Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 55401-55421 [2017-25063]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 82, Number 223 (Tuesday, November 21, 2017)]
[Notices]
[Pages 55401-55421]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-25063]


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NUCLEAR REGULATORY COMMISSION

[NRC-2017-0220]


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 October 24, 2017 to November 6, 2017. The 
last biweekly notice was published on November 7, 2017.

DATES: Comments must be filed by December 21, 2017. A request for a 
hearing must be filed by January 22, 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 Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0220. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. 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: OWFN-2-A13, 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: Shirley Rohrer, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-5411, email: Shirley.Rohrer@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2017-0220, facility name, unit 
number(s), plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0220.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in 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-2017-0220, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at https://www.regulations.gov as well as enter 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 Sec.  50.92 of title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment 
period such that failure to act in a timely way would result, for 
example in derating or shutdown of the facility. If the Commission 
takes action prior to the expiration of either the comment period or 
the notice period, it will publish in the Federal Register a notice of

[[Page 55402]]

issuance. If the Commission makes a final no significant hazards 
consideration determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.

A. Opportunity To Request a Hearing and Petition for Leave To Intervene

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and petition for leave to intervene 
(petition) with respect to the action. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested persons should consult a 
current copy of 10 CFR 2.309. The NRC's regulations are accessible 
electronically from the NRC Library on the NRC's Web site at https://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of 
the regulations is available at the NRC's Public Document Room, located 
at One White Flint North, Room O1-F21, 11555 Rockville Pike (First 
Floor), Rockville, Maryland 20852. If a petition is filed, the 
Commission or a presiding officer will rule on the petition and, if 
appropriate, a notice of a hearing will be issued.
    As required by 10 CFR 2.309(d) the petition should specifically 
explain the reasons why intervention should be permitted with 
particular reference to the following general requirements for 
standing: (1) The name, address, and telephone number of the 
petitioner; (2) the nature of the petitioner's right under the Act to 
be made a party to the proceeding; (3) the nature and extent of the 
petitioner's property, financial, or other interest in the proceeding; 
and (4) the possible effect of any decision or order which may be 
entered in the proceeding on the petitioner's interest.
    In accordance with 10 CFR 2.309(f), the petition must also set 
forth the specific contentions which the petitioner seeks to have 
litigated in the proceeding. Each contention must consist of a specific 
statement of the issue of law or fact to be raised or controverted. In 
addition, the petitioner must provide a brief explanation of the bases 
for the contention and a concise statement of the alleged facts or 
expert opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to the specific sources and 
documents on which the petitioner intends to rely to support its 
position on the issue. The petition must include sufficient information 
to show that a genuine dispute exists with the applicant or licensee on 
a material issue of law or fact. Contentions must be limited to matters 
within the scope of the proceeding. The contention must be one which, 
if proven, would entitle the petitioner to relief. A petitioner who 
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene. 
Parties have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that party's admitted 
contentions, including the opportunity to present evidence, consistent 
with the NRC's regulations, policies, and procedures.
    Petitions must be filed no later than 60 days from the date of 
publication of this notice. Petitions and motions for leave to file new 
or amended contentions that are filed after the deadline will not be 
entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in 
accordance with the filing instructions in the ``Electronic Submissions 
(E-Filing)'' section of this document.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to establish when the hearing is held. If the final determination is 
that the amendment request involves no significant hazards 
consideration, the 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 Web site at https://www.nrc.gov/site-help/e-submittals.html. Participants may not submit 
paper copies of their filings unless they seek an exemption in

[[Page 55403]]

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 hearing.docket@nrc.gov, 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 Web site at https://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 Web site at https://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 Web site at https://www.nrc.gov/site-help/e-submittals.html, by 
email to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-
7640. The NRC 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.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant (PNP), Van Buren County, Michigan
    Date of amendment request: August 31, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17248A389.
    Description of amendment request: The proposed amendment would 
revise the PNP Site Emergency Plan (SEP) for the permanently shut down 
and defueled condition. The proposed PNP SEP changes would revise the 
shift staffing and Emergency Response Organization (ERO) staffing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the PNP SEP do not impact the function 
of plant structures, systems, or components (SSCs). The proposed 
changes do not affect accident initiators or precursors, nor does it 
alter design assumptions. The proposed changes do not prevent the 
ability of the on-shift staff and augmented ERO to perform their 
intended functions to mitigate the consequences of any accident or 
event that will be credible in the permanently shut down and 
defueled condition. The proposed changes only remove positions that 
will no longer be credited in the PNP SEP.
    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?

[[Page 55404]]

    Response: No.
    The proposed changes reduce the number of on-shift and augmented 
ERO positions commensurate with the hazards associated with a 
permanently shut down and defueled facility. The proposed changes do 
not involve installation of new equipment or modification of 
existing equipment, so that no new equipment failure modes are 
introduced. Also, the proposed changes do not result in a change to 
the way that the equipment or facility is operated so that no new 
accident initiators are created.
    Therefore, the proposed 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.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed changes are 
associated with the PNP SEP and do not impact operation of the plant 
or its response to transients or accidents. The change does not 
affect the Technical Specifications. The proposed changes do not 
involve a change in the method of plant operation, and no accident 
analyses will be affected by the proposed changes. Safety analysis 
acceptance criteria are not affected by the proposed changes. The 
revised PNP SEP will continue to provide the necessary response 
staff with the proposed changes.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William Dennis, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 
10601.
    NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company (EGC), LLC, Docket Nos. STN 50-456 and STN 
50-457, Braidwood Station, Units 1 and 2, Will County, Illinois and 
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 
2, Ogle County, Illinois
    Date of amendment request: September 1, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17244A093.
    Description of amendment request: The amendments would modify the 
licensing basis by the addition of a license condition to allow for the 
implementation of the provisions of 10 CFR, Section 50.69, ``Risk-
informed categorization and treatment of structures, systems and 
components for nuclear power reactors.''
    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 will permit the use of a risk-informed 
categorization process to modify the scope of SSCs [structures, 
systems, and components] subject to NRC [Nuclear Regulatory 
Commission] special treatment requirements and to implement 
alternative treatments per the regulations. The process used to 
evaluate SSCs for changes to NRC special treatment requirements and 
the use of alternative requirements ensures the ability of the SSCs 
to perform their design function. The potential change to special 
treatment requirements does not change the design and operation of 
the SSCs. As a result, the proposed change does not significantly 
affect any initiators to accidents previously evaluated or the 
ability to mitigate any accidents previously evaluated. The 
consequences of the accidents previously evaluated are not affected 
because the mitigation functions performed by the SSCs assumed in 
the safety analysis are not being modified. The SSCs required to 
safely shut down the reactor and maintain it in a safe shutdown 
condition following an accident will continue to perform their 
design functions.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    The proposed change will permit the use of a risk-informed 
categorization process to modify the scope of SSCs subject to NRC 
special treatment requirements and to implement alternative 
treatments per the regulations. The proposed change does not change 
the functional requirements, configuration, or method of operation 
of any SSC. Under the proposed change, no additional plant equipment 
will be installed.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change will permit the use of a risk-informed 
categorization process to modify the scope of SSCs subject to NRC 
special treatment requirements and to implement alternative 
treatments per the regulations. The proposed change does not affect 
any Safety Limits or operating parameters used to establish the 
safety margin. The safety margins included in analyses of accidents 
are not affected by the proposed change.
    The regulation requires that there be no significant effect on 
plant risk due to any change to the special treatment requirements 
for SSCs and that the SSCs continue to be capable of performing 
their design basis functions, as well as to perform any beyond 
design basis functions consistent with the categorization process 
and results.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: David J. Wrona.
Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 
and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and 
Lancaster Counties, Pennsylvania
    Date of amendment request: August 30, 2017, as supplemented by 
letter dated October 24, 2017. Publicly-available versions are in ADAMS 
under Accession Nos. ML17243A014 and ML17297B521, respectively.
    Description of amendment request: The amendments would modify the 
licensing basis by the addition of a license condition to allow for the 
implementation of the provisions of 10 CFR 50.69, ``Risk-informed 
categorization and treatment of structures, systems and components for 
nuclear power reactors.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with NRC staff edits shown in 
square brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will permit the use of a risk-informed 
categorization process to modify the scope of [structures, systems, 
and components] SSCs subject to NRC special treatment requirements 
and to implement alternative treatments per the regulations. The 
process used to evaluate SSCs for changes to NRC special treatment

[[Page 55405]]

