[Federal Register Volume 82, Number 232 (Tuesday, December 5, 2017)]
[Notices]
[Pages 57469-57478]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-25901]



[[Page 57469]]

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NUCLEAR REGULATORY COMMISSION

[NRC-2017-0225]


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 November 7, 2017, to November 17, 2017. The 
last biweekly notice was published on November 21, 2017.

DATES: Comments must be filed by January 4, 2018. A request for a 
hearing must be filed by February 5, 2018.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0225. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; 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: OWFN-2-A13, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Paula Blechman, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-2422, email: [email protected].

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2017-0225, facility name, unit 
number(s), plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0225.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in this 
document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2017-0225, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

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

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

[[Page 57470]]

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

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and petition for leave to intervene 
(petition) with respect to the action. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested persons should consult a 
current copy of 10 CFR 2.309. The NRC's regulations are accessible 
electronically from the NRC Library on the NRC's Web site at 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 Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit 
paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at

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[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 Web site 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 Web site 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 Web site at http://www.nrc.gov/site-help/e-submittals.html, by 
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m. 
and 6 p.m., Eastern Time, Monday through Friday, excluding government 
holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
stating why there is good cause for not filing electronically and 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff. Participants filing adjudicatory documents in this 
manner are responsible for serving the document on all other 
participants. Filing is considered complete by first-class mail as of 
the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service. A presiding officer, having granted an 
exemption request from using E-Filing, may require a participant or 
party to use E-Filing if the presiding officer subsequently determines 
that the reason for granting the exemption from use of E-Filing no 
longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the 
Commission or the presiding officer. If you do not have an NRC-issued 
digital ID certificate as described above, click cancel when the link 
requests certificates and you will be automatically directed to the 
NRC's electronic hearing dockets where you will be able to access any 
publicly-available documents in a particular hearing docket. 
Participants are requested not to include personal privacy information, 
such as social security numbers, home addresses, or personal phone 
numbers in their filings, unless an NRC regulation or other law 
requires submission of such information. For example, in some 
instances, individuals provide home addresses in order to demonstrate 
proximity to a facility or site. With respect to copyrighted works, 
except for limited excerpts that serve the purpose of the adjudicatory 
filings and would constitute a Fair Use application, participants are 
requested not to include copyrighted materials in their submission.
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2 (HBRSEP), Darlington County, South Carolina
    Date of amendment request: September 27, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17270A041.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to reflect the addition of a 
second qualified offsite power circuit. In addition, the proposed 
amendment requests approval to change the Updated Final Safety Analysis 
Report (UFSAR) to allow for the use of automatic load tap changers 
(LTCs) on the new (230 kilovolt (kV)) and the replacement (115kV) 
startup transformers.
    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 revises TS 3.8.1 to reflect the addition of 
a second qualified offsite circuit at HBRSEP. The proposed change 
modifies the TS 3.8.1 LCO [Limiting Condition for Operation], 
Conditions, Required Actions and Completion Times to be more 
consistent with NUREG-1431 [``Standard Technical Specifications--
Westinghouse Plants'']. The AC [alternating current] power systems 
are not an initiator of any accident previously evaluated. As a 
result, the probability of an accident previously evaluated is not 
increased. The consequences of an accident with the proposed LCO 
requiring two qualified offsite circuits between the offsite 
transmission network and the onsite emergency AC Electrical Power 
Distribution System to be operable are no different than the 
consequences of an accident in Modes 1, 2, 3, and 4 with the 
existing LCO that requires the single qualified offsite circuit to 
be operable. The additional 230kV startup transformer will improve 
the reliability and availability of offsite power to the emergency

