[Federal Register Volume 83, Number 39 (Tuesday, February 27, 2018)]
[Notices]
[Pages 8509-8523]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-03727]


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NUCLEAR REGULATORY COMMISSION

[NRC-2018-0031]


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 January 30, 2018, to February 12, 2018. The 
last biweekly notice was published on February 13, 2018.

DATES: Comments must be filed by March 29, 2018. A request for a 
hearing must be filed by April 30, 2018.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0031. Address 
questions about NRC dockets to Jennifer Borges; telephone: 301-287-
9127; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: May Ma, Office of Administration, Mail 
Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Paula Blechman, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-2242, email: [email protected].

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2018-0031, facility name, unit 
number(s), plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
     Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0031.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in DAMS) 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-2018-0031, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

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

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

[[Page 8510]]

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

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

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and petition for leave to intervene 
(petition) with respect to the action. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested persons should consult a 
current copy of 10 CFR 2.309. The NRC's regulations are accessible 
electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of 
the regulations is available at the NRC's Public Document Room, located 
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. If a petition is filed, the 
Commission or a presiding officer will rule on the petition and, if 
appropriate, a notice of a hearing will be issued.
    As required by 10 CFR 2.309(d) the petition should specifically 
explain the reasons why intervention should be permitted with 
particular reference to the following general requirements for 
standing: (1) The name, address, and telephone number of the 
petitioner; (2) the nature of the petitioner's right under the Act to 
be made a party to the proceeding; (3) the nature and extent of the 
petitioner's property, financial, or other interest in the proceeding; 
and (4) the possible effect of any decision or order which may be 
entered in the proceeding on the petitioner's interest.
    In accordance with 10 CFR 2.309(f), the petition must also set 
forth the specific contentions which the petitioner seeks to have 
litigated in the proceeding. Each contention must consist of a specific 
statement of the issue of law or fact to be raised or controverted. In 
addition, the petitioner must provide a brief explanation of the bases 
for the contention and a concise statement of the alleged facts or 
expert opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to the specific sources and 
documents on which the petitioner intends to rely to support its 
position on the issue. The petition must include sufficient information 
to show that a genuine dispute exists with the applicant or licensee on 
a material issue of law or fact. Contentions must be limited to matters 
within the scope of the proceeding. The contention must be one which, 
if proven, would entitle the petitioner to relief. A petitioner who 
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene. 
Parties have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that party's admitted 
contentions, including the opportunity to present evidence, consistent 
with the NRC's regulations, policies, and procedures.
    Petitions must be filed no later than 60 days from the date of 
publication of this notice. Petitions and motions for leave to file new 
or amended contentions that are filed after the deadline will not be 
entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in 
accordance with the filing instructions in the ``Electronic Submissions 
(E-Filing)'' section of this document.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to establish when the hearing is held. If the final determination is 
that the amendment request involves no significant hazards 
consideration, the Commission may issue the amendment and make it 
immediately effective, notwithstanding the request for a hearing. Any 
hearing would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of the amendment unless the Commission finds an imminent 
danger to the health or safety of the public, in which case it will 
issue an appropriate order or rule under 10 CFR part 2.
    A State, local governmental body, Federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission no later 
than 60 days from the date of publication of this notice. The petition 
must be filed in accordance with the filing instructions in the 
``Electronic Submissions (E-Filing)'' section of this document, and 
should meet the requirements for petitions set forth in this section, 
except that under 10 CFR 2.309(h)(2) a State, local governmental body, 
or Federally recognized Indian Tribe, or agency thereof does not need 
to address the standing requirements in 10 CFR 2.309(d) if the facility 
is located within its boundaries. Alternatively, a State, local 
governmental body, Federally-recognized Indian Tribe, or agency thereof 
may participate as a non-party under 10 CFR 2.315(c).
    If a hearing is granted, any person who is not a party to the 
proceeding and is not affiliated with or represented by a party may, at 
the discretion of the presiding officer, be permitted to make a limited 
appearance pursuant to the provisions of 10 CFR 2.315(a). A person 
making a limited appearance may make an oral or written statement of 
his or her position on the issues but may not otherwise participate in 
the proceeding. A limited appearance may be made at any session of the 
hearing or at any prehearing conference, subject to the limits and 
conditions as may be imposed by the presiding officer. Details 
regarding the opportunity to make a limited appearance will be provided 
by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

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

[[Page 8511]]

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

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

    Date of amendment request: October 9, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17283A248.
    Description of amendment request: The amendment would revise 
Limiting Condition for Operation (LCO) 3.10.1, to expand its scope to 
include provisions for temperature excursions greater than 200 degrees 
Fahrenheit ([deg]F) as a consequence of inservice leak and hydrostatic 
testing, and as a consequence of scram time testing initiated in 
conjunction with an inservice leak or hydrostatic test, while 
considering operational conditions to be in Mode 4. This change is 
consistent with NRC approved Technical Specification Task Force (TSTF) 
Improved Standard Technical Specification Change Traveler, TSTF-484, 
``Use of TS 3.10.1 for Scram Time Testing Activities,'' Revision 0.
    The NRC staff issued a Notice of Availability for TSTF-484 in the 
Federal Register on October 27, 2006 (71 FR 63050). The staff also 
issued a Federal Register notice on August 21, 2006 (71 FR 48561), that 
provided a model safety evaluation and a model no significant hazards 
consideration (NSHC) determination that licensees could reference in 
their plant-specific application. In its application dated October 9, 
2017, the licensee affirmed the applicability of the model NSHC 
determination for Fermi 2.
    Basis for proposed no NSHC determination: As required by 10 CFR 
50.91(a), the licensee affirmed the applicability of the model NSHC, 
which is presented below:

    Criterion 1: The proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 8512]]

    Technical Specifications currently allow for operation at 
greater than 200[emsp14][deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact the probability or 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.
    Criterion 2: The proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Technical Specifications currently allow for operation at 
greater than 200[emsp14][deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. No new operational conditions beyond those currently allowed by 
LCO 3.10.1 are introduced. The changes do not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements or eliminate any existing requirements. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are consistent with the safety analysis assumptions 
and current plant operating practice. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Criterion 3: The proposed change does not involve a significant 
reduction in a margin of safety.
    Technical Specifications currently allow for operation at 
greater than 200[emsp14][deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact any margin of safety. Allowing completion of 
inspections and testing and supporting completion of scram time 
testing initiated in conjunction with an inservice leak or 
hydrostatic test prior to power operation results in enhanced safe 
operations by eliminating unnecessary maneuvers to control reactor 
temperature and pressure. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the above 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: Jon P. Christinidis, DTE Energy, Expert 
Attorney--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-
1279.
    NRC Branch Chief: David J. Wrona.

