[Federal Register Volume 83, Number 99 (Tuesday, May 22, 2018)]
[Notices]
[Pages 23728-23742]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-10565]


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NUCLEAR REGULATORY COMMISSION

[NRC-2018-0096]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, 
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is 
publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued, and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from April 24, 2018, to May 7, 2018. The last 
biweekly notice was published on May 8, 2018.

DATES: Comments must be filed by June 21, 2018. A request for a hearing 
must be filed by July 23, 2018.

ADDRESSES: You may submit comments by any of the following methods:
     Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0096. Address 
questions about NRC dockets to Jennifer Borges; telephone: 301-287-
9127; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: May Ma, Office of Administration, Mail 
Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001.
    For additional direction on obtaining information and submitting 
comments,

[[Page 23729]]

see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

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

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC 2018-0096, facility name, unit 
number(s), plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
     Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0096.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in this 
document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2018-0096, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov, as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

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

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

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

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and 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

[[Page 23730]]

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

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing and petition for leave to intervene (petition), any 
motion or other document filed in the proceeding prior to the 
submission of a request for hearing or petition to intervene, and 
documents filed by interested governmental entities that request to 
participate under 10 CFR 2.315(c), must be filed in accordance with the 
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 
46562; August 3, 2012). The E-Filing process requires participants to 
submit and serve all adjudicatory documents over the internet, or in 
some cases to mail copies on electronic storage media. Detailed 
guidance on making electronic submissions may be found in the Guidance 
for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit 
paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [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

[[Page 23731]]

on the NRC's public website at http://www.nrc.gov/site-help/e-submittals.html, by email to [email protected], or by a toll-free 
call at 1-866-672-7640. The NRC Electronic Filing Help Desk is 
available between 9 a.m. and 6 p.m., Eastern Time, Monday through 
Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
stating why there is good cause for not filing electronically and 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff. Participants filing adjudicatory documents in this 
manner are responsible for serving the document on all other 
participants. Filing is considered complete by first-class mail as of 
the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service. A presiding officer, having granted an 
exemption request from using E-Filing, may require a participant or 
party to use E-Filing if the presiding officer subsequently determines 
that the reason for granting the exemption from use of E-Filing no 
longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the 
Commission or the presiding officer. If you do not have an NRC-issued 
digital ID certificate as described above, click cancel when the link 
requests certificates and you will be automatically directed to the 
NRC's electronic hearing dockets where you will be able to access any 
publicly available documents in a particular hearing docket. 
Participants are requested not to include personal privacy information, 
such as social security numbers, home addresses, or personal phone 
numbers in their filings, unless an NRC regulation or other law 
requires submission of such information. For example, in some 
instances, individuals provide home addresses in order to demonstrate 
proximity to a facility or site. With respect to copyrighted works, 
except for limited excerpts that serve the purpose of the adjudicatory 
filings and would constitute a Fair Use application, participants are 
requested not to include copyrighted materials in their submission.
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

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

    Date of amendment request: January 10, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18010A344.
    Description of amendment request: The amendments would modify the 
licensing basis to allow for the implementation of the provisions of 10 
CFR 50.69, ``Risk-informed characterization and treatment of 
structures, systems, and components for nuclear reactors.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, 
550 South Tryon Street, M/C DEC45A, Charlotte, NC 28202.
    NRC Acting Branch Chief: Brian W. Tindell.

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

    Date of amendment request: February 1, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18033B768.

[[Page 23732]]

    Description of amendment request: The amendment would revise the 
licensing basis to allow for the implementation of the provisions of 10 
CFR 50.69, ``Risk-informed categorization and treatment of structures, 
systems and components for nuclear power reactors.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara Nichols, Deputy General Counsel, Duke 
Energy Corporation, 550 South Tryon St., M/C DEC45A, Charlotte, NC 
28202.
    NRC Acting Branch Chief: Brian W. Tindell.

Entergy Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1 (RBS), West Feliciana Parish, Louisiana

