Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 10911-10927 [2018-04827]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 83, Number 49 (Tuesday, March 13, 2018)]
[Notices]
[Pages 10911-10927]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-04827]


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NUCLEAR REGULATORY COMMISSION

[NRC-2018-0045]


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 February 13 to February 26, 2018. The last 
biweekly notice was published on February 27, 2018.

DATES: Comments must be filed by April 12, 2018. A request for a 
hearing must be filed by May 14, 2018.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2018-0045. Address 
questions about NRC dockets to Jennifer Borges; telephone: 301-287-
9127; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: May Ma, Office of Administration, Mail 
Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

[[Page 10912]]


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

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2018-0045, facility name, unit 
number(s), plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
     Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2018-0045.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [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-2018-0045, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

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

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

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

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and petition for leave to intervene 
(petition) with respect to the action. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested persons should consult a 
current copy of 10 CFR 2.309. The NRC's regulations are accessible 
electronically from the NRC Library on the NRC's website at https://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of 
the regulations is available at the NRC's Public Document Room, located 
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. If a petition is filed, the 
Commission or a presiding officer will rule on the petition and, if 
appropriate, a notice of a hearing will be issued.
    As required by 10 CFR 2.309(d) the petition should specifically 
explain the reasons why intervention should be permitted with 
particular reference to the following general requirements for 
standing: (1) The name, address, and telephone number of the 
petitioner; (2) the nature of the petitioner's right under the Act to 
be made a party to the proceeding; (3) the nature and extent of the 
petitioner's property, financial, or other interest in the proceeding; 
and (4) the possible effect of any decision or order which may be 
entered in the proceeding on the petitioner's interest.
    In accordance with 10 CFR 2.309(f), the petition must also set 
forth the specific contentions which the petitioner seeks to have 
litigated in the proceeding. Each contention must consist of a specific 
statement of the issue of law or fact to be raised or controverted. In 
addition, the petitioner must provide a brief explanation of the bases 
for the contention and a concise statement of the alleged facts or 
expert opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to the specific sources and 
documents on which the petitioner intends to rely to support its 
position on the issue. The petition must include sufficient information 
to show that a genuine dispute exists with the applicant or licensee on 
a material issue of law or fact. Contentions must be limited to matters 
within the scope of the proceeding. The contention must be one which, 
if proven, would entitle the

[[Page 10913]]

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 https://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 https://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a 
digital ID certificate and a docket has been created, the participant 
can then submit adjudicatory documents. Submissions must be in Portable 
Document Format (PDF). Additional guidance on PDF submissions is 
available on the NRC's public website at https://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the 
time the document is submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
document on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before adjudicatory documents are 
filed so that they can obtain access to the documents via the E-Filing 
system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic 
Filing Help Desk through the ``Contact Us'' link located on the NRC's 
public website at https://www.nrc.gov/site-help/e-submittals.html, by 
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC

[[Page 10914]]

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

    Date of amendment request: September 14, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17261B255.
    Description of amendment request: The amendments would modify 
Catawba Nuclear Station, Units 1 and 2, Technical Specification (TS) 
3.7.8, ``Nuclear Service Water System (NSWS).'' Specifically, the 
proposed change would add a new Condition D for one NSWS pond return 
header being inoperable due to the NSWS being aligned for single pond 
return header operation with a Completion Time (CT) of 30 days. This 
would involve isolating one train of the NSWS pond return piping at the 
Auxiliary Building wall and maintaining the discharge crossover lines 
open between trains in the Auxiliary Building and Emergency Diesel 
Generator Buildings. This provides a common safety-related discharge 
path through the single remaining in-service pond return line. This 
alignment, single pond return header operation, allows a pond return 
header to be removed from service while a flow path is maintained 
through both trains of NSWS supplied equipment to the Standby Nuclear 
Service Water Pond (SNSWP).
    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 single pond return header operation configuration 
for NSWS operation and the associated proposed TS and TS Bases 
changes have been evaluated to assess their impact on plant 
operation and to ensure that the design basis safety functions of 
safety related systems are not adversely impacted. During single 
pond return header operation, the operating NSWS header will be able 
to discharge all required NSWS flow from safety related components. 
PRA [Probabilistic Risk Assessment] has demonstrated that due to the 
limited proposed time in the single pond return header 
configuration, the resultant plant risk remains acceptable.
    The purpose of this amendment request is to ultimately 
facilitate inspections and modifications of the NSWS Pond Return 
buried piping between the Auxiliary Building and the Discharge to 
the SNSWP. Therefore, NRC approval of this request will ultimately 
help to enhance the long-term structural integrity of the NSWS and 
will help to ensure the system's reliability for many years.
    In general, the NSWS serves as an accident mitigation system and 
cannot by itself initiate an accident or transient situation. The 
only exception is that the NSWS piping can serve as a source of 
floodwater to safety related equipment in the Auxiliary Building or 
in the diesel generator buildings in the event of a leak or a break 
in the system piping. The probability of such an event is not 
significantly increased as a result of this proposed request. Safety 
related NSWS piping is tested and inspected in accordance with all 
applicable in-service testing and in-service inspection 
requirements. Given the negligible influence of flooding events on 
the NSWS for the submittal configuration (i.e., no dominant 
contribution from floods), it is judged that the analyses assessing 
the influence of these events provide an acceptable evaluation of 
the contribution of the flood risk for the requested CT of 30 days. 
The proposed 30 day TS Required Action CT has been evaluated for 
risk significance and the results of this evaluation have been found 
acceptable. The probabilities of occurrence of accidents presented 
in the UFSAR [Updated Final Safety Analysis Report] will not 
increase as a result of implementation of this change. Because the 
PRA analysis supporting the proposed change yielded acceptable 
results, the NSWS will maintain its required availability in 
response to accident situations. Since NSWS availability is 
maintained, the response of the plant to accident situations will 
remain acceptable and the consequences of accidents presented in the 
UFSAR will not increase.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Implementation of this amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. The proposed request does not affect the basic operation 
of the NSWS or any of the systems that it supports. These include 
the Emergency Core Cooling System, the Containment Spray System, the 
Containment Valve Injection Water System, the Auxiliary Feedwater 
System, the Component Cooling Water System, the Control Room Area 
Ventilation System, the Control Room Area Chilled Water System,

[[Page 10915]]

