[Federal Register Volume 83, Number 118 (Tuesday, June 19, 2018)]
[Notices]
[Pages 28456-28467]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-12506]



[[Page 28456]]

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NUCLEAR REGULATORY COMMISSION

[NRC-2018-0114]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, 
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is 
publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued, and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from May 22, 2018, to June 4, 2018. The last 
biweekly notice was published on June 5, 2018.

DATES: Comments must be filed by July 19, 2018. A request for a hearing 
must be filed by August 20, 2018.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0114. Address 
questions about NRC dockets 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: May Ma, Office of Administration, Mail 
Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Janet Burkhardt, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-1384, email: [email protected].

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2018-0114, 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 http://www.regulations.gov and search for Docket ID NRC-2018-0114.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in this 
document
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2018-0114, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 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.

[[Page 28457]]

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 http://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 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

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[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 http://www.nrc.gov/site-help/e-submittals.html, by 
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m. 
and 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.
    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 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

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

    Date of amendment request: January 23, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18023A896.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3.6.4.1, ``Secondary Containment,'' 
Surveillance Requirement (SR) 3.6.4.1.2, for Brunswick Steam Electric 
Plant, Units 1 and 2. The proposed changes are based on Technical 
Specifications Task Force (TSTF) Traveler TSTF-551, Revision 3, 
``Revise Secondary Containment Surveillance Requirements'' (ADAMS 
Accession No. ML16277A226).
    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 addresses conditions during which Secondary 
Containment SR 3.6.4.1.2 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 change is no different than the 
consequences of an accident while utilizing the existing eight hour 
Completion Time for an inoperable Secondary Containment. As a 
result, the consequences of an accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not alter the protection system design, 
create new failure

[[Page 28459]]

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.
    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 addresses conditions during which Secondary 
Containment SR 3.6.4.1.2 is not met. The allowance for both an inner 
and outer Secondary Containment door to be open simultaneously for 
entry and exit does not affect the safety function of the Secondary 
Containment as the doors are promptly closed after entry or exit, 
thereby restoring the Secondary Containment boundary.
    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: Kathryn B. Nolan, Deputy General Counsel, 
550 South Tryon Street, Mail Code DEC45A, Charlotte, NC 28202.
    NRC Acting Branch Chief: Brian W. Tindell.

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

    Date of amendment request: January 23, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18023A899.
    Description of amendment request: The amendments would revise the 
Technical Specifications to adopt Technical Specifications Task Force 
(TSTF) Traveler TSTF-208, Revision 0, ``Extension of Time to Reach Mode 
2 in LCO [Limiting Condition for Operation] 3.0.3.''
    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 change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The time frame to take response action in accordance with LCO 
3.0.3 is not an initiating condition for any accident previously 
evaluated. The proposed change does not authorize the addition of 
any new plant equipment or systems, nor does it alter the 
assumptions of any accident analyses. The small increase in the time 
allowed to reach Mode 2 would not place the plant in any 
significantly increased probability of an accident occurring. The 
unit would already be preparing for a plant shutdown condition 
because of the 1 hour requirement to initiate shutdown actions. 
There is no change in the time period to reach Mode 3. The Mode 3 
Condition is the point at which the plant reactor core is no longer 
critical (i.e., Hot Shutdown).
    Therefore, since there is no change to the time period to reach 
the Hot Shutdown condition, the small change in the time to reach 
Mode 2 status does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to the allowed time to reach Mode 2 in LCO 
3.0.3 does not require any modification to the plant or change 
equipment operation. The proposed change will not introduce failure 
modes that could result in a new accident, and the change does not 
alter assumptions made in the safety analysis. The proposed change 
will not alter the design configuration, or method of operation of 
plant equipment beyond its normal functional capabilities. The 
proposed change does not create any new credible failure mechanisms, 
malfunctions, or accident initiators.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from those that have been 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to the allowed time to reach Mode 2 in LCO 
3.0.3 does not alter or exceed a design basis or safety limit. There 
is no change being made to safety analysis assumptions or the safety 
limits that would adversely affect plant safety as a result of the 
proposed change. Margins of safety are unaffected by the proposed 
change and the applicable requirements of 10 CFR 50.36(c)(2)(ii) and 
10 CFR 50, Appendix A will continue to be met.
    Therefore, the proposed change does not involve any reduction in 
a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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 B. Nolan, Deputy General Counsel, 
550 South Tryon Street, M/C DEC45A, Charlotte, NC 28202.
    NRC Acting Branch Chief: Brian W. Tindell.

