Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 6218-6242 [2018-02636]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 83, Number 30 (Tuesday, February 13, 2018)]
[Notices]
[Pages 6218-6242]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-02636]


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NUCLEAR REGULATORY COMMISSION

[NRC-2018-0021]


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 13, 2018, to January 29, 2018. The 
last biweekly notice was published on January 30, 2018.

DATES: Comments must be filed by March 15, 2018. A request for a 
hearing must be filed by April 16, 2018.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking website: Go to https://www.regulations.gov and search for Docket ID NRC-2018-0021. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
9127; email: [email protected]. For

[[Page 6219]]

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-3-D1, 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: Shirley Rohrer, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-5411, email: [email protected].

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2018-0021, facility name, unit 
number(s), plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
     Federal Rulemaking website: Go to https://www.regulations.gov and search for Docket ID NRC-2018-0021.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in this 
document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2018-0021, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

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

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

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

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

[[Page 6220]]

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

[[Page 6221]]

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

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

    Date of amendment request: November 15, 2017. A publicly available 
version is in ADAMS under Accession No. ML17331A484.
    Description of amendment request: The amendments would revise fire 
protection license condition 2.B.(6) to allow, as a performance-based 
method, certain currently-installed thermal insulation materials to be 
retained and allow future use of these insulation materials in limited 
applications subject to appropriate engineering reviews and controls, 
as a deviation from the National Fire Protection Association Standard 
805, Chapter 3, Section 3.3, Prevention.
    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.
    A fire hazards evaluation was performed for the areas of the 
plant where the identified insulation materials are installed. The 
fire hazards evaluation demonstrates that these materials do not 
contribute appreciably to the spread of fire, nor represent a 
secondary combustible beyond those currently analyzed in the Fire 
Probabilistic Risk Analysis (FPRA) due to the limited applications 
where these materials are installed. Therefore, it is concluded that 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The identified installations of the insulation materials were 
evaluated against the fire scenarios supporting the FPRA. In all 
instances, the supporting analyses and existing fire scenarios were 
found to be bounding. Expanded zones of fire influence would not 
fail additional FPRA targets, or there were no FPRA credited targets 
in the area. Therefore, it is concluded that this change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The limited installations of the insulation materials do not 
compromise post-fire safe shutdown capability as previously 
designed, reviewed, and considered. Essential fire protection safety 
functions are maintained and are capable of being performed. Because 
the insulation materials do not compromise post-fire safe shutdown 
capability as previously designed, reviewed, and considered, it is 
concluded that this change does not involve a significant reduction 
in a margin of safety.

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

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

    Date of amendment request: October 19, 2017, as supplemented by 
letter dated January 11, 2018. Publicly-available versions are in ADAMS 
under Accession Nos. ML17292B648 and ML18011A911, respectively.
    Description of amendment request: The amendment would revise the 
HNP 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).

[[Page 6222]]

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 Nuclear Regulatory 
Commission (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 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 HNP 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 of 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.]
    Therefore, the proposed amendment, for both the conditions 
described herein and any future application of the methodology, does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment, including any future use of the 
methodology, 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 UFSAR accident analysis will 
continue to meet requirements for the scope and type of accidents 
that require analysis.
    Therefore, the proposed amendment, for both the conditions 
described herein and any future application of the methodology, does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment does not exceed or alter any controlling 
numerical value for a parameter established in the UFSAR or 
elsewhere in the HNP 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 
maintains diversity and 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 proposed amendment, for both the conditions 
described herein and any future application of the methodology, does 
not involve a significant reduction in a 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: Lara Nichols, Deputy General Counsel, Duke 
Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, 
Charlotte, NC 28202.
    NRC Branch Chief: Douglas A. Broaddus.

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

    Date of amendment request: December 6, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17340B321.
    Description of amendment request: The amendment would revise 
Technical Specification 3/4.3.2 Table 4.3-2, ``Engineered Safety 
Features Actuation System [ESFAS] Instrumentation Surveillance 
Requirements.'' The amendment would remove from Note 3 of the table the 
exemption from testing ESFAS relays K114, K305, and K313 at power.
    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 remove the Technical Specification 
Table 4.3-2 Note 3 exemption for testing relays K305, K313, and K114 
at power. The Technical Specification Table 4.3-2 Note 3 exemption 
allowed the K305, K313, and K114 to not be tested during power 
operation. The K305 and K313 relays are associated with the Main 
Steam Isolation Signal (MSIS). The K114 relays are associated with 
the Containment Spray Actuation Signal (CSAS). The removal of the 
exemption from testing during power operation means the impacted 
relays will be tested more frequently improving the ability to 
identify failed components.
    The removal of the Technical Specification Table 4.3-2 Note 3 
exemption for testing relays K305, K313, and K114 means these relays 
will be tested more frequently. This testing frequency will be 
consistent with the other Technical Specification Table 4.3-2 
subgroup relays that do not have an exemption. The probability of an 
operator choosing the wrong subgroup relay during testing is no 
different for this change as it is for the existing Technical 
Specification Table 4.3-2 subgroup relays that are already tested on 
this same frequency. Thus, there will be no significant increase in 
the probability of an operator error causing an accident.
    The change will also eliminate a potential single failure 
vulnerability associated with MSIS (relays K305 and K313) and CSAS 
(relay K114). The elimination of the single failure potential will 
lower the probability of an accident due to the spurious actuation 
of the MSIS or CSAS.
    The change uses a parallel 2 out of 2 with second 2 out of 2 to 
ensure no single failure of one actuation path would prevent the 
other actuation path from completing its function. This ensures no 
additional failure mode would prevent required equipment from 
actuating and increasing accident consequences.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change will remove the Technical Specification 
Table 4.3-2 Note 3 exemption for testing relays K305, K313, and 
K114. The K305, K313, and K114 relays are part of the Engineered 
Safety Features Actuation System (ESFAS). The ESFAS is used for 
accident mitigation but an inadvertent actuation could cause an 
accident. The K305 and K313 relays are associated with the MSIS. The 
K114 relays are associated with the CSAS. The potential failures of 
the main steam isolation and containment spray systems have been 
evaluated in the Waterford 3 Updated Final Safety Analysis Report 
(UFSAR). The potential accidents are as follows:
     Loss of External Load which could be caused by closure 
of the Main Steam Isolation Valves (MSIVs) (UFSAR Section 15.2, 
Decrease in Heat Removal by the Secondary System).
     Loss of normal Feedwater Flow which could be caused by 
the closure of the Main Feedwater Isolation Valves (UFSAR Section 
15.2, Decrease in Heat Removal by the Secondary System).

[[Page 6223]]

     Asymmetric Steam Generator Transient which could be 
caused by the closure of one MSIV (UFSAR Section 15.9.1.1, 
Asymmetric Steam Generator Transient).
     Loss of component cooling to Reactor Coolant Pumps 
(RCPs) which could be caused by the closure of the RCP Component 
Coolant Water valve. This could lead to RCP seal assembly damage and 
the possibility for a loss of coolant accident (UFSAR Section 15.6, 
Decrease In Reactor Coolant System Inventory).
     Inadvertent containment spray which could be caused by 
actuation of one train of containment spray (UFSAR Section 
6.2.1.1.3, Design Evaluation--Containment Pressure--Temperature 
Analysis).
    The removal of the exemption from testing during power operation 
means the impacted relays will be tested more frequently thereby 
improving the ability to identify failed components; however, they 
will be tested at power. The ESFAS K305, K313, and K114 relay test 
logic is designed to test the relays at power and not actuate the 
end devices which could adversely impact the plant. Any failures 
that could actuate plant equipment would continue to be bounded by 
the existing UFSAR accidents; therefore, no new accident is being 
created.
    The ESFAS is used for accident mitigation. The removal of the 
exemption from testing during power operation means the impacted 
relays will be tested more frequently thereby improving the ability 
to identify failed components. This lowers the possibility of the 
ESFAS equipment not being available when needed. This also means 
that with the ESFAS equipment available, this change does not create 
the possibility of a different kind of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will remove the Technical Specification 
Table 4.3-2 Note 3 exemption for testing relays K305, K313, and 
K114. The removal of the exemption from testing during power 
operation means the impacted relays will be tested more frequently 
thereby improving the ability to identify failed components. The 
more frequent testing will improve the margin of safety.
    The change will also eliminate a potential single failure 
vulnerability associated with MSIS (relays K305 and K313) and CSAS 
(relay K114). The elimination of the single failure potential will 
improve the margin of safety by reducing the potential of an 
accident due to the spurious actuation of the MSIS or CSAS.

    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.