the Auxiliary Building Filtered Ventilation Exhaust System, or the 
Diesel Generators. During proposed single pond return header 
operation, the NSWS will remain capable of fulfilling all of its 
design basis requirements.
    No new accident causal mechanisms are created as a result of NRC 
approval of this amendment request. No changes are being made to the 
plant, which will introduce any new type of accident outside those 
assumed in the UFSAR.
    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 the margin of safety?
    Response: No.
    Implementation of this amendment will not involve a significant 
reduction in any margin of safety. Margin of safety is related to 
the confidence in the ability of the fission product barriers to 
perform their design functions during and following an accident 
situation. These barriers include the fuel cladding, the reactor 
coolant system, and the containment system. The performance of these 
fission product barriers will not be impacted by implementation of 
this proposed TS amendment. During single pond return header 
operation, the NSWS and its supported systems will remain capable of 
performing their required functions. No safety margins will be 
impacted.
    The PRA analysis conducted for this proposed amendment 
demonstrated that the impact on overall plant risk remains 
acceptable during single pond return header operation. Therefore, 
there is not a significant reduction in the margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, Duke Energy 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kate B. Nolan, Deputy General Counsel, Duke 
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: October 3, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17277A855.
    Description of amendment request: The amendments would revise 
Surveillance Requirement (SR) 3.8.4.5 contained in Technical 
Specification (TS) 3.8.4, ``DC [Direct Current] Sources--Operating.''
    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 to the battery charger amperage requirements 
of SR 3.8.4.5 contained in TS 3.8.4 does not impact the physical 
function of plant structures, systems, or components (SSC) or the 
manner in which SSCs perform their design function. The proposed 
change does not authorize the addition of any new plant equipment or 
systems, nor does it alter the assumptions of any accident analyses. 
The DC electrical power system, including the battery chargers, is 
not an initiator of any accident sequence analyzed in the Updated 
Final Safety Analysis Report. Rather, the DC electrical power system 
supports operation of equipment used to mitigate accidents. 
Specifically, the purpose of the battery chargers is to continuously 
maintain their respective battery in a charged standby condition 
while providing power to the system loads. The proposed change does 
not adversely affect accident initiators or precursors, nor does it 
alter the design assumptions, conditions, and configuration or the 
manner in which the plant is operated and maintained.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change to the battery charger amperage requirements 
of SR 3.8.4.5 contained in TS 3.8.4 does not require any 
modification to the plant or change equipment operation. The 
proposed change will not introduce failure modes that could result 
in a new accident, and the change does not alter assumptions made in 
the safety analysis. Performance of battery testing is not a 
precursor to any accident previously evaluated. The proposed change 
will not alter the design configuration, or method of operation of 
plant equipment beyond its normal functional capabilities. The 
proposed change does not create any new credible failure mechanisms, 
malfunctions, or accident initiators.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from those that have been 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to the battery charger amperage requirements 
of SR 3.8.4.5 contained in TS 3.8.4 does not alter or exceed a 
design basis or safety limit. There is no change being made to 
safety analysis assumptions or the safety limits that would 
adversely affect plant safety as a result of the proposed change. 
Margins of safety are unaffected by the proposed change and the 
applicable requirements of 10 CFR 50.36(c)(2)(ii) and 10 CFR 50, 
Appendix A will continue to be met.
    Therefore, the proposed change does not involve any reduction in 
a margin of safety.

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

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

    Date of amendment request: December 12, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17346B280.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) 3.6.4.1, ``Secondary 
Containment,'' Surveillance Requirement (SR) 3.6.4.1.1. The proposed 
changes are based on Technical Specifications Task Force (TSTF) 
Traveler TSTF-551, Revision 3, ``Revise Secondary Containment 
Surveillance Requirements'' (ADAMS Accession No. ML16277A226).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change addresses conditions during which the 
secondary containment SRs are not met. The secondary containment is 
not an initiator of any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
increased. The consequences of an accident previously evaluated 
while utilizing the

[[Page 10916]]

proposed changes are no different than the consequences of an 
accident while utilizing the existing four hour Completion Time for 
an inoperable secondary containment. In addition, the proposed Note 
for SR 3.6.4.1.1 provides an alternative means to ensure the 
secondary containment safety function is met. As a result, the 
consequences of an accident previously evaluated are not 
significantly increased.
    Therefore, the proposed 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 analyzed?
    Response: No.
    The proposed change does not alter the protection system design, 
create new failure modes, or change any modes of operation. The 
proposed change does not involve a physical alteration of the plant; 
and no new or different kind of equipment will be installed. 
Consequently, there are no new initiators that could result in a new 
or different kind of accident.
    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 change addresses conditions during which the 
secondary containment SRs are not met. Conditions in which the 
secondary containment vacuum is less than the required vacuum are 
acceptable provided the conditions do not affect the ability of the 
SGT [Standby Gas Treatment] System to establish the required 
secondary containment vacuum under post-accident conditions within 
the time assumed in the accident analysis. This condition is 
incorporated in the proposed change by requiring an analysis of 
actual environmental and secondary containment pressure conditions 
to confirm the capability of the SGT System is maintained within the 
assumptions of the accident analysis. Therefore, the safety function 
of the secondary containment is not affected. The allowance for both 
an inner and outer secondary containment door to be open 
simultaneously for entry and exit does not affect the safety 
function of the secondary containment as the doors are promptly 
closed after entry or exit, thereby restoring the secondary 
containment boundary.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

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

    Date of amendment request: December 18, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17352B255.
    Description of amendment request: The amendment would revise the 
Environmental Protection Plan to incorporate the terms and conditions 
of the Incidental Take Statement included in the Biological Opinion 
issued to Energy Northwest by the National Marine Fisheries Service.
    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 changes are administrative in nature and would in no way 
affect the initial conditions, assumptions, or conclusions of 
Columbia's accident analyses. In addition, the proposed changes 
would not affect the operation or performance of any equipment 
assumed in the accident analyses.
    Therefore there is no 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 analyzed?
    Response: No.
    The changes are administrative in nature and would in no way 
impact or alter the configuration or operation of the facility and 
would create no new modes of operation.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The changes are administrative in nature and would in no way 
affect plant or equipment operation or the accident analysis.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

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

    Date of amendment request: December 11, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17354A007.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) Limiting Condition for Operation 3.7.13, 
``Spent Fuel Pit Storage,'' and TS 4.3, ``Fuel Storage.'' Specifically, 
the proposed changes would (1) resolve a nonconservative TS associated 
with TS Limiting Condition for Operation 3.7.13, (2) negate the need 
for the associated compensatory measures, and (3) remove credit for the 
installed Boraflex panels as a neutron absorber in the criticality 
analysis of record. The proposed changes in the criticality analysis of 
record would instead credit empty cells, rod cluster control assemblies 
(RCCAs), and neutron leakage along the outer two storage rows of the 
spent fuel pit (SFP).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment was evaluated for impact on the following 
previously evaluated events and accidents:

--Multiple Misloads
--Misplaced Assembly
--Dropped Assembly
--Misloaded Assembly
--Over Temperature
--Seismic
--Boron Dilution
--Fuel Handling Accident
--Loss of Spent Fuel Pool Cooling

Multiple Misloads, Misplaced Assembly, Dropped Assembly, Misloaded 
Accidents

    Operation in accordance with the proposed Technical 
Specifications will not significantly increase the probability of 
multiple misloads, misplaced assembly, dropped assembly and 
misloaded assembly accidents because:
    a. There are no changes to the equipment for fuel handling or 
how fuel assemblies are handled, including how fuel assemblies are 
inserted into and removed from SFP storage locations. There are no 
changes to how RCCAs will be handled, including how

[[Page 10917]]

RCCAs are inserted into, or removed from, a fuel assembly.
    b. The processes and procedures that are currently in place are 
sufficiently robust. The proposed Technical Specifications utilize 
the same basic fuel assembly classification and storage location 
concepts as those currently in place. However, they do represent a 
minimal increase in complexity:

--The current TS for fuel storage are complex because the Boraflex 
neutron absorber built into the SFP racks has degraded. To address 
this degradation the SFP is divided into four irregularly shaped 
Regions (Region 1-1, Region 1-2, Region 2-1, and Region 2-2). In 
addition to the four regions there are six special locations known 
as peripheral locations in Region 2-2 which are treated as suitable 
for storage of fuel otherwise designated for Region 1-1 or 1-2. 
These regions are graphically depicted in the current TS Figure 
3.7.13-5.

