Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 10590-10605 [2017-02795]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 82, Number 29 (Tuesday, February 14, 2017)]
[Notices]
[Pages 10590-10605]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-02795]


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NUCLEAR REGULATORY COMMISSION

[NRC-2017-0038]


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 14 to January 30, 2017. The last 
biweekly notice was published on January 31, 2017.

DATES: Comments must be filed by March 16, 2017. A request for a 
hearing must be filed by April 17, 2017.

ADDRESSES: You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0038. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

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

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2017-0038, facility name, unit 
number(s), plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0038.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the

[[Page 10591]]

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 pdr.resource@nrc.gov. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in this 
document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2017-0038, 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 posts all comment submissions at https://www.regulations.gov as well as entering 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 submissions 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 Web site 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

[[Page 10592]]

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 by April 
17, 2017. The petition must be filed in accordance with the filing 
instructions in the ``Electronic Submissions (E-Filing)'' section of 
this document, and should meet the requirements for petitions set forth 
in this section, except that under 10 CFR 2.309(h)(2) a State, local 
governmental body, or Federally-recognized Indian Tribe, or agency 
thereof does not need to address the standing requirements in 10 CFR 
2.309(d) if the facility is located within its boundaries. 
Alternatively, a State, local governmental body, Federally-recognized 
Indian Tribe, or agency thereof may participate as a non-party under 10 
CFR 2.315(c).
    If a hearing is granted, any person who is not a party to the 
proceeding and is not affiliated with or represented by a party may, at 
the discretion of the presiding officer, be permitted to make a limited 
appearance pursuant to the provisions of 10 CFR 2.315(a). A person 
making a limited appearance may make an oral or written statement of 
his or her position on the issues but may not otherwise participate in 
the proceeding. A limited appearance may be made at any session of the 
hearing or at any prehearing conference, subject to the limits and 
conditions as may be imposed by the presiding officer. Details 
regarding the opportunity to make a limited appearance will be provided 
by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing and petition for leave to intervene (petition), any 
motion or other document filed in the proceeding prior to the 
submission of a request for hearing or petition to intervene, and 
documents filed by interested governmental entities that request to 
participate under 10 CFR 2.315(c), must be filed in accordance with the 
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 
46562, August 3, 2012). The E-Filing process requires participants to 
submit and serve all adjudicatory documents over the internet, or in 
some cases to mail copies on electronic storage media. Detailed 
guidance on making electronic submissions may be found in the Guidance 
for Electronic Submissions to the NRC and on the NRC Web site at 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 hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to (1) request a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign submissions and access the E-Filing 
system for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a petition or 
other adjudicatory document (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at 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 Web site 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 Web site at https://www.nrc.gov/site-help/e-submittals.html, by 
email to MSHD.Resource@nrc.gov, 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

[[Page 10593]]

all other participants. Filing is considered complete by first-class 
mail as of the time of deposit in the mail, or by courier, express 
mail, or expedited delivery service upon depositing the document with 
the provider of the service. A presiding officer, having granted an 
exemption request from using E-Filing, may require a participant or 
party to use E-Filing if the presiding officer subsequently determines 
that the reason for granting the exemption from use of E-Filing no 
longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the 
Commission or the presiding officer. If you do not have an NRC-issued 
digital ID certificate as described above, click cancel when the link 
requests certificates and you will be automatically directed to the 
NRC's electronic hearing dockets where you will be able to access any 
publicly available documents in a particular hearing docket. 
Participants are requested not to include personal privacy information, 
such as social security numbers, home addresses, or personal phone 
numbers in their filings, unless an NRC regulation or other law 
requires submission of such information. For example, in some 
instances, individuals provide home addresses in order to demonstrate 
proximity to a facility or site. With respect to copyrighted works, 
except for limited excerpts that serve the purpose of the adjudicatory 
filings and would constitute a Fair Use application, participants are 
requested not to include copyrighted materials in their submission.
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

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

    Date of amendment request: July 20, 2016. Publicly-available 
version is in ADAMS under Accession No. ML16209A222.
    Description of amendment request: The proposed amendment requests 
to revise the Technical Specifications (TSs) associated with dry spent 
fuel storage cask loading and unloading requirements for ONS. 
Specifically, the license amendment request would revise TS 3.7.12, 
``Spent Fuel Pool Boron Concentration''; TS 3.7.18, ``Dry Spent Fuel 
Storage Cask Loading and Unloading''; and TS 4.4, ``Dry Spent Fuel 
Storage Cask Loading and Unloading,'' to remove certain TS requirements 
that no longer pertain to the ONS Independent Spent Fuel Storage 
Facility general license, due to changes in 10 CFR 50.68, ``Criticality 
accident requirements.''
    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 edits in square 
brackets:

    1. Does the proposed change [amendment] involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed changes to Technical Specifications (TSs) 3.7.12, 
3.7.18 and 4.4, do not modify the method of nuclear fuel storage or 
handling at Oconee Nuclear Station (ONS), or make any physical 
changes to the facility design, material, or construction standards. 
The proposed change revises the criticality requirements contained 
in the TSs, as allowed by 10 CFR 50.68(c), that are redundant to 
regulatory requirements provided in 10 CFR part 72 and the Nuclear 
Regulatory Commission (NRC)-approved Certificate of Compliance (CoC) 
for the spent fuel dry shielded canisters utilized at ONS. The 
proposed change to the TS requirements neither result[s] in 
operation that will increase the probability of initiating an 
analyzed event nor alter[s] assumptions relative to mitigation of an 
accident or transient event. The change has no effect on the process 
variables, structures, systems, and components that must be 
maintained consistent with the safety analyses and licensing basis.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to TSs 3.7.12, 3.7.18 and 4.4, do not 
modify the method of nuclear fuel storage or handling at ONS, nor 
make any physical changes to the facility design, material, or 
construction standards. The change does not alter the plant 
configuration (no new or different type of equipment will be 
installed) or make changes in the methods governing normal plant 
operation. The proposed change to the ONS TS requirements does not 
adversely impact the results of the ONS safety analyses and is 
compliant with the current licensing basis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any kind of accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to TS 3.7.12, 3.7.18 and 4.4, do not modify 
the method of nuclear fuel storage or handling at ONS, nor make any 
physical changes to the facility design, material, or construction 
standards. The proposed changes comply with NRC approved regulations 
and the station's Part 72 and 50 licensing basis.
    Therefore, the proposed TS change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Vice President Nuclear & 
EHS Legal Support, Duke Energy Corporation, 526 South Church Street--
EC07H, Charlotte, NC 28202-1802.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: July 21, 2016. Publicly-available 
version is in ADAMS under Accession No. ML16209A223.
    Description of amendment request: The proposed amendment requests 
to revise the Technical Specifications (TSs) for ONS. Specifically, the 
license amendment request would revise TS 2.1.1.1, ``Reactor Core 
Safety Limits (SLs),'' and TS 5.6.5, ``Core Operating Limits Report 
(COLR),'' to allow the use of the COPERNIC fuel performance code.
    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 edits in square 
brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change adds a limit on maximum local fuel pin 
centerline temperature to [the] ONS Technical Specifications that is 
based on a[n] NRC reviewed and approved fuel performance