[[Page 57472]]

buses by increasing the amount of available offsite power sources 
from one to two. The two qualified offsite circuits are designed to 
mitigate the consequences of previously evaluated accidents. The 
proposed change to TS 3.8.1 would not change any of the previously 
evaluated accidents in the UFSAR.
    The proposed change will also allow operation of the LTCs on the 
115kV and 230kV startup transformers in automatic mode. The only 
accident previously evaluated where the probability of an accident 
is potentially affected by the proposed change is a loss of offsite 
power (LOOP). Failure of a LTC while in the automatic mode of 
operation that results in decreased voltage to the safety related 
buses could cause a LOOP if voltage decreased below the degraded 
grid voltage relay (DGVR) setpoint. The three postulated failure 
scenarios are: (1) Failure of a primary microcontroller that results 
in rapidly decreasing voltage supplied to the safety related buses; 
(2) failure of a primary microcontroller to respond to decreasing 
grid voltage; and (3) the backup microcontroller overrides the 
primary microcontroller when not required. For the first scenario, a 
backup microcontroller is provided for each LTC, which makes this 
failure unlikely. For the second scenario, operators would have 
ample time to address the condition utilizing identified procedures 
since grid voltage changes typically occur relatively slowly. In 
addition, the frequency of occurrence of all of these failure modes 
is small, based on the operating history of similar equipment at 
other plants. Furthermore, in all of the above potential failure 
modes, operators can take manual control of the LTC to mitigate the 
effects of the failure. Thus, the probability of a LOOP will not be 
significantly increased by operation of the LTCs in the automatic 
mode. The proposed change to allow operation of the LTCs in 
automatic mode has no effect on the consequences of a LOOP, since 
the emergency diesel generators (EDGs) provide power to safety 
related equipment following a LOOP. The design and function of the 
EDGs are not affected by the proposed change. The LTCs are each 
equipped with a backup microcontroller, which inhibits gross 
improper action of the LTC in the event of primary microcontroller 
failure. Additionally, the operator has procedurally identified 
actions available to prevent a sustained high voltage condition from 
occurring. Damage due to overvoltage is time-dependent, requiring a 
sustained high voltage condition. Therefore, damage to safety 
related equipment is unlikely, and the consequences of previously 
evaluated accidents 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 revises TS 3.8.1 to reflect the addition of 
a second qualified offsite circuit at HBRSEP. The proposed change 
modifies the TS 3.8.1 LCO, Conditions, Required Actions and 
Completion Times to be more consistent with NUREG-1431. The proposed 
change also will allow operation of the LTCs on the 115kV and 230kV 
startup transformers in automatic mode. All aspects of the proposed 
change involve electrical transformers that provide offsite power to 
safety-related equipment for accident mitigation. The proposed 
change does not alter the design, physical configuration or mode of 
operation of any other plant structure, system or component. No 
physical changes are being made to any other portion of the plant, 
so no new accident causal mechanisms are being introduced. The 
proposed change also does not result in any new mechanisms that 
could initiate damage to the reactor or its principal safety 
barriers (i.e., fuel cladding, reactor coolant system or primary 
containment).
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises TS 3.8.1 to reflect the addition of 
a second qualified offsite circuit at HBRSEP. The proposed change 
modifies the TS 3.8.1 LCO, Conditions, Required Actions and 
Completion Times to be more consistent with NUREG-1431. The new 
230kV startup transformer will improve the reliability and 
availability of offsite power to the emergency buses by increasing 
the amount of available offsite power sources from one to two. 
Another improvement to the HBRSEP electrical system configuration as 
a result of the proposed change is that each emergency bus will be 
normally aligned to independent startup sources and will not require 
a fast bus transfer on a unit trip. This reduces the risk of loss of 
power to the emergency buses caused by power transfer and/or 
equipment failures. The margin of safety is increased with the 
proposed change to revise TS 3.8.1 to reflect the additional 
qualified offsite circuit.
    The proposed change will also allow operation of the LTCs on the 
115kV and 230kV startup transformers in automatic mode. The inputs 
or assumptions of any of the analyses that demonstrate the integrity 
of the fuel cladding, reactor coolant system or containment during 
accident conditions are unaffected by this proposed change. The 
allowable values for the degraded voltage protection function are 
unchanged and will continue to ensure that the degraded voltage 
protection function actuates when required, but does not actuate 
prematurely to unnecessarily transfer safety related loads from 
offsite power to the EDGs. Automatic operation of the LTCs increases 
the margin of safety by reducing the potential for transferring 
loads to the EDGs during an undervoltage or overvoltage event on the 
offsite power sources.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, 
Charlotte, NC 28202.
    NRC Branch Chief: Undine Shoop.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant (PNP), Van Buren County, Michigan
    Date of amendment request: November 1, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17306A086.
    Description of amendment request: The proposed amendment would 
revise the PNP renewed facility operating license (RFOL) to change the 
full compliance implementation date for the fire protection program 
transition license condition. Specifically, the licensee is requesting 
additional time for completion of the required modifications necessary 
to achieve full compliance with 10 CFR 50.48(c), ``National Fire 
Protection Association Standard NFPA 805.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the PNP RFOL to change the full 
compliance implementation date for the fire protection program 
transition license condition to allow additional time for completion 
of the required modifications necessary to achieve full compliance 
with 10 CFR 50.48(c) is administrative in nature. This change does 
not alter accident analysis assumptions, add any initiators, or 
affect the function of plant systems or the manner in which systems 
are operated, maintained, modified, tested, or inspected. The 
proposed change does not require any plant modifications which 
affect the performance capability of the structures, systems, and 
components relied upon to mitigate the consequences of postulated 
accidents, and have no impact on 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.
    2. Does the proposed change create the possibility of a new or 
different kind of