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

    Date of amendment request: May 2, 2017, as supplemented by letters 
dated July 20 and November 21, 2017. Publicly-available versions are in 
ADAMS under Accession Nos. ML17122A116, ML17201Q132, and ML17325A588, 
respectively.
    Description of amendment request: The amendments would modify CNS 
Technical Specifications (TSs) to extend the Completion Time (CT) of TS 
3.8.1, ``AC [Alternating Current] Sources--Operating,'' Required Action 
B.6 (existing Required Action B.4, numbered as B.6) for an inoperable 
emergency diesel generator (DG) from 72 hours to 14 days. A conforming 
change is also proposed to extend the maximum CT of TS 3.8.1 Required 
Actions A.3 and B.4. To support this request, the licensee will add a 
supplemental power source (i.e., two supplemental diesel generators 
(SDGs) per station) with the capability to power any emergency bus. The 
SDGs will have the capacity to bring the affected unit to cold 
shutdown. Additionally, the amendments would modify TS 3.8.1 to add new 
two limiting conditions for operation (LCOs), TS LCO 3.8.1.c and TS LCO 
3.8.1.d, to ensure that at least one train of shared components has an 
operable emergency power supply.
    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 involves extending the TS CT for an 
inoperable DG at CNS [. . .]. The proposed change also involves 
adding a new Required Action to TSs to ensure that at least one 
train of shared components at CNS [. . .] has an operable emergency 
power supply whenever one DG is inoperable. The DGs at both stations 
are safety related components which provide a backup electrical 
power supply to the onsite emergency power distribution system. The 
proposed change does not affect the design of the DGs, the 
operational characteristics or function of the DGs, the interfaces 
between the DGs and other plant systems or the reliability of the 
DGs. The DGs are not accident initiators; the DGs are designed to 
mitigate the consequences of previously evaluated accidents 
including a loss of offsite power. Extending the CT for a single DG 
would not affect the previously evaluated accidents since the 
remaining DGs supporting the redundant engineered safety feature 
systems would continue to be available to perform the accident 
mitigation functions. Thus, allowing a DG to be inoperable for an 
additional 11 days for performance of maintenance or testing does 
not increase the probability of a previously evaluated accident.
    Deterministic and probabilistic risk assessment techniques 
evaluated the effect of the proposed TS change to extend the CT for 
an inoperable DG on the availability of an electrical power supply 
to the plant emergency safeguards feature systems. These assessments 
concluded that the proposed CNS [. . .] TS change does not involve a 
significant increase in the risk of power supply unavailability.
    There is a small incremental risk associated with continued 
operation for an additional 11 days with one DG inoperable; however, 
the calculated impact provides risk metrics consistent with the 
acceptance guidelines contained in Regulatory Guides 1.177 and 
1.174. The remaining operable DGs and paths are adequate to supply 
electrical power to the onsite emergency power distribution system. 
A DG is required to operate only if both offsite power sources fail 
and there is an event which requires operation of the plant 
engineered safety features such as a design basis accident. The 
probability of a design basis accident occurring during this period 
is low.
    The consequences of previously evaluated accidents will remain 
the same during the proposed 14 day CT as during the current CNS [. 
. .] 72 hour CT. The ability of the remaining TS required DGs to 
mitigate the consequences of an accident will not be affected since 
no additional failures are postulated while equipment is inoperable 
within the TS CT.
    Regarding the proposed change to add Required Action to ensure 
that at least one train of shared components has an operable 
emergency power supply, there is no change to how or under what 
conditions offsite circuits or DGs are operated nor are there any 
changes to acceptable operating parameters. Power source operability 
requirements for shared components are being moved from the TS Bases 
to TS with the proposed change. The proposed change will ensure that 
at least one train of shared components has an operable emergency 
power supply whenever a DG is inoperable.
    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 involves extending the TS CT for an 
inoperable DG at CNS [. . .]. The proposed change also involves 
adding a new Required Action to TSs to ensure that at least one 
train of shared components at CNS [. . .] has an operable emergency 
power supply whenever one DG is inoperable.
    The proposed change does not involve a change in the CNS [. . .] 
plant design, plant

[[Page 8513]]

configuration, system operation or procedures involved with the DGs. 
The proposed change allows a DG to be inoperable for additional 
time. Equipment will be operated in the same configuration and 
manner that is currently allowed and designed for. The functional 
demands on credited equipment is unchanged. There are no new failure 
modes or mechanisms created due to plant operation for an extended 
period to perform DG maintenance or testing. Extended operation with 
an inoperable DG does not involve any modification to the 
operational limits or physical design of plant systems. There are no 
new accident precursors generated due to the extended CT.
    Regarding the proposed change to add Required Action to ensure 
that at least one train of shared components has an operable 
emergency power supply, there is no change to how or under what 
conditions offsite circuits or DGs are operated nor are there any 
changes to acceptable operating parameters. Power source operability 
requirements for shared components are being moved from the TS Bases 
to TS with the proposed change. The proposed change will ensure that 
at least one train of shared components has an operable emergency 
power supply whenever a DG is inoperable. This change does not alter 
the nature of events postulated in the Updated Final Safety Analysis 
Report nor does it introduce any unique precursor mechanisms.
    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.
    The proposed change involves extending the TS CT for an 
inoperable DG at CNS [. . .]. The proposed change also involves 
adding a new Required Action to TSs to ensure that at least one 
train of shared components at CNS [. . .] has an operable emergency 
power supply whenever one DG is inoperable.
    Currently, if an inoperable DG is not restored to operable 
status within 72 hours at CNS [. . .], TS 3.8.1, requires the units 
to be in Mode 3 (i.e., Hot Standby) within a CT of 6 hours, and to 
be in Mode 5 (i.e., Cold Shutdown) within a CT of 36 hours. The 
proposed TS changes will allow steady state plant operation at 100 
percent power for an additional 11 days for performance of DG 
planned reliability improvements and preventive and corrective 
maintenance.
    Deterministic and probabilistic risk assessment techniques 
evaluated the effect of the proposed TS change to extend the CT for 
an inoperable DG on the availability of an electrical power supply 
to the plant emergency safeguards feature systems. These assessments 
concluded that the proposed CNS [. . .] TS change does not involve a 
significant increase in the risk of power supply unavailability.
    The DGs continue to meet their design requirements; there is no 
reduction in capability or change in design configuration. The DG 
response to loss of offsite power, loss of coolant accident, station 
blackout or fire scenarios is not changed by this proposed 
amendment; there is no change to the DG operating parameters. In the 
extended CT, as in the existing CT, the remaining operable DGs and 
paths are adequate to supply electrical power to the onsite 
emergency power distribution system. The proposed change to extend 
the CT for an inoperable DG does not alter a design basis safety 
limit; therefore, it does not significantly reduce the margin of 
safety. The DGs will continue to operate per the existing design and 
regulatory requirements.
    The proposed TS changes (i.e., the inoperable DG CT extension 
request and proposed change to add Required Action to ensure that at 
least one train of shared components has an operable emergency power 
supply) do not alter the plant design nor do they change the 
assumptions contained in the safety analyses. The standby AC power 
system is designed with sufficient redundancy such that a DG may be 
removed from service for maintenance or testing. The remaining DGs 
are capable of carrying sufficient electrical loads to satisfy the 
Updated Final Safety Analysis Report requirements for accident 
mitigation or unit safe shutdown. The proposed change does not 
impact the redundancy or availability requirements of offsite power 
circuits or change the ability of the plant to cope with a station 
blackout. Therefore, based on the considerations given above, the 
proposed changes do not involve a significant reduction in the 
margin of safety.
    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: 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 Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2 (MNS), Mecklenburg County, North 
Carolina