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: December 13, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17360A159.
    Description of amendment request: The amendments would revise 
technical specifications (TSs) to adopt Technical Specification Task 
Force (TSTF)-542, Reactor Pressure Vessel Water Inventory Control (RPV 
WIC).
    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 replaces existing TS requirements related to 
OPDRVs [operations with a potential for draining the reactor vessel] 
with new requirements on RPV WIC water inventory control] that will 
protect Safety Limit 2.1.1.3. Draining of RPV water inventory in 
Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., refueling) is not an 
accident previously evaluated and, therefore, replacing the existing 
TS controls to prevent or mitigate such an event with a new set of 
controls has no effect on any accident previously evaluated. RPV 
water inventory control in Mode 4 or Mode 5 is not an initiator of 
any accident previously evaluated. The existing OPDRV controls or 
the proposed RPV WIC controls are not mitigating actions assumed in 
any accident previously evaluated.
    The proposed change reduces the probability of an unexpected 
draining event (which is not a previously evaluated accident) by 
imposing new requirements on the limiting time in which an 
unexpected draining event could result in the reactor vessel water 
level dropping to the top of the active fuel (TAF). These controls 
require cognizance of the plant configuration and control of 
configurations with unacceptably short drain times. These 
requirements reduce the probability of an unexpected draining event. 
The current TS requirements are only mitigating actions and impose 
no requirements that reduce the probability of an unexpected 
draining event.
    The proposed change reduces the consequences of an unexpected 
draining event (which is not a previously evaluated accident) by 
requiring an Emergency Core Cooling System (ECCS) subsystem to be 
operable at all times in Modes 4 and 5. The current TS requirements 
do not require any water injection systems, ECCS or otherwise, to be 
operable in certain conditions in Mode 5. The change in requirement 
from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does 
not significantly affect the consequences of an unexpected draining 
event because the proposed Actions ensure equipment is available 
within the limiting drain time that is as capable of mitigating the 
event as the current requirements. The proposed controls provide 
escalating compensatory measures to be established as calculated 
drain times decrease, such as verification of a second method of 
water injection and additional confirmations that secondary 
containment and/or filtration would be available if needed.
    The proposed change reduces or eliminates some requirements that 
were determined to be unnecessary to manage the consequences of an 
unexpected draining event, such as automatic initiation of an ECCS 
subsystem and control room ventilation. These changes do not affect 
the consequences of any accident previously evaluated since a 
draining event in Modes 4 and 5 is not a previously evaluated 
accident and the requirements are not needed to adequately respond 
to a draining event.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    The proposed change replaces existing TS [technical 
specification] requirements related to OPDRVs with new requirements 
on RPV WIC that will protect Safety Limit 2.1.1.3. The proposed 
change will not alter the design function of the equipment involved. 
Under the proposed change, some systems that are currently required 
to be operable during OPDRVs would be required to be available 
within the limiting drain time or to be in service depending on the 
limiting drain time. Should those systems be unable to be placed 
into service, the consequences are no different than if those 
systems were unable to perform their function under the current TS 
requirements.
    The event of concern under the current requirements and the 
proposed change is an unexpected draining event. The proposed change 
does not create new failure mechanisms, malfunctions, or accident 
initiators that would cause a draining event or a new or different 
kind of accident not previously evaluated or included in the design 
and licensing bases.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change replaces existing TS requirements related to 
OPDRVs with new requirements on RPV WIC. The current requirements do 
not have a stated safety basis

[[Page 6224]]

and no margin of safety is established in the licensing basis. The 
safety basis for the new requirements is to protect Safety Limit 
2.1.1.3. New requirements are added to determine the limiting time 
in which the RPV water inventory could drain to the top of the fuel 
in the reactor vessel should an unexpected draining event occur. 
Plant configurations that could result in lowering the RPV water 
level to the TAF within one hour are now prohibited. New escalating 
compensatory measures based on the limiting drain time replace the 
current controls. The proposed TS establish a safety margin by 
providing defense-in-depth to ensure that the Safety Limit is 
protected and to protect the public health and safety. While some 
less restrictive requirements are proposed for plant configurations 
with long calculated drain times, the overall effect of the change 
is to improve plant safety and to add safety margin.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: December 15, 2017. A publicly available 
version is in ADAMS under Accession No. ML17349A027.
    Description of amendment request: The amendment would revise the 
Nine Mile Point Nuclear Station, Unit 1, Technical Specifications (TSs) 
by replacing existing requirements related to ``operations with a 
potential for draining the reactor vessel'' (OPDRVs) with new 
requirements on reactor pressure vessel water (RPV) inventory control 
(WIC). The proposed changes are based on Technical Specifications Task 
Force (TSTF) Improved Standard Technical Specifications Change Traveler 
TSTF-542, Revision 2, ``Reactor Pressure Vessel Water Inventory 
Control'' (ADAMS Accession No. ML16074A448).
    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 replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will ensure RPV water 
level remains above -10 inches indicator scale. Draining of RPV 
water inventory in the cold shutdown and refueling conditions is not 
an accident previously evaluated; therefore, replacing the existing 
TS controls to prevent or mitigate such an event with a new set of 
controls has no effect on any accident previously evaluated. RPV 
water inventory control in the cold shutdown or refueling condition 
is not an initiator of any accident previously evaluated. The 
existing OPDRV controls or the proposed RPV WIC controls are not 
mitigating actions assumed in any accident previously evaluated.
    The proposed changes reduce the probability of an unexpected 
draining event (which is not a previously evaluated accident) by 
imposing new requirements on the limiting time in which an 
unexpected draining event could result in the reactor vessel water 
level dropping to -10 inches indicator scale. These controls require 
cognizance of the plant configuration and control of configurations 
with unacceptably short drain times. These requirements reduce the 
probability of an unexpected draining event. The current TS 
requirements are only mitigating actions and impose no requirements 
that reduce the probability of an unexpected draining event.
    The proposed changes reduce the consequences of an unexpected 
draining event (which is not a previously evaluated accident) by 
requiring a Core Spray subsystem to be operable at all times in the 
cold shutdown and refueling conditions. The change in requirement 
from two Core Spray subsystems to one Core Spray subsystem in the 
cold shutdown or refueling conditions does not significantly affect 
the consequences of an unexpected draining event because the 
proposed Actions ensure equipment is available within the limiting 
drain time that is as capable of mitigating the event as the current 
requirements. The proposed controls provide escalating compensatory 
measures to be established as calculated drain times decrease, such 
as verification of a second method of water injection and additional 
confirmations that containment and/or filtration would be available 
if needed.
    The proposed changes reduce or eliminate some requirements that 
were determined to be unnecessary to manage the consequences of an 
unexpected draining event, such as automatic initiation of a Core 
Spray subsystem and control room ventilation. These changes do not 
affect the consequences of any accident previously evaluated since a 
draining event in the cold shutdown or refueling condition is not a 
previously evaluated accident and the requirements are not needed to 
adequately respond to a draining event.
    Therefore, the proposed changes do 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 replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will maintain RPV water 
level above -10 inches indicator scale. The proposed changes will 
not alter the design function of the equipment involved. Under the 
proposed changes, some systems that are currently required to be 
operable during OPDRVs would be required to be available within the 
limiting drain time or to be in service depending on the limiting 
drain time. Should those systems be unable to be placed into 
service, the consequences are no different than if those systems 
were unable to perform their function under the current TS 
requirements.
    The event of concern under the current requirements and the 
proposed change is an unexpected draining event. The proposed 
changes do not create new failure mechanisms, malfunctions, or 
accident initiators that would cause a draining event or a new or 
different kind of accident not previously evaluated or included in 
the design and licensing bases.
    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 proposed changes replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC. The current requirements do 
not have a stated safety basis and no margin of safety is 
established in the licensing basis. The safety basis for the new 
requirements is to maintain RPV water level above -10 inches 
indicator scale. New requirements are added to determine the 
limiting time in which the RPV water inventory could drain to the 
top of the fuel in the reactor vessel should an unexpected draining 
event occur. Plant configurations that could result in lowering the 
RPV water level to -10 inches indicator scale within one hour are 
now prohibited. New escalating compensatory measures based on the 
limiting drain time replace the current controls. The proposed TS 
establish a safety margin by providing defense-in-depth to maintain 
RPV water level above -10 inches indicator scale to protect the 
public health and safety. While some less restrictive requirements 
are proposed for plant configurations with long calculated drain 
times, the overall effect of the change is to improve plant safety 
and to add safety margin.
    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.

[[Page 6225]]

    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: James G. Danna.

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of amendment request: November 10, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17314A024.
    Description of amendment request: The amendment would make changes 
to the organization, staffing, and training requirements contained in 
Section 6.0, ``Administrative Controls,'' of the Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Technical Specifications (TSs) and 
define two new positions for Certified Fuel Handler and Non-Certified 
Operator in Section 1.0, ``Definitions,'' to reflect the permanently 
defueled condition.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes would not take effect until TMI-1 has 
permanently ceased operation and certified a permanently defueled 
condition. The proposed changes would revise the TMI-1 TS by 
deleting or modifying certain portions of the TS administrative 
controls described in Section 6.0 of the TS that are no longer 
applicable to a permanently shutdown and defueled facility. 
Additionally, the ``Certified Fuel Handler'' and ``Non-Certified 
Operator'' would be added to Section 1.0 of the TS to define these 
positions that are applicable to permanently shutdown and defueled 
facility. These changes are administrative in nature.
    The proposed changes do not involve any physical changes to 
plant Structures, Systems, and Components (SSCs) or the manner in 
which SSCs are operated, maintained, modified, tested, or inspected. 
The proposed changes do not involve a change to any safety limits, 
limiting safety system settings, limiting control settings, limiting 
conditions for operation, surveillance requirements, or design 
features.
    The changes do not directly affect the design of SSCs necessary 
for safe storage of spent irradiated fuel or the methods used for 
handling and storage of such fuel in the Spent Fuel Pool (SFP). The 
proposed changes are administrative in nature and do not affect any 
accidents applicable to the safe management of spent irradiated fuel 
or the permanently shutdown and defueled condition of the reactor.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the TS definitions and administrative 
controls have no impact on facility plant Structures, Systems, and 
Components (SSCs) affecting the safe storage of spent irradiated 
fuel, or on the methods of operation of such SSCs, or on the actual 
handling and storage of spent irradiated fuel. The proposed changes 
do not result in different or more adverse failure modes or 
accidents than previously evaluated because the reactor will be 
permanently shutdown and defueled and TMI-1 will no longer be 
authorized to operate the reactor.
    The proposed changes do not affect systems credited in the 
accident analyses at TMI-1. The proposed changes will continue to 
require proper control and monitoring of safety significant 
parameters and activities.
    The proposed changes do not result in any new mechanisms that 
could initiate damage to the remaining relevant safety barriers in 
support of maintaining the plant in a permanently shutdown and 
defueled condition (e.g., fuel cladding and SFP cooling). Since 
extended operation in a defueled condition will be the only 
operation allowed, and therefore bounded by the existing analyses, 
such a condition does not create the possibility of a new or 
different kind of accident.
    The proposed changes do not alter the protection system design, 
create new failure modes, or change any modes of operation. The 
proposed changes do 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 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 proposed changes involve TS administrative controls once the 
TMI-1 facility has been permanently shutdown and defueled. As 
specified in 10 CFR 50.82(a)(2), the 10 CFR 50 license for TMI-1 
will no longer authorize operation of the reactor or emplacement or 
retention of fuel into the reactor vessel following submittal of the 
certifications required by 10 CFR 50.82(a)(1). As a result, the 
occurrence of certain design basis postulated accidents are no 
longer considered credible when the reactor is permanently defueled.
    The proposed changes are limited to those portions of the 
administrative TSs that are related to the safe storage and 
maintenance of spent irradiated fuel. The proposed TS changes do not 
affect plant design, hardware, system operation, or procedures for 
accident mitigation systems. There is no change in the established 
safety margins for these systems. The requirements that are proposed 
to be added, revised and/or deleted from the TMI-1 TS are not 
credited in the existing accident analysis for the applicable 
postulated accidents; therefore, they do not contribute to the 
margin of safety associated with the accident analysis. Certain 
postulated design basis accidents (DBAs) involving the reactor are 
no longer possible because the reactor will be permanently shutdown 
and defueled and TMI-1 will no longer be authorized to operate the 
reactor.
    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: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: James G. Danna.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: December 20, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17355A019.
    Description of amendment request: The amendment would revise 
technical specification (TS) requirements related to direct current 
(DC) electrical systems, specifically limiting conditions for operation 
3.8.4, 3.8.5, and 3.8.6. The proposed amendment would also add a new 
Battery and Monitoring Maintenance Program to TS Section 5.5, 
``Programs and Manuals.'' The proposed changes are consistent with 
Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision 
2, ``DC Electrical Rewrite--Update to TSTF-360.''
    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 restructure the Technical Specifications 
(TS) for the direct current (DC) electrical power system and are 
consistent with TSTF-500, Revision 2, ``DC Electrical Rewrite--
Update to TSTF-360.'' The proposed changes modify TS Actions 
relating to battery and battery charger