    Date of amendment request: January 29, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18029A187.
    Description of amendment request: The proposed change would modify 
the RBS Updated Safety Analysis Report (USAR) and Technical 
Requirements Manual to relocate the reactor core isolation cooling 
(RCIC) piping injection point from the reactor vessel head spray nozzle 
to the feedwater line using the residual heat removal (RHR) shutdown 
cooling return line.
    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.
    Basis: The relocation of the RCIC injection point from the 
reactor vessel head spray nozzle to the `A' Feedwater line via the 
`A' RHR shutdown cooling return line does not adversely affect the 
design function of an System, Structure, or Component (SSC) or a 
method of performing or controlling a design function of an SSC as 
described in the USAR so there is no change to the likelihood of 
occurrence of a malfunction of a structure, system, or component 
important to safety previously evaluated in the USAR. There is no 
impact to the likelihood of occurrence of a malfunction of a 
structure, system, or component because there are no structures 
systems or components changed or affected by the scope of this 
evaluation.
    Inadvertent initiation of RCIC may be categorized as either a 
Decrease in Reactor Coolant Temperature event or an Increase in 
Reactor Coolant Inventory event. River Bend Transient Safety 
Analysis Design Report, 6224.302-000-035A, states that three systems 
were considered that could introduce a cold water perturbation 
(Decrease in Reactor Coolant Temperature Event) at operating 
pressures: RCIC, High Pressure Core Spray (HPCS), and the feedwater 
system. This report qualifies improper startup of HPCS or RCIC as 
events that would produce no significant power transients. The 
proposed change relocated the injection point of the RCIC flow from 
the reactor head (RPV [reactor pressure vessel]) to the feedwater 
line (FWS). This change will reduce the effects of steam quenching. 
However, the effect of steam quenching is not credited in any of the 
safety analysis. The only portion of the RCIC system operation that 
is credited is water injection at the required flow rate, and the 
design function as described in the USAR of the RCIC system is to 
maintain or supplement the reactor vessel water inventory. The 
source of water for the Inadvertent RCIC injection remains the same. 
The destination of the water for the Inadvertent RCIC injection is 
still the RPV. The ability of the rerouted equipment to satisfy the 
RCIC design function is not reduced from the original design 
requirement to inject 600 gpm [gallons per minute] into the RPV. 
This is maintained by the RCIC flow controller. The entry location 
from the RPV head spray to the feedwater line has no impact to the 
consequences of an inadvertent initiation of RCIC. As the 
consequences of an inadvertent initiation of RCIC are unchanged, the 
consequences of this event remain quantitatively bounded by the Loss 
of Feedwater Heating event described in section 15.1.1 of the USAR 
for the Decrease in Reactor Coolant Temperature category and bounded 
by the Inadvertent HPCS Startup for the Increase in Reactor Coolant 
Inventory category.
    Changing the injection point of RCIC does not increase the 
probability or consequences of an inadvertent RCIC injection. All 
affected piping, fittings, and valve pressure boundaries are 
qualified to the appropriate fluid transients and operational 
conditions in accordance with the design and licensing basis. No 
instrument setpoints were changed as a result of this modification. 
The RCIC system's modes of operation are not changed or affected by 
this modification. Therefore there is no change in the frequency of 
an inadvertent initiation of RCIC event. There is no change in the 
frequency of inadvertent initiation of RCIC by this modification, so 
there is no impact to the probability of any previously evaluated 
accident.

[[Page 23733]]

    Therefore, it is concluded that this change does not 
significantly increase 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.
    Basis: The spurious start of RCIC accident is evaluated in the 
USAR as Event 9 ``Inadvertent HPCS Pump Start (Moderator Temperature 
Decrease) as shown in USAR Appendix 15A. The Inadvertent HPCS Pump 
Start event bounds the inadvertent operation of RCIC event and is 
quantitatively analyzed in accordance with Reg Guide 1.70 rev. 3. 
This event may be classified as either a Decrease in Core Coolant 
Temperature event or an Increase in Reactor Coolant Inventory Event, 
however was categorized as an Increase in Reactor Coolant Inventory 
Event in the RBS USAR as this is the initial effect of this event. 
No new accident is created by the scope of this modification because 
all aspects of the existing Decrease in Core Coolant Temperature and 
Increase in Reactor Coolant Inventory events and their relationship 
to the spurious start of RCIC remain applicable.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Basis: The proposed change does not change any accident 
analyses. The proposed change does not exceed or alter a design 
basis or safety limit; therefore it does not significantly reduce 
the margin of safety.
    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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anna Vinson Jones, Senior Counsel--Entergy 
Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington 
DC 20001.
    NRC Branch Chief: Robert J. Pascarelli.

Entergy Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: February 28, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18067A115.
    Description of amendment request: The amendment would modify the 
River Bend Station Technical Specifications (TSs) by relocating 
specific surveillance frequencies to a licensee-controlled program with 
the adoption of Technical Specifications Task Force (TSTF) Traveler 
TSTF-425, Revision 3, ``Relocate Surveillance Frequencies to Licensee 
Control--RITSTF [Risk-Informed TSTF] Initiative 5b.'' Additionally, the 
change would add a new program, the Surveillance Frequency Control 
Program (SFCP), to TS Chapter 5.0, ``Administrative Controls.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change relocates the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
significantly increased. The systems and components required by the 
technical specifications for which the surveillance frequencies are 
relocated are still required to be operable, meet the acceptance 
criteria for the surveillance requirements, and be capable of 
performing any mitigation function assumed in the accident analysis. 
As a result, the consequences of any accident previously evaluated 
are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the 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 design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, Entergy 
will perform a probabilistic risk evaluation using the guidance 
contained in NRC approved NEI [Nuclear Energy Institute] 04-10, Rev. 
1 in accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology 
provides reasonable acceptance guidelines and methods for evaluating 
the risk increase of proposed changes to surveillance frequencies 
consistent with Regulatory Guide 1.177.
    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: Anna Vinson Jones, Senior Counsel--Entergy 
Services, Inc., 101 Constitution Avenue NW, Suite 200 East, Washington, 
DC 20001.
    NRC Branch Chief: Robert J. Pascarelli.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One (ANO), Unit Nos. 1 and 2, Pope County, Arkansas

    Date of amendment request: March 29, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18088B412.
    Description of amendment request: The amendments would revise the 
ANO, Units 1 and 2, currently approved Emergency Plan Emergency Action 
Level (EAL) scheme, which is based on the Nuclear Energy Institute 
(NEI) guidance established in NEI 99-01, Revision 5, ``Methodology for 
Development of Emergency Action Levels,'' by adopting the EAL schemes 
based on the guidance provided in NEI 99-01, Revision 6, ``Development 
of Emergency Action Levels for Non-Passive Reactors.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or