    Date of amendment request: May 2, 2017, as supplemented by letters 
dated July 20 and November 21, 2017. Publicly-available versions are in 
ADAMS under Accession Nos. ML17122A116, ML17201Q132, and ML17325A588, 
respectively.
    Description of amendment request: The amendments would modify MNS 
Technical Specifications (TSs) to extend the Completion Time (CT) of TS 
3.8.1, ``AC [Alternating Current] Sources--Operating,'' Required Action 
B.6 (existing Required Action B.4, numbered as B.6) for an inoperable 
emergency diesel generator (DG) from 72 hours to 14 days. A conforming 
change is also proposed to extend the maximum CT of TS 3.8.1 Required 
Actions A.3 and B.4. To support this request, the licensee will add a 
supplemental power source (i.e., two supplemental diesel generators 
(SDGs) per station) with the capability to power any emergency bus. The 
SDGs will have the capacity to bring the affected unit to cold 
shutdown. Additionally, the amendments would modify TS 3.8.1 to add new 
two limiting conditions for operation (LCOs), TS LCO 3.8.1.c and TS LCO 
3.8.1.d, to ensure that at least one train of shared components has an 
operable emergency power supply.
    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 involves extending the TS CT for an 
inoperable DG at [. . .] MNS. The proposed change also involves 
adding a new Required Action to TSs to ensure that at least one 
train of shared components at [. . .] MNS has an operable emergency 
power supply whenever one DG is inoperable. The DGs at both stations 
are safety related components which provide a backup electrical 
power supply to the onsite emergency power distribution system. The 
proposed change does not affect the design of the DGs, the 
operational characteristics or function of the DGs, the interfaces 
between the DGs and other plant systems or the reliability of the 
DGs. The DGs are not accident initiators; the DGs are designed to 
mitigate the consequences of previously evaluated accidents 
including a loss of offsite power. Extending the CT for a single DG 
would not affect the previously evaluated accidents since the 
remaining DGs supporting the redundant engineered safety feature 
systems would continue to be available to perform the accident 
mitigation functions. Thus, allowing a DG to be inoperable for an 
additional 11 days for performance of maintenance or testing does 
not increase the probability of a previously evaluated accident.
    Deterministic and probabilistic risk assessment techniques 
evaluated the effect of the proposed TS change to extend the 
[completion time] CT for an inoperable DG on the availability of an 
electrical power supply to the plant emergency safeguards feature 
systems. These assessments concluded that the proposed [. . .] MNS 
TS change does not involve a significant increase in the risk of 
power supply unavailability.

[[Page 8514]]

    There is a small incremental risk associated with continued 
operation for an additional 11 days with one DG inoperable; however, 
the calculated impact provides risk metrics consistent with the 
acceptance guidelines contained in Regulatory Guides 1.177 and 
1.174.
    The remaining operable DGs and paths are adequate to supply 
electrical power to the onsite emergency power distribution system. 
A DG is required to operate only if both offsite power sources fail 
and there is an event which requires operation of the plant 
engineered safety features such as a design basis accident. The 
probability of a design basis accident occurring during this period 
is low.
    The consequences of previously evaluated accidents will remain 
the same during the proposed 14 day CT as during the current [. . .] 
MNS 72 hour CT. The ability of the remaining TS required DGs to 
mitigate the consequences of an accident will not be affected since 
no additional failures are postulated while equipment is inoperable 
within the TS CT.
    Regarding the proposed change to add Required Action to ensure 
that at least one train of shared components has an operable 
emergency power supply, there is no change to how or under what 
conditions offsite circuits or DGs are operated nor are there any 
changes to acceptable operating parameters. Power source operability 
requirements for shared components are being moved from the TS Bases 
to TS with the proposed change. The proposed change will ensure that 
at least one train of shared components has an operable emergency 
power supply whenever a DG is inoperable.
    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 involves extending the TS CT for an 
inoperable DG at [. . .] MNS. The proposed change also involves 
adding a new Required Action to TSs to ensure that at least one 
train of shared components at [. . .] MNS has an operable emergency 
power supply whenever one DG is inoperable.
    The proposed change does not involve a change in the [. . .] MNS 
plant design, plant configuration, system operation or procedures 
involved with the DGs. The proposed change allows a DG to be 
inoperable for additional time. Equipment will be operated in the 
same configuration and manner that is currently allowed and designed 
for. The functional demands on credited equipment is unchanged. 
There are no new failure modes or mechanisms created due to plant 
operation for an extended period to perform DG maintenance or 
testing. Extended operation with an inoperable DG does not involve 
any modification to the operational limits or physical design of 
plant systems. There are no new accident precursors generated due to 
the extended CT.
    Regarding the proposed change to add Required Action to ensure 
that at least one train of shared components has an operable 
emergency power supply, there is no change to how or under what 
conditions offsite circuits or DGs are operated nor are there any 
changes to acceptable operating parameters. Power source operability 
requirements for shared components are being moved from the TS Bases 
to TS with the proposed change. The proposed change will ensure that 
at least one train of shared components has an operable emergency 
power supply whenever a DG is inoperable. This change does not alter 
the nature of events postulated in the Updated Final Safety Analysis 
Report nor does it introduce any unique precursor mechanisms.
    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.
    The proposed change involves extending the TS CT for an 
inoperable DG at [. . .] MNS. The proposed change also involves 
adding a new Required Action to TSs to ensure that at least one 
train of shared components at [. . .] MNS has an operable emergency 
power supply whenever one DG is inoperable.
    Currently, if an inoperable DG is not restored to operable 
status within 72 hours at [. . .] MNS, TS 3.8.1, requires the units 
to be in Mode 3 (i.e., Hot Standby) within a CT of 6 hours, and to 
be in Mode 5 (i.e., Cold Shutdown) within a CT of 36 hours. The 
proposed TS changes will allow steady state plant operation at 100 
percent power for an additional 11 days for performance of DG 
planned reliability improvements and preventive and corrective 
maintenance.
    Deterministic and probabilistic risk assessment techniques 
evaluated the effect of the proposed TS change to extend the CT for 
an inoperable DG on the availability of an electrical power supply 
to the plant emergency safeguards feature systems. These assessments 
concluded that the proposed [. . .] MNS TS change does not involve a 
significant increase in the risk of power supply unavailability.
    The DGs continue to meet their design requirements; there is no 
reduction in capability or change in design configuration. The DG 
response to loss of offsite power, loss of coolant accident, station 
blackout or fire scenarios is not changed by this proposed 
amendment; there is no change to the DG operating parameters. In the 
extended CT, as in the existing CT, the remaining operable DGs and 
paths are adequate to supply electrical power to the onsite 
emergency power distribution system. The proposed change to extend 
the CT for an inoperable DG does not alter a design basis safety 
limit; therefore, it does not significantly reduce the margin of 
safety. The DGs will continue to operate per the existing design and 
regulatory requirements.
    The proposed TS changes (i.e., the inoperable DG CT extension 
request and proposed change to add Required Action to ensure that at 
least one train of shared components has an operable emergency power 
supply) do not alter the plant design nor do they change the 
assumptions contained in the safety analyses. The standby AC power 
system is designed with sufficient redundancy such that a DG may be 
removed from service for maintenance or testing. The remaining DGs 
are capable of carrying sufficient electrical loads to satisfy the 
Updated Final Safety Analysis Report requirements for accident 
mitigation or unit safe shutdown. The proposed change does not 
impact the redundancy or availability requirements of offsite power 
circuits or change the ability of the plant to cope with a station 
blackout. Therefore, based on the considerations given above, the 
proposed changes do not involve a significant reduction in the 
margin of safety.
    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: 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.