[[Page 10594]]

code, and does not require a physical change to plant systems, 
structures or components. Plant operations and analysis will 
continue to be in accordance with the ONS licensing basis. The peak 
fuel centerline temperature is the basis for protecting the fuel and 
is consistent with the safety analysis.
    The proposed change also adds a topical report for a[n] NRC 
reviewed and approved fuel performance code to the list of topical 
reports in [the] ONS Technical Specifications, which is 
administrative in nature and has no impact on a plant configuration 
or system performance relied upon to mitigate the consequences of an 
accident. The list of topical reports in the Technical 
Specifications used to develop the core operating limits does not 
impact either the initiation of an accident or the mitigation of its 
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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change adds a limit on maximum local fuel pin 
centerline temperature to [the] ONS Technical Specifications that is 
based on a[n] NRC reviewed and approved fuel performance code, and 
does not require a physical change to plant systems, structures or 
components. Specifying maximum local fuel pin centerline temperature 
ensures that the fuel design limits are met. Operations and analysis 
will continue to be in compliance with NRC regulations. The addition 
of a new fuel pin centerline melt temperature versus burnup 
relationship does not affect any accident initiators that would 
create a new accident.
    The proposed change also adds a topical report for a[n] NRC 
reviewed and approved fuel performance code to the list of topical 
reports in [the] ONS Technical Specifications, which is 
administrative in nature and has no impact on a plant configuration 
or on system performance. The proposed change updates the list of 
NRC-approved topical reports used to develop the core operating 
limits. There is no change to the parameters within which the plant 
is normally operated. The possibility of a new or different kind of 
accident is not created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to TS 2.1.1.1 adds a limit on maximum local 
fuel pin centerline temperature that is based on an NRC reviewed and 
approved fuel performance code, and does not require a physical 
change to plant systems, structures or components. Plant operations 
and analysis will continue to be in accordance with [the] ONS 
licensing basis.
    Adding the local fuel pin centerline temperature and burnup 
relationship defined by the NRC reviewed and approved fuel 
performance code to the ONS Technical Specifications, does not 
impact the safety margins established in the ONS licensing basis.
    The proposed change also adds a topical report for a[n] NRC 
reviewed and approved fuel performance code to the list of topical 
reports in [the] ONS Technical Specifications, which is 
administrative in nature and does not amend the cycle specific 
parameters presently required by the Technical Specifications. The 
individual Technical Specifications continue to require operation of 
the plant within the bounds of the limits specified in the Core 
Operating Limits Report. The proposed change to the list of 
analytical methods referenced in the Core Operating Limits Report 
does not impact the margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Vice President Nuclear & 
EHS Legal Support, Duke Energy Corporation, 526 South Church Street--
EC07H, Charlotte, NC 28202-1802.
    NRC Branch Chief: Michael T. Markley.

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

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

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

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

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

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

Duke Energy Carolinas, LLC, Docket Nos. 52-018 and 52-019, William 
States Lee III Nuclear Station, Units 1, and 2 (WLS), Cherokee County, 
South Carolina

    Date of amendment request: April 29, 2016, as supplemented by 
letters dated October 3, 2016, and January 16, 2017. Publicly-available 
versions are in ADAMS under Accession Nos. ML16120A076, ML16277A521, 
and ML17017A210, respectively.
    Description of amendment request: The NRC staff previously made a 
proposed determination that the amendment request dated April 29, 2016, 
involves no significant hazards considerations (81 FR 43650; July 5, 
2016). Subsequently, by letter dated January 16, 2017, the licensee 
provided additional information that expanded the scope of the 
amendment request as originally noticed. Accordingly, this notice 
supersedes the previous notice in its entirety.
    The amendments would (1) consolidate the Emergency Operations 
Facilities (EOFs) for BSEP, HNP, and RNP with the Duke Energy Progress, 
LLC (Duke Energy) corporate EOF in Charlotte, North Carolina; (2) 
decrease the frequency for a multi-site drill from once per 6 years to 
once per 8 years; (3) allow the multi-site drill performance with sites 
other than CNS, MNS, or ONS, (4) change the BSEP, HNP, and RNP 
augmentation times to be consistent with those of the sites currently 
supported by the Duke Energy corporate EOF, and (5) decrease the 
frequency of the unannounced augmentation drill at BSEP from twice per 
year to once per year. The January 16, 2017, letter also acknowledges 
the addition of WLS to the Duke Energy corporate EOF with the issuance 
of the WLS operating license on December 19, 2016 (ADAMS Accession No. 
ML16333A329).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes relocate the BSEP, HNP, and RNP EOFs from 
their present onsite or near-site locations to the established 
corporate EOF in Charlotte, North Carolina, changes the required 
response times for supplementing onsite personnel in response to a 
radiological emergency, decreases the multi-site drill frequency, 
allows the multi-site drill to be performed with sites other than 
ONS, MNS, or CNS, and decreases the frequency of augmentation drills 
at BSEP. The functions and capabilities of the relocated EOFs will 
continue to meet the applicable regulatory requirements. It has

[[Page 10595]]

been evaluated and determined that the change in response time does 
not significantly affect the ability to supplement the onsite staff. 
In addition, analysis shows that the onsite staff can acceptably 
respond to an event for longer than the requested time for augmented 
staff to arrive. The proposed changes have no effect on normal plant 
operation or on any accident initiator or precursors, and do not 
impact the function of plant structures, systems, or components 
(SSCs). The proposed changes do not alter or prevent the ability of 
the emergency response organization to perform its intended 
functions to mitigate the consequences of an accident or 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 change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes only impact the implementation of the 
affected stations' emergency plans by relocating their onsite or 
near-site EOFs to the established corporate EOF in Charlotte, North 
Carolina, changing the required response time of responders who 
supplement the onsite staff, decreasing the multi-site drill 
frequency, allowing the multi-site drill to be performed with sites 
other than ONS, MNS, or CNS, and decreasing the frequency of 
augmentation drills at BSEP. The functions and capabilities of the 
relocated EOFs will continue to meet the applicable regulatory 
requirements. It has been evaluated and determined that the change 
in response time does not significantly affect the ability to 
supplement the onsite staff. In addition, analysis shows that the 
onsite staff can acceptably respond to an event for longer than the 
requested time for augmented staff to arrive. The proposed changes 
will not change the design function or operation of SSCs. The 
changes do not impact the accident analysis. The changes do not 
involve a physical alteration of the plant, a change in the method 
of plant operation, or new operator actions. The proposed changes do 
not introduce failure modes that could result in a new accident, and 
the changes do not alter assumptions made in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes only impacts the implementation of the 
affected stations' emergency plans by relocating their onsite or 
near-site EOFs to the established corporate EOF in Charlotte, North 
Carolina, changing the required response time of responders who 
supplement the onsite staff, decreasing the multi-site drill 
frequency, allowing the multi-site drill to be performed with sites 
other than ONS, MNS, or CNS, and decreasing the frequency of 
augmentation drills at BSEP. The functions and capabilities of the 
relocated EOFs will continue to meet the applicable regulatory 
requirements. It has been evaluated and determined that the change 
in response time does not significantly affect the ability to 
supplement the onsite staff. In addition, analysis shows that the 
onsite staff can acceptably respond to an event for longer than the 
requested time for augmented staff to arrive. Margin of safety is 
associated with confidence in the ability of the fission product 
barriers (i.e., fuel cladding, reactor coolant system pressure 
boundary, and containment structure) to limit the level of radiation 
dose to the public. The proposed changes are associated with the 
emergency plans and do not impact operation of the plant or its 
response to transients or accidents. The changes do not affect the 
Technical Specifications. The changes do not involve a change in the 
method of plant operation, and no accident analyses will be affected 
by the proposed changes. Safety analysis acceptance criteria are not 
affected. The emergency plans will continue to provide the necessary 
response staff for emergencies as demonstrated by staffing and 
functional analyses including the necessary timeliness of performing 
major tasks for the functional areas of the emergency plans.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, 
Charlotte, NC 28202.
    NRC Branch Chief: Benjamin G. Beasley.

