and the overall TS 
system safety function would still be available for the majority of 
anticipated challenges. The risk impact of the proposed TS changes 
was assessed following the three-tiered approach recommended in 
[Regulatory Guide] RG 1.177. A bounding risk assessment was 
performed to justify the proposed TS changes. This application of 
LCO 3.0.9 is predicated upon the licensee's performance of a risk 
assessment and the management of plant risk. The net change to the 
margin of safety is insignificant. Therefore, this change does not 
involve a significant reduction in a margin of safety.

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

Exelon Generation Company, LLC (EGC), Docket Nos. STN 50-456 and STN 
50-457, Braidwood Station, Units 1 and 2, Will County, Illinois

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert 
Cliffs Nuclear Power Plant (Calvert Cliffs), Unit Nos. 1 and 2, Calvert 
County, Maryland

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

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

Exelon Generation Company, LLC and Exelon FitzPatrick, LLC, Docket No. 
50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New 
York

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois


[[Page 45542]]


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

Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine 
Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York

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

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

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

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

    Date of amendment request: June 26, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19178A304.
    Description of amendment request: Except for Calvert Cliffs, the 
proposed amendments would revise the technical specifications (TS) for 
high radiation area administrative controls. The proposed amendments 
for Calvert Cliffs would add TS requirements for high radiation area 
administrative controls.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes are administrative in nature and only 
related to the control of access to high radiation areas for 
controlling dose to plant personnel. The proposed changes do not 
impact any accident initiators and do not require any plant 
modifications which affect the performance capability of the 
structures, systems and components relied upon to mitigate the 
consequences of postulated accidents; therefore, there is no impact 
to the probability or consequences of an accident previously 
evaluated.
    Based on the above, EGC concludes that the proposed changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Will operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Response: No.
    The proposed amendments involve changes to radiological program 
controls for access to high radiation areas, which are 
administrative in nature and do not impact physical plant systems. 
These proposed changes do not alter accident analysis assumptions, 
add any initiators, or affect the function of plant systems or the 
manner in which systems are operated, maintained, modified, tested, 
or inspected. The proposed changes do not require any plant 
modifications which affect the performance capability of the 
structures, systems and components relied upon to mitigate the 
consequences of postulated accidents.
    Based on the above discussion, EGC concludes that the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed changes are administrative in nature and only 
related to the control of access to high radiation areas to minimize 
dose to plant personnel. The proposed changes are intended to 
provide clarity and/or flexibility with respect to the 
administration and programmatic controls while retaining adequate 
margin of safety for minimizing dose to site personnel consistent 
with the requirements of 10 CFR 20, ``Standards for Protection 
Against Radiation,'' and the guidance of [Regulatory Guide] RG 8.38, 
``Control of Access to High and Very High Radiation Areas in Nuclear 
Power Plants,'' published in May 2006. Since there are no associated 
physical plant changes, the ability of the plant to respond to and 
mitigate accidents is unchanged by the proposed changes.
    Based on the above, EGC concludes that the proposed changes do 
not involve a significant reduction in a margin of safety.

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

Exelon Generation Company (EGC), LLC, Docket No. 50-461, Clinton Power 
Station (CPS), Unit No. 1, DeWitt County, Illinois and Docket Nos. 50-
237 and 50-249, Dresden Nuclear Power Station (DNPS), Units 2 and 3, 
Grundy County, Illinois

    Date of amendment request: June 18, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19169A146.
    Description of amendment request: The proposed amendments would 
revise the CPS, Unit No. 1, and DNPS, Units 2 and 3, technical 
specifications (TSs) associated with TS 3.5.2, ``Reactor Pressure 
Vessel (RPV) Water Inventory Control (WIC),'' and TS 3.8.2, ``AC 
Sources--Shutdown,'' surveillance requirements considered no longer 
necessary following NRC-approved licensing activity at these sites. For 
each site, a change to TS 3.3.5.2, ``Reactor Pressure Vessel (RPV) 
Water Inventory Control Instrumentation,'' is proposed to support 
instrumentation functions. Additionally, edits are proposed to RPVWIC-
related TSs to add consistency and clarity. For DNPS, Units 2 and 3 
only, a change to TS 3.6.1.3, ``Primary Containment Isolation Valves,'' 
is proposed to support Mode 4 and 5 operations.
    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 modifies existing TS requirements related to 
the maintenance of RPV inventory in Modes 4 and 5. Draining of RPV 
water inventory in Modes 4 and 5 is not an accident previously 
evaluated and, therefore, replacing the existing TS controls to 
prevent or mitigate such an event with a modified set of controls 
has no effect on any accident previously evaluated. RPV water 
inventory control in Mode 4 or Mode 5 is not an initiator of any 
accident previously evaluated. The existing and the proposed RPV WIC 
controls are not mitigating actions assumed in any accident 
previously evaluated.
    The proposed changes do not affect the probability of an 
unexpected draining event (which is not a previously evaluated 
accident) or the limiting time in which an unexpected draining event 
could result in the reactor vessel water level dropping to the TAF 
[top of active fuel]. The current TS requirements are only 
mitigating actions and impose no requirements that reduce the 
probability of an unexpected draining event.

