Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 32875-32890 [2017-14743]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 82, Number 136 (Tuesday, July 18, 2017)]
[Notices]
[Pages 32875-32890]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-14743]


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NUCLEAR REGULATORY COMMISSION

[NRC-2017-0158]


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 June 20,

[[Page 32876]]

2017 to July 3, 2017. The last biweekly notice was published on July 5, 
2017.

DATES: Comments must be filed by August 17, 2017. A request for a 
hearing must be filed by September 18, 2017.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0158. 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: TWFN-8-D36M, 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: Lynn Ronewicz, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-1927, email: lynn.ronewicz@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2017-0158, 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-2017-0158.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to 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-2017-0158, 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

[[Page 32877]]

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

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing and petition for leave to intervene (petition), any 
motion or other document filed in the proceeding prior to the 
submission of a request for hearing or petition to intervene, and 
documents filed by interested governmental entities that request to 
participate under 10 CFR 2.315(c), must be filed in accordance with the 
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 
46562, August 3, 2012). The E-Filing process requires participants to 
submit and serve all adjudicatory documents over the internet, or in 
some cases to mail copies on electronic storage media. Detailed 
guidance on making electronic submissions may be found in the Guidance 
for Electronic Submissions to the NRC and on the NRC's 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

[[Page 32878]]

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 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.

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

    Date of amendment request: June 14, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17165A555.
    Description of amendment request: The amendments would modify the 
completion date for implementation of Milestone 8 of the Cyber Security 
Plan (CSP). The proposed amendments would extend the CSP Milestone 8 
completion date from September 30, 2017, to December 31, 2017.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the PVNGS Cyber Security Plan 
implementation schedule is administrative in nature. This proposed 
change does not alter accident analysis assumptions, add any 
initiators, or affect the function of plant systems or the manner in 
which systems are operated, maintained, modified, tested, or 
inspected. The proposed change does not require any plant 
modifications which affect the performance capability of the 
structures, systems, and components (SSCs) relied upon to mitigate 
the consequences of postulated accidents, and has no impact on the 
probability or consequences of an accident previously evaluated.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change to the PVNGS Cyber Security Plan 
implementation schedule is administrative in nature. This proposed 
change does not alter accident analysis assumptions, add any 
initiators, or affect the function of plant systems or the manner in 
which systems are operated, maintained, modified, tested, or 
inspected. The proposed change does not require any plant 
modifications which affect the performance capability of the SSCs 
relied upon to mitigate the consequences of postulated accidents, 
and does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Plant safety margins are established through limiting conditions 
for operation, limiting safety systems settings, and safety limits 
specified in the [T]echnical [S]pecifications [TSs]. The proposed 
change to the PVNGS Cyber Security Plan implementation schedule is 
administrative in nature. Since the proposed change is 
administrative in nature, there are no changes to these established 
safety margins.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety as defined in the basis for any TS.

    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

[[Page 32879]]

proposes to determine that the request for amendments involves no 
significant hazards consideration.
    Attorney for licensee: Michael G. Green, Senior Regulatory Counsel, 
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, 
Phoenix, AZ 85072-2034.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: March 30, 2017, as supplemented by 
letter dated May 11, 2017. Publicly-available versions are in ADAMS 
under Accession Nos. ML17095A530 and ML17139D352, respectively.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TSs) in accordance with the NRC-approved 
Technical Specifications Task Force (TSTF) Standard Technical 
Specification Change Traveler TSTF-448, Revision 3, ``Control Room 
Habitability,'' with variations from the TSTF to account for plant-
specific configuration and licensing basis differences. The amendments 
would modify the TSs for the control room ventilation system (CRVS) 
booster fans and would establish a control room envelop (CRE) 
habitability program in TS 5.5, ``Programs and Manuals.'' The NRC staff 
issued ``Notice of Availability of Technical Specification Improvement 
to Modify Requirements Regarding Control Room Envelope Habitability 
Using the Consolidated Line Item Improvement Process,'' associated with 
TSTF-448, Revision 3, in the Federal Register on January 17, 2007 (72 
FR 2022). The notice included a model safety evaluation, a model no 
significant hazards consideration determination, and a model license 
amendment request. In its application dated March 30, 2017, as 
supplemented by letter dated May 11, 2017, the licensee affirmed the 
applicability of the model no significant hazards consideration 
determination, which is presented in the following section.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee affirmed 
the applicability of the model no significant hazards consideration, 
which is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the [CRVS], which is 
a mitigation system designed to minimize unfiltered air leakage into 
the CRE and to filter the CRE atmosphere to protect the CRE 
occupants in the event of accidents previously analyzed. An 
important part of the [CRVS] is the CRE boundary. The [CRVS] is not 
an initiator or precursor to any accident previously evaluated. 
Therefore, the probability of any accident previously evaluated is 
not increased. Performing tests to verify the operability of the CRE 
boundary and implementing a program to assess and maintain CRE 
habitability ensure that the [CRVS] is capable of adequately 
mitigating radiological consequences to CRE occupants during 
accident conditions, and that the [CRVS] will perform as assumed in 
the consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Accident Previously 
Evaluated

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the [CRVS], or its functioning during accident conditions as assumed 
in the licensing basis analyses of design basis accident 
radiological consequences to CRE occupants. No new or different 
accidents result from performing the new surveillance or following 
the new program. The proposed change does not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a significant change in the methods governing 
normal plant operation. The proposed change does not alter any 
safety analysis assumptions and is consistent with current plant 
operating practice.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition.
    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.