[[Page 23734]]

consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the ANO EALs do not involve any physical 
changes to plant equipment or systems and do not alter the 
assumptions of any accident analyses. The proposed changes do not 
adversely affect accident initiators or precursors and do not alter 
design assumptions, plant configuration, or the manner in which the 
plant is operated and maintained. The proposed changes do not 
adversely affect the ability of structures, systems or components 
(SSCs) to perform intended safety functions in mitigating the 
consequences of an initiating event within the assumed acceptance 
limits.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
The changes do not challenge the integrity or performance of any 
safety-related systems. No plant equipment is installed or removed, 
and the changes do not alter the design, physical configuration, or 
method of operation of any plant SSC. Because EALs are not accident 
initiators and no physical changes are made to the plant, no new 
causal mechanisms are introduced.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from an accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is associated with the ability of the fission 
product barriers (i.e., fuel cladding, reactor coolant system 
pressure boundary, and containment structure) to limit the level of 
radiation dose to the public. The proposed changes do not impact 
operation of the plant and no accident analyses are affected by the 
proposed changes. The changes do not affect the Technical 
Specifications or the method of operating the plant. Additionally, 
the proposed changes will not relax any criteria used to establish 
safety limits and will not relax any safety system settings. The 
safety analysis acceptance criteria are not affected by these 
changes. The proposed changes will not result in plant operation in 
a configuration outside the design basis. The proposed changes do 
not adversely affect systems that respond to safely shut down the 
plant and to maintain the plant in a safe shutdown condition.
    Therefore, the changes do not involve a significant reduction in 
a margin of safety.

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

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

    Date of amendment request: February 1, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18036A227.
    Description of amendment request: The proposed amendments would 
revise the Braidwood Station licensing basis for protection from 
tornado-generated missiles by identifying the TORMIS Computer Code as 
the methodology used for assessing tornado-generated missile protection 
of unprotected plant structures, systems, and components (SSCs).
    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 NRC TORMIS Safety Evaluation Report [ADAMS Accession No. 
ML080870291] states the following:
    ``The current Licensing criteria governing tornado missile 
protection are contained in [NUREG-0800] Standard Review Plan (SRP) 
Section 3.5.1.4, [Missiles Generated by Natural Phenomena] and 3.5.2 
[Structures, Systems and Components [SSCs]] to be Protected from 
Externally Generated Missiles]. These criteria generally specify 
that safety-related systems be provided positive tornado missile 
protection (barriers) from the maximum credible tornado threat. 
However, SRP Section 3.5.1.4 includes acceptance criteria permitting 
relaxation of the above deterministic guidance, if it can be 
demonstrated that the probability of damage to unprotected essential 
safety-related features is sufficiently small.''
    As permitted by these SRP sections, the combined probability 
will be maintained below an allowable level, i.e., an acceptance 
criterion threshold, which reflects an extremely low probability of 
occurrence. SRP Section 2.2.3, ``Evaluation of Potential 
Accidents,'' established this threshold as approximately 1.0E-06 per 
year if, ``when combined with reasonable qualitative arguments, the 
realistic probability can be shown to be lower.'' The Braidwood 
Station analysis approach assumes that if the sum of the individual 
probabilities calculated for tornado missiles striking and damaging 
portions of safety-significant SSCs is greater than or equal to 
1.0E-06 per year per unit, then installation of tornado missile 
protection barriers would be required for certain components to 
lower the total cumulative damage probability below the acceptance 
criterion of 1.0E-06 per year per unit. Conversely, if the total 
cumulative damage probability remains below the acceptance criterion 
of 1.0E-06 per year per unit, no additional tornado missile 
protection barriers would be required for any of the unprotected 
safety-significant components.
    With respect to the probability of occurrence or the 
consequences of an accident previously evaluated in the UFSAR 
[Updated Final Safety Analysis Report], the possibility of a tornado 
impacting the Braidwood Station site and causing damage to plant 
SSCs is a licensing basis event currently addressed in the UFSAR. 
The change being proposed (i.e., the use of the TORMIS methodology 
for assessing tornado-generated missile protection of unprotected 
plant SSCs), does not affect the probability of a tornado strike on 
the site; however, from a licensing basis perspective, the proposed 
change does affect the probability that missiles generated by a 
tornado will strike and damage certain safety-significant plant 
SSCs. There are a defined number of safety-significant components 
that could theoretically be struck and damaged by tornado-generated 
missiles. The probability of tornado-generated missile hits on these 
``important'' systems and components is calculated using the TORMIS 
probabilistic methodology. The combined probability of damage for 
unprotected safety-significant equipment will be maintained below 
the acceptance criterion of 1.0E-06 per year per unit to ensure 
adequate equipment remains available to safely shutdown the 
reactors, and maintain overall plant safety, should a tornado strike 
occur. Consequently, the proposed change does not constitute a 
significant increase in the probability of occurrence or the 
consequences of an accident based on the extremely low probability 
of damage caused by tornado-generated missiles and the commensurate 
extremely low probability of a radiological release.
    Finally, the use of the TORMIS methodology will have no impact 
on accident initiators or precursors; does not alter the accident 
analysis assumptions or the manner in which the plant is operated or 
maintained; and does not affect the probability of operator error.
    Based on the above discussion, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The impact of a tornado strike on the Braidwood Station site is 
a licensing basis event that is explicitly addressed in the

[[Page 23735]]