Each one of these regions has its own rules for fuel placement which 
are identified in the TS.
--The current Technical Specifications determine a minimum required 
burnup for each fuel assembly based on initial enrichment, burnup, 
and cooling time with individual fuel assembly storage location 
within the SFP restricted based on this minimum required burnup. The 
minimum required burnup is determined for each of the four regions 
(1-1, 1-2, 2-1, and 2-2) that utilize a total of ten curves. The 
proposed assembly categorization is slightly more complex due to the 
following:
    [cir] The minimum required burnup is dependent on the averaged 
assembly peaking factor in addition to the initial enrichment, 
burnup, and cooling time.
    [cir] the minimum required burnup is used to determine the 
reactivity category of each fuel assembly.
    [cir] the minimum required burnup is adjusted, as necessary, to 
account for hafnium inserts, a reconstituted fuel assembly with 
missing stainless steel replacement rods, and a maximum burnup 
average boron concentration in excess of 950 ppm [parts per 
million].
--The current Technical Specifications restrict acceptable SFP 
storage locations to Regions 1-1, 1-2, 2-1 and 2-2 based on minimum 
required burnup. The proposed Technical Specifications are minimally 
more complex due to the following:
    [cir] Acceptable storage locations are defined by fuel assembly 
category and a base configuration is specified. There are five 
reactivity categories. Certain cell locations in Region 2 require 
that Category 5 fuel assemblies contain a full length RCCA.
    [cir] the base configurations in Region 1 and Region 2 may be 
changed in accordance with certain well-defined criteria. An example 
of a change to a base configuration is that a checkerboard area may 
be formed in Region 2 where all four sides of the checkerboard are 
rows of empty cells.

    The minimal increase in complexity of current and future fuel 
categorization and SFP storage restrictions is offset by the 
significant number of fuel assemblies that have been pre-categorized 
in TS Tables 3.7.13-2 and Table 3.7.13-3. The minimal increase is 
also offset by the use of two curves to determine the minimum 
required burnup (instead of the 10 currently used).
    Operation in accordance with the proposed TS will not 
significantly increase the consequences of multiple misloads, 
misplaced assembly, dropped assembly and misloaded assembly 
criticality accidents because the proposed CSA [criticality safety 
analysis] demonstrates that the acceptance criteria continue to be 
met for each of these accidents.

Over Temperature Accident

    Operation in accordance with the proposed TS will not 
significantly increase the probability of an over temperature 
accident because the proposed change does not alter the manner in 
which the IP2 spent fuel cooling loop is designed, operated, or 
maintained.
    Operation in accordance with the proposed TS will not 
significantly increase the consequences of an over temperature 
accident because the proposed CSA demonstrates that the acceptance 
criteria continue to be met for this accident.

Seismic Event

    Operation in accordance with the proposed TS will not 
significantly increase the probability of a seismic event because 
there are no elements of the proposed changes that influence the 
occurrence of any natural event.
    Operation in accordance with the proposed TS will not 
significantly increase the consequences of a seismic event because 
the proposed changes do not significantly alter the physical 
arrangement of the spent fuel racks and do not increase the 
allowable number of fuel assemblies to be stored in the pool. The 
proposed TS changes require two cell blockers to be in place. These 
cell blockers have been evaluated and they have a negligible effect 
on the seismic response of the SFP racks. In addition, the proposed 
TS changes allow for the placement of miscellaneous non-actinide 
materials, for example, empty or full trash baskets in fuel 
positions of any category, in Water Holes and in 50% Water Holes. 
The placement of miscellaneous materials in the identified locations 
has been evaluated and has a negligible effect on the seismic 
response of the SFP racks.

Boron Dilution Accident

    Operation in accordance with the proposed TS will not 
significantly increase the probability of a boron dilution event 
because the proposed change does not alter the manner in which the 
IP2 spent fuel cooling system or any other plant system is designed, 
operated, or maintained, or otherwise increase the likelihood of 
adding significant quantities of unborated water into the spent fuel 
pit.
    Operation in accordance with the proposed TS will not 
significantly increase the consequences of a boron dilution event 
because the TS minimum soluble boron concentration remains unchanged 
at 2000 ppm and the boron concentration required to ensure 
keff less than or equal to 0.95 has been evaluated at 700 
ppm. The proposed CSA demonstrates that the acceptance criteria 
continue to be met for this accident.

Fuel Handling Accident

    Operation in accordance with the proposed TS will not 
significantly increase the probability of a[n] FHA [fuel handling 
accident] because the individual fuel assemblies will be moved using 
the same equipment, procedures, and other administrative controls 
(i.e. fuel move sheets) that are currently used.
    Operation in accordance with the proposed TS will not 
significantly increase the consequences of a[n] FHA because the 
radiological source term of a single fuel assembly will remain the 
same.

Loss of Spent Fuel Pool Cooling

    Operation in accordance with the proposed TS will not 
significantly increase the probability of a loss of spent fuel pit 
cooling because the proposed change does not alter the manner in 
which the IP2 spent fuel cooling loop is designed, operated, or 
maintained.
    Operation in accordance with the proposed TS will not 
significantly increase the consequences of a loss of spent fuel pit 
cooling because the proposed change credits empty cells whereas the 
thermal design basis for the spent fuel pit cooling loop provides 
for all fuel pit rack locations to be filled at the end of a full 
core discharge. The proposed TS changes require two cell blockers to 
be in place. These cell blockers have been evaluated and they have a 
negligible effect on the thermal response to a loss of spent fuel 
pool cooling. In addition, the proposed TS changes allow for the 
placement of miscellaneous non-actinide materials, for example, 
empty or full trash baskets in fuel positions of any category, in 
Water Holes and in 50% Water Holes. The placement of miscellaneous 
materials in the identified locations has been evaluated and has a 
negligible effect on the thermal response to a loss of spent fuel 
pool cooling.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Operation in accordance with the proposed TS do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. No new modes of operation are introduced by 
the proposed changes. The proposed changes will not create any 
failure mode not bounded by previously evaluated accidents.
    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.
    Operation in accordance with the proposed TS does not involve a 
significant reduction in a margin of safety.

[[Page 10918]]

    The margin of safety required by 10 CFR 50.68(b)(4) remains 
unchanged. The evaluations in the CSA confirm that operation in 
accordance with the proposed amendment continues to meet the 
required subcriticality margins for both normal operations and 
accident conditions. In addition, the SFP seismic and thermal 
margins are essentially unchanged.
    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: Bill Glew, Associate General Counsel, 
Entergy Services, Inc., 639 Loyola Avenue, 22nd Floor, New Orleans, LA 
70113.
    NRC Branch Chief: James G. Danna.

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3 (IP3), Westchester County, New York