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

    Date of amendment request: November 20, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17326A387.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) to replace the current pressure-
temperature limits for heatup, cooldown, and the inservice leak 
hydrostatic tests for the reactor coolant system presented in TS 3.4.9 
that expire at 32 Effective Full Power Years (EFPY) with limitations 
that extend out to 54 EFPY.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will revise the pressure-temperature (P-T) 
limits for heatup, cooldown, and inservice leak hydrostatic test 
limitations for the Reactor Coolant System

[[Page 8515]]

(RCS) to a maximum of 54 Effective Full Power Years (EFPY) in 
accordance with 10 CFR 50, Appendix G. This is the end of the period 
of extended operation for the renewed ANO-2 operating License. The 
P-T limits were developed in accordance with the requirements of 10 
CFR 50, Appendix G, utilizing the analytical methods and flaw 
acceptance criteria of Topical Report WCAP-14040, Revision 4, and 
American Society of Mechanical Engineers (ASME) Code, Section XI, 
Appendix G. These methods and criteria are the previously NRC 
approved standards for the preparation of P-T limits. Updating the 
P-T limits for additional EFPYs maintains the level of assurance 
that reactor coolant pressure boundary integrity will be maintained, 
as specified in 10 CFR 50, Appendix G.
    The proposed changes do not adversely affect accident initiators 
or precursors, and do not alter the design assumptions, conditions, 
or configuration of the plant or the manner in which the plant is 
operated and maintained. The ability of structures, systems, and 
components to perform their intended safety functions is not altered 
or prevented by the proposed changes, and the assumptions used in 
determining the radiological consequences of previously evaluated 
accidents are not affected.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes implement methodologies that have been 
approved by the NRC (provided that any conditions/limitations are 
satisfied). The P-T limits will ensure the protection consistent 
with assuring the integrity of the reactor coolant pressure boundary 
as was previously evaluated. Reactor coolant pressure boundary 
integrity will continue to be maintained in accordance with 10 CFR 
50, Appendix G, and the assumed accident performance of plant 
structures, systems and components will not be affected. These 
changes do not involve any physical alteration of the plant (i.e., 
no new or different type of equipment will be installed), and 
installed equipment is not being operated in a new or different 
manner. Thus, no new failure modes are introduced.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from an accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not affect the function of the reactor 
coolant pressure boundary or its response during plant transients. 
By calculating the P-T limits using NRC-approved methodology, 
adequate margins of safety relating to reactor coolant pressure 
boundary integrity are maintained. The proposed changes do not alter 
the manner in which safety limits, limiting safety system settings, 
or limiting conditions for operation are determined. These changes 
will ensure that protective actions are initiated and the 
operability requirements for equipment assumed to operate for 
accident mitigation are not affected.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

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

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

    Date of amendment request: December 14, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17348A150.
    Description of amendment request: The amendment would revise ANO-2 
Technical Specification (TS) 3.3.3.6, ``Post-Accident 
Instrumentation,'' to ensure that both Category 1 and Type A Regulatory 
Guide (RG) 1.97, Revision 3, ``Instrumentation for Light-Water-Cooled 
Nuclear Power Plants to Assess Plant and Environs Conditions During and 
Following an Accident,'' instrumentation is included in the 
specification (unless already addressed within another specification) 
and gains greater consistency with NUREG-1432, Revision 4, ``Standard 
Technical Specifications for Combustion Engineering Plants.''
    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 PAM [Post-Accident Monitoring] instrumentation is not an 
initiator of any design basis accident or event and, therefore, the 
proposed change does not increase the probability of any accident 
previously evaluated. The proposed change ensures required 
instrumentation is included in and controlled by the station TSs and 
does not change the response of the plant to any accidents.
    The proposed change does not adversely affect accident 
initiators or precursors, nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The removal and addition of 
specific instrumentation within ANO-2 TS 3.3.3.6 is consistent with 
the ANO-2 SAR [Safety Analysis Report], Table 7.5-3 RG 1.97 
variables classified as Type A or Category 1 variables. 
Modifications to the TS Actions associated with inoperable 
instrumentation are consistent with the current ANO-2 licensing 
basis or act to improve consistency with NUREG 1432. The proposed 
change does not adversely affect the ability of structures, systems, 
and components (SSCs) to perform the associated intended safety 
function to mitigate the consequences of an initiating event within 
the assumed acceptance limits. The proposed change does not affect 
the source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of any 
accident previously evaluated. Further, the proposed change does not 
increase the types and amounts of radioactive effluent that may be 
released offsite, nor significantly increase individual or 
cumulative occupational/public radiation exposures.
    Instrumentation that does not meet the RG 1.97 inclusion 
criteria as established in NUREG-1432 are removed from the TS; 
however, the instrumentation remains applicable to other RG 1.97 
criteria and is maintained accordingly. Instrumentation added to the 
ANO-2 PAM TS does not change the manner in which the instrumentation 
is currently maintained since these instruments are currently 
designated as Type A and/or Category 1 variables in the ANO-2 SAR. 
However, including these instruments within the TSs will now require 
different mitigating actions during periods of inoperability, which 
may include a plant shutdown, establishment of alternate monitoring 
methods, and/or submittal of a special report to the NRC.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not result in a change in the manner in 
which the plant is operated during post-accident conditions and does 
not change the established mitigating actions associated with any 
necessary response to a DBA [design-basis accident]. The proposed 
change continues to ensure important instrumentation remains 
available to station operators such that currently established 
mitigating actions are not impacted. The change does not involve a 
physical alteration of the plant (i.e., no new or different type of 
equipment will be installed) or a change in the methods governing 
normal or post-accident plant operation. The change does not alter 
assumptions made in the safety analysis.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from an accident previously evaluated.

