[Federal Register Volume 84, Number 117 (Tuesday, June 18, 2019)]
[Notices]
[Pages 28339-28352]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2019-12573]


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NUCLEAR REGULATORY COMMISSION

[NRC-2019-0135]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to the Atomic Energy Act of 1954, as amended (the 
Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this 
regular biweekly notice. The Act requires the Commission to publish 
notice of any amendments issued, or proposed to be issued, and grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license or combined license, as 
applicable, upon a determination by the Commission that such amendment 
involves no significant hazards consideration, notwithstanding the 
pendency before the Commission of a request for a hearing from any 
person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from May 21, 2019 to June 3, 2019. The last 
biweekly notice was published on June 4, 2019.

DATES: Comments must be filed by July 18, 2019. A request for a hearing 
must be filed by August 19, 2019.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2019-0135. Address 
questions about NRC dockets IDs in Regulations.gov to Jennifer Borges; 
telephone: 301-287-9127; email: [email protected]. For technical 
questions, contact the individual(s) listed in the FOR FURTHER 
INFORMATION CONTACT section of this document.
     Mail comments to: Office of Administration, Mail Stop: 
TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, ATTN: Program Management, Announcements and Editing Staff.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Beverly Clayton, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-3475, email: [email protected].

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2019-0135 facility name, unit 
number(s), plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
     Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2019-0135.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-

[[Page 28340]]

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

B. Submitting Comments

    Please include Docket ID NRC-2019-0135 facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

II. Background

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

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in section 50.92 of title 10 of the Code 
of Federal Regulations (10 CFR), this means that operation of the 
facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment 
period such that failure to act in a timely way would result, for 
example in derating or shutdown of the facility. If the Commission 
takes action prior to the expiration of either the comment period or 
the notice period, it will publish in the Federal Register a notice of 
issuance. If the Commission makes a final no significant hazards 
consideration determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.

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

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and petition for leave to intervene 
(petition) with respect to the action. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested persons should consult a 
current copy of 10 CFR 2.309. The NRC's regulations are accessible 
electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of 
the regulations is available at the NRC's Public Document Room, located 
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. If a petition is filed, the 
Commission or a presiding officer will rule on the petition and, if 
appropriate, a notice of a hearing will be issued.
    As required by 10 CFR 2.309(d) the petition should specifically 
explain the reasons why intervention should be permitted with 
particular reference to the following general requirements for 
standing: (1) The name, address, and telephone number of the 
petitioner; (2) the nature of the petitioner's right under the Act to 
be made a party to the proceeding; (3) the nature and extent of the 
petitioner's property, financial, or other interest in the proceeding; 
and (4) the possible effect of any decision or order which may be 
entered in the proceeding on the petitioner's interest.
    In accordance with 10 CFR 2.309(f), the petition must also set 
forth the specific contentions which the petitioner seeks to have 
litigated in the proceeding. Each contention must consist of a specific 
statement of the issue of law or fact to be raised or controverted. In 
addition, the petitioner must provide a brief explanation of the bases 
for the contention and a concise statement of the alleged facts or 
expert opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to the specific sources and 
documents on which the petitioner intends to rely to support its 
position on the issue. The petition must include sufficient information 
to show that a genuine dispute exists with the applicant or licensee on 
a material issue of law or fact. Contentions must be limited to matters 
within the scope of the proceeding. The contention must be one which, 
if proven, would entitle the petitioner to relief. A petitioner who 
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene. 
Parties have the opportunity

[[Page 28341]]

to participate fully in the conduct of the hearing with respect to 
resolution of that party's admitted contentions, including the 
opportunity to present evidence, consistent with the NRC's regulations, 
policies, and procedures.
    Petitions must be filed no later than 60 days from the date of 
publication of this notice. Petitions and motions for leave to file new 
or amended contentions that are filed after the deadline will not be 
entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in 
accordance with the filing instructions in the ``Electronic Submissions 
(E-Filing)'' section of this document.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to establish when the hearing is held. If the final determination is 
that the amendment request involves no significant hazards 
consideration, the Commission may issue the amendment and make it 
immediately effective, notwithstanding the request for a hearing. Any 
hearing would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of the amendment unless the Commission finds an imminent 
danger to the health or safety of the public, in which case it will 
issue an appropriate order or rule under 10 CFR part 2.
    A State, local governmental body, Federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission no later 
than 60 days from the date of publication of this notice. The petition 
must be filed in accordance with the filing instructions in the 
``Electronic Submissions (E-Filing)'' section of this document, and 
should meet the requirements for petitions set forth in this section, 
except that under 10 CFR 2.309(h)(2) a State, local governmental body, 
or Federally-recognized Indian Tribe, or agency thereof does not need 
to address the standing requirements in 10 CFR 2.309(d) if the facility 
is located within its boundaries. Alternatively, a State, local 
governmental body, Federally-recognized Indian Tribe, or agency thereof 
may participate as a non-party under 10 CFR 2.315(c).
    If a hearing is granted, any person who is not a party to the 
proceeding and is not affiliated with or represented by a party may, at 
the discretion of the presiding officer, be permitted to make a limited 
appearance pursuant to the provisions of 10 CFR 2.315(a). A person 
making a limited appearance may make an oral or written statement of 
his or her position on the issues but may not otherwise participate in 
the proceeding. A limited appearance may be made at any session of the 
hearing or at any prehearing conference, subject to the limits and 
conditions as may be imposed by the presiding officer. Details 
regarding the opportunity to make a limited appearance will be provided 
by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing and petition for leave to intervene (petition), any 
motion or other document filed in the proceeding prior to the 
submission of a request for hearing or petition to intervene, and 
documents filed by interested governmental entities that request to 
participate under 10 CFR 2.315(c), must be filed in accordance with the 
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 
46562; August 3, 2012). The E-Filing process requires participants to 
submit and serve all adjudicatory documents over the internet, or in 
some cases to mail copies on electronic storage media. Detailed 
guidance on making electronic submissions may be found in the Guidance 
for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit 
paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [email protected], or by 
telephone at 301-415-1677, to (1) request a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign submissions and access the E-Filing 
system for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a petition or 
other adjudicatory document (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public website at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a 
digital ID certificate and a docket has been created, the participant 
can then submit adjudicatory documents. Submissions must be in Portable 
Document Format (PDF). Additional guidance on PDF submissions is 
available on the NRC's public website at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the 
time the document is submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
document on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before adjudicatory documents are 
filed so that they can obtain access to the documents via the E-Filing 
system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic 
Filing Help Desk through the ``Contact Us'' link located on the NRC's 
public website at http://www.nrc.gov/site-help/e-submittals.html, by 
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m. 
and 6 p.m., Eastern Time, Monday through Friday, excluding government 
holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper

[[Page 28342]]

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.