Entergy Nuclear Operations, Inc. (ENO), Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: May 15, 2017. A publicly-available 
version is in ADAMS under Package Accession No. ML17139D261.
    Description of amendment request: The proposed amendment would 
replace the Permanently Defueled Emergency Plan and its associated 
Permanently Defueled Emergency Action Level (EAL) Technical Bases 
Document with the Independent Spent Fuel Storage Installation (ISFSI) 
Emergency Plan and its associated ISFSI EAL Technical Bases Document, 
for the Vermont Yankee Nuclear Power Station (VY). The proposed changes 
would reflect the complete removal of all fuel from the spent fuel pool 
(SFP) and permit specific reductions in the size and makeup of the 
Emergency Response Organization due to the elimination of the design-
basis accident related to the spent fuel (fuel handling accident). As 
described in the Post Shutdown Decommissioning Activities Report, spent 
fuel will remain in the SFP until it meets the criteria for transfer, 
the existing ISFSI is expanded, and the spent fuel can be safely 
transferred in an efficient manner to the expanded ISFSI, an activity 
that is currently scheduled for completion in late 2018.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 32880]]


    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 VY facility operating 
license by revising the emergency plan and EAL scheme. VY has 
permanently ceased power operations and is permanently defueled. The 
proposed amendment is conditioned on all spent nuclear fuel being 
removed from wet storage in the SFP and placed in dry storage within 
the ISFSI. Occurrence of postulated accidents associated with spent 
fuel stored in a SFP is no longer credible in a SFP devoid of fuel. 
The proposed amendment has no effect on plant structures, systems, 
or components (SSC) and therefore can neither affect the capability 
of any plant SSC to perform its design function nor increase the 
likelihood of the malfunction of any plant SSC. The proposed 
amendment would have no effect on any of the previously evaluated 
accidents in the VY Defueled Safety Analysis Report or the Holtec 
HI-STORM 100 Final Safety Analysis Report.
    Because VY has permanently ceased power operations, the 
generation of fission products has largely ceased and the remaining 
source term continues to decay. This source term decay continues to 
significantly reduce the consequences of previously evaluated 
postulated accidents. Furthermore, previously generated source term 
materials such as reactor water cleanup resins have been removed 
from the site in accordance with applicable regulations and 
permitting requirements.
    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 constitutes a revision of the emergency 
planning function commensurate with the ongoing and anticipated 
reduction in radiological source term at VY.
    The proposed amendment does 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 amendment. Similarly, the proposed amendment 
would not physically change any SSC 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 amendment does 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 license for VY no longer authorizes 
operation of the reactor or emplacement or retention of fuel into 
the reactor vessel, as specified in 10 CFR 50.82(a)(2), the 
postulated accidents associated with reactor operation are no longer 
credible. In addition, with all spent nuclear fuel transferred out 
of wet storage from the SFP and placed in dry storage within the 
ISFSI, a fuel handling accident is no longer credible during dry 
storage of spent nuclear fuel. Therefore, there are no credible 
events that would result in radiological releases beyond the site 
boundary exceeding the exposure levels in U.S. EPA's ``Protective 
Action Guide and Planning Guidance for Radiological Incidents,'' 
dated January 2017.
    The proposed amendment does not involve a change in the plant's 
design, configuration, or operation. The proposed amendment does not 
affect either the way in which the plant SSCs perform their safety 
function or their design margins. Because there is no change to the 
physical design of the facility, there is no change to these 
margins.
    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: Ms. Susan Raimo, Senior Counsel, Entergy 
Services, Inc., 101 Constitution Ave. NW., Suite 200 East, Washington, 
DC 20001.
    NRC Branch Chief: Bruce Watson.

Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247, and 50-
286, Indian Point Nuclear Generating Unit Nos. 1, 2, and 3, Westchester 
County, New York

    Date of amendment request: April 28, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17129A612.
    Description of amendment request: The amendments would modify the 
completion date for implementation of Milestone 8 of the Cyber Security 
Plan (CSP). The proposed amendments would extend the CSP Milestone 8 
full implementation date from December 31, 2017, to December 31, 2022.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the CSP Implementation Schedule does not 
alter accident analysis assumptions, add any initiators, or affect 
the function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected. The proposed 
change does not require any plant modifications which affect the 
performance capability of the structures, systems, and components 
relied upon to mitigate the consequences of postulated accidents and 
has no impact on the probability or consequences of an accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to the CSP Implementation Schedule does not 
alter accident analysis assumptions, add any initiators, or affect 
the function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected. The proposed 
change does not require any plant modifications which affect the 
performance capability of the structures, systems, and components 
relied upon to mitigate the consequences of postulated accidents and 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Plant safety margins are established through limiting conditions 
for operation, limiting safety system settings, and safety limits 
specified in the technical specifications. The proposed change to 
the CSP Implementation Schedule does not involve these items. In 
addition, the milestone date delay for full implementation of the 
CSP has no substantive impact because other measures have been taken 
which provide adequate protection during this period of time. 
Because there is no change to established safety margins as a result 
of this change, the proposed change does not involve a significant 
reduction in a margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeanne Cho, Assistant General Counsel, 
Entergy

[[Page 32881]]

Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.
    NRC Branch Chief: James G. Danna.

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

    Date of amendment request: April 7, 2017. A publicly-available 
version is in ADAMS under Package Accession No. ML17104A039.
    Description of amendment request: The amendment would revise 
Technical Specification 3.5.4, ``Refueling Water Storage Tank (RWST),'' 
such that the non-seismically qualified piping of the Boric Acid 
Recovery System be connected to the RWST seismic piping. This change 
will only be applicable until the end of the Indian Point Nuclear 
Generating Unit No. 2 Refueling Outage 2R23.
    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 use of the non-seismic Boric Acid Recovery System (BARS) to 
recirculate and filter the RWST water does not involve any changes 
or create any new interfaces with the reactor coolant system or main 
steam system piping. Therefore, the connection of the BARS 
Purification Loop to the RWST would not affect the probability of 
these accidents occurring. The BARS is not credited for safe 
shutdown of the plant or accident mitigation. Administrative 
controls ensure that the BARS can be isolated as necessary and in 
sufficient time to assure that the RWST volume will be adequate to 
perform the safety function as designed. Since the RWST will 
continue to perform its safety function and overall system 
performance is not affected, the consequences of the accident are 
not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The design of the RWST and the SFP [Spent Fuel Pool] 
Purification Loop has been revised to allow recirculation and 
purification using the BARS for a short period of time (not to 
exceed 30 days per fuel cycle) for the next fuel cycle. The BARS 
takes RWST water in and processes it out without additional 
connections that could affect other systems and without an impact 
from its installation. Procedures for the operation of the plant, 
including the BARS, will not create the possibility of a new or 
different type of accident. Contingent upon manual operator action, 
a BARS line break will not result in a loss of the RWST safety 
function. Similarly, an active or passive failure in the BARS will 
not affect safety related structures, systems or components.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The SFP Purification Loop and recirculation and purification of 
the RWST water using the BARS is not credited for safe shutdown of 
the plant or accident mitigation. RWST volume will be maximized 
prior to purification and timely operator action can be taken to 
isolate the non-seismic system from the RWST to assure it can 
perform its function. This will result in no significant reduction 
in the margin of safety.
    Therefore the proposed change does not significantly reduce the 
margin of safety.