[[Page 8516]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
and assumptions are not impacted by the proposed change. The 
proposed change will not result in plant operation in a 
configuration outside the design basis. The proposed change ensures 
appropriate PAM instrumentation is controlled by the station TSs and 
that specified remedial action will be taken when required 
instrumentation is inoperable. The proposed change continues to 
support the operator ability to monitor and control vital systems 
during post-accident conditions.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

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

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

    Date of amendment request: November 3, 2017, as supplemented by 
letters dated December 6, 2017, and January 22, 2018. Publicly-
available versions are in ADAMS under Accession Nos. ML17307A440, 
ML17340B025, and ML18022A598, respectively.
    Description of amendment request: The amendment would revise the 
GGNS Updated Final Safety Analysis Report (UFSAR) to incorporate the 
Tornado Missile Risk Evaluator (TMRE) methodology contained in Nuclear 
Energy Institute (NEI) 17-02, Revision 1, ``Tornado Missile Risk (TMRE) 
Industry Guidance Document,'' September 2017 (ADAMS Accession No. 
ML17268A036). This methodology can only be applied to discovered 
conditions where tornado missile protection is not currently provided, 
and cannot be used to avoid providing tornado missile protection in the 
plant modification process.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment is to incorporate the TMRE methodology 
into the GGNS UFSAR. The TMRE methodology is an alternative 
methodology for determining whether protection from tornado-
generated missiles is required. The methodology can only be applied 
to discovered conditions where tornado missile protection was not 
provided, and cannot be used to avoid providing tornado missile 
protection in the plant modification process.
    The proposed amendment does not involve an increase in the 
probability of an accident previously evaluated. The relevant 
accident previously evaluated is a Design Basis Tornado impacting 
the GGNS site. The probability of a Design Basis Tornado is driven 
by external factors and is not affected by the proposed amendment. 
There are no changes required to any of the previously evaluated 
accidents in the UFSAR.
    The proposed amendment does not involve a significant increase 
in the consequences or a Design Basis Tornado. [The methodology as 
proposed does not alter any input assumptions or results of the 
accident analyses. Instead, it reflects a methodology to more 
realistically evaluate the probability of unacceptable consequences 
of a Design Basis Tornado. As such, there is no significant increase 
in the consequence of an accident previously evaluated. A similar 
consideration would apply in the event additional non-conforming 
conditions are discovered in the future.]
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed amendment is to incorporate the TMRE methodology 
into the GGNS UFSAR. The TMRE methodology is an alternative 
methodology for determining whether protection from tornado-
generated missiles is required. The methodology can only be applied 
to discovered conditions where tornado missile protection was not 
provided, and cannot be used to avoid providing tornado missile 
protection in the plant modification process.
    The proposed amendment will involve no physical changes to the 
existing plant, so no new malfunctions could create the possibility 
of a new or different kind of accident. The proposed amendment makes 
no changes to conditions external to the plant that could create the 
possibility of a new or different kind of accident. The proposed 
change will not create the possibility of a new or different kind of 
accident due to new accident precursors, failure mechanisms, 
malfunctions, or accident initiators not considered in the design 
and licensing bases. The existing Updated Final Safety Analysis 
Report accident analysis will continue to meet requirements for the 
scope and type of accidents that require analysis.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident than those 
previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed amendment is to incorporate the TMRE methodology 
into the GGNS UFSAR. The TMRE methodology is an alternative 
methodology for determining whether protection from tornado-
generated missiles is required. The methodology can only be applied 
to discovered conditions where tornado missile protection was not 
provided, and cannot be used to avoid providing tornado missile 
protection in the plant modification process.
    The change does not exceed or alter any controlling numerical 
value for a parameter established in the UFSAR or elsewhere in the 
GGNS licensing basis related to design basis or safety limits. The 
change does not impact any UFSAR Chapter 6 or 15 Safety Analyses, 
and those analyses remain valid. The change does not reduce 
diversity or redundancy as required by regulation or credited in the 
UFSAR. The change does not reduce defense-in-depth as described in 
the UFSAR.
    Therefore, the changes associated with this license amendment 
request do not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's modified 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 B. Glew, Associate General Counsel, 
Entergy Services, Inc., 440 Hamilton Avenue, White Plains, New York 
10601.
    NRC Branch Chief: Douglas A. Broaddus.

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

    Date of amendment request: December 21, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17355A184.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TSs) pertaining to the Engineered Safety 
Features Actuation System instrumentation to resolve non-conservative 
actions associated with the containment ventilation isolation and the 
control room ventilation isolation functions. In addition, the 
amendments would revise the control room

[[Page 8517]]