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

    Date of amendment request: March 27, 2019. A publicly available 
version is in ADAMS under Accession No. ML19086A315.
    Description of amendment request: The proposed amendment would 
remove License Condition 2.C.(11), ``Shield Wall Deferral (Section 
12.3.2, SSER #4, License Amendment #7)'' and its related Attachment 3, 
``List of Shield Walls'' from Columbia's Renewed Facility Operating 
License, as these items are outdated and no longer applicable to 
Columbia's operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment involves the removal of an outdated 
license condition. The proposed amendment does not impact any 
accident initiators, analyzed events, or assumed mitigation of 
accident or transient events. The proposed change does not involve 
the addition or removal of any equipment or any design changes to 
the facility. The proposed change does not affect any plant 
operations, design functions, or analyses that verify the capability 
of structures, systems, and components (SSCs) to perform a design 
function. The proposed change does not change any of the accidents 
previously evaluated in the Final Safety Analysis Report (FSAR). The 
proposed change does not affect SSCs, operating procedures, and 
administrative controls that have the function of preventing or 
mitigating any of these accidents.
    Therefore, the proposed change does 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 amendment only involves the removal of an outdated 
license condition. No actual plant equipment or accident analyses 
will be affected by the proposed change. The proposed change will 
not change the design function or operation of any SSCs. The 
proposed change will not result in any new failure mechanisms, 
malfunctions, or accident initiators not considered in the design 
and licensing bases. The proposed amendment does not impact any 
accident initiators, analyzed events, or assumed mitigation of 
accident or transient events.
    Therefore, this proposed change does not create the possibility 
of an accident of a new or different kind than previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment only involves the removal of an outdated 
license condition. The proposed change does not involve any physical 
changes to the plant or alter the manner in which plant systems are 
operated, maintained, modified, tested, or inspected. The proposed 
change does not alter the manner in which safety limits, limiting 
safety system settings or limiting conditions for operation are 
determined. The safety analysis acceptance criteria are not affected 
by this change. The proposed change will not result in plant 
operation in a configuration outside the design basis. The proposed 
change does not adversely affect systems that respond to safely 
shutdown the plant and to maintain the plant in a safe shutdown 
condition.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247, and 50-
286, Indian Point Nuclear Generating (Indian Point) Unit Nos. 1, 2, and 
3, Westchester County, New York

    Date of amendment request: April 15, 2019. A publicly available 
version is in ADAMS under Accession No. ML19105B278.
    Description of amendment request: The amendments would revise the 
Indian Point Site Emergency Plan (SEP) for the permanently shutdown and 
defueled condition.
    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 IPEC [Indian Point Energy Center] 
SEP do not impact the

[[Page 28343]]

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 
[Emergency Response Organization] 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 IPEC 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?
    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 
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 IPEC 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 IPEC 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: Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: James G. Danna.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 and 3 (Indian Point 2 and 3 or IP2 
and IP3), Westchester County, New York

    Date of amendment request: April 15, 2019. A publicly available 
version is in ADAMS under Accession No. ML19105B236.
    Description of amendment request: The amendments propose changes to 
the staffing and training requirements for the Indian Point staff 
contained in Section 5.0, ``Administrative Controls,'' of the Indian 
Point 2 and Indian Point 3 Technical Specifications (TSs). Additional 
changes are also proposed to Section 1.1, ``Definitions''; Section 4.0, 
``Design Features''; and Section 5.0, ``Administrative Controls,'' that 
are no longer applicable to a permanently defueled facility once Indian 
Point 2, and subsequently Indian Point 3, are permanently defueled.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would not take effect until IP2 has 
permanently ceased operation and entered a permanently defueled 
condition and the Certified Fuel Handler Training and Retraining 
Program is approved by the NRC. The proposed amendment would modify 
the IP2 TS by deleting the portions of the TS that are no longer 
applicable to a permanently defueled facility, while modifying the 
other sections to correspond to the permanently defueled condition.
    The deletion and modification of provisions of the 
administrative controls do not directly affect the design of 
structures, systems, and components (SSCs) necessary for safe 
storage of irradiated fuel or the methods used for handling and 
storage of such fuel in the spent fuel pool. The changes to the 
administrative controls are administrative in nature and do not 
affect any accidents applicable to the safe management of irradiated 
fuel or the permanently shutdown and defueled condition of the 
reactor. Thus, the consequences of an accident previously evaluated 
are not increased.
    In a permanently defueled condition, the only credible accidents 
are the fuel handling accident (FHA) and those involving radioactive 
waste systems remaining in service. The probability of occurrence of 
previously evaluated accidents is not increased, because extended 
operation in a defueled condition will be the only operation 
allowed. This mode of operation is bounded by the existing analyses. 
Additionally, the occurrence of postulated accidents associated with 
reactor operation is no longer credible in a permanently defueled 
reactor. This significantly reduces the scope of applicable 
accidents.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes have no impact on facility SSCs affecting 
the safe storage of irradiated fuel, or on the methods of operation 
of such SSCs, or on the handling and storage of irradiated fuel 
itself. The administrative removal or modifications of the TS that 
are related only to administration of the facility cannot result in 
different or more adverse failure modes or accidents than previously 
evaluated because the reactor will be permanently shutdown and 
defueled and IP2 will no longer be authorized to operate the reactor 
or retain or place fuel in the reactor vessel.
    The proposed changes to the IP2 TS do not affect systems 
credited in the accident analysis for the FHA or radioactive waste 
system upsets at IP2. The proposed TS will continue to require 
proper control and monitoring of safety significant parameters and 
activities.
    The proposed amendment does not result in any new mechanisms 
that could initiate damage to the remaining relevant safety barriers 
for defueled plants (fuel cladding and spent fuel cooling). Extended 
operation in a defueled condition will be the only operation 
allowed, and it is bounded by the existing analyses, such a 
condition does not create the possibility of a new or different kind 
of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Since the 10 CFR part 50 license for IP2 will no longer 
authorize operation of the reactor or emplacement or retention of 
fuel into the reactor vessel once the certifications required by 10 
CFR 50.82(a)(1) are docketed, as specified in 10 CFR 50.82(a)(2), 
the occurrence of postulated accidents associated with reactor 
operation is no longer credible. The only remaining credible 
accidents are a FHA and those involving radioactive waste systems 
remaining in service. The proposed amendment does not adversely 
affect the inputs or assumptions of any of the design basis analyses 
that impact these analyzed conditions.
    The proposed changes are limited to those portions of the TS 
that are not related to the safe storage of irradiated fuel. The 
requirements that are proposed to be revised or deleted from the IP2 
TS are not credited in the existing accident analysis for the 
remaining applicable postulated accident;