[[Page 57473]]

accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the PNP RFOL to change the full 
compliance implementation date for the fire protection program 
transition license condition to allow additional time for completion 
of the required modifications necessary to achieve full compliance 
with 10 CFR 50.48(c) is administrative in nature. This proposed 
change does not alter accident analysis assumptions, add any 
initiators, or affect the function of plant systems or the manner in 
which systems are operated, maintained, modified, tested, or 
inspected. The proposed change does not require any plant 
modifications which affect the performance capability of the 
structures, systems, and components relied upon to mitigate the 
consequences of postulated accidents and does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes to the PNP RFOL to change the full 
compliance implementation date for the fire protection program 
transition license condition to allow additional time for completion 
of the required modifications necessary to achieve full compliance 
with 10 CFR 50.48(c) is administrative in nature. Plant safety 
margins are established through limiting conditions for operation, 
limiting safety system settings, and safety limits specified in the 
technical specifications. Because there is no change to established 
safety margins as a result of this change, the proposed change does 
not involve a significant reduction in a 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: William Glew, Associate General Counsel 
Nuclear, Entergy Services, Inc., 440 Hamilton Ave., White Plains, NY 
10601.
    NRC Branch Chief: David J. Wrona.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
1 (ANO-1), Pope County, Arkansas
    Date of amendment request: October 2, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17275A910.
    Description of amendment request: The amendment would revise the 
ANO-1 Technical Specification (TS) 3.7.5, ``Emergency Feedwater (EFW) 
System,'' Bases to stipulate the conditions in which the TS 3.7.5, 
Condition A, 7-day Completion Time should apply to the ANO-1 turbine-
driven EFW pump steam supply valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The EFW system 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 to clarify the conditions in which the current 7-day 
Completion Time for an inoperable steam supply path to turbine-
driven EFW pump does not change the response of the plant to any 
accidents, since single failure criterion is not applicable when 
complying with associated TS Actions.
    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 proposed change does not 
adversely affect the ability of structures, systems, and components 
(SSCs) to perform their 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.
    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 EFW system provides plant protection. Absent a single 
failure (which is not assumed while in compliance with TS Actions), 
the EFW system will continue to supply water to the Steam Generators 
(SGs) to remove decay heat and other residual heat by delivering at 
least the minimum required flow rate to the SGs, as required. There 
are no design changes associated with the proposed change. The 
change to the associated TS Bases does not change any existing 
accident scenarios, nor create any new or different accident 
scenarios.
    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 plant operation. In addition, 
the change clarifies the application of the current 7-day Completion 
Time for an inoperable steam supply path to the turbine-driven EFW 
pump and does not impose any new or different requirements or 
eliminate any existing requirements. The change does not alter 
assumptions made in the safety analysis. The proposed change is 
consistent with the safety analysis assumptions, which does not 
assume an EFW system single failure when complying with TS Actions, 
and current plant operating practice.
    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 change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by these changes. The proposed change will not 
result in plant operation in a configuration outside the design 
basis. The associated TS will continue to limit the time in which 
one steam supply path to the turbine-driven EFW pump may be 
inoperable.
    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.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: July 28, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17209A755.
    Description of amendment request: The requested amendment proposes 
changes to combined license (COL) Appendix A, plant-specific Technical 
Specifications (TS) to make them consistent with the remainder of the 
design licensing basis and the TS. Specifically, the requested 
amendment proposes changes to COL Appendix A, the Technical 
Specification updates for reactivity controls and other