[[Page 6226]]

inoperability. The DC electrical power system, including associated 
battery chargers, is not an initiator of any accident sequence 
analyzed in the Updated Safety Analysis Report (USAR). Rather, the 
DC electrical power system supports equipment used to mitigate 
accidents. The proposed changes to restructure TS and change 
surveillances for batteries and chargers to incorporate the updates 
included in TSTF-500, Revision 2, will maintain the same level of 
equipment performance required for mitigating accidents assumed in 
the USAR. Operation in accordance with the proposed TS would ensure 
that the DC electrical power system is capable of performing its 
specified safety function as described in the USAR. Therefore, the 
mitigating functions supported by the DC electrical power system 
will continue to provide the protection assumed by the analysis. The 
relocation of preventive maintenance surveillances, and certain 
operating limits and actions, to a licensee-controlled battery 
monitoring and maintenance program will not challenge the ability of 
the DC electrical power system to perform its design function. 
Appropriate monitoring and maintenance that are consistent with 
industry standards will continue to be performed. In addition, the 
DC electrical power system is within the scope of 10 CFR 50.65, 
``Requirements for monitoring the effectiveness of maintenance at 
nuclear power plants,'' which will ensure the control of maintenance 
activities associated with the DC electrical power system.
    The integrity of fission product barriers, plant configuration, 
and operating procedures as described in the USAR will not be 
affected by the proposed changes. Therefore, the consequences of 
previously analyzed accidents will not increase by implementing 
these changes. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes involve restructuring the TS for the DC 
electrical power system. The DC electrical power system, including 
associated battery chargers, is not an initiator to any accident 
sequence analyzed in the USAR. Rather, the DC electrical power 
system supports equipment used to mitigate accidents. The proposed 
changes to restructure the TS and change surveillances for batteries 
and chargers to incorporate the updates included in TSTF-500, 
Revision 2, ``DC Electrical Rewrite--Update to TSTF-360,'' will 
maintain the same level of equipment performance required for 
mitigating accidents assumed in the USAR. Administrative and 
mechanical controls are in place to ensure the design and operation 
of the DC systems continues to meet the plant design basis described 
in the USAR. Therefore, operation of the facility in accordance with 
this proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The equipment margins will be maintained in 
accordance with the plant-specific design bases as a result of the 
proposed changes. The proposed changes will not adversely affect 
operation of plant equipment. These changes will not result in a 
change to the setpoints at which protective actions are initiated. 
Sufficient DC capacity to support operation of mitigation equipment 
is ensured. The changes associated with the new battery maintenance 
and monitoring program will ensure that the station batteries are 
maintained in a highly reliable manner. The equipment fed by the DC 
electrical sources will continue to provide adequate power to 
safety-related loads in accordance with analysis assumptions.
    TS changes made in accordance with TSTF-500, Revision 2, ``DC 
Electrical Rewrite--Update to TSTF-360,'' maintain the same level of 
equipment performance stated in the USAR and the current TSs. 
Therefore, the proposed changes do not involve a significant 
reduction of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: David J. Wrona.

NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Manitowoc County, Wisconsin

    Date of amendment request: August 31, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17243A201.
    Description of amendment request: The proposed amendment would 
modify the licensing basis, by the addition of a License Condition, to 
allow for the implementation of the provisions of 10 CFR part 50.69, 
``Risk-Informed Categorization and Treatment of Structures, Systems, 
and Components (SSCs) for Nuclear Power Plants.'' The provisions of 10 
CFR 50.69 allow adjustment of the scope of equipment subject to special 
treatment controls (e.g., quality assurance, testing, inspection, 
condition monitoring, assessment, and evaluation). For equipment 
determined to be of low safety significance, alternative treatment 
requirements can be implemented in accordance with this regulation. For 
equipment determined to be of high safety significance, requirements 
will not be changed or will be enhanced.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

[[Page 6227]]

margin. The safety margins included in analyses of accidents are not 
affected by the proposed change. The regulation requires that there 
be no significant effect on plant risk due to any change to the 
special treatment requirements for SSCs and that the SSCs continue 
to be capable of performing their design basis functions, as well as 
to perform any beyond design basis functions consistent with the 
categorization process and results.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Steven Hamrick, Managing Attorney--Nuclear 
Florida Power & Light Company, LAW/WAS, 801 Pennsylvania Ave. NW #220, 
Washington, DC 20004.
    NRC Branch Chief: David J. Wrona.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
No. 1, Rockingham County, New Hampshire

    Date of amendment request: December 1, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17339A428.
    Description of amendment request: The amendment would revise 
certain 18-month surveillance requirements previously performed while 
shut down to be performed during power operations. The amendment would 
also revise the administrative controls portion of the technical 
specifications (TSs) to replace plant-specific titles with generic 
titles and modify TSs 6.1.2, 6.2.2, 6.2.4, and Table 6.2-1 to be 
consistent with NUREG-1431, ``Standard Technical Specifications, 
Westinghouse 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 technical specification (TS) surveillance requirements and 
administrative controls associated with the proposed changes to the 
TS are not initiators of any accidents previously evaluated, so the 
probability of accidents previously evaluated is unaffected by the 
proposed changes. The proposed change does not alter the design, 
function, or operation of any plant structure, system, or component 
(SSC). The capability of any operable TS-required SSC to perform its 
specified safety function is not impacted by the proposed change. As 
a result, 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 change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed change does not challenge the integrity or 
performance of any safety-related systems. No plant equipment is 
installed or removed, and the changes do not alter the design, 
physical configuration, or method of operation of any plant SSC.
    No physical changes are made to the plant, so no new causal 
mechanisms are introduced. Therefore, the proposed changes to the TS 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not challenge the integrity or 
performance of any safety-related systems. No plant equipment is 
installed or removed, and the changes do not alter the design, 
physical configuration, or method of operation of any plant SSC. No 
physical changes are made to the plant, so no new causal mechanisms 
are introduced. Therefore, the proposed changes to the TS 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 
the 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 changes 
do not alter any safety analyses assumptions, safety limits, 
limiting safety system settings, or method of operating the plant. 
The changes do not adversely affect plant operating margins or the 
reliability of equipment credited in the safety analyses. With the 
proposed change, each DC electrical train remains fully capable of 
performing its safety function. 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: Steve Hamrick, Acting Managing Attorney, 
Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-
0420.
    NRC Branch Chief: James G. Danna.

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

    Date of amendment request: July 28, 2017, as supplemented by 
January 23, 2108, letter. Publicly-available versions are in ADAMS 
under Accession No. ML17209A755, and ML18023A440, respectively.
    Description of amendment request: The requested amendment proposes 
changes to combined license Appendix A, plant-specific Technical 
Specifications (TS) to make them consistent with the remainder of the 
design, licensing basis, and the TS. The U.S. Nuclear Regulatory 
Commission (NRC) staff previously noticed this amendment request in the 
Federal Register on December 5, 2017 (82 FR 57473). However, due to 
administrative errors that were inadvertently introduced, the NRC staff 
is noticing this amendment request again.
    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's edits in 
square brackets:

    An evaluation to determine whether or not a significant hazards 
consideration is involved with the proposed amendment was completed 
by focusing on the three standards set forth in 10 CFR 50.92, 
``Issuance of amendment,'' as discussed below. However, to provide 
for ease of review, similar changes have been grouped into 
categories to facilitate the significant hazards evaluations 
required by 10 CFR 50.92. Generic significant hazards evaluations 
are provided for the More Restrictive Changes and a specific 
significant hazards evaluation for each Clarification or Less 
Restrictive change. In regards to obvious editorial or 
administrative changes (e.g., formatting, page rolls, punctuation, 
etc.), an explicit discussion was not always provided, but is 
considered to be addressed by the applicable generic significant 
hazards evaluation.