requirements and the use of alternative requirements ensures the 
ability of the SSCs to perform their design function. The potential 
change to special treatment requirements does not change the design 
and operation of the SSCs. As a result, the proposed change does not 
significantly affect any initiators to accidents previously 
evaluated or the ability to mitigate any accidents previously 
evaluated. The consequences of the accidents previously evaluated 
are not affected because the mitigation functions performed by the 
SSCs assumed in the safety analysis are not being modified. The SSCs 
required to safely shut down the reactor and maintain it in a safe 
shutdown condition following an accident will continue to perform 
their design functions.
    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 will permit the use of a risk-informed 
categorization process to modify the scope of SSCs subject to NRC 
special treatment requirements and to implement alternative 
treatments per the regulations. The proposed change does not change 
the functional requirements, configuration, or method of operation 
of any SSC. Under the proposed change, no additional plant equipment 
will be installed.
    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 will permit the use of a risk-informed 
categorization process to modify the scope of SSCs subject to NRC 
special treatment requirements and to implement alternative 
treatments per the regulations. The proposed change does not affect 
any Safety Limits or operating parameters used to establish the 
safety margin. The safety margins included in analyses of accidents 
are not affected by the proposed change. The regulation requires 
that there be no significant effect on plant risk due to any change 
to the special treatment requirements for SSCs and that the SSCs 
continue to be capable of performing their design basis functions, 
as well as to perform any beyond design basis functions consistent 
with the categorization process and results.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL 
60555.
    NRC Branch Chief: James G. Danna.
Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 
and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and 
Lancaster Counties, Pennsylvania
    Date of amendment request: September 29, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17275A069.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) requirements related to the direct current 
(DC) electrical power system. The proposed changes are based on 
Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 
2, ``DC Electrical Rewrite--Update to TSTF-360.''
    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 restructures the TS for the direct current 
(DC) electrical power system. The proposed changes add actions to 
specifically address battery charger inoperability. The DC 
electrical power system, including associated battery chargers, is 
not an initiator of any accident sequence analyzed in the Updated 
Final Safety Analysis Report (UFSAR). Operation in accordance with 
the proposed TS ensures that the DC electrical power system is 
capable of performing its function as described in the UFSAR. 
Therefore, the mitigative functions supported by the DC electrical 
power system will continue to provide the protection assumed by the 
analysis, and the probability of previously analyzed accidents will 
not increase by implementing these changes.
    The relocation of preventive maintenance surveillances, and 
certain operating limits and actions, to a newly created licensee-
controlled Battery Monitoring and Maintenance Program will not 
challenge the ability of the DC electrical power system to perform 
its design function. Appropriate monitoring and maintenance, 
consistent with industry standards, will continue to be performed. 
In addition, the DC electrical power system is within the scope of 
10 CFR 50.65, ``Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants,'' which will ensure the control 
of maintenance activities associated with the DC electrical power 
system.
    The integrity of fission product barriers, plant configuration, 
and operating procedures as described in the UFSAR will not be 
affected by the proposed changes. Therefore, the consequences of 
previously analyzed accidents will not increase by implementing 
these changes.
    Therefore, the proposed 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 involves restructuring the TS for the DC 
electrical power system. The DC electrical power system, including 
associated battery chargers, is not an initiator to any accident 
sequence analyzed in the UFSAR. Rather, the DC electrical power 
system is used to supply equipment used to mitigate an accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed changes will not adversely affect 
operation of plant equipment. These changes will not result in a 
change to the setpoints at which protective actions are initiated. 
Sufficient DC capacity to support operation of mitigation equipment 
is ensured. The changes associated with the new battery maintenance 
and monitoring program will ensure that the station batteries are 
maintained in a highly reliable manner. The equipment fed by the DC 
electrical sources will continue to provide adequate power to safety 
related loads in accordance with analysis assumptions.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL 
60555.
    NRC Branch Chief: James G. Danna.
Exelon Generation Company, LLC, Docket No. 50-333, James A. FitzPatrick 
Nuclear Power Plant, Oswego County, New York
    Date of amendment request: October 2, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17275A520.

[[Page 55406]]

    Description of amendment request: The amendment would revise the 
James A. FitzPatrick Nuclear Power Plant Technical Specifications (TSs) 
to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-542, 
Revision 2, ``Reactor Pressure Vessel Water Inventory Control'' (ADAMS 
Accession No. ML16074A448). Specifically, the licensee proposed changes 
to replace TS requirements related to operations with a potential for 
draining the reactor vessel (OPDRVs) with new requirements on reactor 
pressure vessel (RPV) water inventory control (WIC) to protect Safety 
Limit 2.1.1.3.
    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 proposed changes replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold 
shutdown) and Mode 5 (i.e., refueling) is not an accident previously 
evaluated, and therefore replacing the existing TS controls to 
prevent or mitigate such an event with a new set of controls has no 
effect on any accident previously evaluated. RPV water inventory 
control in Mode 4 or Mode 5 is not an initiator of any accident 
previously evaluated. The existing OPDRV controls or the proposed 
RPV WIC controls are not mitigating actions assumed in any accident 
previously evaluated.
    The proposed changes reduce the probability of an unexpected 
draining event (which is not a previously evaluated accident) by 
imposing new requirements on the limiting time in which an 
unexpected draining event could result in the reactor vessel water 
level dropping to the top of the active fuel (TAF). These controls 
require cognizance of the plant configuration and control of 
configurations with unacceptably short drain times. These 
requirements reduce the probability of an unexpected draining event. 
The current TS requirements are only mitigating actions and impose 
no requirements that reduce the probability of an unexpected 
draining event.
    The proposed changes reduce the consequences of an unexpected 
draining event (which is not a previously evaluated accident) by 
requiring an Emergency Core Cooling System (ECCS) subsystem to be 
operable at all times in Modes 4 and 5. The current TS requirements 
do not require any water injection systems, ECCS or otherwise, to be 
Operable in certain conditions in Mode 5. The change in requirement 
from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does 
not significantly affect the consequences of an unexpected draining 
event because the proposed Actions ensure equipment is available 
within the limiting drain time that is as capable of mitigating the 
event as the current requirements. The proposed controls provide 
escalating compensatory measures to be established as calculated 
drain times decrease, such as verification of a second method of 
water injection and additional confirmations that containment and/or 
filtration would be available if needed.
    The proposed changes reduce or eliminate some requirements that 
were determined to be unnecessary to manage the consequences of an 
unexpected draining event, such as automatic initiation of an ECCS 
subsystem and control room ventilation. These changes do not affect 
the consequences of any accident previously evaluated since a 
draining event in Modes 4 and 5 is not a previously evaluated 
accident and the requirements are not needed to adequately respond 
to a draining event.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.1.3. The proposed changes will not alter the design 
function of the equipment involved. Under the proposed changes, some 
systems that are currently required to be operable during OPDRVs 
would be required to be available within the limiting drain time or 
to be in service depending on the limiting drain time. Should those 
systems be unable to be placed into service, the consequences are no 
different than if those systems were unable to perform their 
function under the current TS requirements.
    The event of concern under the current requirements and the 
proposed changes are an unexpected draining event. The proposed 
changes do not create new failure mechanisms, malfunctions, or 
accident initiators that would cause a draining event or a new or 
different kind of accident not previously evaluated or included in 
the design and licensing bases.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC. The current requirements do 
not have a stated safety basis and no margin of safety is 
established in the licensing basis. The safety basis for the new 
requirements is to protect Safety Limit 2.1.1.3. New requirements 
are added to determine the limiting time in which the RPV water 
inventory could drain to the top of the fuel in the reactor vessel 
should an unexpected draining event occur. Plant configurations that 
could result in lowering the RPV water level to the TAF within one 
hour are now prohibited. New escalating compensatory measures based 
on the limiting drain time replace the current controls. The 
proposed TS establish a safety margin by providing defense-in-depth 
to ensure that the Safety Limit is protected and to protect the 
public health and safety. While some less restrictive requirements 
are proposed for plant configurations with long calculated drain 
times, the overall effect of the change is to improve plant safety 
and to add safety margin.
    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: Donald P. Ferraro, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Suite 305, 
Kennett Square, PA 19348.
    NRC Branch Chief: James G. Danna.
Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Nuclear Generating Unit Nos. 3, and 4, Miami-Dade County, Florida
    Date of amendment request: August 23, 2017, as supplemented by 
letter dated October 19, 2017. Publicly-available versions are in ADAMS 
under Accession Nos. ML17235B008 and ML17292A789, respectively.
    Description of amendment request: The amendments would modify the 
Technical Specifications (TSs) to relocate the Explosive Gas Monitoring 
Instrumentation, Explosive Gas Mixture, and Gas Decay Tanks System 
requirements to licensee-controlled documents and establish a Gas Decay 
Tank Explosive Gas and Radioactivity Monitoring Program. The proposed 
amendments also relocate the Standby Feedwater System requirements to 
licensee-controlled documents and modify related Auxiliary Feedwater 
(AFW) System requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

[[Page 55407]]

    The proposed license amendments modify the Turkey Point TS by 
relocating the Explosive Gas Monitoring Instrumentation, Explosive 
Gas Mixture, Gas Decay Tanks and Standby Feedwater System 
requirements to licensee controlled documents, by relatedly 
modifying the AFW System requirements and by establishing a Gas 
Decay Tank Explosive Gas and Radioactivity Monitoring Program. The 
proposed changes are administrative in nature and do not alter any 
plant equipment or the manner in which plant equipment is operated 
and maintained. All equipment limitations, applicable methodologies 
and surveillances are maintained by the proposed changes. In 
addition, the proposed changes to the AFW System requirements 
enhance plant safety. As such, the proposed changes cannot affect 
the initiators, the likelihood or the expected outcomes of any 
analyzed accidents.
    Therefore, facility operation in accordance with the proposed 
changes would 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 license amendments modify the Turkey Point TS by 
relocating the Explosive Gas Monitoring Instrumentation, Explosive 
Gas Mixture, Gas Decay Tanks and Standby Feedwater System 
requirements to licensee controlled documents, by relatedly 
modifying the AFW System requirements and by establishing a Gas 
Decay Tank Explosive Gas and Radioactivity Monitoring Program. The 
proposed changes neither install or remove plant equipment nor alter 
any plant equipment design, configuration, or method of operation. 
Hence, no new failure mechanisms are introduced as a result of the 
proposed changes.
    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.
    The proposed license amendments modify the Turkey Point TS by 
relocating the Explosive Gas Monitoring Instrumentation, Explosive 
Gas Mixture, Gas Decay Tanks and Standby Feedwater System 
requirements to licensee controlled documents, by relatedly 
modifying the AFW System requirements and by establishing a Gas 
Decay Tank Explosive Gas and Radioactivity Monitoring Program. The 
proposed changes neither involve changes to safety analyses 
assumptions, safety limits, or limiting safety system settings nor 
adversely impact plant operating margins or the reliability of 
equipment credited in safety analyses.
    Therefore, operation of the facility in accordance with the 
proposed changes will not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB, 
Juno Beach, FL 33408-0420.
    NRC Branch Chief: Undine Shoop.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center (DAEC), Linn County, Iowa
    Date of amendment request: September 5, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17248A284.
    Description of amendment request: The proposed amendment would 
revise DAEC Technical Specifications 3.5.1, ``ECCS [emergency core 
cooling system]-Operating.'' The proposed change would decrease the 
nitrogen supply requirement for the Automatic Depressurization System 
(ADS) in Surveillance Requirement (SR) 3.5.1.3 from 100 days to 30 
days.
    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 modifies a SR for verification of the 
nitrogen supply for the ADS accumulators. Accidents are initiated by 
the malfunction of plant equipment, or the catastrophic failure of 
plant structures, systems or components. The performance of this 
surveillance is not a precursor to any accident previously evaluated 
and does not change the manner in which the ADS operates. Technical 
evaluation of the change concluded that a 30-day nitrogen supply is 
more than adequate to ensure that the reactor is depressurized, so 
the consequences of an accident remain unchanged.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequence of a previously evaluated 
accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve physical alterations to the 
plant. No new or different type of equipment will be installed, and 
there are no physical modifications required to existing installed 
equipment associated with the proposed change. The proposed change 
does not create any failure mechanism, malfunction or accident 
initiator not already considered in the design and licensing basis.
    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.
    Although the proposed change will decrease the required supply 
of nitrogen for the ADS accumulators from 100 days to 30 days, the 
assessment above has shown that the reactor would be depressurized 
within 3 days following any postulated accident or event that would 
create a hostile environment in the drywell. Once initial 
depressurization is completed, long term core cooling can be assured 
without ADS.
    Therefore, the proposed change will 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, P.O. Box 14000, Juno Beach, 
FL 33408-0420.
    NRC Branch Chief: David J. Wrona.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center, Linn County, Iowa
    Date of amendment request: August 31, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17243A469.
    Description of amendment request: The proposed amendment would 
modify the licensing basis by the addition of a license condition to 
allow for the implementation of the provisions of 10 CFR, part 50.69, 
``Risk-Informed Categorization and Treatment of Structures, Systems, 
and Components (SSCs) for Nuclear Power Reactors.''
    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 will permit the use of a risk-informed 
categorization process to modify the scope of SSCs subject to NRC 
special treatment requirements and to implement alternative 
treatments per the regulations. The process used to evaluate SSCs 
for changes to NRC special treatment