UFSAR. The proposed change simply involves recognition of the 
acceptability of using an analysis tool (i.e., the TORMIS 
methodology) to perform probabilistic tornado missile damage 
calculations in accordance with approved regulatory guidance. The 
proposed change does not result in the creation of any new accident 
precursors; does not result in changes to any existing accident 
scenarios; and does not introduce any operational changes or 
mechanisms that would create the possibility of a new or different 
kind of accident.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident than those previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The existing Braidwood Station licensing basis regarding tornado 
missile protection of safety-significant SSCs assumes that missile 
protection barriers are provided for safety-significant SSCs; or the 
unprotected component is assumed to be unavailable post-tornado.
    The results of the Braidwood Station TORMIS analysis have 
demonstrated that there is an extremely low probability, below an 
established regulatory acceptance limit, that these ``important'' 
SSCs could be struck and subsequently damaged by tornado-generated 
missiles. The change in licensing basis from protecting safety-
significant SSCs from tornado missiles, to demonstrating that there 
is an extremely low probability that safety-significant SSCs will be 
struck and damaged by tornado-generated missiles, does not 
constitute a significant decrease in the margin of safety.
    Therefore, the proposed change to use the TORMIS methodology 
does not involve a significant reduction in the margin of safety.
    Based on the above, EGC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92, and accordingly, a finding of 
``no significant hazards consideration'' is justified.

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

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

    Date of amendment request: March 15, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18074A308.
    Brief description of amendment request: The proposed amendments 
would revise Prairie Island Nuclear Generating Plant Technical 
Specifications (TSs) by relocating specific surveillance frequencies to 
a licensee-controlled program with implementation of Nuclear Energy 
Institute (NEI) 04-10, ``Risk-Informed Technical Specification 
Initiative 5b, Risk-Informed Method for Control of Surveillance 
Frequencies,'' Revision 1. The changes are consistent with Technical 
Specifications Task Force (TSTF) Traveler TSTF-425, ``Relocate 
Surveillance Frequencies to Licensee Control--Risk Informed Technical 
Specifications Task Force (RITSTF) Initiative 5b,'' Revision 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change relocates the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
SFCP [surveillance frequency control program]. Surveillance 
frequencies are not an initiator to any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not significantly increased. The systems and components 
required by the technical specifications for which the surveillance 
frequencies are relocated are still required to be operable, meet 
the acceptance criteria for the surveillance requirements, and be 
capable of performing any mitigation function assumed in the 
accident analysis. As a result, the consequences of any accident 
previously evaluated are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, NSPM will 
perform a probabilistic risk evaluation using the guidance contained 
in NRC approved NEI 04-10, Rev. 1 in accordance with the TS SFCP. 
NEI 04-10, Rev. 1, methodology provides reasonable acceptance 
guidelines and methods for evaluating the risk increase of proposed 
changes to surveillance frequencies consistent with Regulatory Guide 
1.177.
    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: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: David J. Wrona.

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

    Date of amendment request: March 28, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18087A323.
    Description of amendment request: The proposed amendment would 
modify the Monticello Nuclear Generating Plant licensing basis by the 
addition of a license condition to allow for the implementation of the 
provisions of 10 CFR 50.69, ``Risk-Informed Categorization and 
Treatment of Structures, Systems and Components for Nuclear Power 
Reactors.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or

[[Page 23736]]

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: David J. Wrona.

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

    Date of amendment request: April 3, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18094A189.
    Description of amendment request: The proposed amendment would 
change Functional Units 17.A and 17.B of Technical Specification (TS) 
Table 4.3-1, ``Reactor Trip System Instrumentation Surveillance 
Requirements.'' The Trip Actuating Device Operational Test (TADOT) 
column of this table would be revised to delete the ``S/U'' frequency 
and replace it with a reference to Table Notation (8), which would 
state, ``Prior to entering MODE 1 whenever the unit has been in MODE 
3.'' The licensee stated that the change would align the surveillance 
requirements and the mode requirement for the Turbine Trip TADOT with 
the TS \3/4\.3.1, Table 3.3-1, ``Reactor Trip System Instrumentation,'' 
channels and interlocks mode requirement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise the surveillance frequency for 
reactor trip functions from a turbine trip event. These changes do 
not alter these functions physically, or how they are maintained. 
Changing the surveillance from ``prior to Startup'' to ``prior to 
entering MODE 1'' will continue to ensure operability of the 
function before the plant is in a condition that would benefit from 
the associated actuation and prior to applicability. Since these 
changes will not affect the ability of these trips to perform the 
initiation of reactor trips when appropriate, the offsite dose 
consequences for an accident will not be impacted. Equally, the 
potential to cause an accident is not affected because no plant 
system or component has been altered by the proposed changes.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes only affect surveillance frequency 
requirements for the turbine trip functions. This does not affect 
any physical features of the plant, or the manner in which these 
functions are utilized. The proposed surveillance frequency will 
require the functions to be verified operable before the turbine 
trip functions are applicable and able to perform their trip 
functions. Changing the surveillance from ``prior to Startup'' to 
``prior to entering MODE 1'' will continue to ensure operability of 
the function before the plant is in a condition that would benefit 
from the associated actuation. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter any plant setpoints or 
functions that are assumed to actuate in the event of postulated 
accidents. The proposed changes do not alter any plant feature and 
only alters the MODE which the surveillance tests must be performed. 
The proposed changes ensure the functionality of the turbine trips 
when assumed in the analysis for accident mitigation. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