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

    Date of amendment request: March 8, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18068A705.
    Description of amendment request: The amendment would update 
Section 15.4.3.1 of the Updated Final Safety Analysis Report for 
Waterford 3, which describes the dose consequence of the worst 
undetectable single fuel assembly misload. The updated analysis would 
reflect the use of Next Generation Fuel and integrated fuel burnable 
absorbers.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the fuel assembly misload event 
analysis. The analysis of the fuel assembly misload event showed 
that the total number of failed fuel rods is less than other 
Waterford 3 Condition 3 events that have already been demonstrated 
to meet the 10 CFR 50.67 acceptance criteria. For Waterford 3, the 
Excess Load with Loss of Alternating Current (LOAC) has this same 
release and fuel failure that has been shown to meet the offsite 
dose requirements. Since the worst undetectable misload has a fuel 
failure less than the excess load with LOAC event, the fuel assembly 
misload event is consistent with the Standard Review Plan 15.4.7 and 
meets the 10 CFR 50.67 requirements.
    This change is only analyzing the consequences of the fuel 
assembly misload event and no changes are being made that would 
impact the probability of the event occurring.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the fuel assembly misload event 
analysis. The proposed change does not involve a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or a change in the methods governing plant operations. 
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

[[Page 28460]]

consequences exceed the consequences of accidents previously 
analyzed.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the fuel assembly misload event 
analysis. The worst undetectable misloads have fuel failure less 
than the excess load with the Excess Load with Loss of Alternating 
Current (LOAC) event; the fuel assembly misload event meets the 10 
CFR 50.67 criteria and is consistent with the Standard Review Plan 
Section 15.4.7 guidance. The new analysis shows more adverse 
consequences than were shown in previous fuel assembly misload event 
analyses, but remains within the regulatory acceptance limits. Since 
the event remains within the 10 CFR 50.67 requirements and is 
bounded by the excess load with LOAC event, this is not a 
significant reduction in margin.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anna Vinson Jones, Senior Counsel, Entergy 
Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington, 
DC 20001.
    NRC Branch Chief: Robert J. Pascarelli.

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, Unit Nos. 2 and 3, Grundy County, Illinois

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

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

    Date of amendment request: April 25, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18116A133.
    Description of amendment request: The amendments would revise the 
technical specification (TS) requirements for inoperable snubbers for 
each facility. The amendments would also make other administrative 
changes to the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration for each site, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change allows a delay time before declaring 
supported Technical Specification (TS) systems inoperable when the 
associated snubber(s) cannot perform its required safety function. 
Entrance into Actions or delaying entrance into Actions is not an 
initiator of any accident previously evaluated. Consequently, the 
probability of an accident previously evaluated is not significantly 
increased. The consequences of an accident while relying on the 
delay time allowed before declaring a TS supported system inoperable 
and taking its Conditions and Required Actions are no different than 
the consequences of an accident under the same plant conditions 
while relying on the existing TS supported system Conditions and 
Required Actions. Therefore, the consequences of an accident 
previously evaluated are not significantly increased by this change. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change allows a delay time before declaring 
supported TS systems inoperable when the associated snubber(s) 
cannot perform its required safety function. The proposed change 
does not involve a physical alteration of the plant (no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change allows a delay time before declaring 
supported TS systems inoperable when the associated snubber(s) 
cannot perform its required safety function. The proposed change 
restores an allowance in the pre-Improved Standard Technical 
Specifications (ISTS) conversion TS that was unintentionally 
eliminated by the conversion. The pre-ISTS TS were considered to 
provide an adequate margin of safety for plant operation, as does 
the post-ISTS conversion TS. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis for each site 
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 Branch Chief: David J. Wrona.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: March 7, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18066A648.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 5.5.12, ``Primary Containment Leakage 
Rate Testing Program,'' to follow guidance developed by the Nuclear 
Energy Institute (NEI) in topical report NEI 94-01, ``Industry 
Guideline for Implementing Performance-Based Option of 10 CFR part 50, 
Appendix J,'' Revision 3-A, dated July 2012, with the conditions and 
limitations specified in NEI 94-01, Revision 2-A, dated October 2008. 
The proposed license amendment would also revise Technical 
Specification 5.5.12 by deleting two of the four listed exceptions to 
program guidelines.
    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 test interval extensions do not involve either a 
physical change to the plant or a change in the way the plant is 
operated or controlled. The containment is designed to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. As such, 
the containment and the testing requirements invoked to periodically 
demonstrate the integrity of the containment exist to ensure the 
plant's ability to mitigate the consequences of an accident, and do 
not involve the prevention or identification of any precursors of an 
accident. Therefore, the proposed extensions do not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The effect resulting from changing the Type A test frequency to 
1 per 15 years, measured as an increase to the total integrated 
plant risk for those accident sequences influenced by Type A 
testing, is

[[Page 28461]]