[[Page 55408]]

requirements and the use of alternative requirements ensures the 
ability of the SSCs to perform their design function. The potential 
change to special treatment requirements does not change the design 
and operation of the SSCs. As a result, the proposed change does not 
significantly affect any initiators to accidents previously 
evaluated or the ability to mitigate any accidents previously 
evaluated. The consequences of the accidents previously evaluated 
are not affected because the mitigation functions performed by the 
SSCs assumed in the safety analysis are not being modified. The SSCs 
required to safely shut down the reactor and maintain it in a safe 
shutdown condition following an accident will continue to perform 
their design functions.
    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 will permit the use of a risk-informed 
categorization process to modify the scope of SSCs subject to NRC 
special treatment requirements and to implement alternative 
treatments per the regulations. The proposed change does not change 
the functional requirements, configuration, or method of operation 
of any SSC. Under the proposed change, no additional plant equipment 
will be installed.
    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 will permit the use of a risk-informed 
categorization process to modify the scope of SSCs subject to NRC 
special treatment requirements and to implement alternative 
treatments per the regulations. The proposed change does not affect 
any Safety Limits or operating parameters used to establish the 
safety margin. The safety margins included in analyses of accidents 
are not affected by the proposed change. The regulation requires 
that there be no significant effect on plant risk due to any change 
to the special treatment requirements for SSCs and that the SSCs 
continue to be capable of performing their design basis functions, 
as well as to perform any beyond design basis functions consistent 
with the categorization process and results.
    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, P.O. Box 14000, Juno Beach, 
FL 33408-0420.
    NRC Branch Chief: David J. Wrona.
NextEra Energy, Point Beach Nuclear Plant (PBNP), LLC, Docket Nos. 50-
266 and 50-301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two 
Creeks, Manitowoc County, Wisconsin
    Date of amendment request: June 23, 2017, as supplemented by letter 
dated August 21, 2017. Publicly-available versions are in ADAMS under 
Accession Nos. ML17174A458, and ML17233A283, respectively.
    Description of amendment request: The amendments would revise the 
Emergency Plan for PBNP to adopt the Nuclear Energy lnstitute's (NEl's) 
revised Emergency Action Level (EAL) scheme described in NEI 99-01, 
Revision 6, ``Development of Emergency Action Levels for Non-Passive 
Reactors,'' which has been endorsed by the NRC.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not impact the physical configuration 
or function of plant structures, systems, or components (SSCs) or 
the manner in which SSCs are operated, maintained, modified, tested, 
or inspected. No actual facility equipment or accident analyses are 
affected by the proposed changes.
    The change revises the NextEra Emergency Action Levels to be 
consistent with the NRC endorsed EAL scheme contained in NEI 99-01, 
Revision 6, ``Methodology for Development of Emergency Action 
Levels,'' but does not alter any of the requirements of the 
Operating License or the Technical Specifications.
    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 involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
The proposed change does not create any new failure modes for 
existing equipment or any new limiting single failures. 
Additionally, the proposed change does not involve a change in the 
methods governing normal plant operation, and all safety functions 
will continue to perform as previously assumed in the accident 
analyses. Thus, the proposed change does not adversely affect the 
design function or operation of any structures, systems, and 
components important to safety. No new accident scenarios, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. The proposed change does not challenge the 
performance or integrity of any safety-related system.
    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 margin of safety associated with the acceptance criteria of 
any accident is unchanged. The proposed change will have no affect 
on the availability, operability, or performance of safety-related 
systems and components. The proposed change will not adversely 
affect the operation of plant equipment or the function of equipment 
assumed in the accident analysis. The proposed amendment does not 
involve changes to any safety analyses assumptions, safety limits, 
or limiting safety system settings. The changes do not adversely 
impact plant operating margins or the reliability of equipment 
credited in the safety analyses.
    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 and Exelon Generation Company, LLC, Docket Nos. 50-272 
and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey
    Date of amendment request: September 27, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17270A076.
    Description of amendment request: The amendments would relocate the 
reactor coolant system pressure isolation valve (RCS PIV) table from 
the technical specifications (TSs) to the technical requirements manual 
(TRM). The request would also remove references to the table and move 
all notes and leakage acceptance criteria from the table to the TS 
surveillance requirements.

[[Page 55409]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the TS will not alter the way any 
structure, system, or component (SSC) functions, and will not alter 
the manner in which the plant is operated. The proposed changes do 
not alter the design of any SSC. The relocation of the RCS PIV valve 
lists from the TS to the TRM is an administrative change. Future 
revisions to the TRM are subject to 10 CFR 50.59. Therefore the 
probability of an accident previously evaluated is not significantly 
increased.
    The proposed changes do not alter the RCS PIV leakage limits 
contained in the TS nor do they alter the frequency for testing of 
the RCS PIV. Therefore, the consequences of an accident previously 
evaluated are not increased.
    Therefore, these proposed changes do not represent a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a modification to the 
physical configuration of the plant or changes in the methods 
governing normal plant operation. The proposed changes will not 
impose any new or different requirement or introduce a new accident 
initiator, accident precursor, or malfunction mechanism. The 
proposed changes are administrative in nature.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes to the RCS PIV TS are administrative in 
nature. The proposed changes do not alter the RCS PIV leakage limits 
contained in the TS nor do they alter the frequency for testing of 
the RCS PIV. The proposed changes will not result in changes to 
system design or setpoints that are intended to ensure timely 
identification of plant conditions that could be precursors to 
accidents or potential degradation of accident mitigation systems.
    The proposed amendment will not result in a design basis or 
safety limit being exceeded or altered. Therefore, since the 
proposed changes do not impact the response of the plant to a design 
basis accident, 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: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: James G. Danna.
South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina
    Date of amendment request: October 6, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17279A715.
    Description of amendment request: The proposed amendment would 
increase the Integrated Leak Rate Test (ILRT) Peak Calculated 
Containment Internal Pressure, Pa, listed in Technical 
Specification (TS) 6.8.4.g, ``Containment Leakage Rate Testing 
Program,'' to remove the reference to Regulatory Guide (RG) 1.163, 
``Performance-Based Containment Leak Test Program,'' dated September 
1995 and ANSI/ANS (American National Standards Institute/American 
Nuclear Society)--56.8-2002, ``Containment System Leakage Testing 
Requirements,'' and to replace the reference of Nuclear Energy 
Institute (NEI) 94-01, Revision 3-A, ``Industry Guideline for 
Implementing Performance-Based option of 10 CFR part 50, Appendix J,'' 
with NEI 94-01, Revision 2-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, with NRC staff edits in square 
brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes involve removal of RG 1.163 and ANSl/ANS-
56.8-2002 references, replacement of NEI 94-01, Revision 3-A with 
NEI 94-01, Revision 2-A, and an increase in the Pa [Peak 
Calculated Containment Internal Pressure] value for containment 
leakage testing. The activity does not involve a physical change to 
the plant or a change in the manner in which 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 reactor 
containment itself and the testing requirements invoked to 
periodically demonstrate the integrity of the reactor 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 changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The integrity of the reactor containment is subject to two types 
of failure mechanisms which can be categorized as (1) activity based 
and (2) time based. Activity based failure mechanisms are defined as 
degradation due to system and/or 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 updated Pa value reflects the updated 
mass and energy release and containment response calculations, 
ensuring a sound technical basis for the local and integrated 
leakage tests.
    To mitigate time-based mechanisms, the design and construction 
requirements of the containment itself combined with the containment 
inspections performed in accordance with ASME [American Society of 
Mechanical Engineers], Section XI and the Maintenance Rule serve to 
provide a high degree of assurance that the containment will not 
degrade in a manner that is detectable only by a Type A test. The 
change to the Pa value is less than 1 psid [per square 
inch differential]. Radiological consequences will continue to be 
evaluated at the Technical Specification allowed leakage, 
La [allowed leakage] of 0.20 percent by weight of air, 
which will not be increased despite the increase in Pa. 
As described in Section 3.5, past leakage testing yielded values 
well under La. Based on the above, neither the reference 
changes nor the Pa change involves a significant increase 
in the consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes involve removal of RG 1.163 and ANSl/ANS-
56.8-2002 references, replacement of NEI 94-01, Revision 3-A with 
NEI 94-01, Revision 2-A, and an increase in the Pa value 
for containment leakage testing. The reactor containment and the 
testing requirements invoked to periodically demonstrate the 
integrity of the reactor containment exist to ensure the plant's 
ability to mitigate the consequences of an accident. There are not 
any accident initiators or precursors affected by the revision. The 
proposed TS change does not involve a physical change to the plant 
or the manner in which the plant is operated or controlled.
    Therefore, the proposed TS 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?