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

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

    Date of amendment request: April 6, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18096B463.
    Description of amendment request: The requested amendments require 
changes to the Updated Final Safety Analysis Report (UFSAR) in the form 
of

[[Page 23737]]

departures from the incorporated plant-specific Design Control Document 
(DCD) Tier 2 information and related changes to the Vogtle Electric 
Generating Plant, Unit Nos. 3 and 4, combined license (COL) and COL 
Appendix C (and corresponding plant-specific DCD Tier 1) information. 
Specifically, the requested amendments include changes to the equipment 
survivability assessment requirements associated with hydrogen burns 
during beyond design-basis accidents as described in the licensing 
basis documents, including COL Condition 2.D(12)(g)9 and plant-specific 
Tier 1 Sections 2.2.3 and 2.3.9.
    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 and clarifications to the locations of 
Hydrogen Igniters 27, 30, 35, 36, 37, and 38 do not adversely affect 
any safety-related structure, system or component (SSC) or function. 
The hydrogen ignition subsystem is designed to mitigate beyond 
design basis hydrogen generation in the containment. With the 
proposed changes, the hydrogen ignition subsystem continues to 
maintain the designed and analyzed beyond design basis functions. 
The hydrogen ignition subsystem maintains its design function to 
maintain containment integrity. The proposed changes also reconcile 
the as-built equipment with the list of equipment on which the 
equipment survivability assessment is performed to provide 
additional assurance containment penetrations and combustible gas 
control components will perform their design functions after a 
hydrogen burn in containment. The changes are to the equipment 
assessed, not to the design functions of the equipment. The changes 
do not involve an interface with any SSC accident initiator or 
initiating sequence of events, and thus, the probabilities of the 
accidents evaluated in the plant-specific UFSAR are not affected. 
The proposed changes do not involve a change to any mitigation 
sequence or the predicted radiological releases due to postulated 
accident conditions, thus, the consequences of the accidents 
evaluated in the UFSAR are not affected.
    The maximum allowable containment vessel leakage rate specified 
in the Technical Specifications is unchanged, and radiological 
material release source terms are not affected; thus, the 
radiological releases in the accident analyses are not affected. The 
proposed changes do not affect the prevention and mitigation of 
other abnormal events (e.g. anticipated operational occurrences, 
earthquakes, floods and turbine missiles), or their safety or design 
analyses. Therefore, the consequences of the accidents evaluated in 
the UFSAR are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that may initiate a new or different kind of accident, 
or alter any SSC such that a new accident initiator or initiating 
sequence of events is created. The proposed changes reconcile the 
as-built equipment with the list of equipment on which the equipment 
survivability assessment is performed to provide additional 
assurance that containment penetrations and combustible gas control 
components will perform their design functions after a hydrogen burn 
in containment. The equipment survivability assessment changes are 
to the equipment assessed, not to the design functions of the 
equipment. The VLS Hydrogen Ignition subsystem does not interface 
with/affect safety-related equipment or a fission product barrier. 
The subsystem is provided to address the production of hydrogen 
following a beyond design basis accident in accordance with 10 CFR 
50.44(c). The hydrogen ignition subsystem is a non-Class 1E 
subsystem and does not interface with any safety-related system; 
thus, no system or design function or equipment qualification is 
affected by the proposed changes. The changes to the hydrogen 
ignition subsystem do not result in a new failure mode, malfunction 
or sequence of events that could affect a radioactive material 
barrier or safety-related equipment. The proposed changes do not 
adversely affect any system or design function or equipment 
qualification as the changes do not modify any SSCs that prevent 
safety functions from being performed. The changes do not introduce 
a new failure mode, malfunction or sequence of events that could 
adversely affect safety or safety-related equipment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes and clarifications to the locations of 
Hydrogen Igniters 27, 30, 35, 36, 37, and 38 maintain the beyond 
design basis function of the hydrogen ignition subsystem. The 
proposed changes also reconcile the as-built equipment with the list 
of equipment on which the equipment survivability assessment is 
performed to provide additional assurance containment penetrations 
and combustible gas control components will perform their design 
functions after a hydrogen burn in containment. The equipment 
survivability assessment changes are to the equipment assessed, not 
to the design functions of the equipment. The proposed changes would 
not affect any safety-related design code, function, design 
analysis, safety analysis input or result, or existing design/safety 
margin. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the requested 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: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.

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

    Date of amendment request: April 13, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18103A249.
    Description of amendment request: The requested amendments require 
changes to combined license (COL) Appendix A, Technical Specifications 
and the Updated Final Safety Analysis Report (UFSAR) in the form of 
departures from the incorporated plant-specific Design Control Document 
Tier 2 information. Specifically, the requested amendments include 
changes to the COL Appendix A, Technical Specifications related to the 
statuses of the remotely operated containment isolation valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This change clarifies that only Class 1E valves in the 
nonessential containment penetration flow paths that receive the 
containment isolation signal (T signal) are part of the [Post-
Accident Monitoring (PAM)] Technical Specifications and adds 
additional valves to the PAM table in the UFSAR. The Normal Residual 
Heat Removal System (RNS), Chemical and Volume Control System (CVS), 
Component Cooling Water System (CCS), and Steam Generator System 
(SGS) have containment isolation valves that do not