[[Page 57474]]

miscellaneous changes, and Updated Final Safety Analysis Report (UFSAR) 
information in various locations.
    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 change does not involve a physical alteration of 
the plant or a change in the methods governing normal plant 
operations. The change applies to a Diverse Actuation System (DAS) 
Manual Controls Mode 6 note for operability of the Automatic 
Depressurization System (ADS) Stage 4 valves that involves revising 
the note from reactor internals in place to upper internals in 
place. In accordance with Limiting Condition for Operation (LCO) 
3.4.13 ADS--Shutdown, Reactor Coolant System (RCS) Open 
Applicability and TS 3.3.9, Engineered Safeguards Actuation System 
Instrumentation, Function 7, the ADS Stage 4 valves are not required 
to be operable in MODE 6 with the upper internals removed. However, 
the reactor internals would still be present. The change involves 
clarification of the note (with no change in required system or 
device function), such that the appropriate configuration in Mode 6 
would be in place and would not conflict with TS 3.4.13 or TS 3.3.9. 
The revised note previously evaluated. As a result, the probability 
of an accident previously evaluated is not affected.
    The consequences of an accident as a result of the revised note 
and associated requirements and actions are no different than the 
consequences of the same accident during the existing ones. As a 
result, the consequences are not affected by this change.
    The proposed change does not alter or prevent the ability of 
structures, systems, and components from performing their intended 
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 an 
accident previously evaluated. Therefore, this 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 revising the existing LCO 3.1.4 
operability to be applicable to Rod Cluster Control Assemblies 
(RCCAs)with accompanying changes in actions and surveillance 
requirements (with no change in required system or device function), 
such that more appropriate, albeit less restrictive, actions would 
be applied. The proposed change does not involve a physical 
alteration of the plant as described in the UFSAR. No new equipment 
is being introduced, and equipment is not being operated in a new or 
different manner. There are no set points, at which protective or 
mitigative actions are initiated, affected by this change. This 
change will not alter the manner in which equipment operation is 
initiated, nor will the function demands on credited equipment be 
changed. No change is being made to the procedures relied upon to 
respond to an off-normal event as described in the UFSAR as a result 
of this change. As such, no new failure modes are being introduced. 
The change does not alter assumptions made in the safety analysis 
and licensing basis. Therefore, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change will not reduce a margin of safety because 
it has no effect on any assumption of the safety analyses. While the 
LCO 3.1.4 for Rod Group Alignment Limits is made less restrictive by 
eliminating the worth of the [Gray Rod Cluster Assemblies (GRCAs)] 
in MODES 1 and 2 with keff >=1, no credit is taken in the 
current design basis for including their trip reactivity worth. As 
such, there is no significant reduction in a margin of safety. 
Therefore, the requested amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant 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. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia
    Date of amendment request: September 12, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17257A177.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 5.5.17, ``Containment Leakage Rate Testing 
Program,'' for the Vogtle Electric Generating Plant, Units 1 and 2, to 
(1) increase the existing Type A integrated leakage rate test interval 
from 10 to 15 years, (2) extend the Type C containment isolation valve 
leaking testing to a 75-month frequency, (3) adopt the use of American 
National Standards Institute/American Nuclear Society 56.8-2002, 
``Containment System Leakage Testing Requirements,'' and (4) adopt a 
more conservative grace interval of 9 months for Type A, B, and C tests 
in accordance with Nuclear Energy Institute (NEI) 94-01, Revision 3-A, 
``Industry Guideline for Implementing Performance-Based Option of 10 
CFR part 50, Appendix J.''
    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 activity involves the revision of Vogtle Electric 
Generating Plant (VEGP), Units 1 and 2, Technical Specification (TS) 
Section 5.5.17, ``Primary Containment Leakage Rate Testing 
Program,'' to allow the extension of the Type A integrated leakage 
rate test (ILRT) containment test interval to 15 years, and the 
extension of the Type C local leakage rate test (LLRT) interval to 
75 months. The current Type A test interval of 120 months (10 years) 
would be extended on a permanent basis to no longer than 15 years 
from the last Type A test. The current Type C test interval of 60 
months for selected components would be extended on a performance 
basis to no longer than 75 months. Extensions of up to nine months 
(total maximum interval of 84 months for Type C tests) are 
permissible only for non-routine emergent conditions.
    The proposed extensions do not involve either a physical change 
to the plant or a change in the manner in which the plant is 
operated or controlled. The containment is designed to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. As such, 
the containment and the testing requirements invoked to periodically 
demonstrate the integrity of the containment exist to ensure the 
plant's ability to mitigate the consequences of an accident, and do 
not involve the prevention or identification of any precursors of an 
accident.
    The change in Type A test frequency to once-per-fifteen years, 
measured as an increase to the total integrated plant risk for those 
accident sequences influenced by Type A testing, based on the 
internal events (IE) probabilistic risk analysis (PRA) is 1.79E-03 
person-rem/year for Unit 1 and Unit 2. Electric Power Research 
Institute (EPRI) Report No. 1009325, Revision 2-A states that a very 
small population is defined as an increase of <=1.0 person-rem per 
year or <=1% of the total population dose, whichever is less 
restrictive for the risk impact assessment of the extended ILRT 
intervals. This is consistent with the Nuclear Regulatory Commission 
(NRC) Final Safety Evaluation for Nuclear Energy Institute (NEI) 94-
01 and EPRI Report No. 1009325. Moreover, the risk