[[Page 55410]]

    Response: No.
    The proposed changes involve removal of RG 1.163 and ANSl/ANS-
56.8-2002 references, replacement of NEI 94-01, Revision 3-A with 
NEI 94-01, Revision 2-A, and an increase in the Pa value 
for containment leakage testing. The proposed TS change does not 
involve a physical change to the plant or a change in the manner in 
which the plant is operated or controlled. Using the same analysis 
methodology as described in WCAP-10325-P-A [Westinghouse LOCA [loss-
of-accident coolant] Mass and Energy Release Model for Containment 
Design], the updated mass and energy release and containment 
response analyses corrected input errors identified in the NSALs 
[Westinghouse Nuclear Safety Advisory Letters] described previously. 
As shown in Figure 1 [October 6, 2017, submittal], the correction of 
these errors resulted in a slightly higher predicted peak pressure 
than that of the current licensing basis but does not pose a 
significant challenge to the design limit.
    The specific requirements and conditions of the Primary 
Containment Leak Rate Testing Program, as defined in the Technical 
Specifications, exist to ensure that the degree of reactor 
containment structural integrity and leak-tightness that is 
considered in the plant safety analysis is maintained. The overall 
containment leak rate limit specified by the Technical Specification 
is maintained. The containment inspections performed in accordance 
with ASME, Section XI and the Maintenance Rule serve to provide a 
high degree of assurance that the containment will not degrade in a 
manner that is detectable only by Type A testing. The combination of 
these factors ensures that the margin of safety that is in 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 will 
continue to be met.
    Therefore, the proposed TS 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 M. Sutton, Morgan, Lewis & Bockius 
LLP, 1111 Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Branch Chief: Michael T. Markley.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: July 28, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17209A759.
    Description of amendment request: The amendment request proposes to 
revise Technical Specification Section 1.1 (TS), Definition of 
Actuation Logic Test, by adding a new TS Section 1.1 Definition of 
Actuation Logic Output Test (ALOT), revising existing Surveillance 
Requirements 3.3.15.1 and 3.3.16.1 and adding new Surveillance 
Requirements 3.3.15.2 and 3.3.16.2 to implement the new ALOT. This 
submittal requests approval of the license amendment that is necessary 
to implement these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(A), licensee has provided 
its analysis of the issue on no significant hazards consideration 
determination, 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.
    There are no design changes associated with the proposed 
amendment. All design, material, and construction standards that 
were applicable prior to this amendment request will continue to be 
applicable.
    The [Processor Module Self-Diagnostic (PMS)] will continue to 
function in a manner consistent with the plant design basis. There 
will be no changes to the PMS operating limits. The existing 
ACTUATION LOGIC TEST Surveillance Requirements are revised such that 
different portions of the PMS logic circuitry are tested on 
appropriate surveillance test frequencies.
    The proposed change will not adversely affect accident 
initiators or precursors or adversely alter the design assumptions, 
conditions, and configuration of the facility, or the manner in 
which the plant is operated and maintained, with respect to such 
initiators or precursors.
    The proposed changes will not alter the ability of structures, 
systems, and components (SSCs) to perform their specified safety 
functions to mitigate the consequences of an initiating event within 
the assumed acceptance limits.
    Accident analysis acceptance criteria will continue to be met 
with the proposed changes. The proposed changes will not affect the 
source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of any 
accident previously evaluated. The proposed changes will not alter 
any assumptions or change any mitigation actions in the radiological 
consequence evaluations in the Updated Final Safety Analysis Report 
(UFSAR).
    The applicable radiological dose acceptance criteria will 
continue to be met.
    The proposed change revises the frequency of testing certain 
portions of the PMS logic circuitry, but does not physically alter 
any safety-related systems.
    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 
different kind of accident from any accident previously evaluated?
    Response: No.
    With respect to any new or different kind of accident, there are 
no proposed design changes nor are there any changes in the method 
by which any safety-related plant SSC performs its specified safety 
function. The proposed change will not affect the normal method of 
plant operation or change any operating parameters. No equipment 
performance requirements will be affected. The proposed change will 
not alter any assumptions made in the safety analyses.
    The proposed change revises the frequency of testing certain 
portions of the PMS logic circuitry. The proposed change does not 
involve a physical modification of the plant.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of this amendment. There will be no adverse effect or 
challenges imposed on any safety-related system as a result of this 
amendment.
    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 existing ACTUATION LOGIC TEST Surveillance Requirements are 
revised such that different portions of the PMS logic circuitry are 
tested on appropriate surveillance test frequencies. The reliability 
of the PMS is such that not testing the Component Interface Module 
(CIM) logic and driver output circuits when the reactor is at power 
will have a net positive impact on Engineered Safety Feature 
Actuation System (ESFAS) availability. There will be a reduction in 
the potential for challenges to the safety systems, coupled with 
less time that the safety systems are unavailable.
    There will be no effect on those plant systems necessary to 
effect the accomplishment of protection functions.
    No instrument setpoints or system response times are affected. 
None of the acceptance criteria for any accident analysis will be 
changed.
    The proposed change will have no impact on the radiological 
consequences of a design basis accident.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

    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.

[[Page 55411]]

    NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: August 18, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17230A365.
    Description of amendment request: The requested amendment proposes 
to depart from approved AP1000 Design Control Document (DCD) Tier 2 
information (text) and involved Tier 2* information (as incorporated 
into the Updated Final Safety Analysis Report (UFSAR) as plant-specific 
DCD information).
    This amendment request proposes increasing the design pressure of 
the main steam (MS) isolation valve (MSIV) compartments from 6.0 to 6.5 
psi and proposes other changes to the licensing basis regarding 
descriptions of the MSIV compartments.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with Nuclear Regulatory 
Commission (NRC) staff's edits in square brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not adversely affect the operation of 
any structures, systems, and components inside or outside the 
auxiliary building that could initiate or mitigate abnormal events, 
e.g., accidents, anticipated operational occurrences, earthquakes, 
floods, tornado missiles, and turbine missiles, or their safety or 
design analyses, evaluated in the UFSAR. The changes do not 
adversely affect any design function of the auxiliary building or 
the structures, systems, and components contained therein. The 
ability of the affected auxiliary building main steam isolation 
valve compartments and adjacent rooms, including the main control 
room, to withstand the pressurization effects from the postulated 
pipe ruptures is not adversely affected by the increase in design 
pressure, since the structures, systems, and components therein 
remain qualified for this service.
    Therefore, the proposed activity 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 might initiate a new or different kind of 
accident, or alter any [structure, system, and component (SSC)] such 
that a new accident initiator or initiating sequence of events is 
created. The proposed changes do not adversely affect the physical 
design and operation of the [in-containment refueling water storage 
tank (IRWST)] injection, drain, containment recirculation, and 
fourth-stage [automatic depressurization system (ADS)] valves, 
including as-installed inspections, and maintenance requirements, as 
described in the UFSAR. Therefore, the operation of the IRWST 
injection, drain, containment recirculation, and fourth-stage ADS 
valves is not adversely affected. These proposed changes do not 
adversely 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 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 margin of safety for the design of the auxiliary building is 
maintained through continued use of approved codes and standards as 
stated in the UFSAR, and adherence to the assumptions used in the 
analyses of this structure and the events associated with this 
structure. The auxiliary building continues to be a seismic Category 
I building with all current structural safety margins maintained. 
The 3-hour fire rating requirements for the impacted auxiliary 
building walls are maintained. The equipment housed in the main 
steam isolation valve compartments continue to be environmentally 
qualified for their intended service in accordance with the approved 
codes and standards stated within the UFSAR. Thus, the requested 
changes will not adversely affect any safety-related equipment, 
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 
requested change, thus, no margin of safety is reduced. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: October 6, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17279A084.
    Description of amendment request: The amendment request proposes to 
depart from Tier 2 information in the Updated Final Safety Analysis 
Report (UFSAR) (which includes the plant-specific Design Control 
Document (DCD) 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. Pursuant to 
the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the 
design as certified in the 10 CFR part 52, Appendix D, design 
certification rule is also requested for the plant-specific DCD Tier 1 
material departures. Specifically, the requested amendment proposes to 
depart from Tier 2 information in UFSAR Subsection 8.3.2.4 describing 
raceway and cable routing criteria and hazard protection, and involves 
related changes to plant-specific Tier 1 Table 3.3-6, inspections, 
tests, analyses, and acceptance criteria information, with 
corresponding changes to the associated COL Appendix C 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 with NRC staff edits in square 
brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Changes 1, 3 and 4 are clarifications only and do not represent 
a change to the minimum required separation distance between 
raceways. Change 2 reduces the required separation distances between 
raceways from those documented in [Institute of Electrical and 
Electronics Engineers (IEEE)] 384-1981. These reduced separation 
distances are based on specific tests performed on the specified 
raceway configurations, and the recommendations from those tests 
contained in the associated report. The NRC staff previously 
reviewed the descriptions of the ten tests documented in this 
report, including the ones applicable to the existing UFSAR 
exceptions, and concluded that they were acceptable, as documented 
in NUREG-1793, ``Final Safety Evaluation Report Related to 
Certification of

[[Page 55412]]