[[Page 28344]]

and as such, do not contribute to the margin of safety associated 
with the accident analysis. Postulated design basis accidents 
involving the reactor are no longer possible because the reactor 
will be permanently shutdown and defueled and IP2 will no longer be 
authorized to operate the reactor or retain or place fuel in the 
reactor vessel.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2 (Indian Point 2 or IP2), Westchester 
County, New York

    Date of amendment request: April 15, 2019. A publicly available 
version is in ADAMS under Accession No. ML19105B241.
    Description of amendment request: The amendment would revise the 
Indian Point 2 Operating License (OL) and revise the Technical 
Specifications (TSs) in Appendix A to Permanently Defueled TSs, the 
Environmental TS Requirements in Appendix B of the OL, and the Inter-
Unit Transfer TSs in Appendix C. The proposed changes would revise 
certain requirements contained within the Indian Point 2 OL and 
Appendices A through C TSs and remove the requirements that would no 
longer be applicable after Indian Point 2 is permanently shut down and 
defueled.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would not take effect until IP2 has 
permanently ceased operation, entered a permanently defueled 
condition, met the decay requirements established in the analysis of 
the Fuel Handling Accident (FHA), implemented NRC approved License 
Amendments regarding fuel storage requirements and administrative 
controls for the permanently defueled condition, and received NRC 
approval of the Certified Fuel Handler Training and Retraining 
Program. The proposed amendment would modify the IP2 OL and TSs in 
Appendices A through C by deleting the portions of the OL and TSs 
that are no longer applicable to a permanently defueled facility, 
while modifying other portions to correspond to the permanently 
defueled condition. These proposed changes are consistent with the 
criteria set forth in 10 CFR 50.36 for the contents of TSs.
    Section 14 of the IP2 Updated Final Safety Analysis Report 
(UFSAR) describes the DBA [design-basis accident] and transient 
scenarios applicable to IP2 during power operations. After the 
reactor is in a permanently defueled condition, the spent fuel pit 
(SFP) and its cooling systems will be dedicated only to spent fuel 
storage. In this condition, the spectrum of credible accidents will 
be much smaller than for an operational plant. After the 
certifications are docketed for IP2 in accordance with 10 CFR 
50.82(a)(1), and the consequent removal of authorization to operate 
the reactor or to place or retain fuel in the reactor vessel in 
accordance with 10 CFR 50.82(a)(2), the majority of the accident 
scenarios previously postulated in the UFSAR will no longer be 
possible and will be removed from the UFSAR under the provisions of 
10 CFR 50.59.
    The deletion of TS definitions and rules of usage and 
application requirements that will not be applicable in a defueled 
condition has no impact on facility structures, systems, and 
components (SSCs) or the methods of operation of such SSCs. The 
deletion of design features and safety limits not applicable to the 
permanently shut down and defueled status of IP2 has no impact on 
the remaining applicable DBAs.
    The removal of LCOs [limiting conditions for operation] or SRs 
[surveillance requirements] that are related only to the operation 
of the nuclear reactor or only to the prevention, diagnosis, or 
mitigation of reactor-related transients or accidents do not affect 
the applicable DBAs previously evaluated since these DBAs are no 
longer applicable in the permanently defueled condition. The safety 
functions involving core reactivity control, reactor heat removal, 
reactor coolant system (RCS) inventory control, and containment 
integrity are no longer applicable at IP2 as a permanently shut down 
and defueled facility. The analyzed accidents involving damage to 
the RCS, main steam lines, reactor core, and the subsequent release 
of radioactive material will no longer be possible at IP2.
    After IP2 permanently ceases operation, the future generation of 
fission products will cease and the remaining source term will 
decay. The radioactive decay of the irradiated fuel following shut 
down of the reactor will have reduced the consequences of the FHA 
below those previously analyzed.
    The SFP water level, boron concentration, and fuel storage TSs 
are retained to preserve the current requirements for safe storage 
of irradiated fuel. SFP cooling and make-up related equipment and 
support equipment (e.g., electrical power systems) are not required 
to be continuously available since there will be sufficient time to 
effect repairs, establish alternate sources of make-up flow, or 
establish alternate sources of cooling in the event of a loss of 
cooling and make-up flow to the SFP.
    The deletion and modification of provisions of the 
administrative controls of the Appendix A TSs and the non-
radiological environmental protection requirements in Appendix B do 
not directly affect the design of SSCs necessary for safe storage of 
irradiated fuel or the methods used for handling and storage of such 
fuel in the SFP. The changes do not affect any accidents applicable 
to the safe management of irradiated fuel or the permanently shut 
down and defueled condition of the reactor.
    The probability of occurrence of previously evaluated accidents 
is not increased, since extended operation in a defueled condition 
will be the only operation allowed, and therefore bounded by the 
existing analyses. Additionally, the occurrence of postulated 
accidents associated with reactor operation will no longer be 
credible in a permanently defueled reactor. This significantly 
reduces the scope of applicable accidents.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the IP2 OL and Appendices A through C 
TSs have no impact on facility SSCs affecting the safe storage of 
irradiated fuel, or on the methods of operation of such SSCs, or on 
the handling and storage of irradiated fuel itself. The removal of 
TSs that are related only to the operation of the nuclear reactor or 
only to the prevention, diagnosis, or mitigation of reactor-related 
transients or accidents, cannot result in different or more adverse 
failure modes or accidents than previously evaluated because the 
reactor will be permanently shut down and defueled and IP2 will no 
longer be authorized to operate the reactor.
    The proposed deletion and modification of requirements of the 
IP2 OL and Appendices A through C TSs do not affect systems credited 
in the accidents that remain applicable at IP2 in the permanently 
defueled condition. The proposed OL and TSs will continue to require 
proper control and monitoring of safety significant parameters and 
activities.
    The Appendix A TSs regarding SFP water level, boron 
concentration, and fuel storage are retained to preserve the current 
requirements for safe storage of irradiated fuel. The restriction on 
the SFP water level is fulfilled by normal operating conditions and 
preserves initial conditions assumed in the analyses of the 
postulated DBA.
    The proposed amendment does not result in any new mechanisms 
that could initiate damage to the remaining relevant safety barriers 
for defueled plants (fuel cladding and spent fuel cooling). Since 
extended operation in a defueled condition will be the only 
operation allowed, and therefore