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

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

    Date of amendment request: April 9, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17100A269.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) Section 4.2.1, ``Fuel Assemblies,'' and 
Section 5.6.3, ``Core Operating Limits Report (COLR),'' to allow the 
use of Optimized ZIRLO\TM\ as an approved fuel rod cladding material. 
In the letter dated April 9, 2017, the licensee also requested an 
exemption from certain requirements of 10 CFR 50.46 and 10 CFR part 50, 
appendix K, in accordance with 10 CFR 50.12, to support the license 
amendments.
    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 allow the use of Optimized 
ZIRLOTM clad nuclear fuel at BVPS. The NRC approved 
topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, 
``Optimized ZIRLO\TM\,'' prepared by Westinghouse Electric Company 
LLC (Westinghouse), which addresses Optimized ZIRLOTM 
fuel rod cladding and demonstrates that Optimized ZIRLOTM 
fuel rod cladding has essentially the same properties as currently 
licensed ZIRLO[supreg] fuel rod cladding. The use of Optimized 
ZIRLOTM fuel rod cladding material will not result in 
adverse changes to the operation or configuration of the facility. 
The fuel cladding itself is not an accident initiator and does not 
affect accident probability. The correction of a typographical 
error, the addition of a word for clarification of the TS, and the 
addition of a registered trademark designator are administration 
changes and do not affect the fuel cladding design. Use of Optimized 
ZIRLOTM meets the fuel design acceptance criteria and 
hence does not significantly affect the consequences of an accident.
    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 use of Optimized ZIRLOTM fuel rod cladding 
material will not result in adverse changes to the operation or 
configuration of the facility. The correction of a typographical 
error, the addition of a word for clarification of the TS, and the 
addition of a registered trademark designator are administration 
changes and do not affect the fuel cladding design. Topical Report 
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A demonstrated that the 
material properties of Optimized ZIRLOTM fuel rod 
cladding are similar to those of ZIRLO[supreg] fuel rod cladding. 
Therefore, Optimized ZIRLOTM fuel rod cladding will 
perform similarly to ZIRLO[supreg] fuel rod cladding, thus 
precluding the possibility of the fuel rod cladding becoming an 
accident initiator and causing a new or different kind of accident.
    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 will not involve a significant reduction 
in the margin of safety. NRC-approved Topical Report WCAP-12610-P-A 
& CENPD-404-P-A, Addendum 1-A, demonstrated that the material 
properties of the Optimized ZIRLOTM fuel rod cladding are 
similar to those of ZIRLO[supreg] fuel rod cladding. Optimized 
ZIRLOTM fuel rod cladding is expected to perform 
similarly to ZIRLO[supreg] fuel rod cladding for normal

[[Page 32882]]

operating and accident scenarios, including both loss-of-coolant 
accident (LOCA) and non-LOCA scenarios. The use of Optimized 
ZIRLOTM fuel rod cladding will not result in adverse 
changes to the operation or configuration of the facility. The 
correction of a typographical error, the addition of a word for 
clarification of the TS, and the addition of a registered trademark 
designator are administration changes that do not affect the fuel 
cladding design.
    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: David W. Jenkins, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, 
Akron, OH 44308.
    NRC Branch Chief: James G. Danna.

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

    Date of amendment request: June 12, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17164A191.
    Description of amendment request: The requested amendments propose 
changes to the Updated Final Safety Analysis Report in the form of 
departures from the plant-specific Design Control Document (DCD) Tier 2 
information, and involve changes to related plant-specific DCD Tier 1 
information, with corresponding changes to the associated combined 
license (COL) Appendix C information. In addition, revisions are 
proposed to COL Appendix A, Technical Specifications. The proposed 
changes revise the COLs concerning standardizing the Protection and 
Safety Monitoring System (PMS) setpoint nomenclature. No changes are 
proposed to setpoint values or PMS alarms and actuations. Pursuant to 
the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the 
design as certified in the 10 CFR part 52, appendix D, Design 
Certification Rule, is also requested for the plant-specific Tier 1 
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:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    No setpoint values or PMS actuations are proposed to be changed 
by this activity. Nor are any values assumed in the safety analysis 
changed. This is an administrative change to standardize the PMS 
setpoint designators. The proposed amendment does not affect the 
prevention and mitigation of abnormal events, e.g., accidents, 
anticipated operation occurrences, earthquakes, floods, turbine 
missiles, and fires or their safety or design analyses. This change 
does not involve containment of radioactive isotopes or any adverse 
effect on a fission product barrier. There is no impact on 
previously evaluated accidents.
    These proposed changes have no adverse impact on the support, 
design, or operation of mechanical and fluid systems. The response 
of systems to postulated accident conditions is not adversely 
affected and remains within response time assumed in the accident 
analysis. 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 requested amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve a new failure mechanism or 
malfunction, which affects [a structure, system, component (SSC)] 
accident initiator, or interface with any SSC accident initiator or 
initiating sequence of events considered in the design and licensing 
bases. There is no adverse effect on radioisotope barriers or the 
release of radioactive materials. The proposed amendment does not 
adversely affect any accident, including the possibility of creating 
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 type of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    No setpoint values or PMS actuations are proposed to be changed 
by this activity. This is an administrative change to standardize 
the PMS setpoint designators. The proposed changes would not affect 
any safety-related design code, function, design analysis, safety 
analysis input or result, or existing design/safety margin. No 
safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the requested changes.
    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.