[[Page 23738]]

close on a T signal because they have an accident mitigation 
function to be open.
    The status of the valves in the essential containment flow paths 
are summarized on one non-safety display screen and are separately 
indicated on the safety display screens within their respective 
systems. Keeping these indications separate from the ``Remotely 
Operated Containment Isolation Valve Status'' which is on the 
Category 1 display allows the operators to quickly verify that the 
nonessential containment flow paths are isolated and then focus on 
the availability of the essential flow paths for their defense-in-
depth capabilities.
    The valve position indications in the essential flow paths that 
penetrate containment are not Post-Accident Monitoring System (PAMS) 
B1 variables. These essential flow paths support accident mitigation 
functions of non-safety systems and may be intentionally opened for 
extended periods of time following an accident. As a result, 
excluding them from the PAMS B1 summary indication will increase the 
value of the summary indication during operation of the essential 
flow paths.
    Furthermore, opening these essential flow paths pose low risk of 
becoming an unmonitored leak path through the containment vessel. 
The valves are isolated when required by separate Protection and 
Safety Monitoring System (PMS) signals that are associated with each 
system's post-accident functions, and the valve position indications 
are designated as PAMS D2 accordingly.
    No structure, system, or component (SSC) or function is changed 
within this activity. Therefore, the proposed amendment does not 
involve a significant increase in the probability of an accident 
previously evaluated.
    The proposed amendment does not affect the prevention and 
mitigation of abnormal events, e.g., accidents, anticipated 
operation occurrences, earthquakes, floods, turbine missiles, and 
fires or their safety or design analyses. This change does not 
involve containment of radioactive isotopes or any adverse effect on 
a fission product barrier. There is no impact on previously 
evaluated accidents.
    Therefore, the proposed changes do not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve a new failure mechanism or 
malfunction, which affects an SSC accident initiator, or interface 
with any SSC accident initiator or initiating sequence of events 
considered in the design and licensing bases. There is no adverse 
effect on radioisotope barriers or the release of radioactive 
materials. The proposed amendment does not adversely affect any 
accident, including the possibility of creating a new or different 
kind of accident from any accident previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different type of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This activity clarifies that only Class 1E valves in the 
nonessential containment penetration flow paths that receive the 
containment isolation signal (T signal) are part of the PAM 
Technical Specifications and adds additional valves to the PAM table 
in the UFSAR.
    The status of the valves in the essential containment flow paths 
are summarized on one non-safety display screen and are separately 
indicated on the safety display screens within their respective 
systems. Keeping these indications separate from the ``Remotely 
Operated Containment Isolation Valve Status'' which is on the 
Category 1 display allows the operators to quickly verify that the 
nonessential containment flow paths are isolated and then focus on 
the availability of the essential flow paths for their defense-in-
depth capabilities.
    The valve position indications in the essential flow paths that 
penetrate containment are not PAMS B1 variables. These essential 
flow paths support accident mitigation functions of non-safety 
systems and may be intentionally opened for extended periods of time 
following an accident. As a result, excluding them from the PAMS B1 
summary indication will increase the value of the summary indication 
during operation of the essential flow paths.
    Furthermore, opening these essential flow paths pose low risk of 
becoming an unmonitored leak path through the containment vessel. 
The valves are isolated when required by separate PMS signals that 
are associated with each system's post-accident functions and the 
valve position indications are designated as PAMS D2 accordingly.
    No SSC or function is changed within this activity. The proposed 
changes would not affect any safety-related design code, function, 
design analysis, safety analysis input or result, or existing 
design/safety margin. No safety analysis or design basis acceptance 
limit/criterion is challenged or exceeded by the requested changes.
    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: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.

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

    Date of amendment request: April 20, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18110A113.
    Description of amendment request: The requested amendments propose 
changes to the Updated Final Safety Analysis Report (UFSAR) in the form 
of departures from the plant-specific Design Control Document (DCD) 
Tier 2 information and involves changes to the plant-specific Tier 1 
information (and associated Combined License (COL) Appendix C 
information). Specifically, the amendment proposes changes to plant-
specific Tier 1 (and COL Appendix C) Table 2.5.2-3, ``PMS Automatically 
Actuated Engineered Safety Features,'' to revise the nomenclature for 
``Auxiliary Spray and Letdown Purification Line Isolation'' and to 
include ``Component Cooling System Containment Isolation Valve 
Closure.'' Pursuant to the provisions of 10 CFR 52.63(b)(1), an 
exemption from elements of the design as certified in the 10 CFR part 
52, Appendix D, design certification rule is also requested for the 
plant-specific DCD Tier 1 material departures.
    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 nomenclature changes reflect the current plant 
design. These changes provide consistency with the approved plant 
design. The changes do not affect the operation of any systems or 
equipment that initiate an analyzed accident or alter any 
structures, systems, and components accident initiator or initiating 
sequence of events. The proposed changes do not result in any 
increase in probability of an analyzed accident occurring. 
Therefore, the requested amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed nomenclature changes reflect the current plant 
design. These changes provide consistency with the approved plant 
design. The proposed changes do not affect plant electrical systems,

[[Page 23739]]

and do not affect the design function, support, design, or operation 
of mechanical and fluid systems. The proposed changes do not result 
in a new failure mechanism or introduce any new accident precursors. 
No design function described in the UFSAR is affected by the 
proposed changes. Therefore, the requested amendment does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed nomenclature changes reflect the current plant 
design. These changes provide consistency with the approved plant 
design. No safety analysis or design basis acceptance limit/
criterion is involved. Therefore, the proposed amendment does not 
involve a significant reduction in a margin of safety.