Valuation for More Restrictive Changes

    This generic category include changes that impose additional 
requirements, decrease allowed outage times, increase the Frequency 
of Surveillances, impose additional Surveillances, increase the 
scope of Specifications to include additional plant equipment, 
broaden the Applicability of Specifications, or provide additional 
actions. These changes have been evaluated to not be detrimental to 
plant safety.
    More restrictive changes are proposed only when such changes are 
consistent with the current Vogtle Electric Generating Plant, Units 
3 and 4 (VEGP) licensing basis; the applicable VEGP safety analyses; 
and good engineering practice such that the availability and 
reliability of the affected equipment is not reduced.
    Changes to the Technical Specifications (TS) requirements 
categorized as More Restrictive are annotated with an ``MR'' in 
Section 2 Discussion of Change (DOC). This affects TS changes L05 
and L08.
    Southern Nuclear Operating Company (SNC) proposes to amend the 
VEGP TS. SNC

[[Page 6228]]

has evaluated each of the proposed TS changes identified as More 
Restrictive in accordance with the criteria set forth in 10 CFR 
50.92, ``Issuance of amendment,'' and has determined that the 
proposed changes do not involve a significant hazards consideration. 
This significant hazards consideration is applicable to each More 
Restrictive change identified in Section 2.
    The basis for the determination that the proposed changes do not 
involve a significant hazards consideration is an evaluation of 
these changes against each of the criteria in 10 CFR 50.92(c). The 
criteria and conclusions of the evaluation are 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 provide more stringent TS requirements. 
These more stringent requirements impose greater operational control 
and conservatism, and as a result, do not result in operations that 
significantly increase the probability of initiating an analyzed 
event, and do not alter assumptions relative to mitigation of an 
accident or transient event. The more restrictive requirements 
continue to ensure process variables, structures, systems, and 
components are maintained consistent with the safety analyses and 
licensing basis. 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 involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in methods governing normal plant operation. The proposed 
changes do impose different Technical Specification requirements. 
However, these changes are consistent with the assumptions in the 
safety analyses and licensing basis. Therefore, the proposed changes 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The imposition of more restrictive requirements either has no 
effect on or increases a margin of plant safety. As provided in the 
discussion of change, each change in this category is, by 
definition, providing additional restrictions to enhance plant 
safety. The changes maintain requirements within the safety analyses 
and licensing basis. Therefore, the proposed changes do not involve 
a significant reduction in a margin of safety.

Evaluation for Clarification Changes

    This category consists of technical changes which revise 
existing requirements such that the design and operation of a system 
correctly reflects how the LCO is applied and how the Action or 
Surveillance Requirement (SR) is carried out. This adds detail and 
clarity to the Technical Specifications (TS) in operating the 
applicable portions of the as designed and licensed plant.
    Technical changes to the TS requirements categorized as 
``Clarification'' are identified with an ``CL'' and an individual 
number in Section 2 Discussion of Change (DOC).
    Southern Nuclear Operating Company (SNC) proposes to amend the 
Vogtle Electric Generating Plant, Units 3 and 4 (VEGP), Technical 
Specifications. SNC has evaluated each of the proposed technical 
changes identified as ``Clarification'' individually in accordance 
with the criteria set forth in 10 CFR 50.92 and has determined that 
the proposed changes do not involve a significant hazards 
consideration.
    The basis for the determination that the proposed changes do not 
involve a significant hazards consideration is an evaluation of 
these changes against each of the criteria in 10 CFR 50.92(c). The 
criteria and conclusions of the evaluation are presented below.
    L09 SNC proposes to amend TS 3.3.19 Diverse Actuation System 
Manual Controls, Note (c) in Table 3.3.19-1 to ``With upper 
internals in place.''
    SNC has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment by focusing on 
the three standards set forth in 10 CFR 50.92, ``Issuance of 
amendment,'' as discussed below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not 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 is not an initiator to any accident 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 of an accident previously evaluated 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 change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change clarifies TS requirements for the DAS manual 
control ADS Stage 4 valves such that they would be in agreement with 
the requirements set forth for the ADS in RCS Shutdown Mode 6. 
However, the proposed change does not involve a physical alteration 
of the plant as described in the [Updated Final Safety Analysis 
Report (UFSAR)]. No new equipment is being introduced, and equipment 
is not being operated in a new or different manner. There are no 
setpoints, 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 change 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 
condition for the manual control of ADS Stage 4 actuation switches 
in Mode 6 has changed, no action is made less restrictive than 
currently approved for any associated actuated device inoperability. 
As such, there is no significant reduction in a margin of safety.
    L10 SNC proposes to amend current TS 3.5.4, ``Passive Residual 
Heat Removal Heat Exchanger PRHR HX--Operating,'' Surveillance 
Requirement (SR) 3.5.4.6 to: Verify both PRHR HX air operated outlet 
valves stroke open and both IRWST gutter isolation valves stroke 
closed.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant or a change in the methods governing normal plant 
operations. The change involves correcting an existing surveillance 
requirement (with no change in required system or device

[[Page 6229]]

function), such that the surveillance requirement complies with the 
In-Containment Refueling Water Storage Tank (IRWST) Gutter Isolation 
valve design and the Passive Residual Heat Removal (PRHR) Heat 
Exchanger (HX) outlet isolation valve design. Revised surveillance 
requirement presentation and compliance with TS actions are not an 
initiator to any accident 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 
surveillance requirement are no different than the consequences of 
the same accident during the existing one. As a result, the 
consequences of an accident previously evaluated 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 change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change clarifies the surveillance requirement such 
that it agrees with the IRWST and PRHR HX isolation valve design. 
However, 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 setpoints, 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 change 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 
surveillance requirement has changed for the IRWST and PRHR HX 
isolation valves, no action is made less restrictive than currently 
approved for any associated actuated device inoperability. As such, 
there is no significant reduction in a margin of safety.

10 CFR 50.92 Evaluations for Less Restrictive Changes

    This category consists of technical changes which revise 
existing requirements such that more restoration time is provided, 
fewer compensatory measures are needed, unnecessary Surveillance 
Requirements (SR) are deleted, or less restrictive surveillance 
requirements are required. This would also include unnecessary 
requirements which are deleted from the Technical Specifications 
(TS) and other technical changes that do not fit a generic category. 
These changes are evaluated individually.
    Technical changes to the TS requirements categorized as ``Less 
Restrictive'' are identified with an ``LR'' and an individual number 
in Section 2 Discussion of Change (DOC).
    Southern Nuclear Operating Company (SNC) proposes to amend the 
Vogtle Electric Generating Plant, Units 3 and 4 (VEGP), Technical 
Specifications. SNC has evaluated each of the proposed technical 
changes identified as ``Less Restrictive'' individually in 
accordance with the criteria set forth in 10 CFR 50.92 and has 
determined that the proposed changes do not involve a significant 
hazards consideration.
    The basis for the determination that the proposed changes do not 
involve a significant hazards consideration is an evaluation of 
these changes against each of the criteria in 10 CFR 50.92(c). The 
criteria and conclusions of the evaluation are presented below.
    L01 SNC proposes to amend TS 1.1 Definitions--Shutdown Margin 
by:
    Changing Shutdown Margin (SDM) definition c. ``In MODE 2 with 
keff<1.0 and MODES 3, 4, and 5, the worth of fully inserted Gray Rod 
Cluster Assemblies (GRCAs) will be included in the SDM 
calculation.'' to ``In MODE 2 with keff<1.0 and in MODES 3, 4, and 
5, the worth of the verified fully inserted Gray Rod Cluster 
Assemblies (GRCAs) which have passed the acceptance criteria for 
GRCA bank worth measurements performed during startup physics 
testing may be included in the SDM calculation.''
    SNC has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment by focusing on 
the three standards set forth in 10 CFR 50.92, ``Issuance of 
amendment,'' as discussed below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant or a change in the methods governing normal plant 
operations. The change proposed involves re-defining whether the 
worth of the Gray Rod Cluster Assemblies (GRCAs) should be included 
in MODE 2 with keff<1.0 and Modes 3, 4, and 5 when calculating the 
appropriate Shutdown Margin (SDM). The worth of the GRCAs for MODE 2 
with keff<1.0 and Modes 3, 4, and 5 is not credited in the safety 
analyses as stated in the NRC Safety Evaluation Report (SER) 
``Westinghouse Electric Company's Final Topical Report Safety 
Evaluation For WCAP-16943, ``Enhanced Gray Rod Cluster Assembly 
Rodlet Design,'' Section 3.0 for ensuring adequate SDM exists.
    The change involves revising the existing SDM definition (with 
no change in required system or device function), such that a more 
appropriate, albeit less restrictive, definition would be applied 
when calculating SDM. The revised SDM definition is not an initiator 
of any accident 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 
definition requirements are no different than the consequences of 
the same accident during the existing one. As a result, the 
consequences of an accident previously evaluated 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.
    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 setpoints, 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 
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 removes the requirement to include the worth 
of the GRCAs when calculating the SDM because they are not credited 
for SDM in MODE 2 with keff<1.0 and in MODES 3, 4, and 5. 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 setpoints, 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

[[Page 6230]]