[[Page 45543]]

    The proposed changes do not affect the consequences of an 
unexpected draining event (which is not a previously evaluated 
accident) or the current requirement to maintain an operable ECCS 
[emergency core cooling system] subsystem at all times in Modes 4 
and 5. The proposed changes do not significantly affect the 
consequences of an unexpected draining event because the proposed 
Actions continue to ensure equipment is available within the 
limiting DRAIN TIME, and are equivalent to the current requirements.
    The proposed changes reduce or eliminate some requirements that 
were determined to be unnecessary to manage the consequences of an 
unexpected draining event, such as the automatic starting of EDGs 
[emergency diesel generators] on ECCS initiation signals. These 
changes do not affect the consequences of any accident previously 
evaluated since a draining event in Modes 4 and 5 is not a 
previously evaluated accident and the requirements proposed for 
elimination are not needed to adequately respond to a draining 
event.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    The proposed changes replace existing TS requirements related to 
RPV WIC with modified requirements that will continue to protect 
Safety Limit 2.1.1.3. The proposed changes will not alter the design 
function of the equipment involved.
    The event of concern under the current requirements and the 
proposed changes is an unexpected draining event. The proposed 
changes do not create new failure mechanisms, malfunctions, or 
accident initiators that would cause an RPV or refueling cavity 
draining event or a new or different kind of accident not previously 
evaluated or included in the design and licensing bases.
    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 changes modify certain existing TS requirements 
related to RPV WIC. The safety basis for the current RPV WIC 
requirements is to protect Safety Limit 2.1.1.3. The new TS 
requirements continue to meet this safety basis in all respects.
    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 Acting Branch Chief: Lisa M. Regner.

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

    Date of amendment request: March 5, 2019, as supplemented by 
letters dated May 23 and July 22, 2019. Publicly-available versions are 
in ADAMS under Accession Nos. ML19064B368, ML19143A347, and 
ML19203A176, respectively.
    Description of amendment request: The proposed amendment would: 
revise the combined main steam isolation valve (MSIV) leakage rate 
limit for all four steam lines in Technical Specification (TS) TS 
3.6.1.3, ``Primary Containment Isolation Valves (PCIVs),'' Surveillance 
Requirement (SR) 3.6.1.3; revise the leakage rate through each MSIV 
leakage path; add a new TS 3.6.2.6, ``Residual Heat Removal (RHR) 
Drywell Spray''; and revise TS 3.6.4.1, ``Secondary Containment,'' to 
address short-duration conditions during which the secondary 
containment pressure may not meet the SR pressure requirement, in 
accordance with Technical Specifications Task Force Traveler (TSTF) 
551, ``Revise Secondary Containment Surveillance Requirements,'' 
Revision 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The increase in the total MSIV leakage rate limit has been 
evaluated in a revision to the radiological consequence analysis of 
the Loss of Coolant Accident (LOCA). Based on the results of the 
analysis, it has been demonstrated that, with the requested change, 
the dose consequences of this limiting Design Basis Accident (DBA) 
are within the acceptance criteria provided by the NRC for use with 
the Alternative Source Term (AST) methodology in 10 CFR 50.67 and 10 
CFR 50, appendix A, GDC [General Design Criteria] 19. Additional 
guidance is provided in Regulatory Guide 1.183, ``Alternative 
Radiological Source Terms for Evaluating Design Basis Accidents at 
Nuclear Power Reactors'' and Standard Review Plan (SRP) Section 
15.0.1.
    The proposed change to the MSIV leakage limit does not involve 
physical change to any plant structure, system, or component. As a 
result, no new failure modes of the MSIVs have been introduced.
    The proposed change does not affect the normal design or 
operation of the facility before the accident; rather, it affects 
leakage limit assumptions that constitute inputs to the evaluation 
of the consequences. The radiological consequences of the analyzed 
LOCA have been evaluated using the plant licensing basis for this 
accident. The resulting doses are slightly higher than the 
previously approved AST doses; with exception of the Control Room 
dose that is slightly lower. However, adequate margin to the 
regulatory limits specified in 10 CFR 50.67 for offsite doses and 10 
CFR 50, Appendix A, GDC 19 for control room operator doses is still 
available. Thus, the results conclude that the control room and 
offsite doses remain within applicable regulatory limits. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    In addition, the proposed change to SR 3.6.4.1.1 addresses 
short-duration conditions during which the secondary containment 
vacuum requirement is 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. 
The consequences of an accident previously evaluated while utilizing 
the proposed changes are no different than the consequences of an 
accident while utilizing the existing four-hour Completion Time 
(i.e., allowed outage time) for an inoperable secondary containment. 
In addition, the proposed change provides an alternative means to 
ensure the secondary containment safety function is met. As a 
result, the consequences of an accident previously evaluated are not 
significantly increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    The change in the MSIV leakage rate limits does not affect the 
design, functional performance, or normal operation of the facility. 
Similarly, it does not affect the design or operation of any 
component in the facility such that new equipment failure modes are 
created. This is supported by operating experience at other EGC 
sites that have increased their MSIV leakage limits. As such the 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    In addition, the proposed change to SR 3.6.4.1.1 does not alter 
the protection system design, create new failure modes, or change 
any modes of operation. The proposed change does not involve a 
physical alteration of the plant; and no new or different kind of 
equipment will be installed. Consequently, there are no new 
initiators that could result in a new or different kind of accident.

[[Page 45544]]

    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This proposed license amendment involves changes in the MSIV 
leakage rate limits. The revised leakage rate limits are used in the 
reanalysis of the LOCA radiological consequences.
    The analysis has been performed using conservative 
methodologies. Safety margins and analytical conservatisms have been 
evaluated and have been found acceptable. The analyzed LOCA event 
has been carefully selected and margin has been retained to ensure 
that the analysis adequately bounds postulated event scenario. The 
dose consequences of this limiting event are within the acceptance 
criteria presented in 10 CFR 50.67 for offsite doses and 10 CFR 50, 
appendix A, GDC 19 for control room operator doses. The margin of 
safety is that provided by meeting the applicable regulatory limits.
    In addition, the proposed change to SR 3.6.4.1.1 addresses 
short-duration conditions during which the secondary containment 
vacuum requirement is not met. Conditions in which the secondary 
containment vacuum is less than the required vacuum are acceptable 
provided the conditions do not affect the ability of the SGT 
[standby gas treatment] System to establish the required secondary 
containment vacuum under post-accident conditions within the time 
assumed in the accident analysis. This condition is incorporated in 
the proposed change by requiring an analysis of actual environmental 
and secondary containment pressure conditions to confirm the 
capability of the SGT System is maintained within the assumptions of 
the accident analysis. Therefore, the safety function of the 
secondary containment is not affected.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center (DAEC), Linn County, Iowa