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

    Date of amendment request: June 9, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17163A174.
    Description of amendment request: The requested amendments propose 
changes to combined license (COL) Appendix C (and plant-specific Tier 
1) Table 2.7.2-2 to revise the minimum chilled water flow rates to the 
supply air handling units serving the Main Control Room and the Class 
1E electrical rooms, and the unit coolers serving the normal residual 
heat removal system and chemical and volume control system pump rooms. 
The proposed COL Appendix C (and plant-specific Design Control Document 
(DCD) Tier 1) changes require additional changes to corresponding Tier 
2 component data information in Updated Final Safety Analysis Report 
Chapters 6 and 9. Because this proposed change requires a departure 
from Tier 1 information in the Westinghouse Electric Company's AP1000 
DCD, the licensee also requested an exemption from the requirements of 
the Generic DCD Tier 1 in accordance with 10 CFR 52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to COL Appendix C (and plant-specific Tier 
1) Table 2.7.2-2, Updated Final Safety Analysis Report (UFSAR) Table 
9.2.7-1, and associated UFSAR design information to identify the 
revised equipment parameters for the nuclear island nonradioactive 
ventilation system (VBS) air handling units (AHUs) and 
radiologically controlled area (RCA) ventilation system (VAS) unit 
coolers and reduced chilled water system (VWS) cooling coil flow 
rates does not adversely impact the

[[Page 32883]]

plant response to any accidents which are previously evaluated. The 
function of the cooling coils to provide chilled water to the VBS 
AHUs and VAS unit coolers is not credited in the safety analysis.
    No safety-related structure, system, component (SSC) or function 
is adversely affected by this change. The change does not involve an 
interface with any SSC accident initiator or initiating sequence of 
events, and thus, the probabilities of the accidents evaluated in 
the plant-specific UFSAR are not affected. The proposed changes do 
not involve a change to the predicted radiological releases due to 
postulated accident conditions, thus, the consequences of the 
accidents evaluated in the UFSAR are not affected. The proposed 
changes do not increase the probability or consequences of an 
accident previously evaluated as the VWS, VBS and VAS do not provide 
safety-related functions and the functions of each system to support 
required room environments are not changed.
    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 to COL Appendix C (and plant-specific Tier 
1) Table 2.7.2-2, UFSAR Table 9.2.7-1, and associated UFSAR design 
information to identify the revised equipment parameters for VBS 
AHUs and VAS unit coolers and reduced VWS cooling coil flow rates do 
not affect any safety-related equipment, and do not add any new 
interfaces to safety-related SSCs. The VWS function to provide 
chilled water is not adversely impacted. The function of the VAS to 
provide ventilation and cooling to maintain the environment of the 
serviced areas within the design temperature range is not adversely 
impacted by this change. No system or design function or equipment 
qualification is affected by these changes as the change does not 
modify the operation of any SSCs. The changes do not introduce a new 
failure mode, malfunction or sequence of events that could affect 
safety or safety-related equipment. Revised equipment parameters, 
including the reduced cooling coil flow rates, do not adversely 
impact the function of associated components.
    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 changes to COL Appendix C (and plant-specific Tier 1) Table 
2.7.2-2, UFSAR Table 9.2.7-1, and associated UFSAR design 
information do not affect any other safety-related equipment or 
fission product barriers. The requested changes will not adversely 
affect compliance with any design code, function, design analysis, 
safety analysis input or result, or design/safety margin. No safety 
analysis or design basis acceptance limit/criterion is challenged or 
exceeded by the requested changes as previously evaluated accidents 
are not impacted.
    Therefore, the proposed amendment does not involve a significant 
reduction in the margin of safety.

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

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

    Date of amendment request: May 24, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17144A408.
    Description of amendment request: The amendments would revise 
Surveillance Requirement 3.3.1.3 to change the thermal power at which 
the surveillance may be performed.
    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 [Technical Specification] does 
not affect the initiators of any analyzed accident. In addition, 
operation in accordance with the proposed amendment to the TS 
ensures that the previously evaluated accidents will continue to be 
mitigated as analyzed. The proposed amendment does not adversely 
affect the design function or operation of any structures, systems, 
and components important to safety.
    The probability or consequences of accidents previously 
evaluated in the UFSAR [Updated Final Safety Analysis Report] are 
unaffected by this proposed amendment because there is no change to 
any equipment response or accident mitigation scenario. There are no 
new or additional challenges to fission product barrier integrity.
    Therefore, it is concluded that 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?
    The proposed amendment does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
The proposed amendment does not create any new failure modes for 
existing equipment or any new limiting single failures. The proposed 
amendment does not involve a change in the methods governing normal 
plant operation and all safety functions will continue to perform as 
previously assumed in accident analyses. Thus, the proposed 
amendment does not adversely affect the design function or operation 
of any structures, systems, and components important to safety.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced due to the proposed amendment. The 
proposed amendment does not challenge the performance or integrity 
of any safety-related system.
    Therefore, it is concluded that the proposed amendment does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed changes involve a significant reduction in 
a margin of safety?
    The margin of safety associated with the acceptance criteria of 
any accident is unchanged. The proposed amendment will have no 
affect on the availability, operability, or performance of the 
safety-related systems and components. No change is being made to 
the requirement to perform the surveillance. The NOTE in the 
surveillance is being changed to clarify when the initial 
surveillance after refueling is to be performed. The Technical 
Specification Limiting Condition for Operation (LCO) limits are not 
being changed.
    The proposed amendment will not adversely affect the operation 
of plant equipment or the function of equipment assumed in the 
accident analysis.
    Therefore, it is concluded that 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: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, 40 Iverness Center 
Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Michael T. Markley.

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: May 5, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17128A120.