0.0318 person-rem/year. EPRI [Electric Power Research Institute] 
Report No. 1009325, Revision 2-A, states that a very small 
population dose is defined as an increase of less than or equal to 
1.0 person-rem per year or less than or equal to 1 percent of the 
total population dose, whichever is less restrictive for the risk 
impact assessment of the extended integrated leak rate test 
intervals. The results of the risk assessment calculation for the 
Type A test extension meet these criteria. The risk impact for the 
integrated leak rate test extension when compared to other severe 
accident risks is negligible.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as: (1) Activity based, 
and (2) time based. Activity based failure mechanisms are defined as 
degradation due to system and component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment combined with the containment inspections performed in 
accordance with [American Society for Mechanical Engineers Boiler 
and Pressure Vessel Code (ASME Code)], Section XI, and Technical 
Specification requirements serve to provide a high degree of 
assurance that the containment would not degrade in a manner that is 
detectable only by a Type A test. Based on the above, the proposed 
test interval extensions do not significantly increase the 
consequences of an accident previously evaluated.
    The proposed amendment also deletes two previously granted 
exceptions to Primary Containment Leakage Rate Testing Program 
guidelines. The exception regarding the performance of a Type A test 
no later than a specified date would be deleted as this Type A test 
has already been performed. Additionally, the exception to use the 
corrections to NEI 94-01, Revision 0, would be deleted as those 
corrections would no longer be in use. These changes to the 
exceptions in Technical Specification 5.5.12 are administrative in 
nature and do not affect the probability or consequences of an 
accident previously evaluated.
    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.
    Containment Type A and Type C testing requirements periodically 
demonstrate the integrity of the containment and exist to ensure the 
plant's ability to mitigate the consequences of an accident. These 
tests do not involve any accident precursors or initiators.
    The proposed change does not involve a physical modification to 
the plant (that is, no new or different type of equipment will be 
installed) nor does it alter the design, configuration, or change 
the manner in which the plant is operated or controlled beyond the 
standard functional capabilities of the equipment.
    The proposed amendment also deletes two previously granted 
exceptions. The exception regarding the performance of a Type A test 
no later than a specified date would be deleted as this Type A test 
has already been performed. Additionally, the exception to use the 
corrections to NEI 94-01, Revision 0, would be deleted as those 
corrections would no longer be in use. These changes to the 
exceptions in Technical Specification 5.5.12 are administrative in 
nature and do not create the possibility of a new or different kind 
of accident from any previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed license amendment does not alter the way safety 
limits, limiting safety system set points, or limiting conditions 
for operation are determined. The specific requirements and 
conditions of the Technical Specification Primary Containment 
Leakage Rate Testing Program exist to ensure that the degree of 
containment structural integrity and leak-tightness that is 
considered in the plant safety analysis is maintained. The overall 
containment leak rate limit specified by Technical Specifications is 
maintained. The design, operation, testing methods and acceptance 
criteria for Type A, B, and C containment leakage tests specified in 
applicable codes and standards would continue to be met, with the 
acceptance of this proposed amendment, since they are not affected 
by implementation of a performance-based containment testing 
program. This ensures that the margin of safety in the plant safety 
analysis is maintained.
    The proposed amendment also deletes two previously granted 
exceptions. The exception regarding the performance of a Type A test 
no later than a specified date would be deleted as this Type A test 
has already been performed. Additionally, the exception to use the 
corrections to NEI 94-01, Revision 0, would be deleted as those 
corrections would no longer be in use. These changes to the 
exceptions in Technical Specification 5.5.12 are administrative in 
nature and do not involve a significant reduction in a margin of 
safety.
    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: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: David J. Wrona.

NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant (PBNP), Unit Nos. 1 and 2, Town of Two Creeks, 
Manitowoc County, Wisconsin

    Date of amendment request: March 30, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18092A239.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 5.5.15, ``Containment Leakage Rate Testing 
Program,'' to require a program in accordance with Nuclear Energy 
Institute (NEI) topical report NEI 94-01, Revision 3-A, ``Industry 
Guideline for Implementing Performance-Based Option of 10 CFR part 50, 
Appendix J.'' This proposed change will allow extension of the Type A 
test interval up to one test in 15 years and extension of the Type C 
test interval up to 75 months, based on acceptable performance history 
as defined in NEI 94-01, Revision 3-A.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, ``Industry Guideline for Implementing 
Performance-Based Option of 10 CFR part 50, Appendix J,'' for 
development of the PBNP performance-based containment testing 
program. NEI 94-01 allows, based on risk and performance, an 
extension of Type A and Type C containment leak test intervals. 
Implementation of these guidelines continues to provide adequate 
assurance that during design basis accidents, the primary 
containment and its components will limit leakage rates to less than 
the values assumed in the plant safety analyses.
    The findings of the PBNP risk assessment confirm the general 
findings of previous studies that the risk impact with extending the 
containment leak rate is small. Per the guidance provided in 
Regulatory Guide 1.174, an extension of the leak test interval in 
accordance with NEI 94-01, Revision 3-A results in an estimated 
change within, the very small change region.
    Since the change is implementing a performance-based containment 
testing program, the proposed amendment does not involve either a 
physical change to the plant or a change in the manner in which the 
plant is operated or controlled. The requirement for containment 
leakage rate acceptance will not be changed by this amendment.