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

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

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

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

    Date of amendment request: August 30, 2016, as supplemented by 
letter dated November 20, 2017.
    Brief description of amendment: The amendment revised the Columbia 
Generating Station Final Safety Analysis Report to reclassify reactor 
water cleanup piping, valves, pumps, and mechanical modules located 
outside of the primary and secondary containment in the radwaste 
building from Quality Group C to Quality Group D.
    Date of issuance: April 17, 2018.
    Effective date: As of its date of issuance and shall be implemented 
from the date of issuance until restart after Refueling Outage 24 
(spring 2019).
    Amendment No.: 248. A publicly-available version is in ADAMS under 
Accession No. ML18075A351; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment. This 
Notice of Issuance is being reissued in its entirety to reflect a 
correction to the ``Effective date'' by letter dated April 27, 2018 
(ADAMS Accession No. ML18109A215).
    Renewed Facility Operating License No. NPF-21: The amendment 
revised the Final Safety Analysis Report.
    Date of initial notice in Federal Register: December 6, 2016 (81 FR 
87968). The supplemental letter dated November 20, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 17, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 25, 2017.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.5.1, ``ECCS [Emergency Core Cooling Systems]--
Operating,'' and deleted the Note associated with Surveillance 
Requirement 3.5.1.2 to reflect the residual heat removal system design 
and ensure the residual heat removal system's operation is consistent 
with the TS 3.5.1 limiting condition for operation requirements.
    Date of issuance: May 2, 2018.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 249. A publicly-available version is in ADAMS under 
Accession No. ML18100A199; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-21: The amendment 
revised the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: November 7, 2017 (82 FR 
51649).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 2, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 28, 2017, as supplemented by 
letter dated February 28, 2018.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.8.1.3, ``Diesel Fuel Oil,'' by relocating the 
current required stored diesel fuel oil numerical volumes from the TSs 
to the TS Bases and replacing them with comparable duration-based 
requirements. In addition, the amendment revised TS 3.8.1.1 and TS 
3.8.1.2, ``AC [Alternating Current] Sources Operating,'' and ``AC 
Sources Shutdown,'' respectively, to relocate the specific numerical 
value for feed tank fuel oil volume to the TS Bases and replace it with 
the feed tank operating time requirement. The changes are consistent 
with Technical Specifications Task Force (TSTF) Traveler TSTF-501, 
Revision 1, ``Relocate Fuel Oil and Lube Oil Volume Values to Licensee 
Control.''

[[Page 23740]]

    Date of issuance: April 26, 2018.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 251. A publicly-available version is in ADAMS under 
Accession No. ML18026B053; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and TSs.
    Date of initial notice in Federal Register: July 5, 2017 (82 FR 
31093). The supplemental letter dated February 28, 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 April 26, 2018.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois, and Docket 
Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle 
County, Illinois

    Date of amendment request: June 30, 2017.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.7.11, ``Control Room Ventilation (VC) Temperature 
Control System,'' to modify the TS Actions for two inoperable VC 
temperature control system trains.
    Date of issuance: April 30, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 195/195; 201/201. A publicly-available version is 
in ADAMS under Accession No. ML18054B436; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and 
NPF-66: The amendments revised the TSs and Licenses.
    Date of initial notice in Federal Register: August 29, 2017 (82 FR 
41068).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 30, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 18, 2017.
    Brief description of amendment: The amendment revised the design 
value for the spent fuel storage pool in Technical Specification 4.3.2, 
``Drainage,'' to an appropriate value, consistent with the original 
design basis.
    Date of issuance: April 30, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No: 217. A publicly-available version is in ADAMS under 
Accession No. ML18072A050; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-62: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: September 12, 2017 (82 
FR 42848).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 30, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 4, 2017, as supplemented by 
letter dated November 6, 2017.
    Brief description of amendments: The amendments revised the non-
destructive examination inspection interval for special lifting devices 
from annually or prior to each use, typically at each refueling outage, 
to a 10-year interval.
    Date of issuance: May 1, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 225 (Unit 1) and 212 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML18100A788; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-42 and DPR-60: The 
amendments revised the Prairie Island Nuclear Generating Plant, Unit 
Nos. 1 and 2, Updated Safety Analysis Report.
    Date of initial notice in Federal Register: September 26, 2017 (82 
FR 44855). The supplemental letter dated November 6, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 1, 2018.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, 
California

    Date of amendment request: September 28, 2017.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.1.4, ``Rod Group Alignment Limits''; TS 3.1.5, 
``Shutdown Bank Insertion Limits''; TS 3.1.6, ``Control Bank Insertion 
Limits''; and TS 3.1.7, ``Rod Position Indication,'' to adopt Technical 
Specifications Task Force (TSTF) Traveler TSTF-547, Revision 1, 
``Clarification of Rod Position Requirements.'' The NRC approved the 
TSTF and issued an associated model safety evaluation by letter dated 
March 4, 2016.
    Date of issuance: April 30, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 232 (Unit 1) and 234 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML18096A054; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: November 7, 2017 (82 FR 
51653).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 30, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 7, 2017, as supplemented by letters 
dated November 1, November 27, December 14, December 19 (four letters), 
and

[[Page 23741]]