[[Page 32884]]

    Description of amendment request: The requested amendments propose 
changes to more clearly define the boundaries and seismic requirements 
for the portion of the fire protection system (FPS) piping that is 
required to remain functional following a safe shutdown earthquake 
(SSE) (i.e., the ``seismic standpipe system''). The proposed changes 
also include the removal of SSE requirements from pipe lines that do 
not need to remain functional following an SSE (specifically, the FPS 
piping that is part of the non-seismic FPS containment spray system and 
the FPS open tray system).
    Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from 
elements of the design as certified in the 10 CFR part 52, appendix D, 
design certification rule is also requested for the plant-specific 
Design Control Document Tier 1 material departures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed modification changes would clarify the boundaries 
for the portion of the nonsafety-related FPS required to remain 
functional following a SSE for manual firefighting in areas with SSE 
equipment, and the addition of two new open-nozzle suppression 
systems with associated system isolation valves to provide adequate 
spray coverage to accommodate the final cable tray location, 
configuration and quantity. These changes do not affect any accident 
initiating event or component failure, thus the probabilities of the 
accidents previously evaluated are not adversely affected. No 
function used to mitigate a radioactive material release and no 
radioactive material release source term is involved, thus the 
radiological releases in the accident analyses are not adversely 
affected. Therefore, the proposed amendment does not involve an 
increase in the probability or 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 clarification of the boundaries for the portion of 
the nonsafety-related FPS required to remain functional following a 
SSE for manual firefighting in areas with equipment required for 
safe shutdown following an SSE 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 changes 
affect the physical design and operation of the FPS, including as-
installed inspections, testing, and maintenance requirements, as 
described in the Updated Final Safety Analysis Report (UFSAR) due to 
the addition of two open-nozzle suppression systems with associated 
system isolation valves. However, the additional open-nozzle 
suppression systems with associated system isolation valves are 
similar in design and function as the existing cable tray 
suppression systems and raceway covers. Therefore, the operation of 
the FPS is not affected. These proposed changes do not adversely 
affect any other SSC design functions or methods of operation in a 
manner that results in a new failure mode, malfunction, or sequence 
of events that affect safety-related or nonsafety-related equipment. 
Therefore, this activity does not allow for a new fission product 
release path, result in a new fission product barrier failure mode, 
or create a new sequence of events that results in significant fuel 
cladding failures.
    Therefore, the requested amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed clarification of the boundaries for the portion of 
the FPS required to remain functional following a SSE, and the 
addition of two new open-nozzle suppression systems with associated 
system isolation valves do not affect any safety or accident 
analysis as the FPS is a nonsafety-related system. The only function 
of the FPS following a design basis earthquake is to provide water 
for hose valves for manual firefighting in safe shutdown equipment 
areas. The proposed changes continue to meet the existing design 
basis, design function, regulatory criterion, or analyses. 
Therefore, the proposed changes satisfy the same design functions in 
accordance with the codes and standards currently stated in the 
UFSAR. These changes do not adversely affect any design code, 
function, design analysis, safety analysis input or result, or 
design/safety margin. No safety analysis or design basis acceptance 
limit/criterion is challenged or exceeded by the proposed changes, 
and no margin of safety is reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: May 9, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17129A608.
    Description of amendment request: The requested amendments propose 
to depart from approved AP1000 Design Control Document (DCD) Tier 2 
information (text, tables, and figures) as incorporated into the 
Updated Final Safety Analysis Report (UFSAR) as plant-specific DCD 
information, and also propose to depart from involved plant-specific 
Tier 1 information (and associated combined license (COL) Appendix C 
information) and from involved plant-specific Technical Specifications 
as incorporated in Appendix A of the COL. Specifically, the proposed 
amendments would revise the licensing basis information to reflect 
design changes to the main control room emergency habitability system 
(VES) to address the main control room envelope temperature response.
    Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from 
elements of the design as certified in the 10 CFR part 52, appendix D, 
design certification rule is also requested for the plant-specific DCD 
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:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that initiate an analyzed accident or alter any 
structures, systems, and components (SSCs) accident initiator or 
initiating sequence of events. The VES design changes involve: (1) 
Addition of an automatic and manual, Class 1E, electrical load shed 
of nonessential nonsafety-related equipment within the main control 
room envelope (MCRE); and (2) adding a description of the 
requirements for maintaining habitability of the MCRE beyond 72 
hours following a Design Basis Accident to the design and licensing 
basis. Neither planned or inadvertent operation nor failure of the 
VES is an accident initiator or part of an initiating sequence of 
events for an accident previously evaluated. For example, if VES 
actuation occurs from a loss of power

[[Page 32885]]