[[Page 28462]]

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William Blair, Managing Attorney--Nuclear, 
Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard, 
Juno Beach, FL 33408-0420.
    NRC Branch Chief: David J. Wrona.

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

    Date of amendment request: April 13, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18103A218.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.8.3.1, ``Distribution--Operating,'' to 
increase the alternating current (AC) inverters allowed outage time 
(AOT) from 24 hours to 7 days. The proposed change is based on 
application of the HCGS probabilistic risk assessment (PRA) in support 
of a risk-informed extension, and on additional considerations and 
compensatory actions.
    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 TS amendment does not affect the design of the AC 
inverters, the operational characteristics or function of the 
inverters, the interfaces between the inverters and other plant 
systems, or the reliability of the inverters. An inoperable AC 
inverter is not considered an initiator of an analyzed event. In 
addition, TS Actions and the associated Allowed Outage Times are not 
initiators of previously evaluated accidents. Extending the Allowed 
Outage Time for an inoperable AC inverter would not have a 
significant impact on the frequency of occurrence of an accident 
previously evaluated. The proposed amendment will not result in 
modifications to plant activities associated with inverter 
maintenance, but rather, provides operational flexibility by 
allowing additional time to perform inverter troubleshooting, 
corrective maintenance, and post-maintenance testing on-line.
    The proposed extension of the Completion Time for an inoperable 
AC inverter will not significantly affect the capability of the 
inverters to perform their safety function, which is to ensure an 
uninterruptible supply of 120-volt AC electrical power to the 
associated power distribution subsystems. An evaluation, using PRA 
methods, confirmed that the increase in plant risk associated with 
implementation of the proposed Allowed Outage Time extension is 
consistent with the NRC's Safety Goal Policy Statement, as further 
described in RG [Regulatory Guide] 1.174 and RG 1.177. In addition, 
a deterministic evaluation concluded that plant defense-in-depth 
philosophy will be maintained with the proposed Allowed Outage Time 
extension.
    There will be no impact on the source term or pathways assumed 
in accidents previously evaluated. No analysis assumptions will be 
changed and there will be no adverse effects on onsite or offsite 
doses as the result of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not involve physical alteration of 
the HCGS. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. There 
is no change being made to the parameters with in which the HCGS is 
operated. There are no setpoints at which protective or mitigating 
actions are initiated that are affected by this proposed action. The 
use of the alternate Class 1E power source for the AC distribution 
panel is consistent with the HCGS plant design. The change does not 
alter assumptions made in the safety analysis. This proposed action 
will not alter the manner in which equipment operation is initiated, 
nor will the functional demands on credited equipment be changed. No 
alteration is proposed to the procedures that ensure the HCGS 
remains with in analyzed limits, and no change is being made to 
procedures relied upon to respond to an off-normal event. As such, 
no new failure modes are being introduced.
    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.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident. These barriers include the fuel 
cladding, the reactor coolant system, and the containment system. 
The proposed change, which would increase the AOT from 24 hours to 7 
days for one inoperable inverter, does not exceed or alter a 
setpoint, design basis or safety limit.
    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.

[[Page 28463]]

    Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: James G. Danna.

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

    Date of amendment request: April 26, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18116A138.
    Description of amendment request: The requested amendment proposes 
changes to combined license (COL) Appendix C, with corresponding 
changes to the associated plant-specific Tier 1 information, and 
involves associated Tier 2 information in the Updated Final Safety 
Analysis Report (UFSAR) (which includes the plant-specific Design 
Control Document (DCD) Tier 2 information). Pursuant to the provisions 
of 10 CFR 52.63(b)(1), also requested is an exemption from elements of 
the design as certified in the 10 CFR part 52, appendix D, design 
certification rule for the plant-specific DCD departures.
    The requested amendment proposes changes to COL Appendix C (and 
plant-specific Tier 1) to reflect a new design of containment sump 
level sensors that affects the acceptance criterion for the detected 
containment sump level change test and the associated minimum 
detectable unidentified leakage rate in plant-specific DCD Tier 2 
information.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change is to the containment sump water level 
instrumentation and its expected [reactor coolant system (RCS)] 
leakage detection capability. The affected equipment is not safety-
related, but the containment sump water level sensors are 
seismically qualified. The change in containment sump level 
monitoring instruments has no adverse effect on the ability to 
detect a 0.5 [gallons per minute (gpm)] leak in containment, and 
therefore, has no adverse effect on design criteria for leak-before-
break. The change does not affect the operation of any systems or 
equipment that initiate an analyzed accident or alter any 
structures, systems, and components (SSC) accident initiator or 
initiating sequence of events.
    Because the containment sump water level monitoring channels are 
still capable of detecting a 0.5 gpm leak in containment, the change 
to the SSC has no effect on plant operations. There is no change to 
plant systems or the response of systems to postulated accident 
conditions. There is no change to the predicted radioactive releases 
due to normal operation or postulated accident conditions.
    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 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 to the 
containment sump water level instrumentation and its expected RCS 
leakage detection capability has no adverse effect on the ability to 
detect a 0.5 gpm leak in containment. The containment sump level 
instrumentation functions are unchanged and leak-before-break design 
criteria are not adversely affected.
    Loss of coolant accidents for a spectrum of pipe sizes and 
locations are already postulated in UFSAR Chapter 15, Section 15.6. 
Breaks in the main steam lines inside containment are also analyzed 
in UFSAR Chapter 15, Section 15.1. Unidentified leakage detection 
and operator action in response to unidentified leakage are not 
postulated for any of the design basis accident analyses described 
in UFSAR Chapter 15.
    Therefore, the requested 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 described change to the containment sump water level 
instrumentation and its expected RCS leakage detection capability is 
proposed to verify that the ability to detect a 0.5 gpm leak in 
containment is maintained. The proposed change does not alter any 
safety-related equipment, applicable design codes, code compliance, 
design function, or safety analysis. By ensuring that the chosen 
equipment can detect a 0.5 gpm leak in containment with the 
described accuracy, guidance in Regulatory Guide 1.45, Revision 0, 
as committed to in the UFSAR, and requirements in the Technical 
Specifications are met which ensures that leak-before-break design 
criteria are not adversely affected. Consequently, no safety 
analysis or design basis acceptance limit/criterion is challenged or 
exceeded by the proposed change, thus the margin of safety is not 
reduced.
    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 hazard 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, Unit Nos. 3 and 4, Burke County, 
Georgia