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 change 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 
SDM calculation defined is made less restrictive by eliminating the 
worth of the GRCAs in MODE 2 with keff<1.0 and in MODES 3, 4, and 5, 
no credit is taken in the safety analyses for including their worth 
as discussed in the NRC Safety Evaluation Report (SER) 
``Westinghouse Electric Company's Final Topical Report Safety 
Evaluation For WCAP-16943, ``Enhanced Gray Rod Cluster Assembly 
Rodlet Design,'' Section 3.0. As such, there is no significant 
reduction in a margin of safety.
    L02 SNC proposes to amend TS 3.1.4 Rod Group Alignment Limits 
by:
    L02A. Change Limiting Condition of Operation (LCO) from ``All 
shutdown and control rods shall be OPERABLE.'' to ``Each rod cluster 
control assembly (RCCA) shall be OPERABLE.''
    L02B. Change LCO AND statement from ``Individual indicated rod 
positions shall be within 12 steps of their group step counter 
demand position.'' to ``Individual indicated rod positions of each 
RCCA and Gray Rod Cluster Assembly shall be within their 12 steps of 
their group step counter demand position.''
    L02C. Delete LCO 3.1.4 note.
    L02D. Change Action Condition A from ``one or more rod(s) 
inoperable.'' to where it now applies to ``One or more RCCA(s) 
inoperable.''
    L02E. Acronym defined in change to Required Action B.1 
Completion Time from ``1 hour with the OPDMS not monitoring 
parameters'' to ``1 hour with the On-Line Power Distribution 
Monitoring System not monitoring parameters.''
    L02F. Add Required Action B.2.3.1 where the Required Action will 
be to ``Perform SR 3.2.5.1'' with a Completion Time of ``Once per 12 
hours,'' OR perform B.2.3, which is renumbered as B.2.3.2.1.
    L02G. Delete Required Action B.2.4 Note, and renumber the 
Required Action to B.2.3.2.2.
    L02H. Delete Required Action B.2.5 Note, and renumber the 
Required Action to B.2.3.2.3.
    L02I. Renumber Required Action B.2.6 to B.2.4.
    L02J. Change SR 3.1.4.2 Note from ``Not applicable to GRCAs'' to 
``Not applicable to Axial Offset (AO) Control Bank RCCAs.''
    L02K. Change SR 3.1.4.2 from ``Verify rod freedom of movement 
(trippability) by moving each rod not fully inserted in the core 
>=10 steps in either direction.'' to ``Verify rod freedom of 
movement (trippability) by moving each RCCA not fully inserted in 
the core >=10 steps in either direction.''
    L02L. Delete the Note to SR 3.1.4.3
    L02M. Change SR 3.1.4.3 from ``Verify rod drop time of each rod 
. . .'' to ``Verify rod drop time of each RCCA . . .''.
    SNC has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment by focusing on 
the three standards set forth in 10 CFR 50.92, ``Issuance of 
amendment,'' as discussed below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant or a change in the methods governing normal plant 
operations. The proposed changes involve revising the existing LCO 
3.1.4 operability to be applicable to 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 
changes involve excluding the Gray Rod Cluster Assemblies (GRCAs) in 
the LCO 3.1.4 Rod Group Alignments LCO since their trip reactivity 
worth is not credited in the shutdown margin assessments in MODES 1 
and 2, nor required by the design basis to be operable. Only the rod 
cluster control assemblies (RCCAs) are required to be operable. The 
maximum rod misalignment is an initial assumption in the safety 
analyses that directly affects core power distributions and 
assumption of available shutdown margin (SDM). Since the GRCAs do 
not have a function to maintain the reactor sub-critical unless they 
are fully inserted, and the reactor is shut down, operability does 
not apply to GRCAs like it does for RCCAs in MODES 1 and 2. The 
design basis function of the GRCAs when the reactor is critical does 
not include a provision of trip reactivity.
    The revised LCO, associated actions and surveillance 
requirements are not an initiator to any accident 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 LCO 
requirements, associated actions, and surveillance requirements are 
no different than the consequences of the same accident during the 
existing ones. As a result, the consequences of an accident 
previously evaluated 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. 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 setpoints, 
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 
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 involves revising the existing LCO 3.1.4 
operability to be applicable to 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 setpoints, 
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 change 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 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.
    L03 SNC proposes to amend TS 3.1.6 Control Bank Insertion Limits 
by changing Note 2. from ``This LCO is not applicable to Gray Rod 
Cluster Assembly (GRCA) banks during GRCA bank sequence exchange 
with On-Line Power Distribution Monitoring System monitoring 
parameters'' to ``This LCO is not applicable to Gray Rod Cluster 
Assembly (GRCA) banks for up to one hour during GRCA bank sequence 
exchange.''
    SNC has evaluated whether or not a significant hazards 
consideration is involved

[[Page 6231]]

with the proposed amendment by focusing on the three standards set 
forth in 10 CFR 50.92, ``Issuance of amendment,'' as discussed 
below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant or a change in the methods governing normal plant 
operations. The proposed change to TS 3.1.6 Control Bank Insertion 
Limits Note 2 is to not require On Line Power Distribution System 
(OPDMS) during GRCA bank sequence exchange and limit the LCO 
applicability exception for one hour after the insertion or sequence 
or overlap limits are violated due to the short duration of the 
sequence exchange. The final mechanical shim (MSHIM) design 
established that the GRCA bank sequence exchange will best be 
accomplished by moving both banks at the same time. The entire 
exchange sequence will only take a few minutes from the time banks 
begin moving. During this short duration, OPDMS is not suited for 
real time monitoring relative to the time constant for the vanadium 
fixed incore detector system. The exchange transient may be 
completed before the OPDMS detects a significant change in the core 
radial power distribution. In addition, it is unlikely there would 
be significant time to take corrective action in response to an 
OPDMS alarm if one occurred during the exchange.
    The revised LCO note exception is not an initiator of any 
accident 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 LCO 
note exception is no different than the consequences of the same 
accident during the existing one. As a result, the consequences of 
an accident previously evaluated 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. 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 setpoints, 
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 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant 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 setpoints, 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 change 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 
proposed change to TS 3.1.6, Note 2 would not require OPDMS be 
functional during GRCA bank sequence exchange for up to one hour, 
OPDMS operability is still required by TS 3.2.5 On-Line Power 
Distribution Monitoring System (OPDMS)--Monitored Parameters. As 
such, there is no significant reduction in a margin of safety.
    L04 SNC proposes to amend TS 3.1.7 Rod Position Indication by 
deleting Required Action B.2 and renumbering the remaining Condition 
B Required Actions.
    SNC has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment by focusing on 
the three standards set forth in 10 CFR 50.92, ``Issuance of 
amendment,'' as discussed below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant or a change in the methods governing normal plant 
operations. The proposed change is to remove Required Action B.2 for 
monitoring and recording Reactor Coolant System (RCS) Tavg (with no 
change in required system or device function), such that more 
appropriate, albeit less restrictive, actions would be applied. 
There are no safety benefits, no acceptance criteria or no actions 
associated with any trends for recording Tavg. Monitoring Tavg 
provides no power distribution information for unmonitored rods that 
isn't already provided by complying with the existing requirements 
of Condition A, and average coolant temperature provides no 
indication of changes in shutdown margin.
    The revised actions are not an initiator of any accident 
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 LCO 
requirements and actions are no different than the consequences of 
the same accident during the existing ones. As a result, the 
consequences of an accident previously evaluated 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. 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 setpoints, 
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 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant 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 setpoints, 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.

[[Page 6232]]

    3. Does the proposed change 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 
required actions of LCO 3.1.7 for Rod Position Indication are made 
less restrictive by deletion of Action B.2 for monitoring Tavg, 
monitoring Tavg provides no power distribution information for 
unmonitored rods that aren't already provided by complying with the 
existing requirements of Condition A. As such, there is no 
significant reduction in a margin of safety.
    L06 SNC proposes to amend TS 3.3.1 ``Reactor Trip System 
Instrumentation,'' Table 3.3.1-1 FUNCTION 12, (page 2 of 2), Passive 
Residual Heat Removal Actuation by deleting SR 3.3.1.9.
    SNC has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment by focusing on 
the three standards set forth in 10 CFR 50.92, ``Issuance of 
amendment,'' as discussed 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 is to delete the Surveillance Requirement 
(SR) 3.3.1.9 Channel Calibration for the passive residual heat 
removal (PRHR) reactor trip system actuation. The PRHR reactor trip 
actuation initiates a reactor trip in the event either of the 
parallel PRHR discharge valves is not fully closed. The proper 
adjustment of the valve position indication contact inputs to the 
breaker position are verified by performance of SR 3.3.1.10 Trip 
Actuating Device Operational Test (TADOT). The revised surveillance 
requirements are not an initiator to any accident previously 
evaluated. The reactor trip from PRHR actuation has not changed, and 
the proper adjustment of the valve position indication contact 
inputs continues to be addressed by current SR 3.3.1.10. As a 
result, the probability of an accident previously evaluated is not 
affected.
    The consequences of an accident as a result of the revised 
surveillance requirements are no different than the consequences of 
the same accident during the existing ones. As a result, the 
consequences of an accident previously evaluated 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.
    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 setpoints, 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 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant 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 setpoints, 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 change 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 
surveillance requirements have been made less restrictive, the 
intent of the deleted surveillance requirement remains covered by an 
existing surveillance requirement. As such, there is no significant 
reduction in a margin of safety.
    L07 SNC proposes to amend TS, Section 3.3.5, ``Reactor Trip 
System Manual Actuation,'' Table 3.3.5-1 ``Reactor Trip System 
Manual Actuation,'' Functions 1. Manual Reactor Trip, 2. Safeguards 
Actuation Input from Engineered Safety Feature Actuation System--
Manual and 4. Core Makeup Tank Actuation Input from Engineered 
Safety Feature Actuation System--Manual for Required Channels to 2 
switches.
    SNC has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment by focusing on 
the three standards set forth in 10 CFR 50.92, ``Issuance of 
amendment,'' as discussed 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 define the required channels operable for 
manual reactor trip based upon the existing design. Required 
channels operable are not an initiator to any accident previously 
evaluated. As a result, the probability of an accident previously 
evaluated is not affected. The consequences of an accident with 
defined number of switches operable for manual reactor trip are no 
different than the consequences of the same accident using the 
existing required channels operable. As a result, the consequences 
of an accident previously evaluated are not affected by this change.
    The proposed change does not alter or prevent the ability of 
structures, systems, and components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits.
    The proposed change does not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated.
    Further, the proposed change does not increase the types or 
amounts of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures. The proposed change is consistent with the 
safety analysis assumptions and resultant consequences.
    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 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 setpoints, 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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to define the required channels operable 
consistent with the plant design does not alter the manner in which 
safety limits, limiting safety system settings or limiting 
conditions for operation are determined. The safety analysis 
acceptance criteria are not affected by this

[[Page 6233]]

change. The proposed change will not result in plant operation in a 
configuration outside of the design basis. Therefore, there is no 
significant reduction in a margin of safety.
    L11 SNC proposes to amend current TS 3.8.3, ``Inverters--
Operating,'' by changing:
    1. Action Condition A. from ``One inverter inoperable.'' to 
``One or two inverter(s) within one division inoperable.''
    2. Second Note in Required Action A.1 from ``Restore inverter to 
OPERABLE status.'' to ``Restore inverter(s) to OPERABLE status.''
    SNC has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment by focusing on 
the three standards set forth in 10 CFR 50.92, ``Issuance of 
amendment,'' as discussed below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant or a change in the methods governing normal plant 
operations. The proposed changes to action conditions to explicitly 
define an inverter division that contains two inoperable inverters 
is not an accident initiator nor do they impact mitigation of the 
consequences of any accident. Therefore, this change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed change does not involve a physical alteration of 
the plant as described in the UFSAR and does not alter the method of 
operation or control of equipment as described in the UFSAR. The 
current assumptions in the safety analysis regarding accident 
initiators and mitigation of accidents are unaffected by this 
change. Plant equipment remains capable of performing mitigative 
functions assumed by the accident analysis. No additional failure 
modes or mechanisms are being introduced and the likelihood of 
previously analyzed failures remains unchanged.
    The integrity of fission product barriers, plant configuration, 
and operating procedures as described in the UFSAR will not be 
affected by this change. Therefore, the consequences of previously 
analyzed accidents will not increase because of this change. 
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 to action conditions to explicitly define 
an inverter division that contains two inoperable inverters 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 setpoints, 
at which protective or mitigative actions are initiated, that are 
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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed change will not reduce a margin of 
safety because it has no such effect on any assumption of the safety 
analyses.
    Operation in accordance with the proposed TS operability ensures 
that the plant response to analyzed events continues to provide the 
margins of safety assumed by the analysis. Appropriate monitoring 
and maintenance, consistent with industry standards, will continue 
to be performed. Therefore, there is no significant reduction in a 
margin of safety.