the AP1000 Standard Design,'' (Initial Report) Subsection 8.3.2.2.
    The reduced separation does not adversely impact the ability to 
safely shutdown the plant, and maintain it shutdown. The referenced 
test report has shown a failure of a faulted cable will not 
propagate to a nearby target cable in way that adversely impacts its 
function.
    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.
    Changes 1, 3 and 4 are clarifications only and do not represent 
a change to the minimum required separation distance between 
circuits. Change 2 reduces the required separation distances between 
circuits from those documented in IEEE 384-1981. This change does 
not result in a new accident initiator or impact a current accident 
initiator.
    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.
    Changes 1, 3 and 4 are clarifications only and do not represent 
a change to the minimum required separation distance between 
circuits. Change 2 reduces the required separation distances between 
circuits from those documented in IEEE 384-1981. These reduced 
separation distances are based on specific tests performed on the 
specified raceway configurations, and the recommendations from those 
tests contained in the associated report. The NRC staff previously 
reviewed the descriptions of the ten tests documented in this 
report, including the ones applicable to the existing UFSAR 
exceptions, and concluded that they were acceptable, as documented 
in NUREG-1793, ``Final Safety Evaluation Report Related to 
Certification of the AP1000 Standard Design,'' (Initial Report) 
Subsection 8.3.2.2.
    The reduced separation does not adversely impact the ability to 
safely shutdown the plant, and maintain it shutdown. The referenced 
test report has shown a failure of a faulted cable will not 
propagate to a nearby target cable in a way that adversely impacts 
its function.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia
    Date of amendment request: July 10, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17191B163.
    Description of amendment request: The amendments would revise the 
technical specifications (TSs) by: (1) Adding a Note to the 
surveillance requirements (SRs) of TS 3.7.7, ``Main Turbine Bypass 
System,'' to clarify that the SRs are not required to be met when the 
limiting condition for operation (LCO) does not require the Main 
Turbine Bypass System to be operable, (2) clarifying that LCO 3.2.3, 
``LINEAR HEAT GENERATION RATE (LHGR),'' also has limits for an 
inoperable Main Turbine Bypass System that are made applicable as 
specified in the Core Operating Limits Report, and (3) deleting an 
outdated footnote for LCO 3.2.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change (1) adds a Note to the Surveillance 
Requirements (SRs) of the Hatch Nuclear Plant (HNP) Unit 1 and Unit 
2 Technical Specifications (TS) 3.7.7 clarifying that the SRs are 
not required to be met when the LCO does not require the Main 
Turbine Bypass System to be Operable, (2) clarifies that LCO 3.2.3, 
``LINEAR HEAT GENERATION RATE'' also has limits for an inoperable 
Main Turbine Bypass System that are made applicable as specified in 
the Core Operating Limits Report, and (3) deletes an outdated 
footnote for LCO 3.2.3. The proposed change does not affect the 
requirement to meet the LCO, nor does it affect the requirements to 
perform the SRs when the Main Turbine Bypass System is being used to 
meet the LCO. This change simply clarifies the existing allowance to 
apply the Main Turbine Bypass System inoperable limits to minimum 
critical power ratio (MCPR) and linear heat generation rate (LHGR) 
in lieu of the requirement for the Main Turbine Bypass System to be 
Operable. The current safety analysis evaluation is unaffected by 
this proposed change. The change regarding the outdated footnote has 
no effect on the actual TS requirements.
    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 (1) adds a Note to the Surveillance 
Requirements (SRs) of the Hatch Nuclear Plant (HNP) Unit 1 and Unit 
2 Technical Specifications (TS) 3.7.7 clarifying that the SRs are 
not required to be met when the LCO does not require the Main 
Turbine Bypass System to be Operable, (2) clarifies that LCO 3.2.3, 
``LINEAR HEAT GENERATION RATE'' also has limits for an inoperable 
Main Turbine Bypass System that are made applicable as specified in 
the Core Operating Limits Report, and (3) deletes an outdated 
footnote for LCO 3.2.3. This change simply clarifies the existing 
allowance to apply the Main Turbine Bypass System inoperable limits 
to minimum critical power ratio (MCPR) and linear heat generation 
rate (LHGR) in lieu of the requirement for the Main Turbine Bypass 
System to be Operable. The change regarding the outdated footnote 
has no effect on the actual TS requirements. The current safety 
analysis evaluation is unaffected by these proposed changes.
    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 (1) adds a Note to the Surveillance 
Requirements (SRs) of the Hatch Nuclear Plant (HNP) Unit 1 and Unit 
2 Technical Specifications (TS) 3.7.7 clarifying that the SRs are 
not required to be met when the LCO does not require the Main 
Turbine Bypass System to be Operable, (2) clarifies that LCO 3.2.3, 
``LINEAR HEAT GENERATION RATE'' also has limits for an inoperable 
Main Turbine Bypass System that are made applicable as specified in 
the Core Operating Limits Report, and (3) deletes an outdated 
footnote for LCO 3.2.3. This change simply clarifies the existing 
allowance to apply the Main Turbine Bypass System inoperable limits 
to minimum critical power ratio (MCPR) and linear heat generation 
rate (LHGR) in lieu of the requirement for the Main Turbine Bypass 
System to be Operable. The applicable safety analyses for TS 3.7.7 
is unaffected by this clarification. The change regarding the 
outdated footnote has no effect on the actual TS requirements.
    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

[[Page 55413]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, 40 Inverness Center 
Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Michael T. Markley.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: September 13, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17256A626.
    Description of amendment request: The requested amendment proposes 
to depart from approved AP1000 Design Control Document (DCD) Tier 2 
information as incorporated into the Updated Final Safety Analysis 
Report (UFSAR) as plant-specific DCD information, and from Technical 
Specifications as incorporated in Appendix A of the Combined License 
(COL). Specifically, the proposed changes revise COL Appendix A 
Technical Specification 3.6.8 to identify the trisodium phosphate (TSP) 
mass value required in the pH adjustment baskets. The TSP mass value 
adjusts the pH of the containment water to >7.0 following a postulated 
accident.
    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 activity revises the mass of trisodium phosphate 
(TSP), which raises the pH of post-accident containment water to 7.0 
or greater following a postulated accident. The change to the TSP 
mass value does not adversely impact the ability to support 
radionuclide retention with high radioactivity in containment and 
helps prevent corrosion of containment equipment during long-term 
floodup conditions. The proposed changes do not adversely impact 
previously evaluated accidents, because pH control capability is 
provided to mitigate already postulated accidents. As described in 
Updated Final Safety Analysis Report (UFSAR) Subsection 
15.6.5.3.1.3, the passive core cooling system (PXS) is assumed to 
provide sufficient TSP to the post-loss-of-coolant accident (LOCA) 
cooling solution to maintain the pH at greater than or equal to 7.0 
following a LOCA. The pH adjustment baskets provide for long-term pH 
control. Long-term pH control is not adversely impacted as the pH 
adjustment baskets contain the required amount of TSP to support pH 
control requirements following a design basis accident (DBA).
    No safety-related structure, system, component (SSC) or function 
is adversely affected by this change. The change does not involve an 
interface with any SSC accident initiator or initiating sequence of 
events, and thus, the probabilities of the accidents evaluated in 
the UFSAR are not affected. The proposed changes do not involve a 
change to the predicted radiological releases due to postulated 
accident conditions, thus, the consequences of the accidents 
evaluated in the UFSAR are not affected.
    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 activity revises the mass of TSP, which raises the 
pH of containment to 7.0 or greater following a postulated accident. 
The proposed activity does not create the possibility of a new or 
different kind of accident as pH adjustment is used to support 
proper containment chemistry requirements following an accident. The 
proposed activity does not adversely affect any safety related 
equipment, and does not add any new interfaces to safety-related 
SSCs that adversely affect safety functions. No system or design 
function or equipment qualification is adversely affected by these 
changes as the changes do not modify any SSCs that prevent safety 
functions from being performed. The capability to maintain a maximum 
containment pH below 9.5 is not adversely impacted by these changes. 
The changes do not introduce a new failure mode, malfunction or 
sequence of events that could adversely affect safety or safety 
related equipment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed activity revises the mass of TSP, which raises the 
pH of containment to 7.0 or greater following a postulated accident. 
The proposed activity does not affect any other safety-related 
equipment or fission product barriers. Containment water pH 
adjustment is not adversely impacted. The requested changes will not 
adversely affect compliance with 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 requested changes as previously 
evaluated accidents are not impacted.
    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 Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, 
Georgia
    Date of amendment request: September 29, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17272A957.
    Description of amendment request: The requested amendment proposes 
to depart from Tier 2* and associated Tier 2 information in the Updated 
Final Safety Analysis Report (UFSAR) (which includes the plant-specific 
DCD Tier 2 information). The requested amendment proposes to depart 
from UFSAR Tier 2* information regarding resolution of human 
engineering deficiencies (HEDs) contained in Westinghouse Electric 
Company's report APP-OCS-GEH-320, ``AP1000 Human Factors Engineering 
Integrated Systems Validation Plan,'' which is incorporated by 
reference into the VEGP Units 3 and 4 UFSAR.
    The proposed changes would revise the licensing basis of the 
combined licenses regarding the process for addressing and re-testing 
of HEDs identified during the integrated system validation (ISV) as 
described in Tier 2* document, APPOCS-GEH-320 ``AP1000 Human Factors 
Engineering Integrated System Validation Plan.'' APPOCS-GEH-320 
references APP-OCS-GEH-420, ``Human Factors Engineering Discrepancy 
Resolution Process,'' which defines the process for tracking, 
resolution, and closure of HEDs. The proposed changes to APP-OCS-GEH-
320 do not impact APP-OCS-GEH-420.
    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 Integrated System Validation (ISV) provides a comprehensive 
human performance-based assessment of the design of the AP1000 
Human-System Interface (HSI) resources, based on their realistic 
operation

[[Page 55414]]