    Date of amendment request: April 27, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18117A464.
    Description of amendment request: The requested amendment proposes 
to depart from Tier 2 information in the Updated Final Safety Analysis 
Report (UFSAR) (which includes the plant-specific Design Control 
Document Tier 2 information) and involves related changes to plant-
specific Tier 1 information, with corresponding changes to the 
associated combined license (COL) Appendix C information. Specifically, 
the amendment, if approved, would revise the Tier 2 information in the 
UFSAR and related changes to Tier 1 and the associated COL Appendix C 
to remove the fire protection system non-safety related containment 
cable spray and install passive fire stops and radiant energy shields. 
The changes to Tier 1 require an exemption, which is included in the 
license amendment request.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not affect the operation or reliability 
of any system, structure or component (SSC) required to maintain a 
normal power operating condition or to mitigate anticipated 
transients without safety-related systems. Testing has demonstrated 
that the passive fire stops prevent propagation of fires along the 
length of cable trays and prevent the propagation of cable tray 
fires to adjacent fire zones. The proposed changes do not affect the 
operation of equipment whose failure could initiate an accident 
previously analyzed. The existence or failure of passive fire stops 
in fire zone 1100 AF 11300B does not affect normal equipment 
operation.

[[Page 28464]]

    The proposed changes do not adversely affect the reliability or 
function of an SSC relied upon to mitigate an accident previously 
analyzed. The existence or failure of passive fire stops in fire 
zone 1100 AF 11300B will not adversely affect passive core cooling 
system (PXS) performance during containment recirculation because 
the passive fire stops are located outside of the zone of influence 
(ZOI) of postulated high energy line breaks, and the passive fire 
stops' material-of-construction complies with in-containment 
refueling water storage tank (IRWST) and containment recirculation 
screens design criteria for debris generation and transport.
    The existing active open nozzle cable tray suppression system is 
not fully automatic, is nonsafety-related, and is not credited in 
the probabilistic risk assessment (PRA). Therefore, replacing the 
active open nozzle cable tray suppression system with passive fire 
stops does not have an impact on PRA calculations and results.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not affect the operation of systems or 
equipment that could 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 use of passive fire stops is 
recognized by Regulatory Guide 1.189. The passive fire stops in 
nonsafety-related open cable trays are more reliable than active 
systems such as the current open nozzle cable tray suppression 
system because they require no mechanical or human action to perform 
their protective function. When protection is required, there is no 
delay for operator or mechanical response. Testing has demonstrated 
that the passive fire stops prevent propagation of fires along the 
length of cable trays and prevent the propagation of cable tray 
fires to adjacent fire zones.
    The existence or failure of passive fire stops in fire zone 1100 
AF 11300B will not adversely affect passive core cooling system 
(PXS) performance during containment recirculation because the 
passive fire stops are located outside of the zone of influence 
(ZOI) of postulated high energy line breaks, and their material-of-
construction complies with in-containment refueling water storage 
tank (IRWST) and containment recirculation screens design criteria 
for debris generation and transport.
    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 do not affect existing safety margins. The 
current open nozzle cable tray suppression system is nonsafety-
related. The use of passive fire stops is recognized by Regulatory 
Guide 1.189. The passive fire stops in nonsafety-related open cable 
trays are more reliable than active systems such as the current open 
nozzle cable tray suppression system because they require no 
mechanical or human action to perform their protective function. 
When protection is required, there is no delay for operator or 
mechanical response. Testing has demonstrated that the passive fire 
stops prevent propagation of fires along the length of cable trays 
and prevent the propagation of cable tray fires to adjacent fire 
zones.
    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 hazard 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, Unit Nos. 3 and 4, Burke County, 
Georgia