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

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

    Date of amendment request: November 17, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17321B080.
    Description of amendment request: The amendment request proposes 
changes to combined license (COL) License Condition and changes to the 
Updated Final Safety Analysis Report (UFSAR) in the form of departures 
from the incorporated plant-specific Design Control Document Tier 2* 
and associated Tier 2 information. Specifically, this amendment request 
involves a change to COL License Condition requirements regarding the 
Natural Circulation (first plant test) using the steam generators and 
the Passive Residual Heat Removal Heat Exchanger (first plant test). A 
COL License Condition is proposed to be revised to include an exception 
that would allow the requirements of a Technical Specification to be 
suspended during performance of the Natural Circulation (first plant 
test) using the steam generators. In addition, a revised Passive 
Residual Heat Removal Heat Exchanger (first plant test) is proposed to 
be performed as part of the Power Ascension Testing requirements 
instead of as part of the Initial Criticality and Low-Power Testing 
requirements as currently specified in a COL License Condition.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not adversely affect the operation of 
any systems or equipment that initiate an analyzed accident or alter 
any structures, systems, and components (SSC) accident initiator or 
initiating sequence of events. The proposed changes do not adversely 
affect the ability of the steam generators, applicable reactor trip 
functions, and the passive residual heat removal heat exchanger to 
perform the required safety function to remove core decay heat 
during forced and natural circulation when necessary to prevent 
exceeding the reactor core and the reactor coolant system design 
limits, and do not adversely affect the probability of inadvertent 
operation or failure of the passive residual heat removal heat 
exchanger. The proposed changes do not result in any increase in 
probability of an analyzed accident occurring, and maintain the 
initial conditions and operating limits required by the accident 
analysis, and the analyses of normal operation and anticipated 
operational occurrences, so that the reactor core and the reactor 
coolant system design limits are not exceeded for events requiring 
emergency core decay heat removal.
    Therefore, the requested amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed 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 ability of the steam generators, applicable reactor trip 
functions, and the passive residual heat removal heat exchanger to 
perform the required safety function to remove core decay heat 
during forced and natural circulation when necessary to prevent 
exceeding the reactor

[[Page 6234]]

core and the reactor coolant system design limits, and do not 
adversely affect the probability of inadvertent operation or failure 
of the passive residual heat removal heat exchanger. The proposed 
changes do not result in the possibility of an accident occurring, 
and maintain the initial conditions and operating limits required by 
the accident analysis, and the analyses of normal operation and 
anticipated operational occurrences, so that the reactor core and 
the reactor coolant system design limits are not exceeded for events 
requiring emergency core decay heat removal.
    These proposed changes do not adversely affect any other SSC 
design functions or methods of operation in a manner that results in 
a new failure mode, malfunction, or sequence of events that affect 
safety related or nonsafety related equipment. Therefore, this 
activity does not allow for a new fission product release path, 
result in a new fission product barrier failure mode, or create a 
new sequence of events that results in significant fuel cladding 
failures.
    Therefore, the requested amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes maintain existing safety margins through 
continued application of the existing requirements of the UFSAR. The 
proposed changes maintain the initial conditions and operating 
limits required by the accident analysis, and the analyses of normal 
operation and anticipated operational occurrences, so that the 
reactor core and the reactor coolant system design limits are not 
exceeded for events requiring emergency core decay heat removal. 
Therefore, the proposed changes satisfy the same safety functions in 
accordance with the same requirements as stated in the UFSAR. These 
changes do not adversely affect any design code, function, design 
analysis, safety analysis input or result, or design/safety margin.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes, and no margin of 
safety is reduced. Therefore, the requested amendment does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant 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: December 21, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17355A416.
    Description of amendment request: The requested amendment proposes 
changes to combined license License Condition 2.D by adding a new 
condition to address the Tier 2* change process. The proposal also 
requests exemptions from the requirements of 10 CFR part 52, Appendix 
D, Paragraphs II.F, VIII.B.6.b, and VIII.B.6.c.
    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 would add a license condition that would 
allow use of the Tier 2 departure evaluation process for Tier 2* 
departures, where such departures would not have more than a minimal 
impact to safety. Changing the criteria by which departures from 
Tier 2* information are evaluated to determine if NRC approval is 
required does not affect the plant itself. Changing these criteria 
does not affect prevention and mitigation of abnormal events, e.g., 
accidents, anticipated operational occurrences, earthquakes, floods 
and turbine missiles, or their safety or design analyses. No safety-
related structure, system, component (SSC) or function is adversely 
affected. The changes neither involve nor interface with any SSC 
accident initiator or initiating sequence of events, and thus, the 
probabilities of the accidents evaluated in the Updated Final Safety 
Analysis Report (UFSAR) are not affected. Because the changes do not 
involve any safety related SSC or function used to mitigate an 
accident, the consequences of the accidents evaluated in the UFSAR 
are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes would add a license condition that would 
allow use of the Tier 2 departure evaluation process for Tier 2* 
departures, where such departures would not have more than a minimal 
impact to safety. The changes do not affect the safety-related 
equipment itself, nor do they affect equipment which, if it failed, 
could initiate an accident or a failure of a fission product 
barrier. No analysis is adversely affected. No system or design 
function or equipment qualification is adversely affected by the 
changes. 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 would result in significant 
fuel cladding failures. In addition, the changes do not result in a 
new failure mode, malfunction or sequence of events that could 
affect safety or safety-related equipment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes would add a license condition that would 
allow use of the Tier 2 departure evaluation process for Tier 2* 
departures, where such departures would not have more than a minimal 
impact to safety.
    The proposed change is not a modification, addition to, or 
removal of any plant SSCs. Furthermore, the proposed amendment is 
not a change to procedures or method of control of the nuclear plant 
or any plant SSCs. The only impact of this activity is the 
application of the current Tier 2 departure evaluation process to 
Tier 2* departures.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendment request: December 21, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17355A177.
    Description of amendment request: The proposed amendment 
establishes Conditions, Required Actions, and Completion Times in the 
Technical Specification (TS) 3.75 for the Condition where one steam 
supply to the turbine driven Auxiliary Feedwater (AFW) pump is 
inoperable concurrent with an inoperable motor driven AFW train. In 
addition, this amendment establishes changes to the TS, that establish 
specific Actions: (1) For when two motor driven AFW trains are 
inoperable at the same time and; (2) for when the turbine

[[Page 6235]]

driven AFW train is inoperable either (a) due solely to one inoperable 
steam supply, or (b) due to reasons other than one inoperable steam 
supply. The licensee stated that the change is consistent with NRC-
approved Technical Specification Task Force (TSTF) Traveler, TSTF-412, 
Revision 3, ``Provide Actions for One Steam Supply to Turbine Driven 
AFW/EFW Pump Inoperable.'' (ADAMS Accession No. ML070100363).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 10.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, by referencing the environmental evaluation included in 
the model safety evaluation published in the Federal Register on July 
17, 2007 (72 FR 39089), which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The Auxiliary/Emergency Feedwater (AFW/EFW) System is not an 
initiator of any design basis accident or event, and therefore the 
proposed changes do not increase the probability of any accident 
previously evaluated. The proposed changes to address the condition 
of one or two motor driven AFW/EFW trains inoperable and the turbine 
driven AFW/EFW train inoperable due to one steam supply inoperable 
do not change the response of the plant to any accidents.
    The proposed changes do 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 changes do 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 changes do 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 changes do 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, the changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not result in a change in the manner in 
which the AFW/EFW System provides plant protection. The AFW/EFW 
System will continue to supply water to the steam generators to 
remove decay heat and other residual heat by delivering at least the 
minimum required flow rate to the steam generators. There are no 
design changes associated with the proposed changes. The changes to 
the Conditions and Required Actions do not change any existing 
accident scenarios, nor create any new or different accident 
scenarios.
    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 changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by these changes. The proposed changes will not 
result in plant operation in a configuration outside the design 
basis.
    Therefore, it is concluded that 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, Inc., 40 Inverness Center 
Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Michael T. Markley.