    Date of amendment request: December 8, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17349A131.
    Description of amendment request: The amendment would allow for a 
one-time extension to the 15-year frequency of the IP3 containment 
leakage rate test (i.e., Integrated Leakage Rate Test (ILRT) or Type A 
test). Specifically, Technical Specification 5.5.15, ``Containment 
Leakage Rate Testing Program,'' would be revised to allow the existing 
ILRT frequency to be extended one time from 15 to 16 years. The next 
required ILRT test would be performed no later than the plant restart 
after the spring 2021 (3R21) refueling outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment involves changes to the IP3 containment 
leakage rate testing program. The proposed amendment does not 
involve a physical change to the plant or a change in the manner in 
which the plant is operated or controlled. The primary containment 
function is to provide an essentially leak tight barrier against the 
uncontrolled release of radioactivity to the environment for 
postulated accidents. As such, the containment itself and the 
testing requirements to periodically demonstrate the integrity of 
the containment exist to ensure the plant's ability to mitigate the 
consequences of an accident do not involve any accident precursors 
or initiators. Therefore, the probability of occurrence of an 
accident previously evaluated is not significantly increased by the 
proposed amendment.
    The proposed amendment adopts the NRC accepted guidelines of NEI 
[Nuclear Energy Institute] 94-01, Revision 3-A, for development of 
the IP3 performance-based testing program for the Type A testing. 
Implementation of these guidelines continues to provide adequate 
assurance that during design basis accidents, the primary 
containment and its components would limit leakage rates to less 
than the values assumed in the plant safety analyses. The potential 
consequences of extending the ILRT interval one-time to 16 years 
have been evaluated by analyzing the resulting changes in risk. The 
increase in risk in terms of person-rem per year within 50 miles 
resulting from design basis accidents was estimated to be acceptably 
small and determined to be within the guidelines published in the 
NRC Final Safety Evaluation for NEI Topical Report (TR) 94-01, 
Revision 3-A. Additionally, the proposed change maintains defense-
in-depth by preserving a reasonable balance among prevention of core 
damage, prevention of containment failure, and consequence 
mitigation. Entergy has determined that the increase in conditional 
containment failure probability due to the proposed change would be 
very small. Therefore, it is concluded that the proposed amendment 
does not significantly increase the consequences of an accident 
previously evaluated. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, for the establishment of a one-time only 16-
year interval for the performance of the containment ILRT. The 
containment and the testing requirements to periodically demonstrate 
the integrity of the containment exist to ensure the plant's ability 
to mitigate the consequences of an accident do not involve any 
accident precursors or initiators. The proposed change does not 
involve a physical change to the plant (i.e., no new or different 
type of equipment will be installed) or a change to the manner in 
which the plant is operated or controlled. 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 amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, for the establishment of a one-time only 16-
year interval for the performance of the containment ILRT. This 
amendment does not alter the manner in which safety limits, limiting 
safety system setpoints, or limiting conditions for operation are 
determined. The specific requirements and conditions of the 
containment leakage rate testing program, as defined in the TS, 
ensure that the degree of primary containment structural integrity 
and leak-tightness that is considered in the plant's safety analysis 
is maintained. The overall containment leakage rate limit specified 
by the TS is maintained, and the Type A, Type B, and Type C 
containment leakage tests would be performed at the frequencies 
established in accordance with the NRC accepted guidelines of NEI 
94-01, Revision 3-A. Containment inspections performed in accordance 
with other plant programs serve to provide a high degree of 
assurance that the containment would not degrade in a manner that is 
not detectable by an ILRT. A risk assessment using the current IP3 
PSA [probabilistic safety analysis] model concluded that extending 
the ILRT test interval one-time from 15 years to 16 years results in 
a very small change to the risk profile. 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: Bill Glew, Associate General Counsel, 
Entergy Services, Inc., 639 Loyola Avenue, 22nd Floor, New Orleans, LA 
70113.
    NRC Branch Chief: James G. Danna.

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

    Date of amendment request: January 9, 2017. A publicly-available 
version is in ADAMS under Accession No. ML18009B037.
    Description of amendment request: The proposed change would 
incorporate a revised alternative source term dose calculation 
resulting from the removal of a reduction factor credit for dual remote 
Control Room outside air intakes that had been previously misapplied. 
This would modify the loss-of-coolant accident (LOCA) dose calculation 
and the subsequent calculation results as described in the CPS Updated 
Safety Analysis Report and would revise the affected CPS Technical 
Specifications.
    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 10919]]

consequences of an accident previously evaluated?
    Response: No.
    The proposed change results in higher Control Room X/Qs 
[atmospheric dispersion values] which are equivalent to reduced 
atmospheric dispersion. The increased Control Room X/Qs, in turn, 
result in higher post-accident Control Room doses. Neither the 
higher X/Qs, nor the resultant increase in the Control Room doses 
affect any initiator or precursor of any accident previously 
evaluated. Therefore, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The proposed change results in an increase in the post-LOCA 
radiological dose to a Control Room occupant. However, the resultant 
post-LOCA Control Room dose remains within the regulatory limits of 
10 CFR 50.67 [, ``Accident source term''] and 10 CFR 50, Appendix A, 
``General Design Criteria for Nuclear Power Plants'' Criterion 19, 
``Control Room.'' Therefore, the proposed change does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    In summary, 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 does not alter the design function of 
operation of the Control Room heating, ventilation, and air-
conditioning (HVAC) system, or the ability of this system to perform 
its design function. The only change is the removal of the Control 
Room dose reduction factor credit taken for providing a dual remote 
Control Room air intake. The proposed change does not alter the 
safety limits, or safety analysis associated with the operation of 
the plant. Accordingly, the change does not introduce any new 
accident initiators. Rather, this proposed change is the result of 
an evaluation of the Control Room doses following the most limiting 
LOCA that can occur at CPS. The proposed change does not introduce 
any new modes of plant operation. As a result, no new failure modes 
are introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The revised post-LOCA dose consequences to a Control Room 
occupant were calculated in accordance with the requirements of 10 
CFR 50.67, [Regulatory Guide (RG)] 1.183, [``Alternative 
Radiological Source Terms for Evaluating Design Basis Accidents at 
Nuclear Power Reactors''] and NRC SRP [Standard Review Plan] Section 
15.0.1, ``Radiological Consequences Analyses Using Alternative 
Source Terms.''
    The margin of safety is considered to be that provided by 
meeting the applicable regulatory limits. The increased Control Room 
X/Qs result in an increase in Control Room dose following the design 
basis LOCA; however, since the Control Room dose following the 
design basis accident remains within the regulatory limits, there is 
not a significant reduction in a margin of safety.
    Therefore, operation of CPS in accordance with the proposed 
change will not involve a significant reduction in a margin of 
safety.

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

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

    Date of amendment request: January 24, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18024A275.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TSs) 3.7.2, ``Diesel Generator Cooling 
Water (DGCW) System''; 3.8.1, ``AC [Alternating Current] Sources-
Operating''; and the associated TS Bases to allow an extended period to 
install isolation valves to support replacing degraded Core Standby 
Cooling System (CSCS) piping.
    The proposed changes modify TS 3.7.2 to include a 7-day Completion 
Time (CT) when one or more required DGCW subsystem(s) are inoperable. 
The proposed changes to TS 3.8.1 include a 7-day CT when a Division 2 
Diesel Generator (DG) and the required opposite unit Division 2 DG are 
inoperable. The proposed changes will only be used during four 
refueling outages, two for Unit 1 prior to July 1, 2024, and two for 
Unit 2 prior to July 1, 2023. The current planned schedule, subject to 
change, is L2R17 (2019), L1R18 (2020), L2R18 (2021), and L1R19 (2022).
    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 previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed change 
does not have a detrimental impact on the integrity of any plant 
structure, system, or component that initiates an analyzed event. No 
active or passive failure mechanisms that could lead to an accident 
are affected. Non-Code line stops required to provide isolation will 
maintain the availability of the online unit's CSCS. The non-Code 
line stops being used to isolate the system during the specified 
refueling outages are being designed to the same or greater pressure 
rating and seismic requirements as the CSCS piping.
    Redundancy is provided by designing the CSCS as multiple 
independent subsystems. Divisional separation between subsystems 
assures that no single failure can affect more than one division's 
subsystem. Therefore, assuming a single failure in any division's 
subsystem including the subsystem shared between units, two other 
divisional subsystems in each unit will remain unaffected. This 
ensures adequate redundancy to supply the minimum required cooling 
water for safe shutdown of the operating unit or mitigate the 
consequences of an accident.
    The proposed limited use of increased CT's of the operating 
unit's CSCS maintains the design basis assumptions. 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 involves the temporary installation of new 
equipment (mechanical line stops) that will be designed and 
installed to the same or greater pressure rating and seismic design 
as the CSCS piping. The currently installed equipment will not be 
operated in a new or different manner. No new or different system 
interactions are created and no new processes are introduced. The 
proposed changes will not introduce any new failure mechanisms, 
malfunctions, or accident initiators not already considered in the 
design and licensing bases.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change does not alter any existing setpoints at 
which protective actions are initiated and no new setpoints or 
protective actions are introduced. The design and operation of the 
CSCS remains unchanged. The proposed change provides a limited 
period to restore inoperable DGCW subsystems and Division 2 DGs 
instead of interrupting plant operations, possibly requiring an 
orderly plant shutdown of the operating unit. The potential to avoid 
a plant transient in conjunction with maintaining availability of 
the DGCW subsystems and Division 2 DGs offsets any risk associated 
with the limited CT. The proposed change