[[Page 28345]]

bounded by the existing analyses, such a condition does not create 
the possibility of a new or different kind of accident.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Because the 10 CFR part 50 license for IP2 will no longer 
authorize operation of the reactor or emplacement or retention of 
fuel in the reactor vessel after the certifications required by 10 
CFR 50.82(a)(1) are docketed for IP2 as specified in 10 CFR 
50.82(a)(2), the occurrence of postulated accidents associated with 
reactor operation are no longer credible. The only remaining 
credible accidents are the FHA and the accidental release of waste 
liquids or waste gas. The proposed amendment does not adversely 
affect the inputs or assumptions of any of the design basis analyses 
that impact the remaining DBAs.
    The proposed amendment would modify the IP2 OL and TSs in 
Appendices A through C by deleting the portions of the OL and TSs 
that are no longer applicable to a permanently defueled facility, 
while modifying other portions to correspond to the permanently 
defueled condition. The requirements that are proposed to be deleted 
from the IP2 OL and Appendix A TSs are not credited in the existing 
accident analyses for the remaining DBAs; and as such, do not 
contribute to the margin of safety associated with the accident 
analyses. Postulated DBAs involving the reactors will no longer be 
possible because the reactor will be permanently shut down and 
defueled and IP2 will no longer be authorized to operate the 
reactor.
    The Appendix A TSs regarding SFP water level, boron 
concentration, and fuel storage are retained to preserve the current 
requirements for safe storage of irradiated fuel.
    Therefore, the proposed amendment does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    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: April 26, 2019. A publicly available 
version is in ADAMS under Accession No. ML19116A196.
    Description of amendment request: The amendments would revise Peach 
Bottom Atomic Power Station, Units 2 and 3, Technical Specification 
(TS) 3.8.1, ``AC [Alternating Current] Power--Operating,'' Required 
Action A.3, to provide a temporary one-time extension of the completion 
time to allow sufficient time to perform physical modifications to 
replace 27 inaccessible electrical cables. These electrical cables are 
reaching the end of their dependable service life and are in need of 
replacement.
    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 temporary one-time change to extend the Completion 
Time for TS 3.8.1, Required Action A.3, will not increase the 
probability of an accident, since the proposed Completion Time 
extension in the time duration that one qualified offsite circuit is 
out of service has no direct physical impact on the plant. The 
proposed inoperable offsite circuit limits the available redundancy 
of the offsite electrical system to a period not to exceed 21 days. 
Therefore, the proposed TS change does not have a direct impact on 
the plant that would make an accident more likely to occur due to 
extended Completion Time. Other sources of offsite and onsite power 
remain available.
    During transients or events which require these systems/
subsystems to be operating, there is sufficient capacity in the 
operable systems/subsystems to support plant operation or shutdown. 
Therefore, failures that are accident initiators will not occur more 
frequently than previously postulated as a result of the proposed 
temporary one-time TS change.
    In addition, the consequences of an accident previously 
evaluated in the Updated Final Safety Analysis Report (UFSAR) will 
not be increased. With one offsite circuit inoperable, the 
consequences of any postulated accidents occurring on Unit 2 or Unit 
3 during the proposed one-time Completion Time extension are bounded 
by the previous analyses as described in the UFSAR. The minimum 
equipment required to mitigate the consequences of an accident and/
or safely shut down the plant will be operable or available during 
the extended Completion Time period of 21 days.
    A risk evaluation has also been performed for the temporary one-
time 21-day Completion Time extension. The evaluation concluded that 
the probability of a Loss of Offsite Power (LOOP) for the proposed 
configuration is very low. Therefore, the proposed change does not 
significantly increase the probability of an accident previously 
evaluated because: (a) The emergency buses continue to be fed from a 
reliable offsite source and; (b) the effect of the proposed 
configuration on the probability of a LOOP is very low.
    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 possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed temporary one-time change to extend the Completion 
Time for TS 3.8.1, Required Action A.3, will not create the 
possibility of a new or different type of accident since it will 
only extend the time period that one of the offsite circuits can be 
out of service; the extension of the time duration for one offsite 
circuit being inoperable has no direct physical impact on the plant 
and does not create any new accident initiators. Other sources of 
offsite and onsite power remain available. The systems involved are 
accident mitigation systems. The possible impacts that the 
inoperable equipment may have on supported systems was previously 
analyzed in the UFSAR. The impact of inoperable support systems was 
also previously assessed, and any accident initiators created by the 
inoperable systems were evaluated. Extending the duration of the 
Completion Time does not create any additional accident initiators 
for the plant.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The existing TS Completion Time limit of seven (7) days for one 
offsite circuit inoperable was established to ensure that sufficient 
safety-related equipment is available for response to all accident 
conditions and that sufficient decay heat removal capability is 
available for a Loss of Coolant Accident (LOCA) coincident with a 
LOOP on one unit and simultaneous safe shutdown of the other unit. 
Although a very slight reduction in the margin of safety might be 
incurred during the proposed one-time extended Completion Time 
period, this slight reduction is judged to be minimal due to the low 
probability of an event occurring during the extended period. Other 
sources of offsite and onsite power remain available and operable 
during the 21-day extended period along with maintaining the 
availability of essential Emergency Core Cooling System (ECCS)/decay 
heat removal capability. The very slight reduction in the margin of 
safety resulting from extending the Completion Time from seven (7) 
days to 21 days when an offsite circuit is inoperable is not 
considered significant, since the remaining operable offsite 
circuit, the emergency Diesel Generators (DGs), the Station Blackout 
(SBO) line, and the FLEX DGs are available and provide an effective 
defense-in-depth plan to support the station electrical plant 
configurations during the extended 21-day Completion Time period.

[[Page 28346]]