ventilation isolation function to no longer credit containment 
radiation monitoring instrumentation, eliminate redundant radiation 
monitoring instrumentation requirements, eliminate select core 
alterations applicability requirements, relocate radiation monitoring 
and reactor coolant system leakage detection requirements within the 
TSs to align with their respective functions, and relocate the spent 
fuel pool area monitoring requirements to licensee-controlled 
documents.
    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 instrumentation associated with the proposed changes to the 
technical specifications (TS) is not an initiator of any accidents 
previously evaluated, so the probability of accidents previously 
evaluated is unaffected by the proposed changes. There is no change 
to any equipment response or accident scenario, with the exception 
of the Control Room isolation on Containment high-radiation 
instrumentation function which impose no additional challenges to 
fission product barrier integrity. The exception is supported by 
revised radiological analyses which demonstrate that the Control 
Room air intake radioactivity monitoring instrumentation provides 
timely automatic isolation of the Control Room ventilation system 
and thereby limits Control Room operator doses to within regulatory 
limits for any design basis accident. The proposed changes also 
eliminate limitations imposed on Containment and Control Room 
ventilation instrumentation during CORE ALTERATIONS since the 
applicable postulated accidents do not result in fuel cladding 
integrity damage. Hence, the capability of any TS-required SSC 
[structure, system, or component] to perform its specified safety 
function is not impacted by the proposed changes and the outcomes of 
accidents previously evaluated are unaffected. Therefore, the 
proposed changes do not result in a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
The changes do not challenge the integrity or performance of any 
safety-related systems. No plant equipment is installed or removed, 
and the changes do not alter the design, configuration, or method of 
operation of any plant SSC with the exception of the Control Room 
isolation on Containment high-radiation instrumentation function 
which is supported by revised accident analyses which demonstrate 
that the radiological consequences remain within applicable 
regulatory limits. The elimination of core alterations applicability 
requirements do not impact the outcome of any applicable postulated 
accident since none result in fuel cladding damage. No physical 
changes are made to the plant, so no new causal mechanisms are 
introduced. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The ability of any operable SSC to perform its designated safety 
function is unaffected by the proposed changes. The proposed change 
do not revise any safety limits or limiting safety system settings. 
The proposed changes revises safety analyses assumptions and the 
method of operating the plant with regard to the Control Room 
isolation on Containment high-radiation instrumentation function. 
The changes are supported by revised accident analyses which 
demonstrate that no adverse impact will result to either the plant 
operating margins or the reliability of equipment credited in the 
safety analyses. The existing margin in dose assessment currently 
afforded Control Room operators during any design basis accident is 
maintained. No other safety margins are impacted by the proposed 
changes. Therefore, the proposed changes do not involve a 
significant reduction in the margin of safety.

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

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

    Date of amendment request: November 10, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17318A240.
    Description of amendment request: The proposed amendment revises 
Technical Specification (TS) 3.6.4.1, ``Secondary Containment,'' 
Surveillance Requirement (SR) 3.6.4.1.2. The SR is modified to 
acknowledge that secondary containment access openings may be open for 
entry and exit.
    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 SR is not met. The secondary containment is 
not an initiator of any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
increased. The consequences of an accident previously evaluated 
while utilizing the proposed changes are no different than the 
consequences of an accident while utilizing the existing four-hour 
Completion Time for an inoperable secondary containment. As a 
result, the consequences of an accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    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 SR is not met. The allowance for both an inner 
and outer secondary containment door to be open simultaneously for 
entry and exit does not affect the safety function of the secondary 
containment as the doors are promptly closed after entry or exit, 
thereby restoring the secondary containment boundary.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

[[Page 8518]]

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

    Date of amendment request: December 19, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17353A189.
    Description of amendment request: The proposed amendment would 
adopt Technical Specifications Task Force (TSTF) traveler TSTF-425, 
``Relocate Surveillance Frequencies to Licensee Control--RITSTF [Risk-
Informed Technical Specifications Task Force Initiative 5b,'' Revision 
3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

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

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: November 30, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17334B211.
    Description of amendment request: The proposed changes include 
changes to the Updated Final Safety Analysis Report (UFSAR) in the form 
of departures from the incorporated plant-specific Design Control 
Document (DCD) Tier 2* and Tier 1 information and related changes to 
the VEGP Units 3 and 4 Combined License (COL) Appendix C information. 
Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from the 
elements of the design as certified in 10 CFR part 52, Appendix D, 
design certification rule is also requested for the plant-specific Tier 
1 material departures. This submittal requests approval of the license 
amendment, necessary to implement these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), licensee has provided 
its analysis of the issue on no significant hazards consideration 
determination, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed consistency and editorial changes to COL Appendix C 
(and associated plant-specific Tier 1) and Tier 2 and Tier 2* 
information in the UFSAR do not involve a technical change, (e.g. 
there is no design parameter or requirement, calculation, analysis, 
function or qualification change). No structure, requirement, 
calculation, analysis, function or qualification change). No 
structure, system, or component (SSC) design or function would be 
affected. No design or safety analysis would be affected. The 
proposed changes do not affect any accident initiating event or 
component failure, thus the probabilities of the accidents 
previously evaluated are not affected. No function used to mitigate 
a radioactive material release and no radioactive material release 
source term is involved, thus the radiological releases in the 
accident analyses 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 consistency and editorial changes to COL Appendix C 
(and associated plant specific Tier 1) and Tier 2 and Tier 2* 
information in the UFSAR do not change the design or functionality 
of safety-related SSCs. The proposed change does not affect plant 
electrical systems, and does not affect the design function, 
support, design, or operation of mechanical and fluid systems. The 
proposed change does not result in a new failure mechanism or 
introduce any new accident precursors. No design function described 
in the UFSAR is affected by the proposed changes.
    Therefore, the 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 consistency and editorial changes to COL Appendix C 
(and associated plant specific Tier 1) and Tier 2 and Tier 2* 
information in the UFSAR do not involve any change to the design as 
described in the COL. There would be no change to an existing design 
basis, design function, regulatory criterion, or analysis. No safety 
analysis or design basis acceptance limit/criterion is involved.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.


[[Page 8519]]


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

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

    Date of amendment request: February 1, 2018. A publicly-available 
version is in ADAMS under Accession No. ML18032A359.
    Description of amendment request: The requested amendment proposes 
changes to relax the minimum gap requirement above grade between the 
nuclear island and the annex building/turbine building and removing the 
minimum gap requirement for the radwaste building from the Inspections, 
Tests, Analyses and Acceptance Criteria. Pursuant to the provisions of 
10 CFR 52.63(b)(1), an exemption from elements of the design as 
certified in the 10 CFR part 52, Appendix D, design certification rule 
is also requested for the plant-specific Design Control Document Tier 1 
material departures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff edits in square 
brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes are to relax the minimum gap requirement 
above grade between the nuclear island and the annex building/
turbine building from a 4 inch gap to a 3 inch gap. The proposed 
changes modify and clarify the gap requirements between the nuclear 
island and the annex building/turbine building and radwaste 
building, respectively. The proposed change deletes the gap 
requirement for the radwaste building from the Inspections, Tests, 
Analyses and Acceptance Criteria (ITAAC) in (COL) [Combined License] 
Appendix C. 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 structure, system or component (SSC) such 
that a new accident initiator or initiating sequence of events is 
created.
    The changes do not impact the support, design, or operation of 
mechanical and fluid systems. The changes do not impact the support, 
design, or operation of any safety-related structures. There is no 
change to plant systems or the response of systems to postulated 
accident conditions. There is no change to the predicted radioactive 
releases due to normal operation or postulated accident conditions. 
The plant response to previously evaluated accidents or external 
events is not adversely affected, nor do the proposed changes create 
any new accident precursors.
    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 to relax the minimum gap requirement 
above grade between the nuclear island and the annex building/
turbine building from a 4 inch gap to a 3 inch gap. The proposed 
changes modify and clarify the gap requirements between the nuclear 
island and the annex building/turbine building and radwaste 
building, respectively. The proposed changes delete the gap 
requirement for the radwaste building from the ITAAC in COL Appendix 
C. 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 adversely affect the design function 
of the nuclear island and adjoining buildings' SSC design functions 
or methods of operation in a manner that results in a new failure 
mode, malfunction, or sequence of events that affect safety-related 
or non-safety-related equipment. This activity does not allow for a 
new fission product release path, result in a new fission product 
barrier failure mode, or create a new sequence of events that result 
in significant fuel cladding failures.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes maintain existing safety margin and provide 
adequate protection through continued application of the existing 
requirements in the UFSAR [Updated Final Safety Analysis Report]. 
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 adversely affect any design code, function, design 
analysis, safety analysis input or result, or design/safety margin. 
No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes.
    Because no safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by these changes, no significant 
margin of safety is reduced.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.