within a simulator driven Main Control Room (MCR). The ISV is part 
of the overall AP1000 Human Factors Engineering (HFE) program. The 
changes to APP-OCS-GEH-320, which is incorporated by reference into 
the UFSAR, clarify the resources and methodology used during re-
testing performed to verify the effectiveness of Human Engineering 
Deficiency (HED) resolution. The ISV Plan does not affect the plant 
itself. Changing APP-OCS-GEH-320 and the UFSAR does not affect 
prevention and mitigation of abnormal events, e.g., accidents, 
anticipated operational occurrences, earthquakes, floods and turbine 
missiles, or their safety or design analyses. No safety-related 
structure, system, component (SSC) or function is adversely 
affected. The changes neither involve nor interface with any SSC 
accident initiator or initiating sequence of events, and thus, the 
probabilities of the accidents evaluated in the UFSAR are not 
affected. Because the changes do not involve any safety-related SSC 
or function used to mitigate an accident, 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 changes to APP-OCS-GEH-320 and the VEGP 3 and 4 UFSAR affect 
only the testing and validation of the MCR design and HSI using a 
plant simulator. Therefore, the changes do not affect the safety-
related equipment itself, nor do they affect equipment which, if it 
failed, could initiate an accident or a failure of a fission product 
barrier. No analysis is adversely affected. No system or design 
function or equipment qualification is adversely affected by the 
changes. 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 would result in significant 
fuel cladding failures. In addition, the changes do not result in a 
new failure mode, malfunction or sequence of events that could 
affect safety or safety related equipment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The changes to APP-OCS-GEH-320 and the UFSAR affect the testing 
and validation of the MCR design and HSI using a plant simulator. 
Therefore, the changes do not affect the assessments or the plant 
itself. These changes do not affect safety-related equipment or 
equipment whose failure could initiate an accident, nor does it 
adversely interface with safety-related equipment or fission product 
barriers. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the requested change.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and based on 
this review it appears that the three standards of 10 CFR 50.92 (c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazard consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.
Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna 
Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
    Date of amendment request: September 20, 2017. A publicly-available 
version is in ADAMS under Package Accession No. ML17265A434.
    Description of amendment request: The amendments would revise 
technical specification (TS) requirements related to ``operations with 
a potential for draining the reactor vessel'' (OPDRVs) with new 
requirements on reactor pressure vessel (RPV) water inventory control 
(WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires 
RPV water level to be greater than the top of active irradiated fuel. 
The proposed changes are based on Technical Specifications Task Force 
(TSTF) Traveler TSTF-542, Revision 2, ``Reactor Pressure Vessel Water 
Inventory Control,'' dated December 20, 2016.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold 
shutdown) and Mode 5 (i.e., refueling) is not an accident previously 
evaluated and, therefore, replacing the existing TS controls to 
prevent or mitigate such an event with a new set of controls has no 
effect on any accident previously evaluated. RPV water inventory 
control in Mode 4 or Mode 5 is not an initiator of any accident 
previously evaluated. The existing OPDRV controls or the proposed 
RPV WIC controls are not mitigating actions assumed in any accident 
previously evaluated.
    The proposed changes reduce the probability of an unexpected 
draining event (which is not a previously evaluated accident) by 
imposing new requirements on the limiting time in which an 
unexpected draining event could result in the reactor vessel water 
level dropping to the top of the active fuel (TAF). These controls 
require cognizance of the plant configuration and control of 
configurations with unacceptably short drain times. These 
requirements reduce the probability of an unexpected draining event. 
The current TS requirements are only mitigating actions and impose 
no requirements that reduce the probability of an unexpected 
draining event.
    The proposed changes reduce the consequences of an unexpected 
draining event (which is not a previously evaluated accident) by 
requiring an Emergency Core Cooling System (ECCS) subsystem to be 
operable at all times in Modes 4 and 5. The current TS requirements 
do not require any water injection systems, ECCS or otherwise, to be 
Operable in certain conditions in Mode 5. The change in requirement 
from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does 
not significantly affect the consequences of an unexpected draining 
event because the proposed Actions ensure equipment is available 
within the limiting drain time that is as capable of mitigating the 
event as the current requirements. The proposed controls provide 
escalating compensatory measures to be established as calculated 
drain times decrease, such as verification of a second method of 
water injection and additional confirmations that containment and/or 
filtration would be available if needed.
    The proposed changes reduce or eliminate some requirements that 
were determined to be unnecessary to manage the consequences of an 
unexpected draining event, such as automatic initiation of an ECCS 
subsystem and the Control Room Emergency Outside Air Supply (CREOAS) 
system. These changes do not affect the consequences of any accident 
previously evaluated since a draining event in Modes 4 and 5 is not 
a previously evaluated accident and the requirements are not needed 
to adequately respond to a draining event.
    The administrative update to delete expired completion time 
notes is purely administrative in nature.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.1.3. The proposed changes will not alter the design 
function of the equipment involved. Under the proposed changes, some 
systems that are currently required to be operable during OPDRVs 
would be required to be available within the

[[Page 55415]]

limiting drain time or to be in service depending on the limiting 
drain time. Should those systems be unable to be placed into 
service, the consequences are no different than if those systems 
were unable to perform their function under the current TS 
requirements. The event of concern under the current requirements 
and the proposed changes are an unexpected draining event. The 
proposed changes do not create new failure mechanisms, malfunctions, 
or accident initiators that would cause a draining event or a new or 
different kind of accident not previously evaluated or included in 
the design and licensing bases.
    The administrative update to delete expired completion time 
notes is purely administrative in nature.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC. The current requirements do 
not have a stated safety basis and no margin of safety is 
established in the licensing basis. The safety basis for the new 
requirements is to protect Safety Limit 2.1.1.3. New requirements 
are added to determine the limiting time in which the RPV water 
inventory could drain to the top of the fuel in the reactor vessel 
should an unexpected draining event occur. Plant configurations that 
could result in lowering the RPV water level to the TAF within one 
hour are now prohibited. New escalating compensatory measures based 
on the limiting drain time replace the current controls. The 
proposed TS establish a safety margin by providing defense-in-depth 
to ensure that the Safety Limit is protected and to protect the 
public health and safety. While some less restrictive requirements 
are proposed for plant configurations with long calculated drain 
times, the overall effect of the change is to improve plant safety 
and to add safety margin.
    The administrative update to delete expired completion time 
notes is purely administrative in nature.
    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: Damon D. Obie, Associate General Counsel, 
Talen Energy Supply, LLC, 835 Hamilton St., Suite 150, Allentown, PA 
18101.
    NRC Branch Chief: James G. Danna.
Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260, and 50-
296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Limestone 
County, Alabama
    Date of amendment request: August 15, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17228A490.
    Description of amendment request: The amendments would revise the 
BFN, Units 1, 2, and 3 Technical Specification (TS) 5.5.12, ``Primary 
Containment Leakage Rate Testing Program,'' by adopting Nuclear Energy 
Institute (NEI) 94-01, Revision 3-A, ``Industry Guideline for 
Implementing Performance-Based Option of 10 CFR part 50, Appendix J,'' 
as the implementation document for the performance-based Option B of 10 
CFR part 50, Appendix J. The proposed changes permanently extend the 
Type A containment integrated leak rate testing (ILRT) interval from 10 
years to 15 years and the Type C local leakage rate testing (LLRT) 
intervals from 60 months to 75 months.
    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 consequence of an accident previously evaluated?
    Response: No.
    The proposed revision to TS 5.5.12 changes the testing period to 
a permanent 15-year interval for Type A testing (10 CFR part 50, 
Appendix J, Option B, ILRT) and a 75-month interval for Type C 
testing (10 CFR part 50, Appendix J, Option B, LLRT). The current 
Type A test interval of 10 years would be extended to 15 years from 
the last Type A test. The proposed extension to Type A testing does 
not involve a significant increase in the consequences of an 
accident because research documented in NUREG-1493, ``Performance-
Based Containment System Leakage Testing Requirements'' 
[``Performance-Based Containment Leak-Test Program''], September 
1995, has found that, generically, very few potential containment 
leakage paths are not identified by Type B and C tests. NUREG-1493 
concluded that reducing the Type A testing frequency to one per 20 
years was found to lead to an imperceptible increase in risk. A high 
degree of assurance is provided through testing and inspection that 
the containment will not degrade in a manner detectable only by Type 
A testing. The last Type A test (performed November 19, 2010 for 
BFN, Unit 1, June 3, 2009 for BFN, Unit 2 and May 12, 2012 for BFN, 
Unit 3) shows leakage to be below acceptance criteria, indicating a 
very leak tight containment. Inspections required by the ASME Code 
[American Society of Mechanical Engineers Boiler and Press Vessel 
Code] Section Xl (Subsection IWE) and Maintenance Rule monitoring 
(10 CFR 50.65, ``Requirements for Monitoring the Effectiveness of 
Maintenance at Nuclear Power Plants'') are performed in order to 
identify indications of containment degradation that could affect 
that leak tightness. Types B and C testing required by TSs will 
identify any containment opening such as valves that would otherwise 
be detected by the Type A tests. These factors show that a Type A 
test interval extension will not represent a significant increase in 
the consequences of an accident.
    The proposed amendment involves changes to the BFN, Units 1, 2, 
and 3, 10 CFR 50 Appendix J Testing Program Plan. The proposed 
amendment does not involve a physical change to the plant or a 
change in the manner in which the units are operated or controlled. 
The primary containment function is to provide an essentially leak 
tight barrier against the uncontrolled release of radioactivity to 
the environment for postulated accidents. As such, the containment 
itself and the testing requirements to periodically demonstrate the 
integrity of the containment exist to ensure the plant's ability to 
mitigate the consequences of an accident, and do not involve any 
accident precursors or initiators. Therefore, the probability of 
occurrence of an accident previously evaluated is not significantly 
increased by the proposed amendment.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, for development of the BFN, Units 1, 2, and 3, 
performance-based leakage testing program. 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 potential consequences of extending the 
ILRT interval from 10 years to 15 years have been evaluated by 
analyzing the resulting changes in risk. The increase in risk in 
terms of person-rem [roentgen equivalent man] per year resulting 
from design basis accidents was estimated to be very small, and the 
increase in the LERF [large early release frequency] resulting from 
the proposed change was determined to be within the guidelines 
published in NRC RG [Regulatory Guide] 1.174. Additionally, the 
proposed change maintains defense-in-depth by preserving a 
reasonable balance among prevention of core damage, prevention of 
containment failure, and consequence mitigation. TVA has determined 
that the increase in CCFP [conditional containment failure 
probability] due to the proposed change would be very small.
    Based on the above discussions, the proposed changes do not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.

[[Page 55416]]