to the plant in a station blackout condition, the additional added 
features including Wall Panel Information System displays would not 
be available regardless of the load shed feature. This condition was 
originally evaluated as part of the AP1000 design certification and 
no changes are proposed to the plant station blackout response. No 
additional re-evaluation of other probability or consequences from 
failures are required to support this change. Therefore, the 
probabilities of the accidents evaluated in the UFSAR are not 
affected.
    The proposed changes do not have an adverse impact on the 
ability of the VES to perform its design functions. The design of 
the VES continues to meet the same regulatory acceptance criteria, 
codes, and standards as required by the UFSAR. In addition, the 
changes maintain the capability of the VES to mitigate the 
consequences of an accident in conformance with the applicable 
regulatory acceptance criteria, and there is no adverse effect on 
any safety-related SSC or function used to mitigate an accident. The 
changes do not affect the prevention and mitigation of other 
abnormal events, e.g., anticipated operational occurrences, 
earthquakes, floods and turbine missiles, or their safety or design 
analyses. Therefore, the consequences of the accidents evaluated in 
the UFSAR are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that may initiate a new or different kind of accident, 
or alter any SSC such that a new accident initiator or initiating 
sequence of events is created. The VES design changes involve: (1) 
Addition of an automatic and manual, Class 1E, electrical load shed 
of nonessential nonsafety-related equipment within the MCRE; and (2) 
adding a description of the requirements for maintaining 
habitability of the MCRE beyond 72 hours following a DBA to the 
design and licensing basis. Although a new failure mode of the VES 
is created by the addition of the MCR Load Shed Panels, neither 
planned nor inadvertent operation nor failure of the VES is an 
accident initiator or part of an initiating sequence of events for a 
new or different kind of accident. In addition, these proposed 
changes do not adversely affect any other VES or SSC design 
functions or methods of operation in a manner that results in a new 
failure mode, malfunction, or sequence of events that affect safety-
related or nonsafety-related equipment. Therefore, this activity 
does not allow for a new fission product release path, result in a 
new fission product barrier failure mode, or create a new sequence 
of events that result in significant fuel cladding failures.
    Therefore, the 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 changes to the VES description and associated COL Appendix A 
Technical Specification changes provide continued verification that; 
the VES design functions to maintain heat loads inside the MCRE 
within design-basis assumptions to limit the heat up of the room, a 
72-hour supply of breathable-quality air for the occupants of the 
MCRE is readily available, and the MCRE pressure boundary is 
maintained at a positive pressure with respect to the surrounding 
areas. The changes support the system's intended design functions 
and continue to meet the regulatory requirements for protecting 
public health and safety.
    The proposed changes also maintain existing safety margins. The 
proposed changes do not adversely affect VES design requirements and 
design functions. The proposed changes maintain existing safety 
margin through continued application of the existing requirements of 
the UFSAR, while adding additional design features and controls that 
maintain VES design functions required to meet the existing safety 
margins. Therefore, the proposed changes satisfy the same design 
functions in accordance with the same codes and standards as stated 
in the UFSAR. These changes do not adversely affect any design code, 
function, design analysis, safety analysis input or result, or 
design/safety margin.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant 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.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of amendment request: April 6, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17097A425.
    Description of amendment request: The amendment would revise the 
Final Safety Analysis Report (FSAR) to allow bypassing of thermal 
overload protection during motor-operated valve surveillance testing.
    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.
    Keeping the thermal overload protection (TOP) devices bypassed 
during surveillance testing does not introduce the possibility of a 
change in the frequency of an accident because failure of a single 
safety-related motor-operated valve (MOV) is not, by itself, an 
initiator of any previously evaluated design basis accident. Valves 
are active components that either position to ``open'' or ``close'' 
as required to fulfill safety functions. As such, safety-related 
MOVs are subject to single active failures, but such failures are 
not accident initiators. (For safety-related systems, redundancy in 
the design ensures that failure of a valve to open or to close on 
demand, as applicable, will not prevent fulfillment of the safety 
function(s). However, the associated safety functions are for 
accident mitigation/response, and while an MOV failure can affect 
such functions (without loss of the overall function), a single MOV 
failure cannot by itself initiate any accident previously evaluated 
in the FSAR.)
    Furthermore, the change does not result in an increase in the 
consequences of an accident previously evaluated in the FSAR. The 
proposed change would permit MOV TOP devices to remain bypassed 
during surveillance stroke testing but not during valve maintenance. 
In regard to the bypassing of TOP devices during testing, the 
potential for valve damage is of greater concern during valve 
maintenance activities (when work has been done on the affected 
valve(s)) than it is for surveillance stroke tests. It may be 
assumed that the low probability of valve damage resulting from--or 
occurring during--surveillance valve stroke tests (with the TOP 
devices bypassed) does not change the single-failure assumptions 
already considered in the plant's design and accident analyses. As 
previously noted, redundancy in the design of safety-related systems 
ensures that failure of a valve to open or close on demand, as 
applicable, will not prevent fulfillment of the safety function(s). 
Accordingly, it may be concluded that the provisions for bypassing 
TOP devices during MOV surveillance testing does not require any 
changes to assumptions regarding MOV availability, single-failure 
protection, or the associated systems' capabilities for performing 
accident mitigation functions. With no changes to such assumptions, 
the proposed change does not result in more than a minimal increase 
in the consequences of an accident previously evaluated in the FSAR.
    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.
    NRC [RG] 1.106, Revision 1 [``Thermal Overload Protection for 
Electric Motors on Motor-Operated Valves''], requires the removal of 
MOV thermal overload relay bypass jumpers during both maintenance 
and

[[Page 32886]]

periodic tests. The regulatory guide's position is that having the 
thermal overload protection enabled during periodic tests of an MOV 
is desired to prevent valve motor damage. The concern is that the 
motor may be damaged if the thermal overload protection is not in 
force.
    Keeping the [TOP] devices bypassed during surveillance testing 
does not introduce the possibility of an accident of a different 
type than any previously evaluated in the FSAR. Although there could 
be a slight increase in the probability of valve damage due to the 
proposed change, any such failure would not be of a different kind 
or nature than what may already be experienced by an MOV. Thus, no 
new failure modes or initiators of a different type of accident are 
introduced. The single active failure of a[n] [MOV] is already 
considered in the accident analysis assumptions described in the 
FSAR, and the failure of a single MOV is not by itself an accident 
initiator.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    No, this change does not affect design basis limits for a 
fission product barrier. No changes to the accident analyses, 
including any associated assumptions, are required or being made for 
the proposed change. Because of redundancy incorporated into the 
plant design (for single-failure protection), the failure of a 
single [MOV] will not result in the loss of any overall safety 
function.
    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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
    NRC Branch Chief: Robert J. Pascarelli.

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

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

Duke Energy Florida, Inc., et al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant, Citrus County, Florida

    Date of amendment request: December 9, 2016.
    Brief description of amendment: The amendment approved the removal 
of the existing cyber security license condition from the facility 
operating license.
    Date of issuance: June 22, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 254. A publicly-available version is in ADAMS under 
Package Accession No. ML17096A279; documents related to this amendment 
are listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. DPR-72: The amendment revised the 
license.
    Date of initial notice in Federal Register: January 31, 2017 (82 FR 
8868).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 22, 2017.
    No significant hazards consideration comments received: No.