    Date of amendment request: April 9, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19101A280.
    Description of amendment request: The amendment would revise the 
DAEC Emergency Plan on-shift and augmented Emergency Response 
Organization (ERO) staffing to support the planned permanent cessation 
of operations and permanent defueling of the DAEC reactor. 
Specifically, the proposed changes would eliminate the on-shift 
positions not needed for the safe storage of spent fuel in the spent 
fuel pool during the initial decommissioning period and eliminate the 
ERO positions not necessary to effectively respond to credible 
accidents. The proposed changes in staffing are commensurate with the 
reduced spectrum of credible accidents for a permanently shut down and 
defueled power reactor facility.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the DAEC Emergency Plan do not impact 
the function of plant Structures, Systems, or Components (SSCs). The 
proposed changes do not involve the modification of any plant 
equipment or affect plant operation. The proposed changes do not 
affect accident initiators or precursors, nor do the proposed 
changes alter design assumptions. The proposed changes do not 
prevent the ability of the on-shift staff and ERO to perform their 
intended functions to mitigate the consequences of any accident or 
event that will be credible in the permanently defueled condition. 
The proposed changes only remove positions that will no longer be 
needed or credited in the Emergency Plan in the permanently defueled 
condition.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes reduce the number of on-shift and ERO 
positions commensurate with the hazards associated with a 
permanently shut down and defueled facility. The proposed changes do 
not involve installation of new equipment or modification of 
existing equipment, so that no new equipment failure modes are 
introduced. Additionally, the proposed changes do not result in a 
change to the way that the equipment or facility is operated so that 
no new accident initiators are created.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    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. The proposed changes are 
associated with the Emergency Plan and 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 the 
proposed changes and margins of safety are maintained. The revised 
Emergency Plan will continue to provide the necessary response staff 
with the proposed changes.
    Therefore, the proposed changes have no impact to 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: Steven Hamrick, Managing Attorney--Nuclear, 
Florida Power Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Acting Branch Chief: Lisa M. Regner.

NextEra Energy Duane Arnold (NEDA), LLC, Docket No. 50-331, Duane 
Arnold Energy Center (DAEC), Linn County, Iowa

    Date of amendment request: June 20, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19176A356.
    Description of amendment request: NEDA requests an amendment to the 
DAEC operating license (OL) and technical specifications (TSs). The 
proposed changes will revise the OL and TSs consistent with the 
permanent cessation of reactor operation and permanent defueling of the 
reactor. The revised OL and TSs will be identified as the DAEC post 
defueled technical specifications (PDTSs). By letter dated January 18, 
2019 (ADAMS Accession No. ML19023A196), NEDA provided formal 
notification to the NRC pursuant

[[Page 45545]]

to 10 CFR 50.82(a)(1)(i) and 10 CFR 50.4(b)(8) of the intention to 
permanently cease power operations at the DAEC in the fourth quarter of 
2020. After the certifications of permanent cessation of power 
operation and of permanent removal of fuel from the DAEC reactor vessel 
are docketed, in accordance with 10 CFR 50.82(a)(1)(i) and (ii) 
respectively, and pursuant to 10 CFR 50.82(a)(2), the 10 CFR 50 license 
will no longer authorize reactor operation or emplacement or retention 
of fuel in the reactor vessel. As a result, certain license conditions 
and TSs may be revised or removed to reflect the permanently defueled 
condition. In general, the changes propose the elimination of items 
applicable in operating conditions where fuel is placed in the reactor 
vessel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes would not take effect until DAEC has 
certified to the NRC that it has permanently ceased operation and 
entered a permanently defueled condition. Because the 10 CFR part 50 
license for DAEC will no longer authorize operation of the reactor, 
or emplacement or retention of fuel into the reactor vessel with the 
certifications required by 10 CFR part 50.82(a)(1) submitted, as 
specified in 10 CFR part 50.82(a)(2), the occurrence of postulated 
accidents associated with reactor operation is no longer credible. 
DAEC's accident analyses are contained in Chapter 15 of the Updated 
Final Safety Analysis Report (UFSAR). In a permanently defueled 
condition, the only credible UFSAR described accident that remains 
is the Fuel Handling Accident (FHA). Other Chapter 15 accidents will 
no longer be applicable to a permanently defueled reactor.
    The UFSAR-described FHA analyses for DAEC shows that, following 
the required decay time after reactor shutdown and provided the SFP 
[spent fuel pool] water level requirement of TS LCO [limiting 
condition for operation] 3.7.8 is met, the dose consequences are 
acceptable without relying on secondary containment or the Standby 
Gas Treatment System. The control building envelop is credited for 
reduction of operator dose. Consequently, the TS requirements for 
the Standby Filter Unit and Control Building Chillers are retained.
    The probability of occurrence of previously evaluated accidents 
is not increased, since safe storage and handling of fuel will be 
the only operations performed, and therefore, bounded by the 
existing analyses. Additionally, the occurrence of postulated 
accidents associated with reactor operation will no longer be 
credible in the permanently defueled condition. This significantly 
reduces the scope of applicable accidents. The deletion of TS 
definitions and rules of usage and application requirements that 
will not be applicable in a defueled condition has no impact on 
facility SSCs [structures, system, and components] or the methods of 
operation of such SSCs. The deletion of design features and safety 
limits not applicable to the permanently shut down and defueled DAEC 
has no impact on the remaining applicable DBA [design-basis 
accident].
    The removal of LCOs or SRs [surveillance requirements] that are 
related only to the operation of the nuclear reactor or only to the 
prevention, diagnosis, or mitigation of reactor-related transients 
or accidents do not affect the applicable DBAs previously evaluated 
since these DBAs are no longer applicable in the permanently 
defueled condition.
    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 changes to delete or modify certain DAEC Operating 
License, TS, and current licensing bases (CLB) have no impact on 
facility SSCs affecting the safe storage of spent irradiated fuel, 
or on the methods of operation of such SSCs, or on the handling and 
storage of the spent irradiated fuel itself. The removal of TS that 
are related only to the operation of the nuclear reactor, or only to 
the prevention, diagnosis, or mitigation of reactor related 
transients or accidents, cannot result in different or more adverse 
failure modes or accidents than previously evaluated because the 
reactor will be permanently shut down and defueled.
    The proposed modification or deletion of requirements of the 
DAEC Operating License, TS, and CLB do not affect systems credited 
in the accident analysis for the remaining credible DBA at DAEC. The 
proposed Operating License and PDTS will continue to require proper 
control and monitoring of safety significant parameters and 
activities. The TS regarding SFP water level and spent fuel storage 
is retained to preserve the current requirements for safe storage of 
irradiated fuel. 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, spent fuel 
racks, SFP integrity, and SFP water level). Since extended operation 
in a defueled condition and safe fuel handling 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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes are to delete or modify certain Operating 
License, TS and CLB once the DAEC facility has been permanently shut 
down and defueled. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50 
license for DAEC will no longer authorize operation of the reactor 
or emplacement or retention of fuel into the reactor vessel 
following submittal of the certifications required by 10 CFR 
50.82(a)(1). As a result, the occurrence of certain design basis 
postulated accidents are no longer considered credible when the 
reactor is permanently defueled.
    The only remaining credible UFSAR described accident is a FHA. 
The proposed changes do not adversely affect the inputs or 
assumptions of any of the design basis analyses that impact the FHA.
    The proposed changes are limited to those portions of the 
Operating License, TS, and CLB that are not related to the safe 
storage of irradiated fuel. The requirements proposed to be revised 
or deleted from the Operating License, TS, and CLB are not credited 
in the existing accident analysis for the remaining postulated 
accident (i.e., FHA); and, as such, do not contribute to the margin 
of safety associated with the accident analysis. Certain postulated 
DBAs involving the reactor are no longer possible because the 
reactor will be permanently shut down and defueled and DAEC will no 
longer be authorized to operate the reactor.
    Therefore, the proposed changes have no impact to 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: Steven Hamrick, Managing Attorney--Nuclear, 
Florida Power Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Acting Branch Chief: Lisa M. Regner.