    The proposed revision to TS 5.5.12 changes the testing period to 
a permanent 15-year interval for Type A testing (10 CFR part 50, 
Appendix J, Option B, ILRT) and a 75-month interval for Type C 
testing (10 CFR part 50, Appendix J, Option B, LLRT). The current 
test interval of 10 years, based on past performance, would be 
extended to 15 years from the last Type A test (performed November 
19, 2010 for BFN, Unit 1, June 3, 2009 for BFN, Unit 2 and May 12, 
2012 for BFN, Unit 3). The proposed extension to Type A and Type C 
test intervals does not create the possibility of a new or different 
type of accident because there are no physical changes being made to 
the plant and there are no changes to the operation of the plant 
that could introduce a new failure mode creating an accident or 
affecting the mitigation of an accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed revision to TS 5.5.12 changes the testing period to 
a permanent 15-year interval for Type A testing (10 CFR part 50, 
Appendix J, Option B, ILRT) and a 75-month interval for Type C 
testing (10 CFR part 50, Appendix J, Option B, LLRT). The current 
test interval of 10 years, based on past performance, would be 
extended to 15 years from the last Type A test (performed November 
19, 2010 for BFN, Unit 1, June 3, 2009 for BFN, Unit 2 and May 12, 
2012 for BFN, Unit 3). The proposed extension to Type A testing will 
not significantly reduce the margin of safety. NUREG-1493, 
``Performance-Based Containment System Leakage Testing 
Requirements'' [``Performance-Based Containment Leak-Test 
Program''], September 1995, generic study of the effects of 
extending containment leakage testing, found that a 20 year 
extension to Type A leakage testing resulted in an imperceptible 
increase in risk to the public. NUREG-1493 found that, generically, 
the design containment leakage rate contributes about 0.1% to the 
individual risk and that the decrease in Type A testing frequency 
would have a minimal effect on this risk since 95% of the potential 
leakage paths are detected by Type C testing. Regular inspections 
required by the ASME Code Section Xl (Subsection IWE) and 
maintenance rule monitoring (10 CFR 50.65, ``Requirements for 
Monitoring the Effectiveness of Maintenance at Nuclear Power 
Plants'') will further reduce the risk of a containment leakage path 
going undetected.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, for development of the BFN, Units 1, 2, and 3, 
performance-based leakage testing program, and establishes a 15-year 
interval for the performance of the primary containment ILRT and a 
75-month interval for Type C testing. The amendment does not alter 
the manner in which safety limits, limiting safety system setpoints, 
or limiting conditions for operation are determined. The specific 
requirements and conditions of the 10 CFR part 50, Appendix J 
Testing Program Plan, as defined in the TS, ensure that the degree 
of primary containment structural integrity and leak-tightness that 
is considered in the plant safety analyses is maintained. The 
overall containment leakage rate limit specified by the TS is 
maintained, and the Type A, B, and C containment leakage tests will 
continue to be performed at the frequencies established in 
accordance with the NRC-accepted guidelines of NEI 94-01, Revision 
3-A.
    Containment inspections performed in accordance with other plant 
programs serve to provide a high degree of assurance that the 
containment will not degrade in a manner that is detectable only by 
an ILRT. This ensures that evidence of containment structural 
degradation is identified in a timely manner. Furthermore, a risk 
assessment using the current BFN, Units 1, 2, and 3, PRA 
[probabilistic risk assessment] model concluded that extending the 
ILRT test interval from 10 years to 15 years results in a very small 
change to the BFN, Units 1, 2, and 3, risk profile.
    Accordingly, 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Dr., WT 6A, Knoxville, TN 37902.
    NRC Branch Chief: Undine Shoop.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2 (SQN), Hamilton County, Tennessee
Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar 
Nuclear Plant, Units 1 and 2 (WBN), Rhea County, Tennessee
    Date of amendment request: August 7, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17219A505.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.2.4, ``Quadrant Power Tilt Ratio 
(QPTR),'' and TS 3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' 
to avoid confusion as to when an incore power distribution measurement 
for QPTR is required. The amendment would also revise the WBN TSs for 
consistency with the existing SQN TSs and Westinghouse Standard TSs in 
NUREG-1431, Revision 4.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
Further, the proposed changes do not increase the types or amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures. The proposed changes do not significantly 
increase the probability of an accident and are consistent with 
safety analysis assumptions and resultant consequences.
    Therefore, the changes do not increase 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 result in a change in the manner in 
which the reactor trip system (RTS) and engineered safety feature 
actuation system (ESFAS) provide plant protection. The RTS and ESFAS 
will continue to have the same setpoints after the proposed changes 
are implemented. There are no design changes associated with the 
change. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by these changes. Redundant RTS and ESFAS trains 
are maintained, and diversity with regard to the signals that 
provide reactor trip and engineered safety features actuation is 
also

[[Page 55417]]

maintained. All signals credited as providing primary or secondary 
protection, and all operator actions credited in the accident 
analyses will remain the same. The proposed changes will not result 
in plant operation in a configuration outside the design basis.
    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Branch Chief: Undine Shoop.

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.

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment requests: December 15, 2016.
    Brief description of amendments: The amendments modified Technical 
Specification (TS) 3.4.10, ``Pressurizer Safety Valves,'' TS 3.7.4, 
``Steam Generator Power Operated Relief Valves (SG PORVs),'' and TS 
3.7.6, ``Condensate Storage System,'' to revise the Completion Times 
for Limiting Condition for Operation (LCO) of TS LCO 3.4.10 Required 
Action B.2, TS LCO 3.7.4 Required Action C.2, and TS LCO 3.7.6 Required 
Action B.2 from 12 to 24 hours. The proposed changes are consistent 
with Technical Specifications Task Force (TSTF) Traveler TSTF-352-A, 
Revision 1, ``Provide Consistent Completion Time to Reach MODE 4.''
    Date of issuance: October 23, 2017.
    Effective date: These license amendments are effective as of its 
date of issuance and shall be implemented within 120 days of issuance.
    Amendment Nos.: 294 (Unit 1) and 290 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17254A144; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the renewed licenses and technical specifications.
    Date of initial notice in Federal Register: April 25, 2017 (82 FR 
19099).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 23, 2017.
    No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment requests: December 15, 2016.
    Brief description of amendments: The amendments modified technical 
specification (TS) limiting condition for operation (LCO) 3.7.5, 
``Auxiliary Feedwater (AFW) System,'' Condition A and Required Action 
A.1. Condition A was revised to include the situation when one turbine-
driven AFW pump is inoperable in MODE 3, immediately following a 
refueling outage, only applicable if MODE 2 has not been entered 
following the refueling outage. Required Action A.1 was revised to 
include the turbine-driven AFW addition to Condition A. The amendments 
are consistent with Technical Specifications Task Force (TSTF) Traveler 
TSTF-340-A, Revision 3, ``Allow 7 day Completion Time for a turbine-
driven AFW pump inoperable.''
    Date of issuance: October 23, 2017.
    Effective date: These license amendments are effective as of its 
date of issuance and shall be implemented within 120 days of issuance.
    Amendment Nos.: 295 (Unit 1) and 291 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17257A297; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the renewed licenses and TSs.
    Date of initial notice in Federal Register: April 25, 2017 (82 FR 
19100).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 23, 2017.
    No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment requests: December 15, 2016.
    Brief description of amendments: The amendments revised Technical 
Specification 3.1.2, ``Core Reactivity,'' to revise the Completion 
Times of Required Actions A.1 and A.2 from 72 hours to 7 days. This 
proposed change is consistent with Technical Specifications Task Force 
(TSTF) Traveler TSTF-142-A, Revision 0, ``Increase the Completion Time 
when the Core Reactivity Balance is Not Within Limit.''
    Date of issuance: October 23, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: 296 (Unit 1) and 292 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17261B290; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the Renewed Licenses and Technical Specifications.

[[Page 55418]]

    Date of initial notice in Federal Register: April 11, 2017 (82 FR 
17457).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 23, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment requests: January 11, 2017.
    Brief description of amendments: The amendments modified Technical 
Specification (TS) 3.8.1, ``AC Sources--Operating,'' to allow greater 
flexibility in performing Surveillance Requirements (SRs) by modifying 
Mode restriction notes in TS SRs 3.8.1.8, 3.8.1.11, 3.8.1.16, 3.8.1.17, 
and 3.8.1.19. This proposed change was consistent with Technical 
Specifications Task Force (TSTF) Traveler TSTF-283-A, Revision 3, 
``Modify Section 3.8 Mode Restriction Notes.''
    Date of issuance: October 25, 2017.
    Effective date: These license amendments are effective as of its 
date of issuance and shall be implemented within 120 days of issuance.
    Amendment Nos.: 300 (Unit 1) and 279 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17269A055; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the renewed facility operating licenses and 
technical specifications.
    Date of initial notice in Federal Register: May 23, 2017 (82 FR 
23620).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 25, 2017.
    No significant hazards consideration comments received: Yes. One 
comment from a member of the public was received, however it was not 
related to the no significant hazards consideration determination nor 
the license amendment request.

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

    Date of amendment requests: January 11, 2017.
    Brief description of amendments: The amendments modified Technical 
Specification (TS) 3.1.8, ``PHYSICS TESTS Exceptions,'' to allow the 
numbers of channels required by the Limiting Condition for Operation 
(LCO) section of TS 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' to be reduced from ``4'' to ``3'' to allow one 
nuclear instrumentation channel to be used as an input to the 
reactivity computer for physics testing without placing the nuclear 
instrumentation channel in a tripped condition. This proposed change is 
consistent with Technical Specifications Task Force (TSTF) Traveler 
TSTF-315-A, Revision 0, ``Reduce plant trips due to spurious signals to 
the NIS [Nuclear Instrumentation System] during physics testing.''
    Date of issuance: October 25, 2017.
    Effective date: These license amendments are effective as of their 
date of issuance and shall be implemented within 120 days of issuance.
    Amendment Nos.: 301 (Unit 1) and 280 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17261B218; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the renewed facility operating licenses and 
technical specifications.
    Date of initial notice in Federal Register: May 23, 2017 (82 FR 
23621).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 25, 2017.
    No significant hazards consideration comments received: Yes. One 
comment from a member of the public was received, however it was not 
related to the proposed no significant hazards consideration 
determination or to the license amendment request.

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

    Date of amendment requests: January 11, 2017.
    Brief description of amendments: The amendments modify the limiting 
condition for operation (LCO) Required Action B.2 for Technical 
Specification (TS) 3.4.10, ``Pressurizer Safety Valves,'' LCO Required 
Action C.2 for TS 3.7.4, ``Steam Generator Power Operated Relief Valves 
(SG PORVs),'' and LCO Required Action G.1 for TS 3.4.12, ``Low 
Temperature Overpressure Protection (LTOP) System.'' Specifically, the 
Completion Times are revised from 12 hours to 24 hours for TS LCO 
3.4.10, Required Action B.2, and TS LCO 3.7.4, Required Action C.2; and 
from 8 hours to 12 hours for TS LCO 3.4.12, Required Action G.1. The 
changes are consistent with Technical Specifications Task Force (TSTF) 
Traveler TSTF-352-A, Revision 1, ``Provide Consistent Completion Time 
to Reach MODE 4.''
    Date of issuance: October 31, 2017.
    Effective date: These license amendments are effective as of their 
date of issuance and shall be implemented within 120 days of issuance.
    Amendment Nos.: 302 (Unit 1) and 281 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17269A198; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Renewed Licenses and Technical Specifications.
    Date of initial notice in Federal Register: May 23, 2017 (82 FR 
23622).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 31, 2017.
    No significant hazards consideration comments received: Yes. One 
comment from a member of the public was received, however it was not 
related to the proposed no significant hazards consideration 
determination or to the license amendment request.