    The proposed TS change to extend the Completion Time does not 
affect the acceptance criteria for any analyzed event, nor is there 
a change to any safety limit. The proposed TS change does not affect 
any Structures, Systems or Components (SSC) or their capability to 
perform their intended functions. The proposed change does not alter 
the manner in which safety limits, limiting safety system settings, 
or limiting conditions for operation are determined. Neither the 
safety analyses nor the safety analysis acceptance criteria are 
affected by this change. The proposed change will not result in 
plant operation in a configuration outside the current design basis. 
The margin of safety is maintained by maintaining the capability to 
supply emergency buses with a redundant, separate, reliable offsite 
power source, and maintaining the onsite power sources in their 
design basis configuration.
    Operations personnel are fully qualified and trained to respond 
to, and mitigate, a Design Basis Accident (DBA), including actions 
needed to ensure decay heat removal systems are available while 
PBAPS [Peach Bottom Atomic Power Station], Units 2 and 3, are in the 
operational electrical configurations described within this 
submittal. Accordingly, existing procedures are in place that 
address safe plant shutdown and decay heat removal for situations 
applicable during the extended one-time Completion Time period.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: June 26, 2018, as supplemented by 
letters dated February 25, 2019, and May 17, 2019 (ADAMS Accession Nos. 
ML18177A044, ML19056A387, and ML19137A070, respectively).
    Description of amendment request: The license amendment request was 
originally noticed in the Federal Register on December 18, 2018 (83 FR 
64894). This notice is being reissued in its entirety to include a 
revised description of the amendment request. The amendment would 
modify Technical Specification 3.3.1, ``Oxygen Concentration,'' to 
require inerting the primary containment to less than 4 percent by 
volume oxygen concentration within 72 hours of entering power operating 
condition. Also, the amendment would add a new requirement to identify 
required actions if the primary containment oxygen concentration 
increases to greater than or equal to four volume percent while in the 
power operating condition.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change modifies the Technical Specifications (TS) 
by adopting containment inerting and de-inerting requirements that 
are consistent with the guidance of NUREG-1433, ``Standard Technical 
Specifications--General Electric BWR/4 Plants, Volume 1, Revision 
4.0,'' published April 2012. The proposed change will allow inerting 
of the primary containment within 24 hours of exceeding 15 percent 
(%) Rated Thermal Power (RTP), and de-inerting 24 hours prior to 
reducing reactor power to less than or equal to 15% RTP. Also, a new 
TS condition will be added to identify required actions if the 
primary containment oxygen concentration increases to greater than 
or equal to 4% by volume while in the power operating condition. The 
proposed change does not alter the physical configuration of the 
plant, nor does it affect any previously analyzed accident 
initiators. The accident analysis assumes that a Loss of Coolant 
Accident (LOCA) occurs at 100% RTP. The consequences of a LOCA at 
less than or equal to 15% RTP would be much less severe, and produce 
less hydrogen than a LOCA at 100% RTP.
    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 adopts the STS [Standard Technical 
Specifications] guidance regarding containment inerting/de-inerting 
requirements. The proposed change introduces no new mode of plant 
operation and does not involve any physical modification to the 
plant. The proposed change is consistent with the current safety 
analysis assumptions. No setpoints are being changed which would 
alter the dynamic response of plant equipment. Accordingly, no new 
failure modes are introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises the Applicability presentation of 
the Oxygen Concentration TS. No safety limits are affected. The 
Oxygen Concentration TS requirements assure sufficient safety 
margins are maintained, and that the design, operation, surveillance 
methods, and acceptance criteria specified in applicable codes and 
standards (or alternatives approved for use by the NRC) will 
continue to be met as described in the plants' licensing basis. The 
proposed change does not adversely affect existing plant safety 
margins or the reliability of the equipment assumed to operate in 
the safety analysis. As such, there are no changes being made to 
safety analysis assumptions, safety limits, or limiting safety 
system settings that would adversely affect plant safety.
    Therefore, the proposed change does not result in a significant 
reduction in a margin of safety.

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

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: April 26, 2019. A publicly available 
version is in ADAMS under Accession No. ML19119A249.
    Description of amendment request: The amendment request proposes 
changes to the Combined License (COL) Numbers NPF-91 and NPF-92 for 
VEGP, Units 3 and 4, and Updated Final Safety Analysis Report (UFSAR) 
in the form of departures from the incorporated plant-specific Design 
Control Document Tier 2 * and Tier 2 information related to the design-
specific pre-operational Automatic Depressurization System (ADS) 
Blowdown Test. The requested amendment involves changes to credit the 
previously completed ADS Blowdown first three plant tests as described 
in the licensing basis documents, including COL Condition 2.D.(2)(a). 
Specifically, the proposed change would revise the COL, License 
Condition 2.D.(2)(a)2, by removing the requirement to perform the ADS 
Slowdown first three plant test during preoperational testing.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 28347]]

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 changes made by the Nuclear Regulatory Commission 
shown in square brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not affect the operation of any systems 
or equipment that initiates an analyzed accident or alter any 
structures, systems, or components (SSC) accident initiator or 
initiating sequence of events. The proposed changes remove the 
requirement to perform the ADS Blowdown first three plant test based 
on the successful completion of the tests at the lead AP1000 units. 
The change does not adversely affect any methodology which would 
increase the probability or consequences of a previously evaluated 
accident.
    The change does not impact the support, design, or operation of 
mechanical or fluid systems. There is no change to plant systems or 
the response of systems to postulated accident conditions. There is 
no change to predicted radioactive releases due to normal operation 
or postulated accident conditions. The plant response to previously 
evaluated accidents or external events is not adversely affected, 
nor does the proposed change create any new accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of a previously 
evaluated accident.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not affect the operation of any systems 
or equipment that may initiate a new or different kind of accident, 
or alter any SSC such that a new accident initiator or initiating 
sequence of events is created.
    The proposed change credits previously completed ADS Blowdown 
first three plant testing based on the successful completion of the 
tests at the lead AP1000 units. The proposed changes do not 
adversely affect any design function of any SSC design functions or 
methods of operation in a manner that results in a new failure mode, 
malfunction, or sequence of events that affect safety-related or 
non-safety-related equipment. This activity does not allow for a new 
fission product release path, result in a new fission product 
barrier failure mode, or create a new sequence of events that result 
in significant fuel cladding failures.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change maintains existing safety margin and 
provides adequate protection through continued application of the 
existing requirement in the UFSAR. The proposed change satisfies the 
same design functions in accordance with the same codes and 
standards as stated in the UFSAR. This change does not adversely 
affect any design code, function, design analysis, safety analysis 
input or result, or design/safety margin. No safety analysis or 
design basis acceptance limit/criterion is challenged or exceeded by 
the proposed change.
    Since no safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by this change, no significant 
margin of safety is reduced.
    Therefore, the proposed 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.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: May 1, 2019. A publicly available 
version is in ADAMS under Accession No. ML19126A309.
    Description of amendment request: The amendments would revise the 
South Texas Project, Units 1 and 2, Technical Specifications in Section 
3.0 and Section 4.0 regarding limiting condition for operation (LCO) 
and surveillance requirement (SR) usage. The proposed changes are 
consistent with the NRC-approved Technical Specifications Task Force 
(TSTF) Traveler TSTF-529, ``Clarify Use and Application Rules,'' using 
the consolidated line item improvement process (ADAMS Accession No. 
ML16062A271). The model safety evaluation was approved by the NRC in a 
letter dated April 21, 2016 (ADAMS Accession No. ML16060A441).
    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 to Technical Specification LCO 3.0.4 has no 
effect on the requirement for systems to be Operable and has no 
effect on the application of Technical Specification actions. The 
proposed change to Technical Specification SR 4.0.3 states that the 
allowance may only be used when there is a reasonable expectation 
the surveillance will be met when performed. Since the proposed 
change does not significantly affect system Operability, the 
proposed change will have no significant effect on the initiating 
events for accidents previously evaluated and will have no 
significant effect on the ability of the systems to mitigate 
accidents previously evaluated.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to the Technical Specifications usage rules 
does not affect the design or function of any plant systems. The 
proposed change does not change the Operability requirements for 
plant systems or the actions taken when plant systems are not 
Operable.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change clarifies the application of Technical 
Specification LCO 3.0.4 and does not result in changes in plant 
operation. Technical Specification SR 4.0.3 is revised to allow 
application of Technical Specification SR 4.0.3 when a Surveillance 
Requirement has not been previously performed if there is reasonable 
expectation that the Surveillance Requirement will be met when 
performed. This expands the use of Technical Specification SR 4.0.3 
while ensuring the affected system is capable of performing its 
safety function. As a result, plant safety is either improved or 
unaffected.
    Therefore, it is concluded that this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Kym Harshaw, Vice President and General 
Counsel, STP Nuclear Operating Company, P.O. Box 289, Wadsworth, TX 
77483.
    NRC Branch Chief: Robert J. Pascarelli.