[[Page 57475]]

impact when compared to other severe accident risks is negligible. 
Therefore, this proposed extension does not involve a significant 
increase in the probability of an accident previously evaluated.
    In addition, as documented in NUREG-1493, ``Performance-Based 
Containment Leak-Test Program,'' dated September 1995, Types B and C 
tests have identified a very large percentage of containment leakage 
paths, and the percentage of containment leakage paths that are 
detected only by Type A testing is very small. The VEGP Type A test 
history supports this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as: (1) Activity based, 
and (2) time based. Activity-based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. The LLRT requirements and administrative controls such 
as configuration management and procedural requirements for system 
restoration ensure that containment integrity is not degraded by 
plant modifications or maintenance activities. The design and 
construction requirements of the containment combined with the 
containment inspections performed in accordance with American 
Society of Mechanical Engineers (ASME) Section XI, and TS 
requirements serve to provide a high degree of assurance that the 
containment would not degrade in a manner that is detectable only by 
a Type A test. Based on the above, the proposed test interval 
extensions do not significantly increase the consequences of an 
accident previously evaluated.
    The proposed amendment also deletes exceptions previously 
granted under TS Amendment Nos. 130 (VEGP-1) and 108 (VEGP-2), to 
allow one-time extensions of the ILRT test frequency for VEGP. These 
exceptions were for activities that would have already taken place 
by the time this amendment is approved; therefore, their deletion is 
solely an administrative action that has no effect on any component 
and no impact on how the unit is operated.
    Therefore, the proposed change does not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment to the TS 5.5.17, Containment Leakage 
Rate Testing Program, involves the extension of the VEGP Type A 
containment test interval to 15 years and the extension of the Type 
C test interval to 75 months. The containment and the testing 
requirements to periodically demonstrate the integrity of the 
containment exist to ensure the plant's ability to mitigate the 
consequences of an accident do not involve any accident precursors 
or initiators. The proposed change does not involve a physical 
change to the plant (i.e., no new or different type of equipment 
will be installed) or a change to the manner in which the plant is 
operated or controlled.
    The proposed amendment also deletes exceptions previously 
granted under TS Amendment Nos. 130 (VEGP-1) and 108 (VEGP-2), to 
allow one-time extensions of the ILRT test frequency for VEGP. These 
exceptions were for activities that would have already taken place 
by the time this amendment is approved; therefore, their deletion is 
solely an administrative action that does not result in any change 
in how the unit is operated.
    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 amendment to TS 5.5.17 involves the extension of 
the VEGP Type A containment test interval to 15 years and the 
extension of the Type C test interval to 75 months for selected 
components. This amendment does not alter the manner in which safety 
limits, limiting safety system set points, or limiting conditions 
for operation are determined. The specific requirements and 
conditions of the TS Containment Leak Rate Testing Program exist to 
ensure that the degree of containment structural integrity and 