    Date of amendment request: April 27, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18117A464.
    Description of amendment request: The requested amendment proposes 
to depart from Tier 2 information in the Updated Final Safety Analysis 
Report (UFSAR) (which includes the plant-specific Design Control 
Document Tier 2 information) and involves related changes to plant-
specific Tier 1 information, with corresponding changes to the 
associated combined license (COL) Appendix C information. Specifically, 
the amendment, if approved, would revise the Tier 2 information in the 
UFSAR and related changes to Tier 1 and the associated COL Appendix C 
to remove the fire protection system non-safety related containment 
cable spray and install passive fire stops and radiant energy shields. 
The changes to Tier 1 require an exemption, which is included in the 
license amendment request.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that initiate an analyzed accident or alter any 
structures, systems, and components (SSC) accident initiator or 
initiating sequence of events.
    The proposed changes do not affect the physical design and 
operation of the Passive Residual Heat Removal Heat Exchanger (PRHR 
HX) or In-containment Refueling Water Storage Tank (IRWST) as 
described in the Updated Final Safety Analysis Report (UFSAR). The 
proposed changes do not affect the probability of inadvertent 
operation or failure. Therefore, the probabilities of the accidents 
previously evaluated in the UFSAR are not affected.
    The proposed changes do not affect the ability of the PRHR HX 
and IRWST to perform their design functions. The designs of the PRHR 
HX and IRWST continue to meet the same regulatory acceptance 
criteria, codes, and standards as required by the UFSAR. In 
addition, the proposed changes maintain the capabilities of the PRHR 
HX and IRWST to mitigate the consequences of an accident and to meet 
the applicable regulatory acceptance criteria.
    The proposed changes do not affect the prevention and mitigation 
of other abnormal events (e.g. anticipated operational occurrences, 
earthquakes, floods and turbine missiles), or their safety or design 
analyses. Therefore, the consequences of the accidents evaluated in 
the UFSAR are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that may initiate a new or different kind of accident, 
or alter any SSC such that a new accident initiator or initiating 
sequence of events is created.
    The proposed changes do not affect any other SSC design 
functions or methods of operation in a manner that results in a new 
failure mode, malfunction, or sequence of events that affect safety-
related or nonsafety related equipment. Therefore, this activity 
does not allow for a new fission product release path, result in a 
new fission product barrier failure mode, or create a new sequence 
of events that result in significant fuel cladding failures.
    Therefore, the requested amendment does not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes maintain existing safety margins. The 
proposed changes verify and maintain the capabilities of the PRHR HX 
and IRWST to perform their design functions. Therefore, the proposed 
changes

[[Page 28465]]

satisfy the same design functions in accordance with the same codes 
and standards as stated in the UFSAR. These changes do not affect 
any design code, function, design analysis, safety analysis input or 
result, or design/safety margin.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes, and no margin of 
safety is reduced.
    Therefore, the requested amendment does not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: March 2, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18075A021.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TSs) consistent with Revision 0 to the 
Technical Specifications Task Force (TSTF) Standard Technical 
Specification Change Document TSTF-490, ``Deletion of E Bar Definition 
and Revision to RCS Specific Activity Tech Spec.'' The proposed 
amendments would adopt TSTF-490 and make the following associated 
changes: (1) Adoption of a TS change to replace the current limits on 
primary coolant gross specific activity with limits on primary coolant 
noble gas activity, and (2) an update of the Alternative Source Term 
(AST) analyses for Surry.
    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:

Criterion 1. The Proposed Changes Do Not Involve a Significant Increase 
in the Probability or Consequences of an Accident Previously Evaluated

    Reactor coolant specific activity is not an initiator for any 
accident previously evaluated, and the allowed time period when 
primary coolant gross activity is not within limits is not an 
initiator for any accident previously evaluated. In addition, the 
current variable limit on primary coolant iodine concentration is 
not an initiator to any accident previously evaluated. Updating the 
Alternative Source Term analyses does not require any changes to any 
plant structures, systems, or components (SSCs) and therefore does 
not affect any accident initiators. As a result, the proposed 
changes do not significantly increase the probability of an 
accident. The proposed TS change will limit primary coolant noble 
gases to concentrations consistent with the accident analyses, and 
the proposed completion time when the limit may be exceeded has no 
impact on the consequences of any design basis accident since the 
consequences of an accident during this time period is the same as 
the consequences of an accident during the existing time periods. 
The revised assessments of the radiological consequences due to 
design basis accidents listed in the Surry Updated Final Safety 
Analysis Report, using the updated AST methodology and proposed 
assumptions and inputs, conclude that the Exclusion Area Boundary 
(EAB), Low Population Zone (LPZ), and Control Room doses are within 
the limits of 10 CFR 50.67 and within the limits of Regulatory Guide 
(RG) 1.183. As a result, the consequences of any accident previously 
evaluated are not significantly increased.