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

    Date of amendment request: June 4, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19157A057.
    Description of amendment request: The amendment would revise the 
Seabrook Technical Specifications (TSs) associated with the emergency 
core cooling system (ECCS) accumulators. Specifically, the proposed 
amendment would modify the TS actions for an inoperable accumulator, 
relocate the actions for inoperable accumulator instrumentation, and 
delete an unnecessary surveillance requirement. The proposed change 
would also delete

[[Page 45546]]

a duplicate surveillance requirement associated with the accumulator 
isolation valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Operability of the ECCS accumulators ensure that a sufficient 
volume of borated water will be immediately forced into the reactor 
core through each of the cold legs in the event the reactor coolant 
system (RCS) pressure falls below the pressure of the accumulators. 
This initial surge of water into the core provides the initial 
cooling mechanism during large RCS pipe ruptures. The proposed 
change does not change the limiting condition for operation (LCO) 
for the accumulators.
    The proposed change deletes a surveillance requirement that 
verifies the accumulator isolation valves automatically open on an 
actuation signal because the technical specifications require 
maintaining the motor-operated valves open and de-energized. In 
addition, the completion times for an inoperable accumulator are 
revised to 24 hours for inoperability due to reasons other than 
boron concentration outside limits and to 72 hours for boron not 
within limits. The consequences of an accident that might occur 
during the revised completion times are no different from those that 
might occur during the current completion times. The change to 
eliminate a duplicate surveillance requirement makes no technical 
changes and is administrative in nature.
    The proposed change does not alter the design, function, or 
operation of any plant structure, system, or component (SSC). The 
capability of any operable TS-required SSC to perform its specified 
safety function is not impacted by the proposed change. As a result, 
the outcomes of accidents previously evaluated are unaffected. 
Therefore, the proposed changes do not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not challenge the integrity or 
performance of any safety-related systems. No plant equipment is 
installed or removed, and the changes do not alter the design, 
physical configuration, or method of operation of any plant system 
or component. No physical changes are made to the plant, so no new 
causal mechanisms are introduced. Therefore, the proposed changes to 
the TS do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The ability of any operable ECCS equipment to perform its 
designated safety function is unaffected by the proposed changes. 
The proposed changes do not alter any safety analyses assumptions, 
safety limits, limiting safety system settings, or method of 
operating the plant. The changes do not adversely affect plant 
operating margins or the reliability of equipment credited in the 
safety analyses. With the proposed change, the ECCS remains capable 
of performing its safety function. Therefore, the proposed changes 
do not involve a significant reduction in the margin of safety.