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

    Date of amendment request: December 2, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16337A249.
    Description of amendment request: The amendment would revise HNP 
Technical Specifications (TSs) to relocate selected figures and values 
from the TSs to the Core Operating Limits Report (COLR), remove all 
references to a specific plant procedure as it pertains to the COLR, 
and adopt Technical Specification Task Force (TSTF)-5, ``Delete Safety 
Limit Violation Notification Requirements,'' Revision 1, which deletes 
duplicate notification, reporting and restart requirements from the 
Administrative Controls section of TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    [Response: No.]
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes are administrative in nature, facilitate 
improved content and presentation of Administrative controls, and 
alter only the format and location of cycle-specific parameter 
figures and limits from the TS to the COLR. This relocation does not 
result in the alteration of the design, material, or construction 
standards that were applicable prior to the change. The proposed 
changes will not result in modification of any system interface that 
would increase the likelihood of an accident since these events are 
independent of the proposed change. The proposed amendment will not 
change, degrade, or prevent actions, or alter any assumptions 
previously made in evaluating the radiological consequences of an 
accident described in the Final Safety Analysis Report (FSAR).
    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 previously evaluated?
    [Response: No.]
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed changes do not involve any change to the configuration 
or method of operation of any plant equipment. Accordingly, no new 
failure modes have been defined for any plant system or component 
important to safety nor has any new limiting single failure been 
identified as a result of the proposed changes. Also, there will be 
no change in types or increase in the amounts of any effluents 
released offsite.
    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 proposed changes do not involve a significant reduction in a 
margin of safety. Previously-approved methodologies will continue to 
be used in determination of cycle-specific core operating limits 
that are present in the COLR. The proposed changes are 
administrative in nature and will not affect the plant design or 
system operating parameters. As such, there is no detrimental impact 
on any equipment design parameter and the plant will continue to be 
operated within prescribed limits.

[[Page 10596]]

    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: November 8, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16313A573.
    Description of amendment request: The proposed amendment would, on 
a one-time basis, extend the Completion Time by 7 days for Technical 
Specification Conditions 3.5.1.A, 3.6.1.5.A, and 3.6.2.3.A. This 
onetime extension will be used to support preventive maintenance, which 
replaces the residual heat removal train A subsystem's pump and motor.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not increase the probability of an 
accident because the residual heat removal (RHR) system cannot 
initiate an accident. The RHR system provides coolant injection to 
the reactor core, cooling of the suppression pool water inventory, 
and drywell sprays following a design basis accident.
    The proposed one time completion time (CT) change for RHR train 
A does not alter the conditions, operating configurations, or 
minimum amount of operating equipment assumed in the safety analysis 
for accident mitigation. No changes are proposed in the manner in 
which the emergency core cooling system (ECCS) provides plant 
protection or which create new modes of plant operation. In 
addition, a probabilistic safety assessment (PSA) evaluation 
concluded that the risk contribution of the increased CT is a very 
small increase in risk. The proposed change in CT will not affect 
the probability of any event initiators. There will be no 
degradation in the performance of, or an increase in the number of 
challenges imposed on, safety related equipment assumed to function 
during an accident situation. There will be no change to normal 
plant operating parameters or accident mitigation performance.
    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 previously evaluated?
    Response: No.
    The proposed amendment will not create the possibility of a new 
or different kind of accident because inoperability of one RHR 
subsystem is not an accident precursor. There are no hardware 
changes nor are there any changes in the method by which any plant 
system performs a safety function. This request does not affect the 
normal method of plant operation. The proposed amendment does not 
introduce new equipment, or new way of operation of the system which 
could create a new or different kind of accident. No new external 
threats, release pathways, or equipment failure modes are created. 
No new accident scenarios, transient precursors, failure mechanisms, 
or limiting single failures are introduced as a result of this 
request.
    Therefore, the implementation of the proposed amendment will not 
create a possibility for an accident of a new or different type than 
those previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Columbia's ECCS is designed with sufficient redundancy such that 
a low pressure ECCS subsystem may be removed from service for 
maintenance or testing. The remaining subsystems are capable of 
providing water and removing heat loads to satisfy the final safety 
analysis report (FSAR) requirements for accident mitigation or plant 
shutdown. A PSA evaluation concluded that the risk contribution of 
the CT extension is very small. There will be no change to the 
manner in which safety limits or limiting safety system settings are 
determined nor will there be any change to those plant systems 
necessary to assure the accomplishment of protection functions. 
There will be no change to post-LOCA peak clad temperatures.
    For these reasons, 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: William A. Horin, Esq., Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: Robert J. Pascarelli.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan

    Date of amendment request: November 9, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16314A027.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.5.10, ``Ventilation Filter 
Testing Program,'' to correct and modify the description of the control 
room ventilation and fuel handling area ventilation systems. In 
addition, the proposed amendment would correct an editorial omission in 
TS Limiting Condition for Operation 3.0.9.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the Palisades Nuclear Plant (PNP) 
Technical Specifications (TS) are editorial or administrative in 
nature. The changes make an editorial correction in the TS, and 
correct and modify the component descriptions in the ventilation 
filter testing program TS. These changes do not alter accident 
analysis assumptions, add any initiators, or affect the function of 
plant systems or the manner in which systems are operated, 
maintained, modified, tested, or inspected. The proposed changes do 
not require any plant modifications which affect the performance 
capability of the structures, systems, and components relied upon to 
mitigate the consequences of postulated accidents, and have no 
impact on the probability or consequences of an accident previously 
evaluated.
    Therefore, the proposed 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 to the PNP TS are editorial or 
administrative in nature. The changes make an editorial correction 
in the TS, and correct and modify the component descriptions within 
the ventilation filter testing program TS. The proposed changes do 
not alter accident analysis assumptions, add any initiators, or 
affect the function of plant systems or the manner in which systems 
are operated, maintained, modified, tested, or inspected. The 
proposed changes do not require any plant modifications which affect 
the performance capability of the structures, systems, and 
components relied upon to mitigate the consequences of postulated

[[Page 10597]]

accidents, and do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Plant safety margins are established through limiting conditions 
for operation, limiting safety system settings, and safety limits 
specified in the technical specifications. The proposed changes to 
the TS are editorial or administrative in nature and do not impact 
any safety margins. Because there is no impact on established safety 
margins as a result of these changes, the proposed change does not 
involve a significant reduction in a margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeanne Cho, Senior Counsel, Entergy 
Services, Inc., 440 Hamilton Ave., White Plains, NY 10601.
    NRC Branch Chief: David J. Wrona.