[[Page 10920]]

does not impact a design basis, limiting safety system setting, or 
safety limit specified in TSs.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: December 15, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17363A067.
    Description of amendment request: The proposed amendment would 
revise the Emergency Plan for the DAEC to adopt the Nuclear Energy 
lnstitute's (NEl's) revised Emergency Action Level (EAL) scheme 
described in NEI 99-01, Revision 6, ``Development of Emergency Action 
Levels for Non-Passive Reactors,'' which has been endorsed by the NRC.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not impact the physical configuration 
or function of plant structures, systems, or components (SSCs) or 
the manner in which SSCs are operated, maintained, modified, tested, 
or inspected. No actual facility equipment or accident analyses are 
affected by the proposed changes.
    The change revises the NextEra Emergency Action Levels to be 
consistent with the NRC endorsed EAL scheme contained in NEI 99-01, 
Revision 6, ``Methodology for Development of Emergency Action 
Levels,'' but does not alter any of the requirements of the 
Operating License or the Technical Specifications.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
The proposed change does not create any new failure modes for 
existing equipment or any new limiting single failures. 
Additionally, the proposed change does not involve a change in the 
methods governing normal plant operation, and all safety functions 
will continue to perform as previously assumed in the accident 
analyses. Thus, the proposed change does not adversely affect the 
design function or operation of any structures, systems, and 
components important to safety.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The proposed change does not challenge the performance or integrity 
of any safety-related system. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety associated with the acceptance criteria of 
any accident is unchanged. The proposed change will have no affect 
[sic] on the availability, operability, or performance of safety-
related systems and components. The proposed change will not 
adversely affect the operation of plant equipment or the function of 
equipment assumed in the accident analysis.
    The proposed amendment does not involve changes to any safety 
analyses assumptions, safety limits, or limiting safety system 
settings. The changes do not adversely impact plant operating 
margins or the reliability of equipment credited in the safety 
analyses.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: December 19, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17353A928.
    Description of amendment request: The proposed amendment would 
separate the Linear Heat Generation Rate (LHGR) requirements and 
actions from the Average Planar Linear Heat Generation Rate (APLHGR) 
requirements and actions contained in Technical Specification (TS) 
3.2.1. The proposed amendment adds new TS 3.2.3, ``Linear Heat 
Generation Rate (LHGR),'' and modifies TS 1.1, ``Definitions,'' TS 
3.4.1, ``Recirculation Loops Operating,'' and TS 5.6.5, ``Core 
Operating Limits Report (COLR),'' to reflect the LHGR change. 
Modifications associated with TS 3.2.1 and the new TS 3.2.3 are also 
being added to the actions for TS 3.3.4.1, ``End of Cycle Recirculation 
Pump Trip (EOC-RPT) Instrumentation,'' and TS 3.7.7, ``The Main Turbine 
Bypass System.''
    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 separation of the LHGR requirements and actions from the 
APLHGR TS is an administrative change. No actions within the TS are 
changed. The addition of the LCO [limiting condition for operation] 
for APLHGR and the proposed LCO for LHGR to the LCO for 3.3.4.1, End 
of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation and the 
LCO for TS 3.7.7, Main Turbine Bypass System reflect within the TS 
requirements APLHGR and LHGR actions which are already occurring via 
the core monitoring processes in place. None of those changes affect 
any plant systems, structures, or components designed for the 
prevention or mitigation of previously evaluated accidents. No new 
equipment is added nor is installed equipment being changed or 
operated in a different manner.
    LHGR limits have been defined to provide sufficient margin 
between the steady-state operating condition and any fuel damage 
condition to accommodate uncertainties and to assure that no fuel 
damage results even during the worst anticipated transient condition 
at any time.
    The proposed change does not modify the limits, change 
assumptions for the accident analysis, or change operation of the 
station. Therefore, the proposed change does not involve an increase 
in the probability or consequences of a previously evaluated 
accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The separation of the LHGR requirements and actions from the 
APLHGR TS is an administrative change. No actions within the TS are 
changed. The addition of the LCO for APLHGR and the proposed LCO for 
LHGR to the LCO for 3.3.4.1, End of Cycle

[[Page 10921]]

Recirculation Pump Trip (EOC-RPT) Instrumentation and the LCO for TS 
3.7.7, Main Turbine Bypass System reflect within the TS requirements 
APLHGR and LHGR actions which are already occurring via the core 
monitoring processes in place. None of those changes affect any 
plant systems, structures, or components designed for the prevention 
or mitigation of previously evaluated accidents. No new equipment is 
added nor is installed equipment being changed or operated in a 
different manner.
    The proposed change does not modify the limits, change 
assumptions for the accident analysis, or change operation of the 
station. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is not affected by the separation of the 
LHGR requirements and actions from the APLHGR TS. Similarly, the 
margin of safety is not affected by the addition of the LCO for 
APLHGR and the proposed LCO for LHGR to the LCO for 3.3.4.1, End of 
Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation and the LCO 
for TS 3.7.7, Main Turbine Bypass System.
    Appropriate measures exist to control the values of these limits 
since it is required by TS that only NRC-approved methods be used to 
determine the limits. The proposed change continues to require 
operation within the core thermal limits as obtained from NRC-
approved reload design methodologies and the actions to be taken if 
a limit is exceeded remain unchanged, again, in accordance with 
existing TS.
    The proposed change does not modify the limits, change 
assumptions for the accident analysis, or change operation of the 
station. Therefore, the proposed change has no impact to the margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William Blair, P.O. Box 14000, Juno Beach, 
FL 33408-0420.

NRC Branch Chief: David J. Wrona.

PSEG Nuclear LLC and Exelon Generation Company, LLC, Docket Nos. 50-272 
and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey

    Date of amendment request: December 18, 2017, as supplemented by 
letter dated February 9, 2018. Publicly-available versions are in ADAMS 
under Accession Nos. ML17352A502 and ML18040A319, respectively.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 3/4.3.1, ``Reactor Trip System 
Instrumentation,'' and TS 3/4.3.2, ``Engineered Safety Feature 
Actuation System Instrumentation,'' to increase the completion times 
and bypass test times at Salem Nuclear Generating Station, Unit Nos. 1 
and 2. The proposed changes are consistent with NRC-approved Technical 
Specifications Task Force (TSTF) Travelers TSTF-411, Revision 1, 
``Surveillance Test Interval Extensions for Components of the Reactor 
Protection System (WCAP-15376-P),'' and TSTF-418, Revision 2, ``RPS 
[Reactor Protection System] and ESFAS [Engineered Safety Feature 
Actuation System] Test Times and Completion Times (WCAP-14333),'' or 
are supported by plant-specific analysis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the completion times and bypass test 
time reduce the potential for inadvertent reactor trips and spurious 
actuations, and therefore do not increase the probability of any 
accident previously evaluated. The proposed changes to the 
completion times and bypass test time do not change the response of 
the plant to any accidents and have an insignificant impact on the 
reliability of the reactor trip system and engineered safety feature 
actuation system (RTS and ESFAS) signals. The RTS and ESFAS will 
remain highly reliable and the proposed changes will not result in a 
significant increase in the risk of plant operation. This is 
demonstrated by showing that the impact on plant safety as measured 
by core damage frequency (CDF) is less than 1.0E-06 per year and the 
impact on large early release frequency (LERF) is less than 1.0E-07 
per year. In addition, for the completion time change, the 
incremental conditional core damage probabilities (ICCDP) and 
incremental conditional large early release probabilities (ICLERP) 
are less than 5.0E-7 and 5.0E-08, respectively. These changes meet 
the acceptance criteria in Regulatory Guides 1.174 and 1.177. 
Therefore, since the RTS and ESFAS will continue to perform their 
functions with high reliability as originally assumed, and the 
increase in risk as measured by CDF, LERF, ICCDP, ICLERP is within 
the acceptance criteria of existing regulatory guidance, there will 
not be a significant increase in the consequences of any accidents.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
The proposed changes are consistent with safety analysis assumptions 
and resultant consequences.
    Therefore, 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.
    The proposed changes do not result in a change in the manner in 
which the RTS and ESFAS provide plant protection. The RTS and ESFAS 
will continue to have the same setpoints after the proposed changes 
are implemented. There are no design changes associated with the 
license amendment. The changes to completion times and bypass test 
time do not change any existing accident scenarios, nor create any 
new or different accident scenarios.
    The proposed changes do not involve a modification to the 
physical configuration of the plant or changes in the methods 
governing normal plant operation. The proposed changes will not 
impose any new or different requirement or introduce a new accident 
initiator, accident precursor, or malfunction mechanism.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by these changes. Redundant RTS and ESFAS trains 
are maintained, and diversity with regard to the signals that 
provide reactor trip and engineered safety features actuation is 
also maintained. All signals credited as primary or secondary, and 
all operator actions credited in the accident analyses will remain 
the same. The proposed changes will not result in plant operation in 
a configuration outside the design basis. The calculated impact on 
risk is insignificant and meets the acceptance criteria contained in 
Regulatory Guides 1.174 and 1.177.
    Therefore, since the proposed changes do not impact the response 
of the plant to a design basis accident, the proposed changes do not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are

[[Page 10922]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: James G. Danna.

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

    Date of amendment request: December 20, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17354A964.
    Description of amendment request: The requested amendment proposes 
changes to Combined License Appendix C (and to plant-specific Tier 1 
information) and associated Tier 2 information to allow a pneumatic 
test to be used in lieu of a hydrostatic test for the Main Control Room 
Emergency Habitability System (VES) consistent with American Society of 
Mechanical Engineers (ASME) Section III.
    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 allow for pneumatic testing of the VES ASME 
Section III components and piping. ASME Section III, ND-6000 
contains the requirements for pressure testing of piping and 
components. ASME Section III, ND-6112.1(a) allows for a pneumatic 
test to be used in lieu of a hydrostatic test when components, 
appurtenances or systems cannot be readily dried and traces of the 
testing medium cannot be tolerated. Due to the design and layout of 
the VES, it may be difficult to dry the system following a 
hydrostatic test. Traces of water could result in sending a slug of 
water through the system or rust to form. Allowing for pneumatic 
testing continues to meet the ASME Section III code. The proposed 
changes do not affect the operation of the VES. The VES maintains 
its design function to maintain control room habitability.
    The proposed changes do not affect the operation of any systems 
or equipment that initiate an analyzed accident or alter any 
structures, systems, and components (SSCs) accident initiator or 
initiating sequence of events. Therefore, the probabilities of 
accidents previously evaluated 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 Updated Final Safety Analysis Report are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that may initiate a new or different kind of accident, 
or alter any SSC such that a new accident initiator or initiating 
sequence of events is created.
    The proposed changes do not affect any other SSC design 
functions or methods of operation in a manner that results in a new 
failure mode, malfunction, or sequence of events that affect safety-
related or nonsafety related equipment. Therefore, this activity 
does not allow for a new fission product release path, result in a 
new fission product barrier failure mode, or create a new sequence 
of events that result in significant fuel cladding failures.
    Therefore, the requested amendment does not create the 
possibility of a new or different 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 allow for pneumatic testing of the VES ASME 
Section III components and piping. The VES ASME Section III 
components and piping continue to meet the ASME Section III code. 
The proposed changes do not have any effect on the ability of the 
safety-related SSCs to perform their design basis functions. The 
proposed changes do not affect the ability of the VES to maintain 
control room habitability.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes, and no margin of 
safety is reduced. Therefore, the requested amendment does not 
involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: January 31, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18031B131.
    Description of amendment request: The requested amendment proposes 
changes to the Technical Specification (TS) 3.4.6, Pressurizer Safety 
Valve, Applicability to require the pressurizer safety valves (PSVs) to 
be operable when the TS 3.4.14, Low Temperature Overpressure Protection 
(LTOP), is not required to be operable. A conforming change to the TS 
3.4.6 Actions is also proposed. Additional TS changes necessary to 
support PSVs operability are proposed for consistency with the TS 3.4.6 
change.
    The request also proposes moving TS Limiting Condition for 
Operation Notes regarding reactor coolant pump starts from TS 3.4.4, 
Reactor Coolant System (RCS) Loops, 3.4.8, Minimum RCS Flow, and 
3.4.14, Low Temperature Overpressure Protection (LTOP), to TS 3.4.3, 
RCS Pressure/Temperature (P/T) Limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that initiate an analyzed accident or alter any 
structures, systems, and components (SSCs) accident initiator or 
initiating sequence of events.
    The proposed changes do not affect the physical design of SSCs 
related to the TS on Engineered Safety Features Actuation System 
(ESFAS), RCS P/T limits, RCS loops, RCS flow, pressurizer, PSVs, 
LTOP, or Reactor Vessel head vent (RVHV), as described in the 
Updated Final Safety Analysis Report (UFSAR). Therefore, the 
operation of the listed functions and components is not affected. 
Therefore, the proposed changes do not affect the probability of an 
accident previously evaluated.
    The proposed changes do not affect the physical design of SSCs 
related to the TS on ESFAS, RCS P/T limits, RCS loops, RCS flow, 
pressurizer, PSVs, LTOP, or RVHV to meet their design functions. The 
design of the functions and components continue to meet the same 
regulatory acceptance criteria, codes, and standards as stated in 
the UFSAR. In addition, the proposed changes maintain the 
capabilities of the ESFAS, RCS P/T

[[Page 10923]]

limits, RCS loops, RCS flow, pressurizer, PSVs, LTOP, or RVHV to 
mitigate the consequences of an accident and to meet the applicable 
regulatory acceptance criteria.
    The proposed changes do not affect the prevention and mitigation 
of other abnormal events (e.g., anticipated operational occurrences, 
earthquakes, floods, and turbine missiles), or their safety or 
design analyses. Therefore, the consequences of the accidents 
evaluated in the UFSAR are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that may initiate a new or different kind of accident, 
or alter any SSC such that a new accident initiator or initiating 
sequence of events is created.
    The proposed changes do not affect any other SSC design 
functions or methods of operation in a manner that results in a new 
failure mode, malfunction, or sequence of events that affect safety-
related or nonsafety related equipment. Therefore, this activity 
does not allow for a new fission product release path, result in a 
new fission product barrier failure mode, or create a new sequence 
of events that result in significant fuel cladding failures.
    Therefore, the requested amendment does not create the 
possibility of a new or different 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 maintain existing safety margins. The 
proposed changes verify and maintain the physical design of SSCs 
related to ESFAS, RCS P/T limits, RCS loops, RCS flow, pressurizer, 
PSVs, LTOP, and RVHV to perform their design functions. Therefore, 
the proposed changes satisfy the same design functions in accordance 
with the same codes and standards as stated in the UFSAR. These 
changes do not affect any design code, function, design analysis, 
safety analysis input or result, or design/safety margin.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes, and no margin of 
safety is reduced. Therefore, the requested amendment does not 
involve a significant reduction in margin of safety.