Southern Nuclear Operating Company, Inc., 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. ML18031B142.
    Description of amendment request: The requested amendment proposes 
to include changes to Combined License (COL) Appendix A, Technical 
Specifications related to fuel management. Specifically, the requested 
amendment proposes improvements to the technical specifications for the 
Rod Position Indication, the Control Rod Drive Mechanism, Power Range 
Neutron Flux Channels and the Mechanical Shim Augmentation.
    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 to clarify proper operation and 
methodology associated with the DRPI [Digital Rod Position 
Indication], Control Rod Gripper Coils, instrumentation associated 
with Quadrant Power Tilt Ratio, or Control or Gray Rods. These 
changes do not affect the operation of this equipment and have no 
adverse impact on their design functions.
    The changes do not involve an interface with any structure, 
system, or component (SSC) accident initiator or initiating sequence 
of events, and thus, the probabilities of the accidents evaluated in 
the plant-specific Updated Final Safety Analysis Report (UFSAR) are 
not affected. The proposed changes do not adversely affect any 
mitigation sequence or the predicted radiological releases due to 
postulated accident conditions, thus, the consequences of the 
accidents evaluated in the UFSAR are not affected.

[[Page 8520]]

    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 verify and maintain the capabilities of the 
DRPI, Control Rod Gripper Coils, instrumentation associated with 
Quadrant Power Tilt Ratio, and Control and Gray Rods to perform 
their design functions. 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. These 
changes are to clarify proper operation and methodology associated 
with this equipment and have no adverse impact on their design 
functions.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not affect existing safety margins. The 
proposed changes verify and maintain the capabilities of the DRPI, 
Control Rod Gripper Coils, instrumentation associated with Quadrant 
Power Tilt Ratio, and Control and Gray Rods 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.
    The proposed changes would not affect any safety-related design 
code, function, design analysis, safety analysis input or result, or 
existing design/safety margin. Because no safety analysis or design 
basis acceptance limit/criterion is challenged or exceeded by the 
requested changes, no margin of safety is significantly reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna 
Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: December 14, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17348B097.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 3.6.4.1, ``Secondary Containment,'' 
Surveillance Requirement (SR) 3.6.4.1.1. The SR would be revised to 
address conditions during which the secondary containment pressure may 
not meet the SR pressure requirements. The proposed changes are based 
on Technical Specifications Task Force (TSTF) Traveler TSTF-551, 
Revision 3, ``Revise Secondary Containment Surveillance Requirements.'' 
Also, the editorial note in SR 3.6.4.1.3 is removed because it is 
redundant to the SR itself and does not alter the requirement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, along with NRC edits in square 
brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change addresses conditions during which the 
secondary containment SR is not met. The secondary containment is 
not an initiator of any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
increased. The consequences of an accident previously evaluated 
while utilizing the proposed 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. 
Additionally, the Note removed from SR 3.6.4.1.3 is editorial 
because it is redundant to the SR itself and does not alter the 
requirement. 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 evaluated?
    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 SR is 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.
    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: Damon D. Obie, Associate General Counsel, 
Talen Energy Supply, LLC, 835 Hamilton St., Suite 150, Allentown, PA 
18101.
    NRC Branch Chief: James G. Danna.

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

    Date of amendment request: September 29, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17272A940.
    Description of amendment request: The amendments would make changes 
to the SQN Emergency Plan to extend staff augmentation times for 
Emergency Response Organization (ERO) functions.
    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:


[[Page 8521]]


    1. Does the proposed change involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    Response: No.
    The proposed removal of maintenance personnel from shift and 
extension in staff augmentation times has no effect on normal plant 
operation or on any accident initiator or precursor and does not 
affect the function of plant structures, systems, or components 
(SCCs). The proposed changes do not alter or prevent the ability of 
the ERO to perform their intended functions to mitigate the 
consequences of an accident or event. The ability of the ERO to 
respond adequately to radiological emergencies has been demonstrated 
as acceptable through a staffing analysis as required by 10 CFR 50 
Appendix E.IV.A.9.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not affect the accident analyses. The 
changes do not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed), a change in 
the method of plant operation, or new operator actions. The proposed 
changes do not introduce failure modes that could result in a new 
accident, and the changes do not alter assumptions made in the 
safety analysis. This proposed change removes maintenance personnel 
from shift and extends the staff augmentation response times in the 
SQN Emergency Plan, which are demonstrated as acceptable through a 
staffing analysis as required by 10 CFR 50 Appendix E.IV.A.9. The 
proposed changes do not alter or prevent the ability of the ERO to 
perform their intended functions to mitigate the consequences of an 
accident or event.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
margin of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed change is 
associated with the SQN Emergency Plan staffing and does not affect 
operation of the plant or its response to transients or accidents. 
The change does not affect the Technical Specifications. The 
proposed changes do not involve a change in the method of plant 
operation, and no accident analyses are affected by the proposed 
changes. Safety analysis acceptance criteria are not affected by 
this proposed change. A staffing analysis and a functional analysis 
were performed for the proposed changes on the timeliness of 
performing major tasks for the functional areas of the SQN Emergency 
Plan. The analysis concluded that removal of maintenance personnel 
from shift and an extension in staff augmentation times would not 
significantly affect the ability to perform the required Emergency 
Plan tasks.
    Therefore, the proposed changes are determined to not adversely 
affect the ability to meet 10 CFR 50.54(q)(2), the requirements of 
10 CFR 50 Appendix E, and the emergency planning standards as 
described in 10 CFR 50.47(b).
    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. Previously Published Notices of Consideration of Issuance of 
Amendments to Facility Operating Licenses and Combined Licenses, 
Proposed No Significant Hazards Consideration Determination, and 
Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Florida Power & Light Company, Docket Nos. 50-250, Turkey Point Nuclear 
Generating Unit No. 3, Miami-Dade County, Florida

    Date of amendment request: December 18, 2017. A publicly-available 
version is in ADAMS under ML17353A492.
    Brief description of amendment request: Revise the Technical 
Specifications to allow a one-time extension of the allowable outage 
time for the Unit 3 Containment Spray System from 72 hours to 14 days.
    Date of publication of individual notice in Federal Register: 
January 30, 2018 (83 FR 4285).
    Expiration date of individual notice: March 1, 2018 (Public 
comments); April 2, 2018 (Hearing requests).