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

    Date of amendment request: October 26, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16300A200.
    Description of amendment request: The proposed change revises TS 
5.5.13, ``Primary Containment Leakage Rate Testing Program,'' to allow 
for the permanent extension of the Type A Integrated Leak Rate Testing 
(ILRT) and Type C Leak Rate Testing frequencies, to change the 
documents used by LSCS to implement the performance-based leakage 
testing program, and to delete the information regarding the 
performance of the next LSCS Type A tests to be performed.
    Additionally, this license amendment request (LAR) proposes to 
delete Condition 2.D.(e) of the LSCS Unit 1 Renewed Facility Operating 
License regarding conducting the third Type A Test of each 10-year 
service period when the plant is shutdown for the 10-year plant 
inservice inspection (ISI). Similarly, this LAR proposes to delete 
Condition 2.D.(c) of the LSCS Unit 2 Renewed Facility Operating License 
regarding conducting the third Type A test of each 10-year service 
period when the plant is shutdown for the 10 year plant ISI.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment to the TS involves the extension of the 
LSCS Type A containment test interval to 15 years and the extension 
of the Type C test interval to 75 months. The current Type A test 
interval of 120 months (10 years) would be extended on a permanent 
basis to no longer than 15 years from the last Type A test. The 
current Type C test interval of 60 months for selected components 
would be extended on a performance basis to no longer than 75 
months. Extensions of up to nine months (total maximum interval of 
84 months for Type C tests) are permissible only for non-routine 
emergent conditions.
    The proposed extension does not involve either a physical change 
to the plant or a change in the manner in which the plant is 
operated or controlled. The containment is designed to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. As such, 
the containment and the testing requirements invoked to periodically 
demonstrate the integrity of the containment exist to ensure the 
plant's ability to mitigate the consequences of an accident, and do 
not involve the prevention or identification of any precursors of an 
accident. The change in dose risk for changing the Type A test 
frequency from three-per-ten years to once-per-fifteen years, 
measured as an increase to the total integrated dose risk for all 
internal events accident sequences for LSCS, is 1.23E-02 person-rem/
yr (0.33%) using the EPRI [Electric Power Research Institute] 
guidance with the base case corrosion included. The change in dose 
risk drops to 3.15E-03 person-rem/yr (0.08%) when using the EPRI 
Expert Elicitation methodology. The values calculated per the EPRI 
guidance are all lower than the acceptance criteria of <=1.0 person-
rem/yr or <1.0% person-rem/yr defined in Section 1.3 of Attachment 3 
of this submittal. The results of the risk assessment for this 
amendment meet these criteria. Moreover, the risk impact for the 
ILRT extension when compared to other severe accident risks is 
negligible. Therefore, this proposed extension does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    As documented in NUREG-1493, Type B and C tests have identified 
a very large percentage of containment leakage paths, and the 
percentage of containment leakage paths that are detected only by 
Type A testing is very small. The LSCS Type A test history supports 
this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as: (1) Activity based 
and (2) time based. Activity based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment combined with the containment inspections performed in 
accordance with ASME [American Society of Mechanical Engineers] 
Section XI and TS requirements serve to provide a high degree of 
assurance that the containment would not degrade in a manner that is 
detectable only by a Type A test. Based on the above, the proposed 
extensions do not significantly increase the consequences of an 
accident previously evaluated.
    The proposed amendment also deletes exceptions previously 
granted to allow one-time extensions of the ILRT test frequency for 
LSCS. These exceptions were for activities that would have already 
taken place by the time this amendment is approved; therefore, their 
deletion is solely an administrative action that has no effect on 
any component and no impact on how the unit is operated.
    Therefore, the proposed change does not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment to the TS involves the extension of the 
LSCS Type A containment test interval to 15 years and the extension 
of the Type C test interval to 75 months. The containment and the 
testing requirements to periodically demonstrate the integrity of 
the containment exist to ensure the plant's ability to mitigate the 
consequences of an accident do not involve any accident precursors 
or initiators. The proposed change does not involve a physical 
change to the plant (i.e., no new or different type of equipment 
will be installed) or a change to the manner in which the plant is 
operated or controlled.
    The proposed amendment also deletes exceptions previously 
granted to allow one-time extensions of the ILRT test frequency for 
LSCS. These exceptions were for activities that would have already 
taken place by the time this amendment is approved; therefore, their 
deletion is solely an administrative action that does not result in 
any change in how the unit is operated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment to TS 5.5.13 involves the extension of 
the LSCS Type A

[[Page 10598]]

containment test interval to 15 years and the extension of the Type 
C test interval to 75 months for selected components. This amendment 
does not alter the manner in which safety limits, limiting safety 
system set points, or limiting conditions for operation are 
determined. The specific requirements and conditions of the TS 
Containment Leak Rate Testing Program exist to ensure that the 
degree of containment structural integrity and leak-tightness that 
is considered in the plant safety analysis is maintained. The 
overall containment leak rate limit specified by TS is maintained.
    The proposed change involves only the extension of the interval 
between Type A containment leak rate tests and Type C tests for 
LSCS. The proposed surveillance interval extension is bounded by the 
15-year ILRT Interval and the 75-month Type C test interval 
currently authorized within NEI [Nuclear Energy Institute] 94-01, 
Revision 3-A. Industry experience supports the conclusion that Type 
B and C testing detects a large percentage of containment leakage 
paths and that the percentage of containment leakage paths that are 
detected only by Type A testing is small. The containment 
inspections performed in accordance with ASME Section XI and TS 
serve to provide a high degree of assurance that the containment 
would not degrade in a manner that is detectable only by Type A 
testing. The combination of these factors ensures that the margin of 
safety in the plant safety analysis is maintained. The design, 
operation, testing methods and acceptance criteria for Type A, B, 
and C containment leakage tests specified in applicable codes and 
standards would continue to be met, with the acceptance of this 
proposed change, since these are not affected by changes to the Type 
A and Type C test intervals.
    The proposed amendment also deletes exceptions previously 
granted to allow one time extensions of the ILRT test frequency for 
LSCS. These exceptions were for activities that would have already 
taken place by the time this amendment is approved; therefore, their 
deletion is solely an administrative action and does not change how 
the unit is operated and maintained. Thus, there is no reduction in 
any margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

South Carolina Electric & Gas Company (SCE&G) and South Carolina Public 
Service Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer 
Nuclear Station (VCSNS) Units 2 and 3, Fairfield County, South Carolina

    Date of amendment request: November 10, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16316A003.
    Description of amendment request: The amendment request proposes to 
add to License Condition 2.D.(1) of the VCSNS Units 2 and 3 combined 
licenses (COLs), an Interim Amendment Request process for changes 
during construction when emergent conditions are present.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff edits in square 
brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would add an Interim Amendment Request 
process to Condition 2.D.(1) of the VCSNS 2 and 3 COLs to allow 
construction to continue, at SCE&G's own risk, in emergent 
conditions, where a non-conforming condition that has little or no 
safety significance is discovered and the work activity cannot be 
adjusted. The Interim Amendment Request process would require SCE&G 
to submit a Nuclear Construction Safety Assessment which (1) 
identifies the proposed change; (2) evaluates whether emergent 
conditions are present; (3) evaluates whether the change would 
result in any material decrease in safety; and (4) evaluates whether 
continued construction would make the non-conforming condition 
irreversible. Only if the continued construction would have no 
material decrease in safety would the NRC issue a determination that 
construction could continue pending SCE&G's initiation of the COL-
ISG-025 PAR [preliminary amendment request]/LAR [license amendment 
request] process. The requirement to include a Nuclear Construction 
Safety Assessment ensures that the proposed amendment would not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated. If the continued construction 
would result a material decrease in safety, then continued 
construction would not be authorized.
    The proposed amendment does not modify the design, construction, 
or operation of any plant structures, systems, or components (SSCs), 
nor does it change any procedures or method of control for any SSCs. 
Because the proposed amendment does not change the design, 
construction, or operation of any SSCs, it does not adversely affect 
any design function as described in the Updated Final Safety 
Analysis Report.
    The proposed amendment does not affect the probability of an 
accident previously evaluated. Similarly, because the proposed 
amendment does not alter the design or operation of the nuclear 
plant or any plant SSCs, the proposed amendment does not represent a 
change to the radiological effects of an accident, and therefore, 
does not involve an increase in the consequences of an accident 
previously evaluated.
    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 amendment would add an Interim Amendment Request 
process to Condition 2.D.(1) of the VCSNS 2 and 3 COLs to allow 
construction to continue, at SCE&G's own risk, in emergent 
conditions, where a non-conforming condition that has little or no 
safety significance is discovered and the work activity cannot be 
adjusted. The Interim Amendment Request process would require SCE&G 
to submit a Nuclear Construction Safety Assessment which (1) 
identifies the proposed change; (2) evaluates whether emergent 
conditions are present; (3) evaluates whether the change would 
result in any material decrease in safety; and (4) evaluates whether 
continued construction would make the non-conforming condition 
irreversible. Only if the continued construction would have no 
material decrease in safety would NRC issue a determination that 
construction could continue pending SCE&G's initiation of the COL-
ISG-025 PAR/LAR process.
    The proposed amendment 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 proposed amendment only adds a new screening 
process and does not change the design, construction, or operation 
of the nuclear plant or any plant operations.
    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 amendment would add an Interim Amendment Request 
process to Condition 2.D.(1) of the VCSNS 2 and 3 COLs to allow 
construction to continue, at SCE&G's own risk, in emergent 
conditions, where a non-conforming condition that has little or no 
safety significance is discovered and the work activity cannot be 
adjusted. The Interim Amendment Request process would require SCE&G 
to submit a Nuclear Construction Safety Assessment which (1) 
identifies the proposed change; (2) evaluates whether emergent 
conditions are present; (3) evaluates whether the change would 
result in any material decrease in safety; and (4) evaluates whether 
continued construction would make the non-conforming condition 
irreversible. Only if the continued