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

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

    Date of amendment request: February 2, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18037B114.
    Description of amendment request: The requested amendment proposes 
departures from the generic AP1000 Design Control Document (DCD) for 
the plant-specific VEGP Combined License (COL) Appendix A Technical 
Specifications (TS) and related departures from generic DCD Tier 2 
information in the Updated Final Safety Analysis Report (UFSAR) (which 
includes the plant-specific DCD Tier 2 information). Specifically, the 
proposed changes would make administrative changes to COL Appendix A, 
TS 5.6.3, for the core operating limits report required documentation 
to include analytical methods which are described elsewhere in the TS 
and in the UFSAR, and make an editorial change to COL Appendix A TS 
5.7.2 for high radiation areas to correct a typographical error.
    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 are administrative and editorial changes 
consistent with the requirements described elsewhere in the TS and 
in the UFSAR, and do not adversely affect the operation of any 
systems or equipment that initiate an analyzed accident or alter any 
structures, systems, and components (SSCs) accident initiator or 
initiating sequence of events. The proposed changes to the 
analytical methods approved for maintaining core operating limits do 
not result in any increase in probability of an analyzed accident 
occurring, and prevent power oscillations and maintain the initial 
conditions and operating limits required by the accident analysis, 
and the analyses of normal operation and anticipated operational 
occurrences, so that fuel design limits are not exceeded for events 
resulting in positive reactivity insertion and reactivity feedback 
effects, and so that the consequences of postulated accidents are 
not changed. The proposed changes do not adversely affect the 
ability of the automatic reactor trips to perform the required 
safety function to trip the reactor when necessary to protect fuel 
design limits, and do not adversely affect the probability of 
inadvertent operation or failure of the automatic reactor trips.
    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 are administrative and editorial changes 
consistent with the requirements described elsewhere in the TS and 
in the UFSAR, and 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 to the 
analytical methods approved for maintaining core operating limits do 
not result in any increase in probability of an analyzed accident 
occurring, and prevent power oscillations and maintain the initial 
conditions and operating limits required by the accident analysis, 
and the analyses of normal operation and anticipated operational 
occurrences, so that fuel design limits are not exceeded for events 
resulting in positive reactivity insertion and reactivity feedback 
effects, and so that the consequences of postulated accidents are 
not changed. The proposed changes do not adversely affect the 
ability of the automatic reactor trips to perform the required 
safety function to trip the reactor when necessary to protect fuel 
design limits, and do not adversely affect the probability of 
inadvertent operation or failure of the automatic reactor trips.
    These proposed changes do not adversely affect any other SSC 
design functions or methods of operation in a manner that results in 
a new failure mode, malfunction, or sequence of events that affect 
safety-related or nonsafety-related equipment. Therefore, this 
activity does not allow for a new fission product release path, 
result in a new fission product barrier failure mode, or create a 
new sequence of events that results in significant fuel cladding 
failures.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes are administrative and editorial changes 
consistent with the requirements described elsewhere in the TS and 
in the UFSAR, and maintain existing safety margins through continued 
application of the existing requirements of the UFSAR. The proposed 
changes maintain the initial conditions and operating limits 
required by the accident analysis, and the analyses of normal 
operation and anticipated operational occurrences, so that the 
existing fuel design limits specified in the UFSAR are not exceeded 
for events resulting in positive reactivity insertion and reactivity 
feedback effects, and so that the consequences of postulated 
accidents are not changed. Therefore, the proposed changes satisfy 
the same safety functions in accordance with the

[[Page 10924]]

same requirements as stated in the UFSAR. These changes do not 
adversely affect any design code, function, design analysis, safety 
analysis input or result, or design/safety margin.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant 
(Watts Bar), Units 1 and 2, Rhea County, Tennessee

    Date of amendment request: January 5, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18008A257.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.6.3, ``Containment Isolation Valves,'' 
and Surveillance Requirement 3.6.3.5 to change the frequency in 
accordance with the Watts Bar Containment Leakage Rate Testing Program, 
which is described in TS 5.7.2.19. The proposed change would allow leak 
rate testing of the containment purge system containment isolation 
valves to be performed at least once every 30 months, as prescribed in 
Regulatory Guide 1.163.
    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 deletes the augmented testing requirement 
for these containment isolation valves and allows the surveillance 
intervals to be set in accordance with the Containment Leakage Rate 
Testing Program. This change does not affect the system function or 
design. The purge valves are not an initiator of any previously 
analyzed accident. Leakage rates do not affect the probability of 
the occurrence of any accident. Operating history has demonstrated 
that the valves do not degrade and cause leakage as previously 
anticipated. Because these valves have been demonstrated to be 
reliable, these valves can be expected to perform the containment 
isolation function as assumed in the accident analyses. The proposed 
changes do not affect the source term, containment isolation, or 
radiological release assumptions used in evaluating the radiological 
consequences of an accident previously evaluated. Further, the 
proposed changes do not increase the types or amounts of radioactive 
effluent that may be released offsite, nor significantly increase 
individual or cumulative occupational/public radiation exposures. 
The proposed changes do not significantly increase the probability 
of an accident and are consistent with safety analysis assumptions 
and resultant consequences.
    Therefore, the changes do not increase the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This change does not involve a physical alteration to the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing any normal plant operation. The 
change does not alter assumptions made in the safety analyses or 
licensing basis. Extending the test intervals has no influence on, 
nor does it contribute in any way to, the possibility of a new or 
different kind of accident or malfunction from those previously 
analyzed. No change has been made to the design, function, or method 
of performing leakage testing. Leakage acceptance criteria have not 
changed. No new accident modes are created by extending the testing 
intervals. No safety-related equipment or safety functions are 
altered as a result of this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The only margin of safety that has the potential of being 
impacted by the proposed change involves the offsite dose 
consequences of postulated accidents, which are directly related to 
the containment leakage rate. The proposed change does not alter the 
method of performing the tests nor does it change the leakage 
acceptance criteria. Sufficient data has been collected to 
demonstrate these resilient seals do not degrade at an accelerated 
rate. Because of this demonstrated reliability, this change will 
provide sufficient surveillance to determine an increase in the 
unfiltered leakage prior to the leakage exceeding that assumed in 
the accident analysis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant, 
Unit 2, Rhea County, Tennessee

    Date of amendment request: October 11, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17284A452.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.3.1, Table 3.3.1-1, ``Reactor Trip 
System (RPS) Instrumentation,'' to increase the values for the nominal 
trip setpoint and the allowable value for Function 14.a, ``Turbine 
Trip--Low Fluid Oil Pressure.'' The proposed changes are due to the 
planned replacement and relocation of the pressure switches from the 
low pressure auto-stop trip fluid oil header to the high pressure 
turbine electrohydraulic control (EHC) oil header. The changes are 
needed due to the higher EHC system operating pressure.
    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 reflects a design change to the turbine 
control system that results in the use of an increased control oil 
pressure system, necessitating a change to the value at which a low 
fluid oil pressure initiates a reactor trip on turbine trip. The low 
fluid oil pressure is an input to the reactor trip instrumentation 
in response to a turbine trip event. The value at which the low 
fluid oil initiates a reactor trip is not an accident initiator. A 
change in the nominal control oil pressure does not introduce any 
mechanisms that would increase the probability of an accident 
previously analyzed. The reactor trip on turbine trip function is 
initiated by the same protective signal as used for the existing 
auto stop low fluid oil system trip signal. There is no change in 
form or function of this signal and the probability or consequences 
of previously analyzed accidents are not impacted.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?