Criterion 2. The Proposed Changes Do Not Create the Possibility of a 
New or Different Kind of Accident From Any Accident Previously 
Evaluated

    The proposed TS change in specific activity limits and the 
updated AST dose consequences analyses do not alter any physical 
part of the plant, (i.e., no new or different type of equipment will 
be installed,) nor do they affect any plant operating parameter or 
create new accident precursors. Therefore, the proposed changes do 
not create the potential for a new or different kind of accident 
from any previously calculated.

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

    The proposed TS change in specific activity limits is consistent 
with the assumptions in -the safety analyses and will ensure the 
monitored values protect the initial assumptions in the safety 
analyses. The proposed changes for radiological events related to 
the computer code used to calculate dose, revised X/Qs for control 
room and offsite receptors (including the computer code and method 
used to determine control room X/Qs for SG releases), the computer 
code used to determine core inventory, the change in FHA [Fuel 
Handling Accident] gap fraction methodology, and removing the LRA 
[Locked Rotor Accident] from the radiological design basis have been 
analyzed and result in acceptable consequences, meeting the criteria 
as specified in 10 CFR 50.67 and RG 1.183. The proposed changes will 
not result in plant operation in a configuration outside the 
analyses or design basis and do not adversely affect systems that 
are required to respond for safe shutdown of the plant and to 
maintain the plant in a safe operating condition. Therefore, the 
changes do not involve a significant reduction in a margin of 
safety.

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

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

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

[[Page 28466]]

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

    Date of amendment request: August 14, 2017.
    Brief description of amendment: The amendment modified Fermi 2 
Technical Specification 5.5.7, ``Ventilation Filter Testing Program 
(VFTP),'' by adopting the format and language of NUREG-1433, ``Standard 
Technical Specifications for General Electric BWR/4 Plants,'' Revision 
4.
    Date of issuance: May 24, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 208. A publicly-available version is in ADAMS under 
Accession No. ML18108A022; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-43: The amendment 
revised the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: September 26, 2017 (82 
FR 44851).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 24, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 29, 2017, as supplemented by 
letters dated January 4, 2018, and January 23, 2018.
    Brief description of amendments: The amendments adopted Technical 
Specifications Task Force (TSTF) Traveler TSTF-542, Revision 2, 
``Reactor Pressure Vessel Water Inventory Control,'' for Brunswick 
Steam Electric Plant, Units 1 and 2. The amendments replaced existing 
technical specification (TS) requirements associated with ``operations 
with the potential for draining the reactor vessel,'' with revised TSs 
providing alternative requirements for reactor pressure vessel water 
inventory control. These alternative requirements protect Safety Limit 
2.1.1.3, which states, ``Reactor vessel water level shall be greater 
than the top of active irradiated fuel.''
    Date of issuance: April 13, 2018.
    Effective date: As of the date of issuance and shall be implemented 
prior to the 2019 Unit 2 refueling outage. This Notice of Issuance 
corrects the effective date of License Amendment No. 283, originally 
noticed in the Federal Register on May 8, 2018 (83 FR 20865).
    Amendment Nos.: 283 (Unit 1) and 311 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML18039A444; documents related 
to this amendment are listed in the Safety Evaluation enclosed with the 
amendments. Amendment Nos. 283 and 311 were corrected by letter dated 
May 23, 2018 (ADAMS Accession No. ML18137A143).
    Renewed Facility Operating License No. DPR-49: The amendments 
revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: September 12, 2017 (82 
FR 42846). The supplemental letters dated January 4, 2018, and January 
23, 2018, 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 April 13, 2018.
    No significant hazards consideration comments received: No.

Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: April 3, 2017, as supplemented by 
letters dated April 3, 2017; May 2, 2017; September 28, 2017; and 
January 8, 2018.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to extend the required frequency of certain 18-
month Surveillance Requirements to 24 months to accommodate a 24-month 
refueling cycle. In addition, the amendment revised certain programs in 
TS Section 5.5, ``Programs and Manuals,'' to change 18-month 
frequencies to 24 months.
    Date of issuance: May 25, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the end of the next refueling outage.
    Amendment No.: 258. A publicly-available version is in ADAMS under 
Accession No. ML18115A150; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-23: The amendment 
revised the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: July 5, 2017 (82 FR 
31092). The supplemental letters dated September 28, 2017, and January 
8, 2018, 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 May 25, 2018.
    No significant hazards consideration comments received: No.