[[Page 10599]]

construction would have no material decrease in safety would the NRC 
issue determination that construction could continue pending SCE&G's 
initiation of the COL-ISG-025 PAR/LAR process.
    The proposed amendment 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 proposed amendment does not alter any design 
function or safety analysis. Consequently, no safety analysis or 
design basis acceptance limit/criterion is challenged or exceeded by 
the proposed amendment, thus the margin of safety is not reduced. 
The only impact of this activity is the addition of an Interim 
Amendment Request process.
    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.

South Carolina Electric & Gas Company and South Carolina Public Service 
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear 
Station, Units 2 and 3, Fairfield, South Carolina

    Date of amendment request: September 15, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16259A315.
    Description of amendment request: The amendment request proposes 
changes to the Updated Final Safety Analysis Report (UFSAR) in the form 
of departures from the incorporated plant-specific Design Control 
Document Tier 2 information and a combined license (COL) License 
Condition which references one of the proposed changes. Additionally, 
the proposed changes to the UFSAR eliminate pressurizer spray line 
monitoring during pressurizer surge line first plant only testing. In 
addition, these proposed changes correct inconsistencies in testing 
purpose, testing duration, and the ability to leave equipment in place 
following the data collection period. These changes involve material 
which is specifically referenced in Section 2.D.(2) of the COL. This 
submittal requests approval of the license amendment necessary to 
implement these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff edits in square 
brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The design functions of the RCS [reactor coolant system] include 
providing an effective reactor coolant pressure boundary. The 
proposed changes for removing the requirement to install temporary 
instrumentation on the pressurizer spray line during the monitoring 
of the pressurizer surge line for thermal stratification and thermal 
cycling during hot functional testing and during the first fuel 
cycle for the first plant only, proposed changes to parameter 
retention requirements, and proposed change to remove the 
pressurizer spray and surge line valve leakage requirement do not 
impact the existing design requirements for the RCS. These changes 
are acceptable as they are consistent with the commitments made for 
the pressurizer surge line monitoring program for the first plant 
only, and do not adversely affect the capability of the pressurizer 
surge line and pressurizer spray lines to perform the required 
reactor coolant pressure boundary design functions.
    These proposed changes to the monitoring of the pressurizer 
surge line for thermal stratification and thermal cycling during hot 
functional testing and during the first fuel cycle for the first 
plant only, proposed changes to parameter retention requirements, 
and proposed change to remove the pressurizer spray and surge line 
valve leakage requirement as described in the current licensing 
basis do not have an adverse effect on any of the design functions 
of the systems. The proposed changes do not affect the support, 
design, or operation of mechanical and fluid systems required to 
mitigate the consequences of an accident. There is no change to 
plant systems or the response of systems to postulated accident 
conditions. There is no change to the predicted radioactive releases 
due to postulated accident conditions. The plant response to 
previously evaluated accidents or external events is not adversely 
affected, nor do the proposed changes create any new accident 
precursors.
    Therefore, the 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 for removing the requirement to install 
temporary instrumentation on the pressurizer spray line during the 
monitoring of the pressurizer surge line for thermal stratification 
and thermal cycling during hot functional testing and during the 
first fuel cycle for the first plant only, proposed changes to 
parameter retention requirements, and proposed change to remove the 
pressurizer spray and surge line valve leakage requirement as 
described in the current licensing basis maintain the required 
design functions, and are consistent with other Updated Final Safety 
Analysis Report (UFSAR) information. The proposed changes do not 
adversely affect the design requirements for the RCS, including the 
pressurizer surge line and pressurizer spray lines. The proposed 
changes do not adversely affect the design function, support, 
design, or operation of mechanical and fluid systems. The proposed 
changes do not result in a new failure mechanism or introduce any 
new accident precursors. No design function described in the UFSAR 
is adversely affected by the proposed changes.
    Therefore, the requested amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    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: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius, LLC, 1111 Pennsylvania Ave. NW., Washington, DC 20004-2514.
    NRC Branch Chief: Jennifer Dixon-Herrity.

South Carolina Electric & Gas Company and South Carolina Public Service 
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear 
Station, Units 2 and 3, Fairfield, South Carolina

    Date of amendment request: November 16, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16323A020.
    Description of amendment request: The amendment request proposes 
changes to plant-specific Tier 1 information, with corresponding 
changes to the associated Combined License (COL) Appendix C 
information, and involves associated Tier 2 information in the Updated 
Final Safety Analysis Report (UFSAR). Specifically, the requested 
amendment proposes clarifications to a plant-specific Tier 1 (and COL 
Appendix C) table and a UFSAR table in regard to the inspections of the 
excore source, intermediate, and power range detectors. Pursuant to the 
provisions of 10 CFR 52.63(b)(1), an exemption from

[[Page 10600]]

elements of the design as certified in the 10 CFR part 52, appendix D, 
design certification rule is also requested for the plant-specific 
Design Control Document Tier 1 material departures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff's edits in 
square brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to specify the inspection of the excore 
source, intermediate, and power range detectors is done to verify 
that aluminum surfaces are contained in stainless steel or titanium, 
and avoids the introduction of aluminum into the post-loss of 
coolant accident (LOCA) containment environment due to detector 
materials. The proposed change does not alter any safety related 
functions. The materials of construction are compatible with the 
post-LOCA conditions inside containment and will not significantly 
contribute to hydrogen generation or chemical precipitates. The 
change does not 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 change does not impact the support, design, or operation of 
mechanical and fluid systems. There is no change to plant systems or 
the response of systems to postulated accident conditions. There is 
no change to the predicted radioactive releases due to normal 
operation or postulated accident conditions. Consequently, the plant 
response to previously evaluated accidents or external events is not 
adversely affected, nor does the proposed change create any new 
accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does 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 change to specify the 
inspection of the excore source, intermediate, and power range 
detectors is done to verify that aluminum surfaces are contained in 
stainless steel or titanium, and avoids the introduction of aluminum 
into the post-LOCA containment environment due to detector 
materials. In addition, the proposed change to the ITAAC 
[inspections, tests, analysis, and acceptance criteria] verified 
materials of construction does not alter the design function of the 
excore detectors. The detector canning materials of construction are 
compatible with the post-LOCA containment environment and do not 
contribute a significant amount of hydrogen or chemical 
precipitates. The change to the ITAAC aligns the inspection with the 
Tier 2 design feature. Consequently, because the excore detectors 
functions are unchanged, there are no adverse effects on accidents 
previously evaluated in the UFSAR.
    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 change to specify the inspection of the excore 
source, intermediate, and power range detectors is done to verify 
that aluminum surfaces are contained in stainless steel or titanium, 
and avoids the introduction of aluminum into the post-LOCA 
containment environment, does not alter any safety-related 
equipment, applicable design codes, code compliance, design 
function, or safety analysis. Consequently, no safety analysis or 
design basis acceptance limit/criterion is challenged or exceeded by 
the proposed change, thus the margin of safety is not reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius, LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Jennifer Dixon-Herrity.