leaktightness that is considered in the plant safety analysis is 
maintained. The overall containment leak rate limit specified by TS 
is maintained.
    The proposed change involves only the extension of the interval 
between Type A containment leak rate tests and Type C tests for 
VEGP. The proposed surveillance interval extension is bounded by the 
15-year ILRT interval and the 75-month Type C test interval 
currently authorized within NEI 94-01, Revision 3-A. Industry 
experience supports the conclusions that Types B and C testing 
detects a large percentage of containment leakage paths and that the 
percentage of containment leakage paths that are detected only by 
Type A testing is small. The containment inspections performed in 
accordance with ASME Section XI and TS serve to provide a high 
degree of assurance that the containment would not degrade in a 
manner that is detectable only by Type A testing. The combination of 
these factors ensures that the margin of safety in the plant safety 
analysis is maintained. The design, operation, testing methods and 
acceptance criteria for Types A, B, and C containment leakage tests 
specified in applicable codes and standards would continue to be 
met, with the acceptance of this proposed change, since these are 
not affected by changes to the Type A and Type C test intervals.
    The proposed amendment also deletes exceptions previously 
granted under TS Amendment Nos. 130 (VEGP-1) and 108 (VEGP-2), to 
allow one-time extensions of the ILRT test frequency for VEGP. This 
exception was for an activity that would have already taken place by 
the time this amendment is approved; therefore, the deletion is 
solely an administrative action and does not change how the unit is 
operated and maintained. Thus, there is no reduction in any margin 
of safety as a result of this administrative change.
    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: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, 40 Inverness Center 
Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Michael T. Markley.
STP Nuclear Operating Company (STPNOC), Docket Nos. 50-498 and 50-499, 
South Texas Project (STP), Units 1 and 2, Matagorda County, Texas
    Date of amendment request: September 18, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17261B272.
    Description of amendment request: The amendment would relocate the 
defined core plane regions where the radial peaking factor limits are 
not applicable, from Technical Specification (TS) 4.2.2.2.f to the Core 
Operating Limits Reports (COLR) for STP Units 1 and 2. The amendment 
would also revise the COLR Administrative Controls TS to add exclusion 
zones to the list of limits found in the COLRs, and to revise the 
description of the methodology used to determine the values. In 
addition, the proposed amendment requests administrative changes to the 
TSs.
    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 relocation of the Fxy exclusion zones to the 
COLRs has no impact on the accidents analyzed in the STPNOC UFSAR 
[Updated Final Safety Analysis Report] and is not an accident 
initiator. Since the change does not impact any conditions that 
would initiate an accident, the probability or consequences of 
previously analyzed events is not increased. The proposed amendment 
does not change the actions to be taken if a core operating limit is 
exceeded and there are no physical changes associated with this 
proposed amendment.
    For each core reload, each accident analysis addressed in the 
STP UFSAR will continue to be examined with respect to changes in 
the cycle-dependent parameters, which are obtained from the use of 
NRC-approved reload design methodologies, to

[[Page 57476]]