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

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

    Date of amendment request: July 30, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19214A046.
    Description of amendment request: The proposed amendment would 
replace ``South Carolina Electric & Gas Company'' with ``Dominion 
Energy South Carolina, Inc.'' or ``DESC'' where appropriate in the 
Renewed Facility Operating License NPF-12.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment is administrative in nature. SCE&G, which 
has been renamed Dominion Energy South Carolina, Inc., will remain 
the licensee authorized to operate and possess VCSNS Unit 1, and its 
functions, powers, resources and management as described in the 
license will not change. The proposed changes do not adversely 
affect accident initiators or precursors, and do not alter the 
design assumptions, conditions, or configuration of the plant or the 
manner in which the plant is operated and maintained. The ability of 
structures, systems, and components to perform their intended safety 
functions is not altered or prevented by the proposed changes, and 
the assumptions used in determining the radiological consequences of 
previously evaluated accidents are not affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment is purely administrative in nature. The 
functions of the licensee will not change. These changes do not 
involve any physical alteration of the plant (i.e., no new or 
different type of equipment will be installed), and installed 
equipment is not being operated in a new or different manner. Thus, 
no new failure modes are introduced. 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 amendment is administrative in nature. SCE&G, which 
has been renamed Dominion Energy South Carolina, Inc., will remain 
the licensee authorized to operate and possess the units, and its 
functions as described in the license will not change. The proposed 
changes do not alter the manner in which safety limits, limiting 
safety system settings, or limiting conditions for operation are 
determined. There are no changes to setpoints at which protective 
actions are initiated, and the operability requirements for 
equipment assumed to operate for accident mitigation are not 
affected. 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: Kathryn M. Sutton, Morgan, Lewis & Bockius 
LLP, 1111 Pennsylvania Avenue NW, Washington, DC 20004.
    NRC Branch Chief: Michael T. Markley.

[[Page 45547]]

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

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 (HNP), Unit Nos. 1 and 2, Appling County, Georgia; and

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant (VEGP), Units 1 and 2, Burke 
County, Georgia

    Date of amendment request: July 15, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19196A222.
    Description of amendment request: The amendments would adopt 
Technical Specification Task Force (TSTF)-563, ``Revise Instrument 
Testing Definitions to Incorporate the Surveillance Frequency Control 
Program.'' TSTF-563 revises the Technical Specification (TS) 
definitions of Channel Calibration and Channel Functional Test in the 
HNP TS, and the definitions of Channel Calibration, Channel Operational 
Test (COT), and Trip Actuating Device Operational Test (TADOT) in the 
FNP and VEGP TS. The HNP, FNP, and VEGP Channel Calibration definition 
and the HNP Channel Functional Test definition currently permit 
performance by means of any series of sequential, overlapping, or total 
channel steps. The FNP and VEGP definitions of COT and TADOT are 
revised to explicitly permit performance by means of any series of 
sequential, overlapping, or total channel steps. The Channel 
Calibration, Channel Functional Test, COT, and TADOT definitions are 
revised to allow the required frequency for testing the components or 
devices in each step to be determined in accordance with the 
Surveillance Frequency Control Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    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 the TS definitions of Channel 
Calibration and Channel Functional Test in the HNP TS, and the 
definitions of Channel Calibration, COT, and TADOT in the FNP and 
VEGP TS to allow the frequency for testing the components or devices 
in each step to be determined in accordance with the Surveillance 
Frequency Control Program. The proposed change also explicitly 
permits the FNP and VEGP COT and TADOT to be performed by any series 
of sequential, overlapping, or total channel steps. All components 
in the channel continue to be tested. The frequency at which a 
channel test is performed is not an initiator of any accident 
previously evaluated, so the probability of an accident is not 
affected by the proposed change. The channels surveilled in 
accordance with the affected definitions continue to be required to 
be operable and the acceptance criteria of the surveillances are 
unchanged. As a result, any mitigating functions assumed in the 
accident analysis will continue to be performed.
    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 revises the TS definitions of Channel 
Calibration and Channel Functional Test in the HNP TS, and the 
definitions of Channel Calibration, COT, and TADOT in the FNP and 
VEGP TS to allow the frequency for testing the components or devices 
in each step to be determined in accordance with the Surveillance 
Frequency Control Program. The proposed change also explicitly 
permits the FNP and VEGP COT and TADOT to be performed by any series 
of sequential, overlapping, or total channel steps. The design 
function or operation of the components involved are not affected 
and there is no physical alteration of the plant (i.e., no new or 
different type of equipment will be installed). No credible new 
failure mechanisms, malfunctions, or accident initiators not 
considered in the design and licensing bases are introduced. The 
change does not alter assumptions made in the safety analysis. The 
proposed change is consistent with the safety analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises the TS definitions of Channel 
Calibration and Channel Functional Test in the HNP TS, and the 
definitions of Channel Calibration, COT, and TADOT in the FNP and 
VEGP TS to allow the frequency for testing the components or devices 
in each step to be determined in accordance with the Surveillance 
Frequency Control Program. The proposed change also explicitly 
permits the FNP and VEGP COT and TADOT to be performed by any series 
of sequential, overlapping, or total channel steps. The Surveillance 
Frequency Control Program assures sufficient safety margins are 
maintained, and that design, operation, surveillance methods, and 
acceptance criteria specified in applicable codes and standards (or 
alternatives approved for use by the NRC) will continue to be met as 
described in the plants' licensing basis. The proposed change does 
not adversely affect existing plant safety margins, or the 
reliability of the equipment assumed to operate in the safety 
analysis. As such, 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 change. Margins of safety are unaffected by method of 
determining surveillance test intervals under an NRC-approved 
licensee-controlled program.
    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: Millicent Ronnlund, Vice President and 
General Counsel, Southern Nuclear Operating Co., Inc., P. O. Box 1295, 
Birmingham, AL 35201-1295.
    NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-424 and 
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of amendment request: July 9, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19190A309.
    Description of amendment request: The amendments would revise the 
actions of Technical Specification (TS) 3.7.7, ``Component Cooling 
Water (CCW) System,'' TS 3.7.8, ``Nuclear Service Cooling Water (NSCW) 
System,'' TS 3.8.1, ``AC Sources--Operating,'' TS 3.8.4, ``DC Sources--
Operating,'' TS 3.8.7, ``Inverters--Operating,'' and TS 3.8.9, 
``Distribution Systems--Operating.'' The proposed license amendments 
modify action end states for the subject TS in conditions where more 
than one safety-related train is inoperable or the electrical power 
system is significantly degraded. Specifically, if the related required 
action statements are not met, instead of requiring the plant to 
achieve hot shutdown (i.e., Mode 4), the end state of cold shutdown 
(i.e., Mode 5) is required.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 45548]]