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. A publicly-available 
version is in ADAMS under Accession No. ML16355A014.
    Description of amendment request: The proposed amendment would 
revise the Facility Operating Licenses and associated Technical 
Specifications (TS) for SONGS, Units 1, 2, and 3, to reflect removal of 
all spent nuclear fuel from the SONGS, Units 2 and 3 spent fuel pools 
(SFPs) and its transfer to dry cask storage within an onsite 
independent spent fuel storage installation (ISFSI). The proposed 
changes would also make conforming changes 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 and defueled status of the 
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.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would modify the SONGS, Units 2 and 3 
facility operating licenses and TS by deleting the portions of the 
licenses and TSs that are no longer applicable to a facility with no 
spent nuclear fuel stored in the SFP, while modifying the remaining 
portions to correspond to all nuclear fuel stored within an ISFSI. 
This amendment becomes effective upon removal of all spent nuclear 
fuel from the SONGS, Units 2 and 3 SFP and its transfer to dry cask 
storage within an ISFSI.
    Additionally, the proposed change would revise the Unit 1 TSs 
for consistency with the proposed changes to the Units 2 and 3 TSs. 
Similar to the changes for Units 2 and 3, the Unit 1 changes delete 
portions of the TSs that are no longer applicable to a facility with 
spent fuel no longer stored in the SFP, while modifying the 
remaining portions to correspond to all nuclear fuel in dry storage. 
The Unit 1 TSs are also proposed to be combined with the Units 2 and 
3 TSs.
    The definition of safety-related Structures, Systems, and 
Components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are 
those relied on to remain functional during and following design 
basis events to assure:
    1. The integrity of the reactor coolant boundary;
    2. The capability to shutdown the reactor and maintain it in a 
safe shutdown condition; or
    3. The capability to prevent or mitigate the consequences of 
accidents which could result in potential offsite exposures 
comparable to the applicable guideline exposures set forth in 10 CFR 
50.43(a)(1) or 100.11.
    The first two criteria (integrity of the reactor coolant 
pressure boundary and safe shutdown of the reactor) are not 
applicable to a plant in a permanently defueled condition. The third 
criterion is related to preventing or mitigating the consequences of 
accidents that could result in potential offsite exposures exceeding 
limits. However, after all nuclear spent fuel assemblies have been 
transferred to dry cask storage within an ISFSI, none of the SSCs at 
SONGS, Units 2 and 3 are required to be relied on for accident 
mitigation. Therefore, none of the SSCs at

[[Page 10601]]

SONGS, Units 2 and 3 meet the definition of a safety-related SSC 
stated in 10 CFR 50.2. The proposed deletion of requirements in the 
TSs is not related to any systems credited in an accident analysis 
at SONGS, Units 2 and 3.
    Chapter 15 of the SONGS, Units 2 and 3 Updated Final Safety 
Analysis Report (UFSAR) described the design basis accidents (DBAs) 
related to the SFP. The majority of these postulated accidents are 
predicated on spent fuel being stored in the SFP. With the removal 
of the spent fuel from the SFP, there are no remaining spent fuel 
assemblies to be monitored and there are no credible accidents that 
require the actions of a Certified Fuel Handler, Shift Manager, or a 
Certified Operator to prevent occurrence or mitigate the 
consequences of an accident.
    With all of the SONGS 1 operating plant above-ground structures 
having been demolished and removed, and all Unit 1 spent fuel having 
been removed from the SFP, there are no remaining design basis 
accidents or transients in Chapter 8 of the Unit 1 Defueled Safety 
Analysis Report (DSAR).
    The proposed changes do not have an adverse impact on the 
remaining decommissioning activities or any of their potential 
consequences.
    The proposed changes related to the relocation of certain 
administrative requirements do not affect operating procedures or 
administrative controls that have the function of preventing or 
mitigating any accidents applicable to the safe management of 
irradiated fuel or decommissioning of the facility.
    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 eliminate the operational requirements and 
certain design requirements associated with the storage of the spent 
fuel in the SFP, and relocate certain administrative controls to the 
Decommissioning Quality Assurance Program or Licensee Controlled 
Specifications (LCS).
    After the removal of the spent fuel from the Units 2 and 3 SFP 
and transfer to the ISFSI, there are no spent fuel assemblies that 
remain in a SFP on site. Coupled with a prohibition against storage 
of fuel in the Units 2 and 3 SFP (the Unit 1 SFP has been 
dismantled), the potential for fuel related accidents is removed. 
The proposed changes do not introduce any new failure modes.
    Therefore, the proposed amendment 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 removal of all spent nuclear fuel from the SFPs into storage 
in casks within an ISFSI, coupled with a prohibition against future 
storage of fuel within the Units 2 and 3 SFPs (the Unit 1 SFP has 
been dismantled), removes the potential for fuel related accidents.
    The design basis and accident assumptions within the SONGS, 
Units 1, 2 and 3 UFSARs and the TSs relating to safe management and 
safe storage of spent fuel in the SFP are no longer applicable. The 
proposed changes do not affect remaining plant operations, systems, 
or components supporting decommissioning activities.
    The proposed deletion of TS requirements is not related to any 
SSCs that will be credited in the accident analysis for an 
applicable postulated accident. As a result, the proposed deletions 
do not affect the margin of safety associated with the accident 
analysis.
    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: Walker A. Matthews, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: Bruce Watson.

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. A publicly-available 
version is in ADAMS under Accession No. ML16355A015.
    Description of amendment request: The proposed amendment would 
revise the Permanently Defueled Emergency Plan into an Independent 
Spent Fuel Storage Facility Installation (ISFSl)-Only Emergency Plan, 
and revise the Emergency Action Level (EAL) scheme into an ISFSl-Only 
EAL scheme, for SONGS, Units 1, 2, and 3. The proposed changes would 
more fully reflect the permanently shutdown and defueled status of the 
facility, as well as the reduced scope of potential radiological 
accidents once all spent fuel has been moved to dry cask storage within 
the onsite SONGS ISFSI, an activity which is currently scheduled for 
completion in 2019.
    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 amendments would modify the SONGS, Units 1, 2 and 3 
licenses by revising the emergency plan and revising the EAL scheme. 
The SONGS units have permanently ceased operation and are 
permanently defueled. The proposed amendments are conditioned on all 
spent nuclear fuel being removed from wet storage in the spent fuel 
pools and placed in dry storage within an ISFSI. Occurrence of 
postulated accidents associated with spent fuel stored in a spent 
fuel pool is no longer credible in a spent fuel pool devoid of such 
fuel. The proposed amendments have no effect on plant systems, 
structures, and components (SSCs) and no effect on the capability of 
any plant SSC to perform its design function. The proposed 
amendments would not increase the likelihood of the malfunction of 
any plant SSC. The proposed amendments would have no effect on any 
of the previously evaluated accidents in the SONGS Updated Final 
Safety Analysis Report (UFSAR).
    Since SONGS has permanently ceased operation, the generation of 
fission products has ceased and the remaining source term continues 
to decay. This continues to significantly reduce the consequences of 
previously postulated accidents.
    Therefore, the proposed amendments do not involve a significant 
increase in the consequences of a previously evaluated accident.
    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 amendments constitute a revision of the emergency 
planning function commensurate with the ongoing and anticipated 
reduction in radiological source term at SONGS.
    The proposed amendments do not involve a physical alteration of 
the plant. No new or different types of equipment will be installed 
and there are no physical modifications to existing equipment as a 
result of the proposed amendments. Similarly, the proposed 
amendments would not physically change any SSCs involved in the 
mitigation of any postulated accidents. Thus, no new initiators or 
precursors of a new or different kind of accident are created. 
Furthermore, the proposed amendments do not create the possibility 
of a new failure mode associated with any equipment or personnel 
failures. The credible events for the ISFSI remain unchanged.
    Therefore, the proposed amendments 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.
    Because the 10 CFR part 50 licenses for SONGS no longer 
authorize operation of the reactors or emplacement or retention of 
fuel into the reactor vessels, as specified in 10 CFR 50.82(a)(2), 
the occurrence of postulated accidents associated with reactor 
operation is