ensure that the transient evaluation of new reloads are bounded by 
previously accepted analyses. This examination, which will be 
conducted per the requirements of 10 CFR 50.59, will ensure that 
future core reloads will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Therefore, there is no impact to the probability or consequences 
of an accident previously evaluated due to the proposed change.
    [The licensee stated that the administrative changes proposed to 
the TSs do not impact the operation of the facility in a manner that 
involves significant hazards considerations.]
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The relocation of the Fxy exclusion zone details from 
the Technical Specifications to the COLRs will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. No safety-related equipment, safety function, 
or plant operation will be altered as a result of this proposed 
change. No new operator actions are created as a result of the 
proposed change. The cycle-specific variables are determined using 
the NRC approved methods and the COLRs are submitted to the NRC to 
allow the staff to continue to trend the values of these limits. The 
Technical Specifications will continue to require operation within 
the core operating limits and appropriate actions will be required 
if these limits are exceeded.
    The relocation of the Fxy exclusion zones to the 
COLRs has no impact on the accidents analyzed in the STPNOC Updated 
Final Safety Analysis Report (UFSAR) and is not an accident 
initiator. Since this change does not impact any conditions that 
would initiate an accident, there is no possibility of a new or 
different kind of accident resulting from this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    [The licensee stated that the administrative changes proposed to 
the TSs do not impact the operation of the facility in a manner that 
involves significant hazards considerations.]
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The relocation of the Fxy exclusion zone details from 
the Technical Specifications to the COLRs will not affect the margin 
of safety. The margin of safety presently provided by the Technical 
Specifications remains unchanged. They will be incorporated into the 
COLR which is submitted to the NRC, therefore appropriate measures 
exist to control the values of these limits. The development of the 
limits for future reloads will continue to conform to those methods 
described in NRC-approved documentation. STPNOC will continue to 
confirm all safety analysis limits remain bounding on a cycle-
specific basis using an NRC-approved methodology. Each core reload 
will involve a Reload Safety Evaluation to assure that operation of 
the unit within the cycle specific limits will not involve a 
significant reduction in the margin of safety.
    The proposed amendment does not affect the design of the 
facility or system operating parameters, does not physically alter 
safety-related systems and does not affect the method in which 
safety-related systems perform their functions.
    Therefore, the proposed change does not impact margin of safety.
    [The licensee stated that the administrative changes proposed to 
the TSs do not impact the operation of the facility in a manner that 
involves significant hazards considerations.]

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

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

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

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: October 26, 2016, as supplemented by 
letters dated February 16, July 17, August 8, September 27, October 3, 
and November 8, 2017.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 5.5.13, ``Primary Containment Leakage Rate Testing 
Program,'' to allow for the permanent extension of the Type A 
Integrated Leak Rate Testing and Type C Leak Rate Testing frequencies, 
to change the documents used by LSCS to implement the performance-based 
leakage testing program, and to delete the information regarding the 
performance of the next LSCS Type A tests to be performed.
    Additionally, the amendments deleted Conditions 2.D.(e) and 
2.D.(c), respectively, of the LSCS Unit 1 and Unit 2 Renewed Facility 
Operating Licenses regarding conducting the third Type A test of each 
10-year service period when the plant is shut down for the 10-year 
inservice inspection.
    Date of issuance: November 16, 2017.
    Effective date: As of the date of its issuance and shall be 
implemented within 60 days from the date of issuance.
    Amendment Nos.: 226 (Unit 1) and 212 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17283A085; 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 TSs and Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: February 14, 2017 (82 
FR 10597). The supplemental letters dated February 16, July 17, August 
8, September 27, October 3, and November 8, 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.

[[Page 57477]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 16, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 17, 2017, as supplemented by 
letters dated March 20, 2017; July 13, 2017; August 8, 2017; August 30, 
2017; and September 15, 2017.
    Brief description of amendments: The amendments revised the Renewed 
Facility Operating Licenses and Technical Specifications to implement a 
measurement uncertainty recapture power uprate. Specifically, the 
amendments authorized an increase in the maximum licensed thermal power 
level from 3,951 megawatts thermal to 4,016 megawatts thermal, which is 
an increase of approximately 1.66 percent.
    Date of issuance: November 15, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendments Nos.: 316 (Unit 2) and 319 (Unit 3). A publicly-
available version is in ADAMS under Accession No. ML17286A013; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2017 (82 FR 
20497). The supplemental letters dated March 20, 2017; July 13, 2017; 
August 8, 2017; August 30, 2017; and September 15, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 15, 2017.
    No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit 1 (FCS), Washington County, Nebraska