[[Page 28348]]

Tennessee Valley Authority (TVA), Docket Nos. 50-390 and 50-391, Watts 
Bar Nuclear Plant (WBN), Units 1 and 2, Rhea County, Tennessee

    Date of amendment request: November 26, 2018, as supplemented by 
letter dated May 13, 2019. Publicly-available versions are in ADAMS 
under Accession Nos. ML18331A134 and ML19134A233, respectively.
    Description of amendment request: The amendments would revise 
technical specifications (TSs) to support performance of 6.9 kiloVolt 
and associated 480 Volt shutdown board (SDBD) maintenance. The proposed 
changes provide operational flexibility for two-unit operation by 
providing sufficient time to perform preventive maintenance on SDBDs 
associated with a defueled unit while the opposite unit is operating in 
Modes 1, 2, 3, or 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 change involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The proposed change modifies the required actions for the 
opposite unit's onsite and offsite AC power sources and electrical 
distribution system. The opposite unit's AC power sources and 
electrical distribution system are required to be operable to 
support the associated unit's required features. In addition, a 
change is proposed to remove the details regarding the required 
input power to the vital inverters. This change will not affect the 
probability of an accident, since the AC power sources, vital 
inverters, and electrical distribution system are not initiators of 
any accident sequence analyzed in the WBN dual-unit Updated Final 
Safety Analysis Report (UFSAR). Rather, the AC power sources, vital 
inverters, and electrical distribution system support equipment used 
to mitigate accidents. The consequences of an analyzed accident will 
not be significantly increased since the minimum requirements for AC 
power sources, vital inverters, and electrical distribution system 
will be maintained to ensure the availability of the required power 
to mitigate accidents assumed in the UFSAR. Operation in accordance 
with the proposed TS will ensure that sufficient AC power sources, 
vital inverters, and electrical distribution subsystems are 
operable, as required to support the unit's required features. 
Therefore, the mitigating functions supported by the AC power 
sources, vital inverters, and electrical distribution system will 
continue to provide the protection assumed by the accident analysis. 
The integrity of fission product barriers, plant configuration, and 
operating procedures as described in the UFSAR will not be affected 
by the proposed changes. Thus, the consequences of previously 
analyzed accidents will not increase by implementing these changes. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No
    The proposed changes involve restructuring the TS for the AC 
electrical power system to provide more flexibility in performing 
maintenance on electrical system components. The AC electrical power 
system is not an initiator to any accident sequence analyzed in the 
UFSAR. Rather, the AC electrical power system supports equipment 
used to mitigate accidents. The proposed changes to modify the 
required actions associated with inoperable opposite unit AC power 
sources and shutdown boards and proposed changes to the details of 
the required power supplies to the vital inverters will maintain the 
same level of equipment performance required for mitigating 
accidents assumed in the UFSAR. Therefore, operation of the facility 
in accordance with this proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The equipment margins will be maintained in 
accordance with the plant-specific design bases as a result of the 
proposed changes. 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 AC capability to support operation of mitigation 
equipment is ensured. The equipment fed by the AC electrical sources 
will continue to provide adequate power to safety-related loads in 
accordance with analysis assumptions. The proposed TS changes 
maintain the same level of equipment performance stated in the UFSAR 
and the current TSs. Therefore, the proposed changes do not involve 
a significant reduction 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.

Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar 
Nuclear Plant, Units 1 and 2, Rhea County, Tennessee

    Date of amendment request: November 29, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18334A389.
    Description of amendment request: The amendments would modify the 
Watts Bar Nuclear Plant, Units 1 and 2, Technical Specification 
requirements related to direct current (DC) electrical systems to be 
consistent with Technical Specifications Task Force (TSTF) Traveler 
TSTF-500, Revision 2, ``DC Electrical Rewrite--Update to TSTF-360'' 
(ADAMS Accession No. ML092670242).
    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 any accident previously 
evaluated?
    Response: No.
    The proposed changes restructure the Technical Specifications 
(TS) for the direct current (DC) electrical power system and are 
consistent with TSTF-500, Revision 2. The proposed changes modify TS 
Actions relating to battery and battery charger inoperability. The 
DC electrical power system, including associated battery chargers, 
is not an initiator of any accident sequence analyzed in the Final 
Safety Analysis Report (FSAR). Rather, the DC electrical power 
system supports equipment used to mitigate accidents. The proposed 
changes to restructure TS and change surveillances for batteries and 
chargers to incorporate the updates included in TSTF-500, Revision 
2, will maintain the same level of equipment performance required 
for mitigating accidents assumed in the FSAR. Operation in 
accordance with the proposed TS would ensure that the DC electrical 
power system is capable of performing its specified safety function 
as described in the FSAR. Therefore, the mitigating functions 
supported by the DC electrical power system will continue to provide 
the protection assumed by the analysis. The relocation of preventive 
maintenance surveillances, and certain operating limits and actions, 
to a 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 
that are 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

[[Page 28349]]

associated with the DC electrical power system.
    The integrity of fission product barriers, plant configuration, 
and operating procedures as described in the FSAR 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 changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed changes involve 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 FSAR. Rather, the DC electrical power 
system supports equipment used to mitigate accidents. The proposed 
changes to restructure the TS and change surveillances for batteries 
and chargers to incorporate the updates included in TSTF-500, 
Revision 2, will maintain the same level of equipment performance 
required for mitigating accidents assumed in the FSAR. 
Administrative and mechanical controls are in place to ensure the 
design and operation of the DC systems continues to meet the plant 
design basis described in the FSAR. Therefore, operation of the 
facility in accordance with this proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The equipment margins will be maintained in 
accordance with the plant-specific design bases because of the 
proposed changes. 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. TS 
changes made in accordance with TSTF-500, Revision 2, maintain the 
same level of equipment performance stated in the FSAR and the 
current TSs. Therefore, the proposed changes do not involve a 
significant reduction [in the margin] of safety.