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

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

[[Page 8522]]

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

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

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina

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

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

    Date of amendment request: July 18, 2017, as supplemented by letter 
dated October 12, 2017.
    Brief description of amendments: The amendments revised the 
technical specifications (TSs) based on Technical Specifications Task 
Force (TSTF) Traveler TSTF-529, ``Clarify Use and Application Rules.'' 
The changes revise and clarify the TS usage rules for completion times, 
limiting conditions for operation, and surveillance requirements.
    Date of issuance: February 1, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: 298 and 294, for the Catawba Nuclear Station Units 
1 and 2; 307 and 286, for the McGuire Nuclear Station, Units 1 and 2; 
407, 409, and 408, for the Oconee Nuclear Station, Units 1, 2, and 3; 
162, for the Shearon Harris Nuclear Power Plant, Unit 1; and 256, for 
the H. B. Robinson Steam Electric Plant, Unit No. 2. A publicly-
available version is in ADAMS under Accession No. ML17340A720; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, NPF-
17, DPR-38, DPR-47, DPR-55, NPF-63, and DPR-23: Amendments revised the 
Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: August 29, 2017 (82 FR 
41067). The supplemental letter dated October 12, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated February 1, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment requests: December 19, 2016, as supplemented by 
letters dated May 25, 2017, and December 12, 2017.
    Brief description of amendments: The amendments modified Technical 
Specification 5.5.2, ``Containment Leakage Rate Testing Program,'' by 
replacing the reference to Regulatory Guide 1.163, ``Performance-Based 
Containment Leak-Test Program,'' with a reference to Nuclear Energy 
Institute (NEI) Topical Report NEI 94-01, Revision 3-A, ``Industry 
Guideline for Implementing Performance-Based Option of 10 CFR part 50, 
Appendix J,'' dated July 2012 and the conditions and limitations 
specified in NEI 94-01, Revisions 2-A, ``Industry Guideline for 
Implementing Performance-Based Option of 10 CFR part 50, Appendix J,'' 
dated October 2008, as the implementation documents used by McGuire to 
implement the performance-based leakage testing program in accordance 
with Option B of 10 CFR part 50, Appendix J. The proposed change would 
also delete the listing of one-time exceptions previously granted to 
Integrated Leak Rate Test frequency.
    Date of issuance: January 31, 2018.
    Effective date: As of its date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: 306 (Unit 1) and 285 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML18009A842; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: May 9, 2017 (82 FR 
21557). The supplemental letters dated May 25 and December 12, 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 January 31, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 29, 2017, as supplemented by 
letter dated January 25, 2018.
    Brief description of amendment: The amendments revised the LSCS 
technical specification (TS) 2.1.1, ``Reactor Core SLs [Safety 
Limits].'' Specifically, this change incorporates revised LSCS, Units 1 
and 2, safety limits for minimum critical power ratio for two 
circulation loop minimum critical power ratio (MCPR) and single 
circulation loop MCPR values for Unit 1 and Unit 2 based on the results 
of the cycle-specific analyses performed by Global Nuclear Fuel (GNF) 
for LSCS Unit 1, Cycle 17, and LSCS Unit 2, Cycle 17.
    Date of issuance: February 6, 2018.
    Effective date: As of the date of issuance and shall be implemented 
as follows:
    Unit 1: Prior to startup from the February 2018 refueling outage 
for Unit 1 (i.e., L1R17) for operation starting in Cycle 18.
    Unit 2: Prior to startup from the February 2018 refueling outage 
for Unit 1 (i.e., L1R17). This will be a mid-Cycle 17 implementation 
for Unit 2.
    Amendment No.: Unit 1-227; Unit 2-213. A publicly-available version 
is in ADAMS under Accession No. ML18008A123; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. NPF-11 and NPF-18: The 
amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: December 5, 2017 (82 FR 
57482). The supplemental letter dated January 25, 2018, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 6, 2018.

[[Page 8523]]

    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of amendment request: May 24, 2017, as supplemented by letter 
dated August 17, 2017.
    Brief description of amendments: The amendments revised 
Surveillance Requirement 3.3.1.3 to change the thermal power at which 
the surveillance may be performed.
    Date of issuance: February 7, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 194 (Unit 1) and 177 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML18012A068; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License No. NPF-68 and NPF-81: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: July 18, 2017 (82 FR 
32883). The supplemental letter dated August 17, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 7, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment requests: March 13, 2017, as supplemented by 
letter dated August 7, 2017.
    Brief description of amendments: The amendments deleted the Note 
associated with Technical Specification (TS) Surveillance Requirement 
(SR) 3.8.1.17 to allow the performance of the SR in Modes 1 through 4.
    Date of issuance: February 2, 2018.
    Effective date: As of its date of issuance and shall be implemented 
no later than 60 days from the date of issuance.
    Amendment Nos.: 340 (Unit 1) and 333 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17296A133; documents related 
to these amendments are listed in the Safety Evaluation (SE) enclosed 
with the amendments.
    Facility Operating License Nos. DPR-77 and DPR-79: The amendments 
revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: July 5, 2017 (82 FR 
31102). The supplemental letter dated August 7, 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 an SE dated February 2, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 20, 2017, as supplemented by 
letter dated September 7, 2017.
    Brief description of amendments: The amendments revised the 
Technical Specification (TS) 3.5, ``Residual Heat Removal (RHR) 
System,'' requirements, as well as the TS 3.13, ``Component Cooling 
System,'' RHR support requirements for consistency with the design 
basis of the RHR system. In addition, an RHR surveillance requirement 
is added in TS Table 4.1-2A, ``Minimum Frequency for Equipment Tests,'' 
to test the RHR system in accordance with the inservice testing 
program, since a TS surveillance does not currently exist for this 
system.
    Date of issuance: February 9, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 291 and 291. A publicly-available version is in 
ADAMS under Accession No. ML17326A225; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License No. DPR-32 and DPR-37: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: March 14, 2017 (82 FR 
13672). The supplemental letter dated September 7, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 9, 2018.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, on February 20, 2018.

    For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2018-03727 Filed 2-26-18; 8:45 am]
 BILLING CODE 7590-01-P