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

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

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

Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit No. 2 (Nine Mile), Oswego County, New York

    Date of amendment request: November 8, 2017.
    Brief description of amendments: The amendments revised the 
technical specification requirements for secondary containment.
    Date of issuance: May 29, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: Clinton--218; LaSalle, Units 1 and 2--228 and 214; 
Limerick, Units 1 and 2--229 and 192; and Nine Mile--169. A publicly-
available version is in ADAMS under Accession No. ML18113A045. 
Documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Facility Operating License Nos. NPF-62, NPF-11, NPF-18, NPF-39, 
NPF-85, and NPF-69: The amendments revised the Facility Operating 
Licenses and Technical Specifications.
    Date of initial notice in Federal Register: December 19, 2017 (82 
FR 60227).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 29, 2018.
    No significant hazards consideration comments received: No.

[[Page 28467]]

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

    Date of amendment request: April 20, 2017, as supplemented by 
letters dated September 14, 2017; February 19, 2018; and May 1, 2018.
    Brief description of amendments: The amendments revised the 
Technical Specifications by replacing the existing requirements related 
to ``operations with a potential for draining the reactor vessel'' with 
new requirements on Reactor Pressure Vessel Water Inventory Control to 
protect Safety Limit 2.1.1.3, which requires reactor vessel water level 
to be greater than the top of active irradiated fuel.
    Date of issuance: May 31, 2018.
    Effective date: As of the date of issuance and shall be implemented 
prior to the commencement of the Unit No. 2 refueling outage (U2R25) in 
February 2019.
    Amendment Nos.: Unit 1--290, Unit 2--235. A publicly-available 
version is in ADAMS under Accession No. ML18123A368; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: The 
amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: August 29, 2017 (82 FR 
41071). The supplemental letters dated September 14, 2017; February 19, 
2018; and May 1, 2018, 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 May 31, 2018.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority (TVA) Docket Nos. 50-259, 50-260, 50-296, 
and 72-052, Browns Ferry Nuclear Plant, Unit Nos. 1, 2, and 3, 
Limestone County, Alabama

TVA Docket Nos. 50-327, 50-328, and 72-034, Sequoyah Nuclear Plant, 
Unit Nos. 1 and 2, Hamilton County, Tennessee

TVA Docket Nos. 50-390, 50-391, and 72-1048, Watts Bar Nuclear Plant, 
Unit Nos. 1 and 2, Rhea County, Tennessee

    Date of amendment request: January 4, 2017, as supplemented by 
letters dated July 7, 2017, and July 27, 2017. (Note: This Notice of 
Issuance corrects the amendments by adding the supplement dated July 
27, 2017, which was inadvertently omitted from the original Federal 
Register notice (January 16, 2018; 83 FR 2234).
    Brief description of amendments: The amendments revised TVA 
Emergency Plans for the above nuclear plants. Specifically, the 
amendments adopted the NRC-endorsed Radiological Emergency Plan 
Emergency Action Level schemes developed by the Nuclear Energy 
Institute (NEI 99-01, Revision 6, ``Development of Emergency Action 
Levels for Non-Passive Reactors'').
    Date of issuance: December 22, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days from the date of its issuance, or July 3, 2018, 
whichever comes later.
    Amendment Nos.: Browns Ferry Nuclear Plant--303 (Unit 1), 327 (Unit 
2), and 287 (Unit 3); Sequoyah Nuclear Plant--339 (Unit 1) and 332 
(Unit 2); and Watts Bar Nuclear Plant--118 (Unit 1) and 18 (Unit 2). A 
publicly-available version is in ADAMS under Accession No. ML17289A032; 
documents related to these amendments are listed in the Safety 
Evaluations enclosed with the amendments. These amendments were 
corrected by letter dated May 29, 2018 (ADAMS Accession No. 
ML18138A452).
    Renewed Facility Operating License Nos. DPR-33, DPR-52, DPR-68, 
DPR-77, and DPR-79, and Facility Operating License Nos, NPF-90 and NPF-
96: The amendments revised the licenses.
    Date of initial notice in Federal Register: June 19, 2017 (82 FR 
27891). The supplemental letters dated July 7, 2017, and July 27, 2017, 
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 December 22, 2017.
    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: April 6, 2017, as supplemented by letter 
dated February 5, 2018.
    Brief description of amendment: The amendment revised the Final 
Safety Analysis Report to clearly describe conformance with NRC 
Regulatory Guide 1.106, Revision 1, ``Thermal Overload Protection for 
Electric Motors on Motor-Operated Valves.''
    Date of issuance: May 30, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 218. A publicly-available version is in ADAMS under 
Accession No. ML18124A026; 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 Final Safety Analysis Report.
    Date of initial notice in Federal Register: July 18, 2017 (82 FR 
32885). The supplemental letter dated February 5, 2018, 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 May 30, 2018.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 6th day of June 2018.

    For the Nuclear Regulatory Commission.
Tara Inverso,
Acting Deputy Director, Division of Operating Reactor Licensing, Office 
of Nuclear Reactor Regulation.
[FR Doc. 2018-12506 Filed 6-18-18; 8:45 am]
 BILLING CODE 7590-01-P