U.S. Department of Transportation, Maritime Administration, Docket No. 
50-238, Nuclear Ship Savannah, Baltimore, Maryland

    Date of amendment request: October 31, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17307A036.
    Description of amendment request: The amendment would revise the 
license to remove a condition that prevents dismantling and disposing 
of the facility without prior approval of the Commission.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes are administrative and do not involve 
modification of any plant equipment or affect basic plant operation.
    The NSS's reactor is not operational and the level of 
radioactivity in the NSS has significantly decreased from the levels 
that existed when the 1976 Possession-only License was issued. No 
aspect of any of the proposed changes is an initiator of any 
accident previously evaluated. Consequently, the probability of an 
accident previously evaluated is not significantly increased.
    Therefore, the proposed changes do 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.
    Both of the proposed changes are administrative and do not 
involve physical alteration of plant equipment that was not 
previously allowed by Technical Specifications. These proposed 
changes do not change the method by which any safety-related system 
performs its function. As such, no new or different types of 
equipment will be installed, and the basic operation of installed 
equipment is unchanged. The methods governing plant operation and 
testing remain consistent with current safety analysis assumptions.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Both of the proposed changes are administrative in nature. No 
margins of safety exist that are relevant to the ship's defueled and 
partially dismantled reactor. As such, there are no changes being 
made to safety analysis assumptions, safety limits or safety system 
settings that would adversely affect plant safety as a result of the 
proposed changes. The proposed changes involve revising the language 
of the license to clearly state previously approved changes, and to 
delete archaic requirements.
    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.

[[Page 6236]]

    Advisor for licensee: Erhard W. Koehler, U.S. Department of 
Transportation, Maritime Administration, 1200 New Jersey Ave. SE, 
Washington, DC 20590.
    NRC Branch Chief: Bruce A. Watson, CHP.

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: November 7, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17317A464.
    Description of amendment request: The amendments would revise the 
Surry Power Station (Surry), Units 1 and 2, Facility Operating License 
Numbers DPR-32 and DPR-37, respectively, in the form of new License 
Conditions, and Technical Specification (TS) 3.16, ``Emergency Power 
System,'' to allow a one-time extension of the Allowed Outage Time 
(AOT) in TS 3.16 Action B.2 from 7 days to 21 days. The requested 
temporary 21-day AOT is needed to replace Reserve Station Service 
Transformer C (RSST-C) and associated cabling during the Surry Unit 2 
fall 2018 refueling outage. The existing RSST-C is original plant 
equipment and is reaching the end of its dependable service life.
    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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed change adds a footnote to TS 3.16, ``Emergency 
Power System,'' to allow a one-time extension of the AOT in TS 3.16 
Action B.2 from 7 days to 21 days to facilitate the replacement of 
RSST-C and associated cabling.
    During the temporary 21-day AOT, the station emergency buses 
will continue to be fed from redundant, separate, reliable offsite 
sources that are capable of supporting the emergency loads under 
worst-case conditions considering a single failure.
    There are two (2) emergency buses for each unit: Buses 1H and 1J 
(Unit 1), and Buses 2H and 2J (Unit 2). While RSST-C is being 
replaced during the temporary 21-day AOT, Buses 1J and 2H will 
continue to be energized from a designated primary offsite source, 
System (Switchyard) Reserve Transformer (SRT) 4. Buses 1H and 2J 
will be energized from Main Step-up Transformer 2, which is the Unit 
2 designated dependable alternate source.
    In both configurations Transfer Bus F is fed through two, in 
series, transformers.
     The normal configuration feeds Transfer Bus F from the 
230 kV switchyard via two (2) transformers (SRT-2 and RSST-C) and 
two (2) breakers. The 230 kV switchyard is connected to ten (10) 
offsite circuits.
     The temporary 21-day AOT configuration feeds Transfer 
Bus F from the 500 kV switch yard via two (2) transformers (Main 
Step-up Transformer 2 and Station Service Transformer 2C) and three 
(3) breakers. The 500 kV switchyard is connected to 3 offsite 
circuits.
    A risk assessment has been performed for the temporary 21-day 
AOT configuration. The assessment concluded that the probability of 
a loss of offsite power for the proposed configuration is very low. 
Thus, the proposed change does not significantly increase the 
probability of an accident previously evaluated because: (a) The 
emergency buses continue to be feed from redundant, separate, 
reliable offsite sources and (b) the effect of the proposed 
configuration on the probability of a loss of offsite power is very 
low.
    There is no increase in the consequences of an accident because 
the emergency buses continue to be fed from redundant, separate, 
reliable offsite circuits and the onsite power sources (i.e., the 
Emergency Diesel Generators) are unaffected.
    The consequences of both a Loss of Offsite Power (LOOP) and a 
Station Blackout (SBO) have been evaluated in the UFSAR. There is no 
change in the station responses to a LOOP or an SBO as a result of 
the extended AOT because RSST-C is not included in designated 
equipment used in the LOOP and SBO coping strategies.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed configuration does not result in a change in the 
manner in which the electrical distribution subsystems downstream of 
RSST-C provide plant protection. During the temporary AOT (21 days 
total), the only change is to substitute the reliable Unit 2 
designated dependable alternate source for a primary offsite power 
source for Emergency Buses 1H and 2J. Other sources of offsite and 
onsite power are unaffected, and other aspects of the offsite and 
onsite power supplies are unchanged and unaffected.
    There are no changes to the other RSSTs or to the supporting 
systems operating characteristics or conditions.
    There is no change in the station responses to a LOOP or an SBO 
because RSST-C is not included in the designated equipment used in 
the LOOP and SSO coping strategies.
    Therefore, the proposed change does 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 TS change does not affect the acceptance criteria 
for any analyzed event, nor is there a change to any safety limit. 
The proposed TS change does not affect any structures, systems or 
components or their capability to perform their intended functions. 
The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. Neither the safety analyses nor the safety 
analysis acceptance criteria are affected by this change. The 
proposed change will not result in plant operation in a 
configuration outside the current design basis as the design basis 
includes use of the Unit 2 dependable alternate source. The proposed 
TS change allows use of the Unit 2 dependable alternate power source 
as the primary source for buses 1H and 2J for a period of up to 21 
days. The margin of safety is maintained by maintaining the 
capability to supply Emergency Buses 1H and 2J with a redundant, 
separate, reliable offsite power source, and maintaining the onsite 
power sources in their design basis configuration. Therefore, the 
proposed change does not involve a significant reduction in margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
    NRC Branch Chief: Michael T. Markley.

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

[[Page 6237]]

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.

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

    Date of amendment: July 1, 2016, as supplemented by letters dated 
June 2 and December 15, 2017.
    Description of amendment request: The amendments revised the 
Technical Specifications for PVNGS, Units 1, 2, and 3, to support the 
implementation of next generation fuel (NGF). In addition to the 
license amendment request, APS requested an exemption from certain 
requirements of 10 CFR 50.46, ``Acceptance criteria for emergency core 
cooling systems [ECCS] for light-water nuclear power reactors,'' and 10 
CFR part 50, Appendix K, ``ECCS Evaluation Models,'' to allow the use 
of Optimized ZIRLOTM as a fuel rod cladding material.
    The proposed change would allow for the implementation of NGF 
including the use of Optimized ZIRLOTM fuel rod cladding 
material. The NGF assemblies contain advanced features to enhance fuel 
reliability, thermal performance, and fuel cycle economics.
    Date of issuance: January 23, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 205 (Unit 1), 205 (Unit 2), and 205 (Unit 3). A 
publicly-available version is in ADAMS under Accession No. ML17319A107; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: 
The amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: October 4, 2016 (81 FR 
68469). The supplemental letters dated June 2 and December 15, 2017, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the NRC staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 23, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 4, 2017.
    Brief description of amendment: The amendment deletes a 
Surveillance Requirement Note associated with TS 3.5.1, ``ECCS 
[Emergency Core Cooling System]--Operating,'' TS 3.5.2, ``ECCS--
Shutdown,'' and TS 3.6.1.7, ``Residual Heat Removal (RHR) Containment 
Spray System,'' to more appropriately reflect the RHR system design, 
and ensure the RHR system operation is consistent with the technical 
specification (TS) Limiting Condition for Operation (LCO) requirements. 
The amendment also adds a Note in the LCO for TS 3.5.1, TS 3.5.2, TS 
3.6.1.7, TS 3.6.1.9, ``Feedwater Leakage Control System,'' and TS 
3.6.2.3, ``Residual Heat Removal (RHR) Suppression Pool Cooling,'' to 
clarify that one of the required subsystems in each of the affected TS 
sections listed above may be inoperable during alignment and operation 
of the RHR system for Shutdown Cooling (i.e., decay heat removal) with 
the reactor steam dome pressure less than the RHR cut in permissive 
value.
    Date of issuance: January 22, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No(s): 215. A publicly-available version is in ADAMS 
under Accession No. ML17324A354; documents related to this amendment 
are listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-62: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: July 5, 2017 (82 FR 
31095).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 22, 2018.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, 
Pennsylvania

    Date of amendment request: December 23, 2013, as supplemented by 
letters dated February 14, 2017; April 27, May 27, June 26, November 6, 
and December 21, 2015; February 24 and May 12, 2016; and January 30, 
April 21, June 23, August 22, October 25, and November 29, 2017.
    Brief description of amendments: The amendments revised the Beaver 
Valley, Unit Nos. 1 and 2, Renewed Facility Operating Licenses (RFOLs) 
to establish and maintain a risk-informed, performance-based fire 
protection program in accordance with the requirements of 10 CFR 
50.48(c).
    Date of issuance: January 22, 2018.
    Effective date: As of the date of issuance and shall be implemented 
consistent with paragraph 2.C.(5) for Unit No. 1, and paragraph 2.F for 
Unit No. 2, of the RFOLs.
    Amendment Nos.: 301 (Unit No. 1) and 190 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML17291A081; 
documents related to these amendments are listed in the safety 
evaluation enclosed with the amendments.
    RFOL Nos. DPR-66 and NPF-73: Amendments revised the RFOLs.
    Date of initial notice in Federal Register: September 9, 2014 (79 
FR 53458). The supplemental letters dated April 27, May 27, June 26, 
November 6, and December 21, 2015; February 24 and May 12, 2016; and 
January 30, April 21, June 23, August 22, October 25, and November 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 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 January 22, 2018.
    No significant hazards consideration comments received: No.