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 requires the plant to be placed in cold 
shutdown instead of hot shutdown when more than one safety-related 
train of the cooling water or electrical distribution systems are 
inoperable or when the electrical power system is significantly 
degraded (e.g., three or more required AC [alternating current] 
sources inoperable). Transitioning the plant from hot shutdown to 
cold shutdown is not an initiator of any accident previously 
evaluated but is assumed in the mitigation of accidents previously 
evaluated. Therefore, the probability of an accident previously 
evaluated is not adversely impacted by the proposed change.
    Component cooling water (CCW) and nuclear service cooling water 
(NSCW) systems and the safety-related electrical power and 
distribution systems are assumed in accident mitigation. SNC 
concludes the proposed change to require the plant be placed in cold 
shutdown instead of hot shutdown is acceptable because placing the 
unit in cold shutdown is considered a safe condition, since most 
design basis accidents and transients either cannot physically occur 
during cold shutdown, or would have significantly reduced plant 
impact and occur much less frequently due to the reduced 
temperatures and pressures in the plant. Therefore, the consequences 
of any accident that assumes the cooling water systems or electrical 
power and distribution systems are not significantly affected by 
this change.
    Consequently, 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 change the design function or 
operation of the cooling water systems or the electrical power and 
distribution systems. No plant modifications or changes to the plant 
configuration or method of operation are involved. The proposed 
change will not introduce new failure modes or effects and will not, 
in the absence of other unrelated failures, lead to an accident 
whose consequences exceed the consequences of accidents previously 
evaluated.
    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 change does not affect any of the controlling 
values of parameters used to avoid exceeding regulatory or licensing 
limits. The proposed change does not exceed or alter the design 
basis or safety limits, or any limiting safety system settings. The 
requirement for the CCW and NSCW systems to perform their designated 
support functions is unaffected. The requirement for the safety-
related electrical power and distribution systems to perform their 
designated support functions is unaffected. The proposed change to 
require the plant be placed in cold shutdown instead of hot shutdown 
is acceptable because placing the unit in cold shutdown is 
considered a safe condition, since most design basis accidents and 
transients either cannot physically occur during cold shutdown, or 
would have significantly reduced plant impact and occur much less 
frequently due to the reduced temperatures and pressures in the 
plant.
    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: Millicent Ronnlund, Vice President and 
General Counsel, Southern Nuclear Operating Co., Inc., P.O. Box 1295, 
Birmingham, AL 35201-1295.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: June 28, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19179A209.
    Description of amendment request: The amendment proposes changes to 
credit previously completed first plant only startup testing described 
in the Updated Final Safety Analysis Report (UFSAR), and related 
changes to the Combined License (COL) Nos. NPF-91 and NPF-92 for VEGP 
Units 3 and 4. Specifically, the proposed changes would revise the COL 
to delete conditions requiring the following tests: Natural Circulation 
(Steam Generator) Test, Rod Cluster Control Assembly (RCCA) Out of Bank 
Measurements, Load follow Demonstration, and the Passive Residual Heat 
Exchanger Test.
    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 does not affect the operation of any systems 
or equipment that initiates an analyzed accident or alter any 
structures, systems, or components (SSC) accident initiator or 
initiating sequence of events. The proposed change involves removing 
the requirement to perform first plant only startup tests including 
the Natural Circulation (Steam Generator) Test, the RCCA Out of Bank 
Measurements, the Load Follow Demonstration, and the Passive 
Residual Heat Exchanger Test. The request is based on the successful 
completion of these tests at the lead AP1000 unit. The change does 
not adversely affect any methodology which would increase the 
probability or consequences of a previously evaluated accident.
    The change does not impact the support, design, or operation of 
mechanical or fluid systems. There is no change to plant systems or 
the response of systems to postulated accident conditions. There is 
no change to predicted radioactive releases due to normal operation 
or postulated accident conditions. The plant response to previously 
evaluated accidents or external events is not adversely affected, 
nor does the proposed change create any new accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of a previously 
evaluated accident.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not affect the operation of any systems 
or equipment that may initiate a new or different kind of accident, 
or alter any SSC such that a new accident initiator or initiating 
sequence of events is created.
    The proposed change credits previously completed first plant 
only startup tests including the Natural Circulation (Steam 
Generator) Test, the RCCA Out of Bank Measurements, the Load Follow 
Demonstration, and the Passive Residual Heat Exchanger Test. The 
request is based on the successful completion of the tests at the 
lead AP1000 unit. The proposed changes do not adversely affect any 
design function of any SSC design functions or methods of operation 
in a manner that results in a new failure mode, malfunction, or 
sequence of events that affect safety-related or non-safety-related 
equipment. This activity does not allow for a new fission product 
release path, result in a new fission product barrier failure mode, 
or create a new sequence of events that result in significant fuel 
cladding failures.
    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 maintains existing safety margin and 
provides adequate protection through continued application of

[[Page 45549]]

the existing requirements in the UFSAR. The proposed change 
satisfies the same design functions in accordance with the same 
codes and standards as stated in the UFSAR. This change does not 
adversely 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. Since no safety analysis or design 
basis acceptance limit/criterion is challenged or exceeded by this 
change, no significant 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: Jennifer L. Dixon-Herrity.