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

    Date of amendment requests: January 11, 2017.
    Brief description of amendments: The amendments modify Technical 
Specification (TS) 3.7.5, ``Auxiliary Feedwater (AFW) System,'' 
Limiting Condition for Operation (LCO) Condition A and Required Action 
A.1. The proposed changes modify Condition A to expand the condition to 
include when one turbine driven AFW pump is inoperable in MODE 3. This 
expanded condition is applicable immediately following a refueling 
outage and only if MODE 2 has not been entered. Required Action A.1 is 
revised to state ``affected equipment'' as opposed to ``steam supply'' 
as a result of the addition of the turbine driven AFW pump to Condition 
A. The changes are consistent with Technical Specifications Task Force 
(TSTF) Traveler TSTF-340-A, Revision 3, ``Allow 7 day Completion Time 
for a turbine-driven AFW pump inoperable.''
    Date of issuance: October 31, 2017.
    Effective date: These license amendments are effective as of their 
date of issuance and shall be

[[Page 55419]]

implemented within 120 days of issuance.
    Amendment Nos.: 304 (Unit 1) and 283 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17277A313; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the renewed facility operating licenses and 
technical specifications.
    Date of initial notice in Federal Register: May 23, 2017 (82 FR 
23621).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 31, 2017.
    No significant hazards consideration comments received: Yes. One 
comment from a member of the public was received, however it was not 
related to the proposed no significant hazards consideration 
determination or to the license amendment request.

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

    Date of application for amendments: January 11, 2017.
    Brief description of amendments: The amendments modify Technical 
Specification (TS) Limiting Condition for Operation (LCO) 3.9.6, 
``Residual Heat Removal (RHR) and Coolant Circulation--Low Water 
Level,'' to add a note which allows all RHR pumps to be secured for 
less than or equal to 15 minutes to support the switching of the 
shutdown cooling loops from one train to another. The changes are 
consistent with Technical Specifications Task Force (TSTF) Travelers 
TSTF-349-A, Revision 1, ``Add Note to LCO 3.9.5 Allowing Shutdown 
Cooling Loops Removal from Operation,'' TSTF-361-A, Revision 2, ``Allow 
standby [Shutdown Cooling] SDC/RHR/[Decay Heat Removal] DHR loop to 
[be] inoperable to support testing,'' and TSTF-438-A, Revision 0, 
``Clarify Exception Notes to be Consistent with the Requirement Being 
Excepted.''
    Date of issuance: October 31, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: Unit 1--303; Unit 2--282. A publicly-available 
version is in ADAMS under Accession No. ML17271A034; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Renewed Licenses and Technical Specifications.
    Date of initial notice in Federal Register: May 23, 2017 (82 FR 
23623).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 31, 2017.
    No significant hazards consideration comments received: Yes. One 
comment from a member of the public was received, however it was not 
related to the proposed no significant hazards consideration 
determination or the license amendment request.

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

    Date of amendment request: December 2, 2016, as supplemented by 
letters dated April 25, May 22, and October 2, 2017.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to (1) relocate cycle-specific parameters to the 
Core Operating Limits Report (COLR) consistent with Technical 
Specification Task Force (TSTF)-339, ``Relocate TS Parameters to 
COLR;'' (2) delete duplicate reporting requirements in the 
Administrative Section of TSs consistent with TSTF-5, ``Delete Safety 
Limit Violation Notification Requirements,'' Revision 1; and (3) delete 
reference to plant procedure PLP-6, ``Technical Specification Equipment 
List Program and Core Operating Limits Report,'' in TSs as it pertains 
to the COLR.
    Date of issuance: November 6, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 161. A publicly-available version is in ADAMS under 
Accession No. ML17250A202; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-63: Amendment revised 
the Facility Operating License and TSs.
    Date of initial notice in Federal Register: February 14, 2017 (82 
FR 10595). The supplemental letters dated April 25, May 22, and October 
2, 2017, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 6, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 8, 2016, as supplemented by 
letter dated July 11, 2017.
    Brief description of amendment: The amendment would, on a one-time 
basis, extend the completion time from 7 days to 14 days for the 
Residual Heat Removal Train A subsystem to operable status associated 
with Technical Specification (TS) 3.5.1, ``ECCS [Emergency Core Cooling 
System]--Operating''; TS 3.6.1.5, ``Residual Heat Removal (RHR) Drywell 
Spray''; and TS 3.6.2.3, ``Residual Heat Removal (RHR) Suppression Pool 
Cooling.'' This amendment will be used to support preventive 
maintenance, which replaces the RHR Train A subsystem's pump and motor.
    Date of issuance: October 30, 2017.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 245. A publicly-available version is in ADAMS under 
Accession No. ML17290A127; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-21: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: February 14, 2017 (82 
FR 10596). The supplemental letter dated July 11, 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 October 30, 2017.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket No. 50-277, 
Peach Bottom Atomic Power Station, Unit 2, York and Lancaster Counties, 
Pennsylvania

    Date of amendment request: May 19, 2017, as supplemented by letter 
dated August 29, 2017.

[[Page 55420]]

    Brief description of amendment: The amendment revised the Technical 
Specifications to decrease the number of safety relief valves and 
safety valves required to be operable when operating at a power level 
less than or equal to 3,358 megawatts thermal. This change is 
applicable only to the current Cycle 22 that is scheduled to end in 
October 2018.
    Date of issuance: October 25, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 5 days.
    Amendment No.: 315. A publicly-available version is in ADAMS under 
Accession No. ML17249A151; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-44: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: July 5, 2017 (82 FR 
31094). The supplemental letter dated August 29, 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 October 25, 2017.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 27, 2016, as 
supplemented by the letters dated July 28, 2017, August 30, 2017, and 
October 19, 2017.
    Brief description of amendments: The amendments revised the 
suppression pool swell design analysis. The new analysis utilizes a 
different computer code and incorporates different analysis assumptions 
than the current analysis. The changes are necessary because the 
current design analysis determining the suppression pool swell response 
to a loss-of-coolant accident was determined to be non-conservative.
    These changes to the suppression pool swell design analysis do not 
require any changes to the LSCS Technical Specifications. Changes to 
the LSCS updated final safety analysis report related to changes to the 
suppression pool swell design analysis shall be made in accordance with 
10 CFR 50.71(e) based on the NRC approval of these changes.
    Date of issuance: October 30, 2017.
    Effective date: These license amendments are effective as of the 
date of its issuance and shall be implemented within 60 days from the 
date of issuance.
    Amendment Nos.: 225 for NPF-11 and 211 for NPF-18. A publicly-
available version is in ADAMS under Accession No. ML17257A304; 
documents related to this amendment are listed in the Safety Evaluation 
enclosed with the amendment.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
approved to revise the LSCS updated final safety analysis report 
related to changes to the suppression pool swell design analysis and 
the Licenses.
    Date of initial notice in Federal Register: March 8, 2017 (82 FR 
13022). The supplements dated July 28, 2017, August 30, 2017, and 
October 19, 2017, contained clarifying information and did not change 
the NRC staff's initial proposed finding of no significant hazards 
consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 30, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 13, 2016, as supplemented by 
letter dated February 17, 2017.
    Brief description of amendment: The amendment revised the Nine Mile 
Point, Unit 2, Technical Specification (TS) safety limit (SL) to 
increase the low pressure isolation setpoint allowable value, which 
will result in earlier main steam line isolation. The revised main 
steam line low pressure isolation capability and the revised SL are 
intended to ensure that Nine Mile Point, Unit 2, remains within the TS 
SLs in the event of a pressure regulator failure maximum demand 
transient.
    Date of issuance: October 31, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment No.: 164. A publicly-available version is in ADAMS under 
Accession No. ML17268A263; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-69: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 28, 2017 (82 FR 
15381). The supplemental letter dated February 17, 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 October 31, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 23, 2017, as supplemented by 
letter dated July 3, 2017.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) by limiting the MODE of applicability 
for the Reactor Protection System, Startup, and Operating Rate of 
Change of Power--High, functional unit trip. Additionally, the 
amendments added new Limiting Condition for Operation (LCO) 3.0.5 and 
relatedly modified LCO 3.0.1 and LCO 3.0.2, to provide for placing 
inoperable equipment under administrative control for the purpose of 
conducting testing required to demonstrate OPERABILITY.
    Date of issuance: November 2, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 243 and 194. A publicly-available version is in 
ADAMS under Accession No. ML17257A015; documents related to this 
amendment are listed in the Safety Evaluation enclosed with the 
amendment.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: March 28, 2017 (82 FR 
15383). The supplemental letter dated July 3, 2017, provided additional 
information that expanded the scope of the application as originally 
noticed and changed the NRC staff's original proposed no significant 
hazards consideration (NSHC) determination as published in the Federal 
Register. Accordingly, the NRC published a second proposed no 
significant hazards consideration determination in the Federal Register 
on September 12, 2017

[[Page 55421]]

(82 FR 42849). This notice superseded the original notice in its 
entirety. It also provided an opportunity to request a hearing by 
November 13, 2017, but indicated that if the Commission makes a final 
NSHC determination, any such hearing would take place after issuance of 
the amendments.
    The Commission's related evaluation of the amendments and final 
NSHC are contained in a Safety Evaluation dated November 2, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 21, 2016.
    Brief description of amendments: The amendments modify the 
Technical Specifications by deleting high-range noble gas effluent 
monitors' requirements and relocating the requirements to the Turkey 
Point Offsite Dose Calculation Manual.
    Date of issuance: October 26, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos: 277 and 272. A publicly-available version is in 
ADAMS under Accession No. ML17228A563. Documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: March 14, 2017 (82 FR 
13666).
    The Commission's related evaluation of the amendments is contained 
in a safety evaluation dated October 26, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 11, 2017.
    Brief description of amendments: The amendments request an 
extension to the time to achieve full compliance with 10 CFR 50.48(c), 
National Fire Protection Association (NPFA) 805, from November 6, 2017, 
to the conclusion of the FNP, Unit 1, Spring 2018 Refueling Outage 
(1R28). The amendments update Attachment S, ``Modification and 
Implementation Items''; of the previously approved NFPA-805 amendment.
    Date of issuance: November 1, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 215 (Unit 1) and 212 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17269A166; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-2 and NPF-8: The 
amendments revised the Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: August 29, 2017 (82 FR 
41059).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 1, 2017.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 14th day of November 2017.

    For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2017-25063 Filed 11-20-17; 8:45 am]
 BILLING CODE 7590-01-P