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

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: March 18, 2019. A publicly available 
version is in ADAMS under Accession No. ML19086A113.
    Description of amendment request: The amendments would revise 
Renewed Facility Operating License Nos. NPF-4 and NPF-7 for the North 
Anna Power Station, Units 1 and 2, respectively, by revising the 
Technical Specification (TS) requirements regarding the Emergency 
Diesel Generators. Specifically, TS 3.8.1, ``AC Sources--Operating,'' 
would be revised to reduce the maximum voltage specified in the 
associated surveillance 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 change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Modifying the maximum steady-state voltage requirement does not 
increase the probability of an accident. Verifying proper operation 
of the EDGs to maintain adequate voltage ensures proper electrical 
and mechanical system function and does not increase the 
consequences of an 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 would provide more restrictive acceptance 
criteria to be applied to existing technical specification 
surveillance tests that demonstrate the capability of the facility 
EDGs to perform their design function. The proposed acceptance 
criteria changes would not create any new failure mechanisms, 
malfunctions, or accident initiators not considered in the design 
and licensing bases. Therefore, the possibility of a new or 
different kind of accident from any previously evaluated has not 
been created.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change involves decreasing maximum voltage test 
acceptance criterion for EDG Surveillance Tests. The conduct of 
surveillance tests on safety-related plant equipment is a means of 
assuring that the equipment is capable of maintaining the margin of 
safety established in the safety analyses for the facility. The 
proposed amendment does not affect EDG performance as described in 
the design basis analyses, including the capability of the EDG to 
maintain required voltage for proper operation of plant safety 
loads. The proposed amendment does not introduce changes to limits 
established in the accident analyses. 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: Mr. W.S. Blair, Senior Counsel, Dominion 
Energy Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
    NRC Branch Chief: Michael T. Markley.

IV. 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.

[[Page 28350]]

    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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: July 31, 2015, as supplemented 
by letters dated April 11, 2016; November 3, 2017; and May 18, June 1, 
September 21, and October 5, 2018.
    Brief description of amendments: The amendments revised certain 
technical specification (TS) requirements related to Completion Times 
for Required Actions to provide the option to calculate a longer, risk-
informed completion time. The allowance is described in a new program, 
``Risk Informed Completion Time Program,'' that was added to TS Section 
5.0, ``Administrative Controls.'' The methodology for using the Risk-
Informed Completion Time Program is described in Nuclear Energy 
Institute (NEI) Report NEI 06-09, ``Risk-Informed Technical 
Specifications Initiative 4b: Risk-Managed Technical Specifications 
(RMTS) Guidelines,'' Revision 0-A.
    Date of issuance: May 29, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 270 days from the date of issuance.
    Amendment Nos.: Unit 1--209; Unit 2--209; Unit 3--209. A publicly-
available version is in ADAMS under Accession No. ML19085A525. 
Documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: 
The amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: December 8, 2015 (80 FR 
76317). By letter dated November 3, 2017, the licensee supplemented its 
application. By supplemental letters dated May 18 and June 1, 2018, the 
licensee provided additional information that expanded the scope of the 
amendment request as originally noticed in the Federal Register. 
Accordingly, the NRC published a second proposed no significant hazards 
consideration determination in the Federal Register on August 14, 2018 
(83 FR 40345), which superseded the original notice in its entirely. 
The supplemental letters dated September 21, and October 5, 2018, 
provided additional information that clarified the application, did not 
expand the scope of the application as noticed, and did not change the 
staff's second proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 29, 2019.

Dairyland Power Cooperative, Docket No.: 50-409, La Crosse Boiling 
Water Reactor, La Crosse County, Wisconsin

    Date of application for amendment: June 27, 2016, supplemented by 
letter dated December 1, 2016, May 31, 2018, and November 15, 2018.
    Brief description of amendment: The amendment revises the La Crosse 
Boiling Water Reactor (LACBWR) license to approve the License 
Termination Plan (LTP). The LACBWR LTP provides the details of the plan 
for characterizing, identifying, and remediating the remaining residual 
radioactivity at the LACBWR site to a level that will allow the site to 
be released for unrestricted use. The LACBWR LTP also describes how the 
licensee will confirm the extent and success of remediation through 
radiological surveys, provide financial assurance to complete 
decommissioning, and ensure the environmental impacts of the 
decommissioning activities are within the scope originally envisioned 
in the associated environmental documents. Decommissioning activities 
at the LACBWR site are scheduled to be complete in 2019, with license 
termination occurring before the end of 2020.
    Date of issuance: May 21, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 75.
    Possession Only License No. DPR-45: The amendment revised the 
Possession Only License.
    Date of initial notice in Federal Register: August 30, 2016 (81 FR 
59663). The supplements dated December 1, 2016, May 31, 2018, and 
November 15, 2018, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not affect the applicability of the NRC's generic no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated May 21, 2019, which is available in the 
Agencywide Documents Access and Management System (ADAMS) at Accession 
No. ML19008A079).
    No significant hazards consideration comments received: Not 
applicable.

Dominion Energy Nuclear Connecticut, Inc., Docket No. 50-423, Millstone 
Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: May 3, 2018, as supplemented by letters 
dated November 29, 2018; March 27, 2019; and May 7, 2019.
    Brief description of amendment: The amendment revised the Technical 
Specifications to reflect the results and constraints of a new 
criticality safety analysis for fuel assembly storage in the Millstone 
Power Station, Unit No. 3, fuel storage racks. Specifically, the 
amendment implemented the following items associated with fuel assembly 
storage: (1) Increased the Technical Specification minimum spent fuel 
pool soluble boron concentration, (2) revised allowed storage patterns 
and initial enrichment/burnup/decay time for fuel assemblies in the 
spent fuel pool to meet keff requirements under normal and 
accident conditions, (3) permitted the storage of any fuel assembly 
with certain enrichment that contains a rod cluster control assembly in 
Region 2 without restriction, and (4) implemented a revised criticality 
analysis for the new fuel storage racks using the updated methods for 
the spent fuel pool criticality analysis for consistency.
    Date of issuance: May 28, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 273. A publicly available version is in ADAMS under 
Accession No. ML19126A000; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-49: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: August 7, 2018 (83 FR 
38735). The supplemental letters dated November 29, 2018; March 27, 
2019; and May 7, 2019, provided additional information that clarified 
the application, did not expand the scope of the application as 
originally noticed, and did not change the NRC staff's original 
proposed no significant hazards