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

    Date of amendment request: July 8, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19189A180.
    Description of amendment request: The amendment request proposes 
changes to the Combined License (COL) Numbers NPF-91 and NPF-92 for 
VEGP Units 3 and 4. The requested amendment proposes changes to 
Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) in COL 
Appendix C, with corresponding changes to the associated plant-specific 
Tier 1 information. Pursuant to the provisions of 10 CFR 52.63(b)(1), 
an exemption from elements of the design as certified in the 10 CFR 
part 52, appendix D, design certification rule is also requested for 
the plant-specific Design Control Document (DCD) Tier 1 material 
departures. Specifically, the requested amendment proposes changes to 
COL Appendix C (and plant-specific Tier 1) to remove a number of 
functional arrangement ITAAC, whose design commitments may be completed 
via other ITAAC or otherwise verified by other means.
    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 non-technical change to COL Appendix C will remove 
a number of functional arrangement ITAAC to improve efficiency of 
the ITAAC completion and closure process. No structure, system, or 
component (SSC) design or function is affected. No design or safety 
analysis is affected. The proposed changes do not affect any 
accident initiating event or component failure, thus the 
probabilities of the accidents previously evaluated are not 
affected. No function used to mitigate a radioactive material 
release and no radioactive material release source term is involved, 
thus the radiological releases in the accident analyses are not 
affected.
    Therefore, the 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 COL Appendix C do not affect the design 
or function of any SSC but will remove a number of functional 
arrangement ITAAC to improve efficiency of the ITAAC completion and 
closure process. The proposed changes would not introduce a new 
failure mode, fault or sequence of events that could result in a 
radioactive material release.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to COL Appendix C will remove a number of 
functional arrangement ITAAC to improve efficiency of the ITAAC 
completion and closure process, and would not affect any design 
parameter, function or analysis. There would be no change to an 
existing design basis, design function, regulatory criterion, or 
analysis. No safety analysis or design basis acceptance limit or 
criterion is involved.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

4. 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.

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

    Date of amendment request: February 5, 2019.
    Brief description of amendments: The amendments correct an 
editorial error in Section 3.0, ``SR [Surveillance Requirement] 
APPLICABILITY,'' specifically, SR 3.0.5. The amendments also modified 
Technical Specifications (TS) 3.5.2, ``ECCS [Emergency Core Cooling 
System]--Operating,'' TS 3.6.6,

[[Page 45550]]

``Containment Spray System,'' TS 3.7.5, ``Auxiliary Feedwater (AFW) 
System,'' TS 3.7.6, ``Component Cooling Water (CCW) System,'' TS 3.7.7, 
``Nuclear Service Water System (NSWS),'' TS 3.7.9, ``Control Room Area 
Ventilation System (CRAVS),'' TS 3.7.11, ``Auxiliary Building Filtered 
Ventilation Exhaust System (ABFVES),'' TS 3.8.1, ``AC [Alternating 
Current] Sources--Operating,'' and TS 3.8.4, ``DC [Direct Current] 
Sources--Operating'' to remove expired TS footnotes.
    Date of issuance: August 8, 2019.
    Effective date: These amendments are effective as of the date of 
issuance and shall be implemented within 120 days of issuance.
    Amendment Nos.: 316 (Unit 1) and 295 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML19184A585; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: The 
amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: April 23, 2019 (84 FR 
16893).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 8, 2019.
    No significant hazards consideration comments received: Yes. One 
comment from a member of the public was received, however it was not 
related to the no significant hazards consideration determination or 
the license amendment request.

Exelon Generation Company, LLC, Docket No. 50-220, Nine Mile Point 
Nuclear Station, Unit 1, Oswego County, New York

    Date of amendment request: June 26, 2018, as supplemented by 
letters dated February 25, 2019, May 17, 2019, and July 30, 2019. 
Publicly-available versions are in ADAMS under Accession Nos. 
ML18177A044, ML19056A387, ML19137A070, and ML19211C702, respectively.
    Brief description of amendment: The amendment revised Technical 
Specification 3.3.1, ``Oxygen Concentration,'' to require inerting the 
primary containment to less than four percent by volume oxygen 
concentration within 72 hours of entering power operating condition. 
Also, the amendment added a new requirement to identify required 
actions, if the primary containment oxygen concentration increases to 
greater than or equal to four volume percent while in the power 
operating condition.
    Date of issuance: July 30, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 237. A publicly-available version is in ADAMS under 
Accession No. ML19176A086; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-63: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    On December 18, 2018, the Nuclear Regulatory Commission (NRC or the 
Commission) staff published a proposed no significant hazards 
consideration (NSHC) determination in the Federal Register (83 FR 
64894) for the proposed amendment. Subsequently, by letters dated 
February 28, 2019, and May 17, 2019, the licensee provided additional 
information that expanded the scope of the amendment request as 
originally noticed in the Federal Register. Accordingly, the NRC 
published a second proposed NSHC determination in the Federal Register 
on June 18, 2019 (84 FR 28346), which superseded the original notice in 
its entirety. The supplemental letter dated July 30, 2019, provided 
additional information that clarified the application, did not expand 
the scope of the application as noticed, and did not change the staff's 
second proposed no significant hazards consideration determination as 
published in the Federal Register.
    Date of initial notice in Federal Register: December 18, 2018 (83 
FR 64894).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 30, 2019.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant (CNP), Unit 1, Berrien County, Michigan

    Date of amendment request: March 7, 2018.
    Brief description of amendment: The amendment approves the use of a 
leak-before-break methodology on designated reactor coolant system 
(RCS) piping segments associated with the CNP, Unit 1, accumulator, 
residual heat removal (RHR), and safety injection (SI) systems. The 
approved methodology provides the CNP, Unit 1, with additional design 
margin for future RCS piping analysis on these systems. The amendment 
also modifies technical specification 3.4.13, ``RCS Operational 
LEAKAGE,'' including adding requirements to meet the RCS operational 
leakage limits as specified in the technical specifications limiting 
conditions for operations 3.4.13.
    Date of issuance: August 1, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 346. A publicly-available version is in ADAMS under 
Accession No. ML19170A362; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-58: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: May 18, 2018 (83 FR 
20862).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 1, 2019.
    No significant hazards consideration comments received: No.