[[Page 10602]]

no longer credible. With all nuclear spent fuel transferred out of 
wet storage from the spent fuel pools and placed in dry storage 
within the ISFSI, a fuel handling accident is no longer credible. 
There are no longer credible events that would result in any 
releases beyond the Exclusion Area Boundary (EAB) exceeding the U.S. 
Environmental Protection Agency (EPA) Protective Action Guideline 
(PAG) exposure levels, as detailed in the EPA's ``Protective Action 
Guide and Planning Guidance for Radiological Incidents,'' Draft for 
Interim Use and Public Comment dated March 2013 (PAG Manual).
    The proposed amendments do not involve a change in the plant's 
design, configuration, or operation. The proposed amendments do not 
affect either the way in which the plant structures, systems, and 
components perform their safety function or their design margins. 
Because there is no change to the physical design of the plant, 
there is no change to any of these margins.
    Therefore, the proposed amendments do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Walker A. Matthews, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: Bruce Watson.

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: August 31, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16244A253.
    Description of amendment request: The amendment request proposes 
changes to the Updated Final Safety Analysis Report (UFSAR) in the form 
of departures from the incorporated plant-specific Design Control 
Document (DCD) Tier 2 information and a combined license (COL) License 
Condition which references one of the proposed changes. Additionally, 
the proposed changes to the UFSAR eliminate pressurizer spray line 
monitoring during pressurizer surge line first plant only testing. In 
addition, these proposed changes correct inconsistencies in testing 
purpose, testing duration, and the ability to leave equipment in place 
following the data collection period. These changes involve material 
which is specifically referenced in Section 2.D.(2) of the COLs. This 
submittal requests approval of the license amendment necessary to 
implement these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff edits in square 
brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The design functions of the RCS [reactor coolant system] include 
providing an effective reactor coolant pressure boundary. The 
proposed changes for removing the requirement to install temporary 
instrumentation on the pressurizer spray line during the monitoring 
of the pressurizer surge line for thermal stratification and thermal 
cycling during hot functional testing and during the first fuel 
cycle for the first plant only, proposed changes to parameter 
retention requirements, and proposed change to remove the 
pressurizer spray and surge line valve leakage requirement do not 
impact the existing design requirements for the RCS. These changes 
are acceptable as they are consistent with the commitments made for 
the pressurizer surge line monitoring program for the first plant 
only, and do not adversely affect the capability of the pressurizer 
surge line and pressurizer spray lines to perform the required 
reactor coolant pressure boundary design functions.
    These proposed changes to the monitoring of the pressurizer 
surge line for thermal stratification and thermal cycling during hot 
functional testing and during the first fuel cycle for the first 
plant only, proposed changes to parameter retention requirements, 
and proposed change to remove the pressurizer spray and surge line 
valve leakage requirement as described in the current licensing 
basis do not have an adverse effect on any of the design functions 
of the systems. The proposed changes do not affect the support, 
design, or operation of mechanical and fluid systems required to 
mitigate the consequences of an accident. There is no change to 
plant systems or the response of systems to postulated accident 
conditions. There is no change to the predicted radioactive releases 
due to postulated accident conditions. The plant response to 
previously evaluated accidents or external events is not adversely 
affected, nor do the proposed changes create any new accident 
precursors.
    Therefore, the 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 for removing the requirement to install 
temporary instrumentation on the pressurizer spray line during the 
monitoring of the pressurizer surge line for thermal stratification 
and thermal cycling during hot functional testing and during the 
first fuel cycle for the first plant only, proposed changes to 
parameter retention requirements, and proposed change to remove the 
pressurizer spray and surge line valve leakage requirement as 
described in the current licensing basis maintain the required 
design functions, and are consistent with other Updated Final Safety 
Analysis Report (UFSAR) information. The proposed changes do not 
adversely affect the design requirements for the RCS, including the 
pressurizer surge line and pressurizer spray lines. The proposed 
changes do not adversely affect the design function, support, 
design, or operation of mechanical and fluid systems. The proposed 
changes do not result in a new failure mechanism or introduce any 
new accident precursors. No design function described in the UFSAR 
is adversely affected by the proposed changes.
    Therefore, the requested amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    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.

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

[[Page 10603]]

hazards consideration determination, and opportunity for a hearing in 
connection with these actions, was published in the Federal Register as 
indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: May 25, 2016, as supplemented by letters 
dated June 15, 2016, and October 18, 2016.
    Brief description of amendment: The amendment revised the Millstone 
Power Station, Unit No. 2, Technical Specifications (TSs) to add the 
evaluation model EMF-2103(P)(A), Revision 3, ``Realistic Large Break 
LOCA Methodology for Pressurized Water Reactors'' (ADAMS Package 
Accession No. ML16286A579), to the TS Section 6.9.1.8.b list of 
analytical methods use to establish core operating limits.
    Date of issuance: January 24, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 332. A publicly-available version is in ADAMS under 
Accession No. ML17025A218; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-65: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: August 30, 2016 (81 FR 
59662). The supplemental letters dated June 15, 2016, and October 18, 
2016, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 24, 2017.
    No significant hazards consideration comments received: No.

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

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: August 16, 2016.
    Brief description of amendments: The amendments modified license 
conditions to reflect the transfer of the Master Decommissioning Trust 
from the Power Authority of the State of New York to Entergy Nuclear 
Operations, Inc., and deletes other conditions so as to apply the 
requirements of 10 CFR 50.75(h)(1).
    Date of issuance: January 30, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 262 (Indian Point Nuclear Generating Unit No. 3); 
313 (James A. FitzPatrick Nuclear Power Plant). A publicly-available 
version is in ADAMS under Accession No. ML17025A288; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the letter dated January 27, 2017 (ADAMS Package Accession No. 
ML16336A488).
    Facility Operating License Nos. DPR-64 and DPR-59: Amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: September 27, 2016 (81 
FR 66305).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 27, 2017.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: August 29, 2016, as supplemented by 
letter dated November 21, 2016.
    Brief description of amendment: The amendment revised Technical 
Specification 5.5.6, ``Primary Containment Leakage Rate Testing 
Program,'' to allow permanent extension of Type A and Type C leak rate 
test intervals through the adoption of Revision 3-A of Nuclear Energy 
Institute (NEI) 94-01 and the limitations and conditions specified in 
Revision 2-A of NEI 94-01 as the guidance documents for implementation 
of performance-based Option B of appendix J to 10 CFR part 50, Option 
B, ``Performance-Based Requirements.'' Based on the guidance in 
Revision 3-A of NEI 94-01, the change allows the maximum interval for 
the Type A primary containment integrated leakage rate test to extend 
from once in 10 years to once in 15 years, and the Type C local leak 
rate test interval to extend to 75 months, provided acceptable 
performance history and other requirements are maintained.
    Date of issuance: January 24, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 132. A publicly-available version is in ADAMS under 
Accession No. ML17009A372; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-59: The amendment 
revised the Renewed Facility Operating License and the Technical 
Specifications.
    Date of initial notice in Federal Register: October 11, 2016 (81 FR 
70178). The supplemental letter dated November 21, 2016, 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 24, 2017.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station (DBNPS), Unit No. 1, Ottawa County, 
Ohio

    Date of application for amendment: February 17, 2016, as 
supplemented by letter dated September 6, 2016.
    Brief description of amendment: The amendment changed the DBNPS 
emergency plan by revising the emergency action level scheme.
    Date of issuance: January 12, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days from the date of issuance.