December 22, 2017, and January 22, 2018.
    Brief description of amendment: The amendment revised the Renewed 
Facility Operating License and Technical Specifications to implement a 
measurement uncertainty recapture power uprate. Specifically, the 
amendment authorized an increase in the maximum licensed thermal power 
level from 3,840 megawatts thermal to 3,902 megawatts thermal, which is 
an increase of approximately 1.6 percent.
    Date of issuance: April 24, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 212. A publicly-available version is in ADAMS under 
Accession No. ML18096A542; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-57: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: October 3, 2017 (82 FR 
46098). The supplemental letters dated November 1, November 27, 
December 14, December 19 (four letters), and December 22, 2017, and 
January 22, 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 April 24, 2018.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-206, 50-361, 
and 50-362, San Onofre Nuclear Generating Station (SONGS), Unit Nos. 1, 
2, and 3, San Diego County, California

    Date of amendment request: December 19, 2016, as supplemented by 
letters dated April 25, 2017, and November 2, 2017.
    Brief description of amendments: The amendments replaced the SONGS, 
Unit Nos. 1, 2, and 3, Physical Security Plan, Training and 
Qualification Plan, and Safeguards Contingency Plan (the Security Plan) 
with an Independent Spent Fuel Storage Installation (ISFSI) Only 
Security Plan. The NRC staff determined that the proposed SONGS ISFSI-
Only Security Plan continues to meet the standards in 10 CFR 72.212, 
``Conditions of general license issued under Sec.  72.210,'' paragraph 
(b)(9). As such, the SONGS ISFSI-Only Security Plan provides reasonable 
assurance that adequate protective measures can and will be taken in 
the event of a design-basis threat of radiological sabotage related to 
the spent fuel. These changes more fully reflect the status of the 
facility, as well as the reduced scope of potential physical security 
challenges at the site once all spent fuel has been moved to dry cask 
storage within the onsite ISFSI, an activity that is currently 
scheduled for completion in 2019.
    Date of issuance: April 23, 2018.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days following Southern California Edison Company's submittal 
of a written certification to the NRC that all spent nuclear fuel 
assemblies have been transferred out of the spent fuel pools and placed 
in storage within the onsite ISFSI.
    Amendment Nos.: 170 (Unit 1), 238 (Unit 2), and 231 (Unit 3). A 
publicly-available version is in ADAMS under Accession No. ML17311A364; 
the Safety Evaluation enclosed with the amendments includes safeguards 
information and is withheld from public disclosure.
    Facility Operating License Nos. DPR-13, NPF-10, and NPF-15: The 
amendments revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: April 4, 2017 (82 FR 
16422).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 5, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 18, 2017, as supplemented by 
letter dated December 15, 2017.
    Description of amendments: The amendments authorized the Southern 
Nuclear Operating Company to depart from the Vogtle Electric Generating 
Plaint Updated Final Safety Analysis Report (UFSAR) Tier 2* and Tier 2 
information regarding changes necessary to reflect an increase in the 
design pressure of the main steam isolation valve (MSIV) compartments 
from 6.0 pounds per square inch (psi) to 6.5 psi and other changes 
regarding descriptions of the MSIV compartments. The Tier 2* changes 
affect Wall 11 information contained in UFSAR Subsections 3H.3.3, 
3H.5.1, and 3H.5.1.3. This change provides additional design margin for 
the MSIV Compartments A and B at the Vogtle Electric Generating Plant.
    Date of issuance: April 18, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 122 (Unit 3) and 121 (Unit 4). Publicly-available 
versions are in ADAMS Package Accession No. ML18085A932, which includes 
the Safety Evaluation that references documents related to these 
amendments.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendments 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: November 21, 2017 (82 
FR 55411). The supplemental letter dated December 15, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in the Safety Evaluation dated April 18, 2018.
    No significant hazards consideration comments received: No.

United States Maritime Administration (MARAD), Docket No. 50-238, 
Nuclear Ship SAVANNAH (NSS), Baltimore, Maryland

    Date of amendment request: October 31, 2017.
    Brief description of amendment: The amendment permits MARAD to 
begin dismantling and disposing of the NSS without prior approval of 
the NRC, consistent with existing decommissioning regulations.
    Date of issuance: April 23, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 15. A publicly-available version is in ADAMS under 
Accession No. ML18081A134.
    Facility Operating License No. NS-1: This amendment revised the 
License.
    Date of initial notice in Federal Register: February 13, 2018 (83 
FR 6235).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 23, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 23, 2017, as supplemented by letters 
dated January 16, 2018, and March 14, 2018.

[[Page 23742]]

    Brief description of amendments: The amendments revised plant 
Technical Specifications Table 3.7-2 and associated Table Notations, 
Table 3.7-4 and Table 4.1-1, reflecting the installation of the Class 
1E 4160V negative sequence voltage (open phase) protective circuitry at 
Surry Power Station, Unit Nos. 1 and 2, to address the potential for a 
consequential open phase condition that could exist on one or two 
phases of a primary offsite power source and that would not currently 
be detected and mitigated by the existing station electrical protection 
scheme.
    Date of issuance: May 3, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 292 (Unit No. 1) and 292 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML18106A007; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: The 
amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: October 10, 2017 (82 FR 
47040). The supplemental letters dated January 16, 2018, and March 14, 
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 3, 2018.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 14th day of May, 2018.

    For the Nuclear Regulatory Commission.
Tara Inverso,
Acting Deputy Director, Division of Operating Reactor Licensing, Office 
of Nuclear Reactor Regulation.
[FR Doc. 2018-10565 Filed 5-21-18; 8:45 am]
 BILLING CODE 7590-01-P