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

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

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

    Date of amendment request: May 29, 2018, as supplemented by letter 
dated March 26, 2019.
    Brief description of amendments: The amendments revised the 
Technical Specifications to include the provisions of Limited Condition 
for Operation (LCO) 3.0.6 in the Standard Technical Specifications. In 
support of this change, the licensee also added a new Safety Function 
Determination Program to the administrative section of the Technical 
Specification; added new notes and actions that direct entering the 
actions for the appropriate supported systems; made changes to LCO 
3.0.2 for Seabrook, St. Lucie, and Turkey Point; and made changes to 
LCO 3.0.1 for Seabrook and Turkey Point.
    Date of issuance: July 31, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos: 161 (Seabrook, Unit No. 1); 249 and 200 (St. Lucie, 
Unit Nos. 1 and 2); and 287 and 281 (Turkey Point, Unit Nos. 3 and 4). 
A publicly-available version is in ADAMS under Accession No. 
ML19148A744; documents related to these amendments are listed in the 
Safety Evaluation enclosed with the amendments.

[[Page 45551]]

    Renewed Facility Operating License Nos. NPF-86, DPR-67, NPF-16, 
DPR-31, and DPR-41: The amendments revised the Renewed Facility 
Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: September 11, 2018 (83 
FR 45985). The supplement dated March 26, 2019, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 31, 2019.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 12, 2018, as supplemented by 
letter dated April 18, 2019.
    Brief description of amendment: The amendment revised the technical 
specifications to delete the note associated with limiting condition 
for operation 3.5.1. The deleted note permitted low pressure coolant 
injection subsystems to be consider operable in certain plant 
conditions.
    Date of issuance: July 30, 2019.
    Effective date: As of the date of issuance and shall be implemented 
90 days of issuance.
    Amendment No.: 202. A publicly-available version is in ADAMS under 
Accession No. ML19162A093; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-22: Amendment revised 
the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 2, 2019 (84 FR 
24). The supplemental letter dated April 18, 2019 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 July 30, 2019.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 18, 2018, as supplemented by letters 
dated July 10, 2018, December 8, 2018, and April 8, 2019.
    Brief description of amendment: The amendments revised the approved 
fire protection program (FPP). Specifically, the amendments deleted 
several modifications which are required as part of PINGP's 
implementation of its risk-informed, performance-based FPP in 
accordance with 10 CFR paragraph 50.48(c), National Fire Protection 
Association Standard 805.
    Date of issuance: July 30, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 228-Unit 1; 216-Unit 2. A publicly-available 
version is in ADAMS under Accession No. ML19140A447; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-42 and DPR-60: The 
amendments revised the Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: August 14, 2018 (83 FR 
40350). The supplemental letters dated July 10, 2018, December 8, 2018, 
and April 8, 2019, 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 July 30, 2019.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 30, 2018.
    Brief description of amendment: The amendment revised Hope Creek 
Technical Specification 3.3.7.4, ``Remote Shutdown System 
Instrumentation and Controls,'' to make the requirements consistent 
with Standard Technical Specification 3.3.3.2, ``Remote Shutdown 
System,'' in NUREG-1433, Volume 1, Revision 4. The amendment increases 
the allowed outage time for inoperable remote shutdown system 
components from 7 days to 30 days. The amendment also deletes Tables 
3.3.7.4-1, 3.3.7.4-2, and 4.3.7.4-1, and relocates these tables to the 
Technical Requirements Manual, where they will be directly controlled 
by the licensee.
    Date of issuance: August 6, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of the date of issuance.
    Amendment No.: 217. A publicly-available version is in ADAMS under 
Accession No. ML19186A205; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-57: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: December 18, 2018 (83 
FR 64897).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 6, 2019.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 14, 2018.
    Brief description of amendments: The amendments revise a license 
condition associated with its approved fire protection program under 10 
CFR 50.48(c), ``National Fire Protection Association Standard (NFPA) 
805.'' Specifically, the plant operating licenses have been revised to 
allow, as a performance-based method, use of thermal insulation 
materials in limited applications subject to appropriate engineering 
reviews and controls, as a deviation from NFPA 805 Chapter 3, Section 
3.3, ``Prevention''.
    Date of issuance: July 30, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 224 (Unit 1) and 221 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML19156A262; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-2 and NPF-8: The 
amendments revised the Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: February 12, 2019 (84 
FR 3510).

[[Page 45552]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 30, 2019.
    No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit No. 1, 
Callaway County, Missouri

    Date of amendment request: September 4, 2018, as supplemented by 
letter dated February 20, 2019.
    Brief description of amendment: The amendment revised Emergency 
Action Levels CA6.1, ``Cold Shutdown/Refueling System Malfunction--
Hazardous event affecting a SAFETY SYSTEM needed for the current 
operating MODE: Alert,'' and SA9.1, ``System Malfunction--Hazardous 
event affecting a SAFETY SYSTEM needed for the current operating MODE: 
Alert.'' In addition, the amendment added a new definition for the term 
``Loss of Safety Function (LOSF)'' and re-definition of the term 
``Visible Damage'' and deleted Initiating Condition HG1 and associated 
EAL HG1.1, ``Hazard--HOSTILE ACTION resulting in loss of physical 
control of the facility: General Emergency,'' within the Callaway 
Plant, Unit No. 1 Radiological Emergency Response Plan.
    Date of issuance: July 30, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 220. A publicly-available version is in ADAMS under 
Accession No. ML19158A290; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-30: The amendment 
revised the Renewed Facility Operating License.
    Date of initial notice in Federal Register: December 4, 2018 (83 FR 
62621). The supplement dated February 20, 2019, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 30, 2019.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 23rd day of August, 2019.

    For the Nuclear Regulatory Commission.
Gregory F. Suber,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2019-18617 Filed 8-28-19; 8:45 am]
 BILLING CODE 7590-01-P