[[Page 10925]]

    Response: No.
    The EHC fluid oil pressure rapidly decreases in response to a 
turbine trip signal. The value at which the low fluid oil pressure 
switches initiates a reactor trip is not an accident initiator. The 
proposed TS change reflects the higher pressure that will be sensed 
after the pressure switches are relocated from the auto stop low 
fluid oil system to the EHC high pressure header. Failure of the new 
switches would not result in a different outcome than is considered 
in the current design basis. Further, the change does not alter 
assumptions made in the safety analysis but ensures that the 
instruments perform as assumed in the accident analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The change involves a parameter that initiates an anticipatory 
reactor trip following a turbine trip. The safety analyses do not 
credit this anticipatory trip for reactor core protection. The 
original pressure switch configuration and the new pressure switch 
configuration both generate the same reactor trip signal. The 
difference is that the initiation of the trip will now be adjusted 
to a different system of higher pressure. This system function of 
sensing and transmitting a reactor trip signal on turbine trip 
remains the same. There is no impact to safety analysis acceptance 
criteria as described in the plant licensing basis because no change 
is made to the accident analysis assumptions.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

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

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

    Date of amendment request: September 14, 2017, as supplemented by 
letter dated December 12, 2017.
    Brief description of amendments: The amendments modified Technical 
Specifications (TSs) to allow temporary changes to TSs 3.5.2, 
``Emergency Core Cooling Systems (ECCS)--Operating''; 3.6.6, 
``Containment Spray System''; 3.7.5, ``Auxiliary Feedwater (AFW) 
System''; 3.7.6, ``Component Cooling Water (CCW) System''; 3.7.7, 
``Nuclear Service Water System (NSWS)''; 3.7.9, ``Control Room Area 
Ventilation System (CRAVS)''; 3.7.11, ``Auxiliary Building Filtered 
Ventilation Exhaust System (ABFVES)''; and 3.8.1, ``[Alternating 
Current] Sources--Operating,'' to permit the ``A'' Train NSWS to be 
inoperable for a total of 14 days to address a non-conforming condition 
on the ``A'' Train supply piping from the Standby Nuclear Service Water 
Pond.
    Date of issuance: February 15, 2018.
    Effective date: These license amendments are effective as of its 
date of issuance and shall be implemented within 120 days of issuance.
    Amendment Nos.: 308 (Unit 1) and 287 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML18030A682; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Renewed Licenses and TSs.
    Date of initial notice in Federal Register: December 19, 2017 (82 
FR 60226).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 15, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 27, 2017.
    Brief description of amendment: The amendment revised the Technical 
Specification (TS) requirements in order to address Generic Letter 
2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay 
Heat Removal, and Containment Spray Systems,'' dated January 11, 2008, 
as described in Technical Specifications Task Force (TSTF) Traveler 
TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas 
Accumulation.''
    Date of issuance: February 16, 2018.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 246. A publicly-available version is in ADAMS under 
Accession No. ML18025A213; 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: June 6, 2017 (82 FR 
26132).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 16, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 14, 2017.
    Brief description of amendment: The amendment revised the ANO-1 
Technical Specification (TS)

[[Page 10926]]

requirements for unavailable barriers by adding Limiting Condition for 
Operation (LCO) 3.0.9, which allows a delay time for entering a 
supported system TS when the inoperability is solely due to an 
unavailable barrier. The change is consistent with Technical 
Specification Task Force (TSTF)-427, Revision 2, ``Allowance for Non 
Technical Specification Barrier Degradation Supported System 
OPERABILITY.'' In addition, the amendment corrected a typographical 
omission on TS page 3.0-3, which was editorial in nature.
    Date of issuance: February 26, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 259. A publicly-available version is in ADAMS under 
Accession No. ML18033A175; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: October 24, 2017 (82 FR 
49236).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 26, 2018.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
2 (ANO-2), Pope County, Arkansas

    Date of amendment request: August 14, 2017.
    Brief description of amendment: The amendment revised the ANO-2 
Technical Specification (TS) requirements for unavailable barriers by 
adding Limiting Condition for Operation (LCO) 3.0.9, which allows a 
delay time for entering a supported system TS when the inoperability is 
solely due to an unavailable barrier. The change is consistent with 
Technical Specification Task Force (TSTF)-427, Revision 2, ``Allowance 
for Non Technical Specification Barrier Degradation Supported System 
OPERABILITY.''
    Date of issuance: February 26, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 309. A publicly-available version is in ADAMS under 
Accession No. ML18051A589; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: October 24, 2017 (82 FR 
49237).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 26, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 28, 2017.
    Brief description of amendments: The amendments revised the Calvert 
Cliffs, Units 1 and 2, Technical Specifications (TSs) to change the low 
level of the refueling water tank to reflect a needed increase in the 
required borated water volume and change the allowable value of the 
refueling water tank level-low function.
    Date of issuance: February 15, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of the end of CC1R24 refueling outage for Calvert 
Cliffs, Unit 1, and within 60 days of the end of CC2R23 refueling 
outage for Calvert Cliffs, Unit 2.
    Amendment Nos.: 323 (Unit 1) and 301 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML18029A195; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: June 19, 2017 (82 FR 
27887).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 15, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 29, 2017.
    Brief description of amendments: The amendments revised the 
Technical Specification (TS) requirements for mode change limitations 
in TS 3.0.4 and TS 4.0.4 based on Technical Specifications Tasks Force 
(TSTF) Improved Standard Technical Specifications Change Traveler, 
TSTF-359, Revision 9, ``Increase Flexibility in Mode Restraints'' 
(ADAMS Accession No. ML031190607).
    Date of issuance: February 20, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 278 (Unit 3) and 273 (Unit 4). A publicly-available 
version is in ADAMS under Accession No. ML18018A559; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: September 12, 2017 (82 
FR 42850).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 20, 2018.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

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

    Date of amendment request: February 13, 2017, as supplemented by 
letter dated August 11, 2017.
    Brief description of amendments: The amendments adopted the NRC-
endorsed Nuclear Energy Institute (NEI) 99-01, Revision 6, 
``Development of Emergency Action Levels for Non-Passive Reactors.''
    Date of issuance: February 16, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within a 365-day period after issuance.
    Amendment Nos.: Salem--322 (Unit No. 1) and 303 (Unit No. 2); Hope 
Creek--210. A publicly-available version is in ADAMS under Accession 
No. ML17355A570; documents related to these amendments are listed in 
the Safety Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-70, DPR-75, and NPF-57: 
The amendments revised the emergency action level technical bases 
documents.
    Date of initial notice in Federal Register: March 28, 2017 (82 FR 
15384). The supplemental letter dated

[[Page 10927]]

August 11, 2017, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the NRC staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 16, 2018.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: May 26, 2016, as supplemented by letter 
dated October 26, 2017.
    Brief description of amendments: The amendments correct a non-
conservative Technical Specification (TS) Surveillance Requirement 
acceptance criterion for the diesel generator steady-state frequency in 
Limiting Condition for Operation 3.8.1, ``AC [Alternating Current] 
Sources--Operating.''
    Date of issuance: February 12, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 341--Unit 1 and 334--Unit 2. A publicly-available 
version is in ADAMS under Accession No. ML18026A810; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-77 and DPR-79. 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: August 2, 2016 (81 FR 
50740). The supplemental letter dated October 26, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 12, 2018.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, on March 6, 2018.

    For the Nuclear Regulatory Commission.
Gregory F. Suber,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2018-04827 Filed 3-12-18; 8:45 am]
 BILLING CODE 7590-01-P