    Date of amendment request: October 25, 2016, as supplemented by 
letter dated September 25, 2017.
    Brief description of amendment: The amendment revised the FCS 
Updated Safety Analysis Report to change the structural design 
methodology for the Auxiliary Building at FCS. Specifically, the 
amendment made the following changes: (1) Use of the ultimate strength 
design method from the industry standard American Concrete Institute 
(ACI) 318-63, ``Publication SP-10, Commentary on Building Code 
Requirements for Reinforced Concrete,'' for normal operating/service 
conditions for future designs and evaluations; (2) use higher concrete 
compressive strength values for Class B concrete, based on original 
strength test data; (3) use higher reinforcing steel yield strength 
values, based on original strength test data; and (4) make minor 
clarifications, including adding a definition of control fluids to the 
dead load section of the Updated Safety Analysis Report.
    Date of issuance: November 17, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 293. A publicly-available version is in ADAMS under 
Accession No. ML17278A607; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Emergency Plan and Emergency Action Level Scheme.
    Date of initial notice in Federal Register: January 17, 2017 (82 FR 
4930).
    The supplemental letter dated September 25, 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 November 17, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 17, 2016, as supplemented by 
letters dated August 7, 2017, and October 18, 2017.
    Brief description of amendments: The amendments revised Technical 
Specification requirements regarding accident monitoring 
instrumentation. Specifically, the amendments modified the list of 
instruments required to be operable based on implementation of 
Regulatory Guide 1.97, Revision 2, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' In addition, allowed outage times 
and required actions for inoperable accident monitoring instrumentation 
channels have been revised to be consistent with NUREG-1431, Revision 
4.0, ``Standard Technical Specifications--Westinghouse Plants.''
    Date of issuance: November 14, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 320 (Unit 1) and 301(Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17227A016; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-70 and DPR-75: The 
amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: January 17, 2017 (82 FR 
4931). The supplemental letters dated August 7, 2017, and October 18, 
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 November 14, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 7, 2017, as supplemented by 
letters dated March 27, 2017, and July 13, 2017.

[[Page 57478]]

    Brief description of amendment: The amendment modified Hope Creek 
Generating Station Technical Specification 6.8.4.f, ``Primary 
Containment Leakage Rate Testing Program,'' to extend the Type A 
reactor containment pressure test interval from one test in 10 years to 
one test in 15 years, and extend the Type C test interval up to 75 
months, with a permissible extension period of 9 months (total of 84 
months) for non-routine emergency conditions.
    Date of issuance: November 8, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 207. A publicly-available version is in ADAMS under 
Accession No. ML17291A209; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-57: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 20, 2016 (81 
FR 92869). The supplemental letters dated March 27, 2017, and July 13, 
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 November 8, 2017.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of amendment request: January 17, 2017, as supplemented by 
letter dated June 29, 2017.
    Brief description of amendments: The amendments change technical 
specifications (TSs) consistent with Technical Specifications Task 
Force (TSTF) Standard Technical Specifications Change Traveler TSTF-
545, Revision 3, ``TS Inservice Testing Program Removal & Clarify SR 
[Surveillance Requirement] Usage Rule Application to Section 5.5 
Testing,'' and TSTF-299, Revision 0, ``Administrative Controls Program 
5.5.2.b Test Interval and Exception.''
    Date of issuance: November 8, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 301 (Unit 1), 325 (Unit 2), and 285 (Unit 3). A 
publicly-available version is in ADAMS under Accession No. ML17277A207; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: April 25, 2017 (82 FR 
19106). The supplemental letter dated June 29, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 8, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 28, 2017.
    Brief description of amendment: The amendment revised the 
completion date for License Condition 2.C.(5) for Watts Bar Nuclear 
Plant, Unit 2, regarding the completion of action to resolve the issues 
identified in Bulletin 2012-01, ``Design Vulnerability in Electric 
Power System'' (ADAMS Accession No. ML12074A115), from December 31, 
2017, to December 31, 2018, to align with the remainder of the 
Tennessee Valley Authority fleet and with the nuclear industry.
    Date of issuance: November 6, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 15 days of issuance.
    Amendment No.: 17. A publicly-available version is in ADAMS under 
Accession No. ML17258A328; documents related to this amendment is 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-96: Amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: July 5, 2017 (82 FR 
31103).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 6, 2017.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 27th day of November 2017.

    For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2017-25901 Filed 12-4-17; 8:45 am]
 BILLING CODE 7590-01-P