[[Page 10604]]

    Amendment No.: 294. A publicly-available version is in ADAMS under 
Accession No. ML16342C946; documents related to this amendment are 
listed in the Safely Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-3: The amendment revised 
the emergency plan.
    Date of initial notice in Federal Register: March 15, 2016 (81 FR 
13843).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 12, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 29, 2016.
    Brief description of amendment: This amendment approves a change to 
the administrative controls associated with the Limiting Condition for 
Operation (LCO) of Technical Specification (TS) 3.5.4, ``Refueling 
Water Storage Tank.''
    Date of issuance: January 18, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 207. A publicly-available version is in ADAMS under 
Accession No. ML16348A200; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-12: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: October 11, 2016 (81 FR 
70183).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 18, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 16, 2016, as supplemented by letter 
dated September 6, 2016.
    Brief description of amendments: The amendments revised the 
scheduled implementation date for Milestone 8 of the SONGS, Units 2 and 
3, Cyber Security Plan to December 31, 2019, in order to more fully 
reflect the permanently shutdown status of the facility and accommodate 
ongoing decommissioning activities.
    Date of issuance: January 23, 2017.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: Unit 2-234 and Unit 3-227: A publicly-available 
version is in ADAMS under Accession No. ML16252A207; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: August 2, 2016 (81 FR 
50735). The supplemental letter dated September 6, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 23, 2017.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph M. 
Farley Nuclear Plant, Unit 2, Houston County, Alabama

    Date of amendment request: September 8, 2016.
    Brief description of amendment: The amendment corrected an error in 
the Joseph M. Farley Nuclear Plant, Unit 2, Renewed Facility Operating 
License No. NPF-8, for Condition 2.C.(23). Specifically, the Unit 2 
referenced date representing the start of the 20-year period of 
extended operation was incorrectly entered as June 25, 2017. The Unit 2 
correct date corresponding to the 20-year period of extended operation 
is March 31, 2021.
    Date of issuance: January 23, 2017.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 204. A publicly-available version is in ADAMS under 
Accession No. ML15329A032; documents related to this amendment is 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-8: Amendment revised the 
Renewed Facility Operating License.
    Date of initial notice in Federal Register: October 25, 2016 (81 FR 
73441).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 23, 2017.
    No significant hazards consideration comments received: No.

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

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

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 (Hatch), Unit Nos. 1 and 2, Appling County, Georgia

    Date of amendment request: March 14, 2016, as supplemented by 
letters dated May 17, 2016, and October 26, 2016.
    Brief description of amendments: The amendments consist of changes 
that insert generic personnel titles in lieu of plant-specific 
personnel titles. In addition, the term ``plant-specific titles'' is 
replaced with ``generic titles'' in Technical Specification (TS) 
5.2.1.a for each plant. Lastly, this change revised the Hatch, Unit 
Nos. 1 and 2, TS 5.1 to be consistent with the corresponding Farley, 
Units 1 and 2, and Vogtle, Units 1 and 2, TS 5.1, and make it 
consistent with the corresponding Improved Standard Technical 
Specifications section.
    Date of issuance: January 13, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Farley--Unit 1 (207) and Unit 2 (203); Vogtle--Unit 
1 (183) and Unit 2 (166); and Hatch--Unit No. 1 (282) and Unit No. 2 
(227). A publicly-available version is in ADAMS under Accession No. 
ML16291A030; documents related to these amendments are listed in the 
Safety Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-2, NPF-8, NPF-68, NPF-
81, DPR-57, and NPF-5: Amendments revised the Renewed Facility 
Operating Licenses and TSs.
    Date of initial notice in Federal Register: May 24, 2016 (81 FR 
32809). The supplemental letters dated May 17,

[[Page 10605]]

2016, and October 26, 2016, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 13, 2017.
    No significant hazards consideration comments received: No.

Susquehanna Nuclear, LLC, Docket No. 50-388, Susquehanna Steam Electric 
Station, Unit 2, Luzerne County, Pennsylvania

    Date of amendment request: January 28, 2016, as supplemented by 
letters April 6, 2016, and October 10, 2016.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.7.1, ``Residual Heat Removal Service Water (RHRSW) 
System and the Ultimate Heat Sink (UHS),'' and TS 3.8.7, ``Distribution 
Systems--Operating,'' to increase the completion time for Conditions A 
and B of TS 3.7.1, and Condition C of TS 3.8.7, from 72 hours to 7 
days, in order to accommodate 480 volt engineered safeguard system load 
center transformer replacements on the Susquehanna Steam Electric 
Station, Unit 1. The change is temporary and will be annotated by a 
note in each TS that specifies the allowance expires on June 15, 2020.
    Date of issuance: January 26, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 248. A publicly-available version is in ADAMS under 
Accession No. ML17004A250; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-22: The amendment revised the 
Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: May 24, 2016 (81 FR 
32810). The supplemental letter dated October 10, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 26, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 12, 2016.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 4.3.1.2, ``Fuel Storage Criticality,'' for Units 1, 
2, and 3, to preclude the placement of fuel in the new fuel storage 
vaults. This TS change removed the existing TS 4.3.1.2 criticality 
criteria wording in its entirety, and replaced it with language that 
specifically restricts the placement of fuel in the new fuel storage 
vaults.
    Date of issuance: January 17, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 296 (Unit 1), 320 (Unit 2), and 280 (Unit 3). A 
publicly-available version is in ADAMS under Accession No. ML16330A158; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: October 11, 2016 (81 FR 
70187).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 17, 2017.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar 
Nuclear Plant, Units 1 and 2, Rhea County, Tennessee

    Date of amendment request: December 8, 2015, as supplemented by 
letters dated March 11, October 13, December 1, and December 8, 2016.
    Brief description of amendment: The amendment revised the Watts Bar 
Nuclear Plant, Units 1 and 2, Technical Specification (TS) 3.8.1, ``AC 
Sources--Operating,'' to extend the Completion Time for one inoperable 
Diesel Generator from 72 hours to 10 days based on the availability of 
a supplemental alternating current power source (specifically, the FLEX 
DG added as part of the mitigating strategies for beyond-design-basis 
events in response to NRC Order EA-12-049). The amendment also made 
clarifying changes to certain TS 3.8.1 Conditions, Required Actions, 
and Surveillance Requirements.
    Date of issuance: January 13, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 110 (Unit 1) and 5 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17006A271; documents related 
to this amendment are listed in the Safety Evaluation enclosed with the 
amendment.
    Facility Operating License Nos. NPF-90 and NPF-96: Amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: May 24, 2016 (81 FR 
32810). The supplement letters dated October 13, November 1, and 
December 8, 2016, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 13, 2017.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 6th day of February 2017.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2017-02795 Filed 2-13-17; 8:45 am]
BILLING CODE 7590-01-P