[[Page 28351]]

consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 28, 2019.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 12, 2018, as supplemented by 
letters dated April 26, October 17, and December 11, 2018.
    Brief description of amendment: The amendment revised the ANO-1 
Technical Specifications and operating license by relocating certain 
surveillance frequencies to a licensee-controlled program, consistent 
with the NRC-approved Technical Specifications Task Force (TSTF) 
Improved Standard Technical Specifications Change Traveler TSTF-425, 
Revision 3, ``Relocate Surveillance Frequencies to Licensee Control--
RITSTF [Risk-Informed TSTF] Initiative 5b.''
    Date of issuance: May 22, 2019.
    Effective date: As of the date of issuance.
    Amendment No.: 264. A publicly available version is in ADAMS under 
Accession No. ML19098A955; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: June 5, 2018 (83 FR 
26102). The supplemental letters dated October 17 and December 11, 
2018, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the NRC staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 22, 2019.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., Cooperative 
Energy, A Mississippi Electric Cooperative, and Entergy Mississippi, 
LLC, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of amendment request: April 10, 2018, as supplemented by 
letters dated October 23, 2018, and March 13, 2019.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to adopt Technical Specifications Task Force 
(TSTF) Traveler TSTF-542, Revision 2, ``Reactor Pressure Vessel Water 
Inventory Control.'' The change replaced existing TS requirements 
related to ``operations with a potential for draining the reactor 
vessel'' with new requirements on reactor pressure vessel water 
inventory control to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 
requires reactor vessel water level to be greater than the top of 
active irradiated fuel.
    Date of issuance: May 23, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment No: 218. A publicly available version is in ADAMS under 
Accession No. ML19084A218; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-29: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: June 5, 2018 (83 FR 
26103). The supplemental letters dated October 23, 2018, and March 13, 
2019, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the NRC staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 23, 2019.
    No significant hazards consideration comments received: No.

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

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
171, 50-277, and 50-278, Peach Bottom Atomic Power Station, Units 1, 2, 
and 3, York and Lancaster Counties, Pennsylvania

    Date of amendment request: May 10, 2018, as supplemented by letters 
dated November 1 and November 29, 2018.
    Brief description of amendments: The amendments revised the 
emergency response organization positions identified in the emergency 
plan for each site.
    Date of issuance: May 24, 2019.
    Effective date: As of the date of issuance and shall be implemented 
on or before December 31, 2019.
    Amendment Nos.: Limerick--235/198 and Peach Bottom--14/325/328. A 
publicly-available version is in ADAMS under Accession No. ML19078A018. 
Documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Facility Operating License Nos. NPF-39, NPF-85, DPR-12, DPR-44, and 
DPR-56: Amendments revised the emergency plans.
    Date of initial notice in Federal Register: July 17, 2018 (83 FR 
33268). The supplemental letters dated November 1 and November 29, 
2018, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the NRC staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a safety evaluation dated May 24, 2019.
    No significant hazards consideration comments received: No.

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: June 29, 2018, as supplemented by letter 
dated October 27, 2018.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3/4.3.1, ``Reactor Trip System Instrumentation''; TS 
3/4.3.2, ``Engineered Safety Feature Actuation System 
Instrumentation''; TS 3/4.7.1.5, ``Main Steam Isolation Valves''; and 
added a new TS for feedwater isolation to better align the TSs with the 
design-basis analyses and the design of the instrumentation.
    Date of issuance: May 31, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 329 (Unit No. 1) and 310 (Unit No. 2). A publicly 
available version is in ADAMS under Accession No. ML19105B171; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-70 and DPR-75: The 
amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: August 28, 2018 (83 FR 
43907). The supplemental letter dated October 27, 2018, provided 
additional information that clarified the

[[Page 28352]]

application, did not expand the scope of the application as originally 
noticed, and did not change the NRC staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 31, 2019.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar 
Nuclear Plant, Units 1 and 2, Rhea County, Tennessee

    Date of amendment request: December 20, 2017, as supplemented by 
letters dated February 15, April 9, and October 4, 2018.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 4.2.1, ``Fuel Assemblies,'' for Unit 2 to allow up 
to 1,792 tritium producing burnable absorber rods in the reactor; and 
revised the Units 1 and 2 TSs related to fuel storage.
    Date of issuance: May 22, 2019.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from the outage where any number of tritium producing 
burnable absorber rods is inserted in the Watts Bar Nuclear Plant, Unit 
2, reactor core not to exceed December 31, 2022.
    Amendment Nos.: 125 (Unit 1) and 27 (Unit 2). A publicly available 
version is in ADAMS under Accession No. ML18347B330; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-90 and NPF-96: The amendments 
revised the Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: June 8, 2018 (83 FR 
26709). The supplement dated October 4, 2018, provided additional 
information that clarified the application, and did not expand the 
scope of the application as originally noticed in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 22, 2019.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Unit 1 (Wolf Creek), Coffey County, Kansas

    Date of amendment request: January 17, 2017, as supplemented by 
letters dated March 22, May 4, July 13, October 18, and November 14, 
2017; January 15, January 29, April 19, June 19, August 9, November 15 
(two letters), and December 6, 2018; and March 5, May 2, and May 15, 
2019.
    Brief description of amendment: The amendment revised the Wolf 
Creek Technical Specifications to replace the existing methodology for 
performing core design, non-loss-of-coolant-accident and loss-of-
coolant accident safety analyses with standard Westinghouse Electric 
Corporation developed and NRC-approved analysis methodologies. In 
addition, the amendment revised the Wolf Creek licensing basis by 
adopting the alternative source term (AST) radiological analysis 
methodology in accordance with 10 CFR 50.67, ``Accident source term.'' 
This amendment represented a full scope implementation of the AST as 
described in Regulatory Guide 1.183, ``Alternative Radiological Source 
Terms for Evaluating Design Basis Accidents at Nuclear Power 
Reactors.''
    Date of issuance: May 31, 2019.
    Effective date: As of the date of issuance and shall be implemented 
during startup (prior to entry into Mode 2) from Refueling Outage 23.
    Amendment No.: 221. A publicly available version is in ADAMS under 
Accession No. ML19100A122; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-42. The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: On July 5, 2017, the 
NRC staff published a proposed no significant hazards consideration 
(NSHC) determination in the Federal Register (82 FR 31084) for the 
proposed amendment. Subsequently by letters dated July 13, October 18, 
and November 14, 2017; January 15, January 29, April 19, June 19, and 
August 9, 2018, the licensee provided additional information that 
expanded the scope of the amendment request as originally noticed in 
the Federal Register. Accordingly, the NRC published a second proposed 
NSHC determination in the Federal Register on October 2, 2018 (83 FR 
49590), which superseded the original notice in its entirety. The 
supplemental letters dated November 15 (two letters) and December 6, 
2018; and March 5, May 2, and May 15, 2019, provided additional 
information that clarified the application, did not expand the scope of 
the application as noticed on October 2, 2018, and did not change the 
NRC staff's proposed NSHC determination published in the Federal 
Register dated October 2, 2018.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 31, 2019.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 10th day of June 2019.

    For the Nuclear Regulatory Commission.
Craig G. Erlanger,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2019-12573 Filed 6-17-19; 8:45 am]
BILLING CODE 7590-01-P