[[Page 6238]]

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: June 20, 2017.
    Brief description of amendment: The amendment revised technical 
specifications (TSs) to delete the list of diesel generator critical 
trips from TS Surveillance Requirement (SR) 3.8.1.13 and clarify that 
the purpose of the SR is to verify that the non-critical automatic 
trips are bypassed.
    Date of issuance: January 18, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 179. A publicly-available version is in ADAMS under 
Accession No. ML17325B690; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-58: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 15, 2017 (82 FR 
38718).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 18, 2018.
    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: June 9, 2017, as supplemented by letter 
dated September 21, 2017.
    Brief description of amendment: The amendment deleted Technical 
Specification (TS) 2.8.3(6), ``Spent Fuel Cask Loading,'' and 
associated Figure 2-11, ``Limiting Burnup Criteria for Acceptable 
Storage in Spent Fuel Cask''; TS 3.2, Table 3-5, item 24, ``Spent Fuel 
Cask Loading''; TS 4.3.1.3, Design Features associated with spent fuel 
casks; and portions of TS 3.2, Table 3-4, item 5, footnote (4) on boron 
concentration associated with cask loading.
    Date of issuance: January 19, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 296. A publicly-available version is in ADAMS under 
Accession No. ML17338A172; 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 renewed facility operating license and TSs.
    Date of initial notice in Federal Register: August 15, 2017 (82 FR 
38718).
    The supplemental letter dated September 21, 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 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 January 19, 2018.
    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: March 6, 2017, as supplemented by 
letters dated May 4, 2017, and September 14, 2017.
    Brief description of amendments: The amendments revised Technical 
Specification 3.6.2.3, ``Containment Cooling System,'' to extend the 
containment fan coil unit allowed outage time from 7 days to 14 days 
for one or two inoperable containment fan coil units.
    Date of issuance: January 18, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 321 (Unit 1) and 302 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17349A108; 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: June 6, 2017 (82 FR 
26136). The supplemental letter dated September 14, 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 18, 2018.
    No significant hazards consideration comment received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia

    Date of amendment request: April 7, 2017.
    Brief description of amendments: The amendment revises the 
requirements of Technical Specification (TS) 3.6.4.1, ``Secondary 
Containment,'' associated with Surveillance Requirement (SR) 3.6.4.1.2. 
Specifically, SR 3.6.4.1.2 verifies that one secondary containment 
access door in each access opening is closed. The amendments would 
allow for brief, inadvertent, simultaneous opening of redundant 
secondary containment access doors during normal entry and exit 
conditions.
    Date of issuance: January 22, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: Unit 1-289, Unit 2-234. A publicly-available 
version is in ADAMS under Accession No. ML17355A440; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: August 29, 2017 (82 FR 
41070).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 22, 2018.
    No significant hazards consideration comments received: No.

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: May 31, 2017, and supplemented by letter 
dated November 16, 2017.
    Description of amendment: The amendment authorizes changes to the 
VEGP Units 3 and 4 Updated Final Safety Analysis Report in the form of 
departures from the plant-specific Design Control Document Tier 2 
information and involves changes to the administrative controls for 
unborated water flow paths to the reactor coolant

[[Page 6239]]

system to support chemical additions during periods when the reactor 
coolant pumps are not in operation. These proposed changes are 
reflected in Appendix A, Technical Specifications.
    Date of issuance: January 9, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 105 (Unit 3) and 104 (Unit 4). A publicly-available 
version is in ADAMS under Accession No. ML17297A349; documents related 
to this amendment are listed in the Safety Evaluation enclosed with the 
amendment.
    Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment 
revised the Facility Combined License.
    Date of initial notice in Federal Register: September 12, 2017 (82 
FR 42853). The supplemental letter dated November 16, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application request 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 the Safety Evaluation dated January 9, 2018.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 15, 2016.
    Brief description of amendments: The amendments replace the SONGS, 
Units 1, 2, and 3 Permanently Defueled Technical Specifications (TS) 
with Independent Spent Fuel Storage Installation (ISFSI) Only TS. These 
changes reflect the removal of all spent nuclear fuel from the SONGS, 
Units 2 and 3, spent fuel pools and its transfer to dry cask storage 
within the onsite ISFSI. The changes also make conforming revisions to 
the SONGS, Unit 1, TS and combine them with the SONGS, Units 2 and 3, 
TS. These changes will more fully reflect the permanently shutdown 
status of the decommissioning facility, as well as the reduced scope of 
structures, systems, and components necessary to ensure plant safety 
once all spent fuel has been permanently moved to the SONGS ISFSI, an 
activity which is currently scheduled for completion in 2019.
    Date of issuance: January 9, 2017.
    Effective date: As of the date Southern California Edison submits a 
written notification to the NRC that all spent nuclear fuel assemblies 
have been transferred out of the SONGS spent fuel pools and placed in 
storage within the onsite independent spent fuel storage installation, 
and shall be implemented within 60 days.
    Amendment Nos.: Unit 1-169, Unit 2-237, and Unit 3-230: A publicly-
available version is in ADAMS under Accession No. ML17345A657; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Facility Operating License Nos. DPR-13, NPF-10, and NPF-15: The 
amendments revise the Facility Operating Licenses.
    Date of initial notice in Federal Register: February 14, 2017 (82 
FR 10600).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 9, 2017.
    No significant hazards consideration comments received: No.

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: January 25, 2017, as supplemented by 
letters dated March 21, 2017; August 4, 2017; and December 4, 2017.
    Brief description of amendments: The amendments revised certain 
surveillance requirements in Technical Specification 3.8.1, ``AC 
[Alternating Current] Sources--Operating.'' The changes are in the use 
of steady-state voltage and frequency acceptance criteria for onsite 
standby power source of the diesel generators, allowing for the use of 
new and more conservative design analysis.
    Date of issuance: January 22, 2018.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 269 (Unit 1) and 251 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17352A711; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: June 6, 2017 (82 FR 
26139). The supplemental letters dated August 4, 2017, and December 4, 
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 22, 2018.
    No significant hazards consideration comments received: No.

IV. Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined Licenses and Final Determination of No Significant Hazards 
Consideration and Opportunity for a Hearing (Exigent Public 
Announcement or Emergency Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual notice of 
consideration of issuance of amendment, proposed no significant hazards 
consideration determination, and opportunity for a hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the

[[Page 6240]]

plant's licensed power level, the Commission may not have had an 
opportunity to provide for public comment on its no significant hazards 
consideration determination. In such case, the license amendment has 
been issued without opportunity for comment. If there has been some 
time for public comment but less than 30 days, the Commission may 
provide an opportunity for public comment. If comments have been 
requested, it is so stated. In either event, the State has been 
consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 51.12(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 
application for amendment, (2) the amendment to Facility Operating 
License or Combined License, as applicable, 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.

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

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

[[Page 6241]]

recognized Indian Tribe, or agency thereof does not need to address the 
standing requirements in 10 CFR 2.309(d) if the facility is located 
within its boundaries. Alternatively, a State, local governmental body, 
Federally-recognized Indian Tribe, or agency thereof may participate as 
a non-party under 10 CFR 2.315(c).
    If a hearing is granted, any person who is not a party to the 
proceeding and is not affiliated with or represented by a party may, at 
the discretion of the presiding officer, be permitted to make a limited 
appearance pursuant to the provisions of 10 CFR 2.315(a). A person 
making a limited appearance may make an oral or written statement of 
his or her position on the issues but may not otherwise participate in 
the proceeding. A limited appearance may be made at any session of the 
hearing or at any prehearing conference, subject to the limits and 
conditions as may be imposed by the presiding officer. Details 
regarding the opportunity to make a limited appearance will be provided 
by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

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

[[Page 6242]]

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

    Date of amendment request: December 28, 2017.
    Description of amendment: The amendment revised a note to Technical 
Specification Surveillance Requirement (SR) 4.1.3.1.2, such that 
Control Element Assembly (CEA) 4 may be excluded from the remaining 
quarterly performances of the SR in Cycle 26. The amendment allows the 
licensee to delay exercising CEA 4 until after repairs can be made 
during the next outage.
    Date of issuance: January 18, 2018.
    Effective date: As of the date of issuance and shall be implemented 
as soon as practicable and prior to the time in which SR 4.1.3.1.2 must 
be completed.
    Amendment No.: 308. A publicly-available version is in ADAMS under 
Accession No. ML18011A064; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Renewed Facility Operating License and Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. Public notice of the proposed amendment was 
published in the Arkansas Democrat-Gazette, located in Little Rock, 
Arkansas, from January 6 through January 7, 2018. The notice provided 
an opportunity to submit comments on the Commission's proposed NSHC 
determination. No comments were received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a safety evaluation dated January 18, 2018.
    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.

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

    Date of amendment request: January 10, 2018, as supplemented by 
letter dated January 17, 2018.
    Description of amendment: The amendment revised Technical 
Specification (TS) 3.3.4, ``Remote Shutdown Instrumentation,'' to make 
a one-time change to TS Table 3.3.4-1, Function 4a, ``RCS Hot Leg 
Temperature Indication,'' to permit the temperature indicator for the 
Reactor Coolant System Loop 3 hot leg to be inoperable for the 
remainder of WBN Unit 2 Operating Cycle 2, the refueling outage for 
which is scheduled to start in spring 2019. The amendment also added a 
condition to the operating license to require implementation of 
compensatory measures described in the application that will remain in 
effect until the temperature indicator is returned to an operable 
condition.
    Date of issuance: January 25, 2018.
    Effective date: As of date of issuance.
    Amendment No.: 19. A publicly-available version is in ADAMS under 
Accession No. ML18022B106; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-96: Amendment revised the 
technical specifications and operating license.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. The Rhea County Herald-News and The Advocate 
& Democrat on January 21, 2018, and The Daily Post-Athenian on January 
22 and January 23, 2018. The notice provided an opportunity to submit 
comments on the Commission's proposed NSHC determination. The 
supplemental letter dated January 17, 2018, 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 notice.
    No comments have been received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a Safety Evaluation dated January 25, 2018.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Branch Chief: Undine Shoop.

    Dated at Rockville, Maryland, this 6th day of February 2018.

    For the Nuclear Regulatory Commission.
Greg A. Casto,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2018-02636 Filed 2-12-18; 8:45 am]
 BILLING CODE 7590-01-P