Duke Energy Florida, Inc., et al., Docket No. 50-302, Crystal River 
Unit 3 (CR-3) Nuclear Generating Plant, Citrus County, Florida

    Date of amendment request: August 31, 2016.
    Brief description of amendment: The amendment approved an amendment 
to the CR-3 Facility Operating License and the Permanently Defueled 
Technical Specifications to reflect removal of all CR-3 spent nuclear 
fuel from the spent fuel pools and its transfer to dry cask storage 
within the independent spent fuel storage installation (ISFSI).
    Date of issuance: June 27, 2017.
    Effective date: The date Duke Energy Florida, LLC submits written 
notification that all spent fuel has been transferred from the spent 
fuel pool to the ISFSI and shall be implemented within 60 days.
    Amendment No.: 255.
    Facility Operating License No. DPR-72: The amendment revised the 
license.
    Date of initial notice in Federal Register: October 25, 2016 (81 FR 
73432).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 27, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 12, 2016, as supplemented by letter 
dated November 17, 2016.
    Brief description of amendment: The amendment reduced the minimum 
reactor dome pressure associated with the critical power correlation 
from 785 pounds per square inch gauge (psig) to 686 psig in Technical 
Specification 2.1.1, ``Reactor Core SLs [Safety Limits],'' and 
associated bases.
    Date of issuance: June 27, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 242. A publicly-available version is in ADAMS under 
Accession No. ML17131A071; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-21: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: The license amendment 
request was originally noticed in the Federal Register on October 25, 
2016 (81 FR 73433). Subsequently, by letter dated November 17, 2016, 
the licensee

[[Page 32887]]

provided additional information that expanded the scope of the 
amendment request as originally noticed in the Federal Register. 
Accordingly, the NRC published a second proposed no significant hazards 
consideration determination in the Federal Register on April 25, 2017 
(82 FR 19102), which superseded the original notice in its entirety.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 27, 2017.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-333, James A. FitzPatrick 
Nuclear Power Plant, Oswego County, New York

    Date of amendment request: May 19, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17139C739.
    Brief description of amendment: The amendment revised the Emergency 
Action Level HU1.5 for James A. FitzPatrick Nuclear Power Plant by 
replacing the phrase ``Lake water level >249.2 ft'' with the phrase ``A 
hazardous event that results in on-site conditions sufficient to 
prohibit the plant staff from accessing the site via personal 
vehicles.''
    Date of issuance: June 30, 2017.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days of issuance.
    Amendment No.: 315. A publicly-available version is in ADAMS under 
Accession No. ML17153A018; 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.
    Date of initial notice in Federal Register: May 30, 2017 (82 FR 
24742).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 30, 2017.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine 
Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York

    Date of amendment request: May 19, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17139C739.
    Brief description of amendments: The amendments revised the 
Emergency Action Level HU1.5 for Nine Mile Point Nuclear Station, Units 
1 and 2, by replacing the phrase ``Lake water level >249.3 ft'' with 
the phrase ``A hazardous event that results in on-site conditions 
sufficient to prohibit the plant staff from accessing the site via 
personal vehicles.''
    Date of issuance: June 30, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 228 (Unit 1) and 162 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17152A320; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-63 and NPF-69: 
Amendments revised the Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: May 30, 2017 (82 FR 
24746).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 30, 2017.
    No significant hazards consideration comments received: Yes. The 
comment is addressed in the Safety Evaluation referenced above.

Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear 
Generating Station (Oyster Creek), Ocean County, New Jersey

    Date amendment request: February 20, 2017.
    Brief description of amendment: The amendment deleted from the 
Oyster Creek facility operating license certain license conditions that 
impose specific requirements on the decommissioning trust fund 
agreement. The provisions of 10 CFR 50.75(h) that specify the 
regulatory requirements for decommissioning trust funds will apply to 
Oyster Creek.
    Date of issuance: June 23, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 291. A publicly-available version is in ADAMS under 
Accession No. ML17067A042; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-16: Amendment revised 
the Facility Operating License.
    Date of initial notice in Federal Register: March 28, 2017 (82 FR 
15381).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 23, 2017.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear 
Power Plant, Wayne County, New York

    Date of amendment request: August 22, 2016.
    Brief description of amendment: The amendment revised Technical 
Specification 4.2.1, ``Reactor Core, Fuel Assemblies,'' and Technical 
Specification 5.6.5, ``Reporting Requirements, Core Operating Limits 
Report (COLR),'' paragraph b, to allow the use of Optimized 
ZIRLOTM fuel cladding material. The amendment is also 
supported by an exemption from certain requirements of 10 CFR 50.46 and 
10 CFR part 50, appendix K,
    Date of issuance: June 21, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 125. A publicly-available version is in ADAMS under 
Accession No. ML17131A066; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: November 8, 2016 (81 FR 
78648).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 21, 2017.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-289 and 50-320, Three 
Mile Island Nuclear Station, Units 1 and 2, Dauphin County, 
Pennsylvania

    Date of amendment request: July 15, 2016, as supplemented by letter 
dated February 13, 2017.
    Brief description of amendment: The amendment approved changes to 
the emergency plan that involve on-shift emergency response staffing 
modifications.
    Date of issuance: June 23, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 291. A publicly-available version is in ADAMS under 
Accession No. ML17137A393; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-50: Amendment revised 
the emergency plan.
    Date of initial notice in Federal Register: The license amendment 
request was originally noticed in the

[[Page 32888]]

Federal Register on October 25, 2016 (81 FR 73435). The supplement 
dated February 13, 2017, expanded the scope of the application as 
originally noticed; therefore, the NRC staff renoticed the application 
in the Federal Register on April 11, 2017 (82 FR 17458).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 23, 2017
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 27, 2016.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting 
Air,'' by removing the current stored diesel fuel oil and lube oil 
numerical volume requirements from the TS and replacing them with 
diesel operating time requirements consistent with NRC-approved 
Revision 1 to Technical Specifications Task Force (TSTF) Improved 
Standard Technical Specifications Change Traveler TSTF-501, ``Relocate 
Stored Fuel Oil and Lube Oil Volume Values to Licensee Control.''
    Date of issuance: June 29, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 177. A publicly-available version is in ADAMS under 
Accession No. ML17163A354; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-58: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 20, 2016 (81 
FR 92869).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 29, 2017
    No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station (CNS), Nemaha County, Nebraska

    Date of amendment request: August 26, 2016.
    Brief description of amendment: The amendment revised the CNS 
Technical Specifications (TSs) to eliminate TS 5.5.6, ``Inservice 
Testing Program,'' to remove requirements duplicated in the American 
Society of Mechanical Engineers Code for Operations and Maintenance of 
Nuclear Power Plants Case OMN-20, ``Inservice Test Frequency.'' A new 
defined term, ``Inservice Testing Program,'' was added to TS Section 
1.1, ``Definitions.'' The licensee stated that the change to the TSs is 
consistent with Technical Specifications Task Force (TSTF) Traveler 
TSTF-545, Revision 3, ``TS Inservice Testing Program Removal & Clarify 
SR [Surveillance Requirement] Usage Rule Application to Section 5.5 
Testing,'' which was made available to the TSTF via NRC letter dated 
December 11, 2015 (ADAMS Accession No. ML15317A071), with no proposed 
technical variations or deviations. However, in some cases, the CNS TSs 
use different section titles or numbering for SRs than the Standard 
Technical Specifications on which TSTF-545 was based. The licensee 
changed the TSTF-545 numbering to be consistent with the CNS TS 
numbering.
    Date of issuance: June 20, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 259. A publicly-available version is in ADAMS under 
Accession No. ML17144A082; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-46: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: November 8, 2016 (81 FR 
78649).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 20, 2017.
    No significant hazards consideration comments received: No.

Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota

    Date of amendment request: August 31, 2016, as supplemented by 
letter dated February 16, 2017.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.8.7 by removing the site-specific Required Actions 
and associated Completion Times, thus reverting to the standard TS 
language contained in NUREG-1431, ``Standard Technical Specifications: 
Westinghouse Plants.''
    Date of issuance: June 20, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 219 (Unit 1) and 206 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17130A716; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-42 and DPR-60: The 
amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: October 25, 2016 (81 FR 
73436). The supplemental letter dated February 16, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 20, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 30, 2016.
    Brief description of amendments: The amendments approved adoption 
of NRC-approved Technical Specifications Task Force (TSTF) Improved 
Standard Technical Specifications Change Traveler TSTF-545, Revision 3, 
``TS Inservice Testing Program Removal & Clarify SR [Surveillance 
Requirement] Usage Rule Application to Section 5.5 Testing,'' dated 
October 21, 2015. Specifically, the amendments deleted the Salem Units 
1 and 2 Technical Specification (TS) Section 6.8.4.j, ``Inservice 
Testing Program,'' and added a new defined term, ``INSERVICE TESTING 
PROGRAM,'' to the TSs. All existing references to the ``Inservice 
Testing Program'' in the Salem Units 1 and 2 TS SRs are replaced with 
``INSERVICE TESTING PROGRAM'' so that the SRs refer to the new 
definition in lieu of the deleted program.
    Date of issuance: June 28, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 319 (Unit No. 1) and 300 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML17165A214; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.

[[Page 32889]]

    Renewed Facility Operating License Nos. DPR-70 and DPR-75: The 
amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: November 8, 2016 (81 FR 
78651).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 28, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 20, 2016.
    Brief description of amendment: The amendment approved adoption of 
NRC-approved Technical Specifications Task Force (TSTF) Improved 
Standard Technical Specifications Change Traveler TSTF-545, Revision 3, 
``TS Inservice Testing Program Removal & Clarify SR [Surveillance 
Requirement] Usage Rule Application to Section 5.5 Testing,'' dated 
October 21, 2015. Specifically, the amendment deleted the Hope Creek 
Technical Specification (TS) Section 6.8.4.i, ``Inservice Testing 
Program,'' and added a new defined term, ``INSERVICE TESTING PROGRAM,'' 
to the TSs. All existing references to the ``Inservice Testing 
Program'' in the Hope Creek TS SRs are replaced with ``INSERVICE 
TESTING PROGRAM'' so that the SRs refer to the new definition in lieu 
of the deleted program.
    Date of issuance: June 28, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 205. A publicly-available version is in ADAMS under 
Accession No. ML17164A355; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-57: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: October 25, 2016 (81 FR 
73437).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 28, 2017.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Alabama Power Company, 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., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, 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: December 1, 2016.
    Brief description of amendments: The amendments modified the 
Technical Specification (TS) requirements in Section 1.3 and Section 
3.0 regarding Limiting Conditions for Operation (LCO) and Surveillance 
Requirement (SR) usage. The changes are consistent with NRC-approved 
Technical Specifications Task Force (TSTF) Traveler TSTF-529, Revision 
4, ``Clarify Use and Application Rules.''
    Date of issuance: June 27, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: Farley--211 (Unit 1) and 208 (Unit 2); Vogtle--187 
(Unit 1) and 168 (Unit 2); and Hatch--285 (Unit No. 1) and 230 (Unit 
No. 2). A publicly-available version is in ADAMS under Accession No. 
ML17137A041; 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: February 28, 2017 (82 
FR 12135).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 27, 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: July 28, 2016.
    Brief description of amendments: The amendments modified the 
technical specifications (TSs) to eliminate Section 5.5.8, ``Inservice 
Testing Program,'' for Farley and Vogtle, and eliminate Section 5.5.6, 
``Inservice Testing Program,'' for Hatch. A new defined term, 
``Inservice Testing Program,'' is added to the TS Definitions section. 
This request is consistent with Technical Specifications Task Force 
(TSTF) Traveler TSTF-545, Revision 3, ``TS Inservice Testing Program 
Removal & Clarify SR [Surveillance Requirement] Usage Rule Application 
to Section 5.5 Testing''.
    Date of issuance: June 30, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: Farley--212 (Unit 1) and 209 (Unit 2); Vogtle--187 
(Unit 1) and 170 (Unit 2); and Hatch--286 (Unit No. 1) and 231 (Unit 
No. 2). A publicly-available version is in ADAMS under Accession No. 
ML17152A218; 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: September 27, 2016 (81 
FR 66309).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 30, 2017.
    No significant hazards consideration comments received: No.

[[Page 32890]]

Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 50-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, 
Georgia

    Date of amendment request: August 29, 2016, as supplemented by 
letter dated February 13, 2017.
    Brief description of amendments: The amendments changed Combined 
License Nos. NPF-91 and NPF-92 for the Vogtle Electric Generating 
Plant, Units 3 and 4. The amendments changed the Updated Final Safety 
Analysis Report (UFSAR) in the form of departures from the incorporated 
plant-specific Design Control Document (DCD) Tier 2* information. 
Specifically, the amendment proposed changes to demonstrate the quality 
and strength of a specific population of welds between stainless steel 
mechanical couplers (couplers) and embedment plates that did not 
receive the nondestructive examinations required by the American 
Institute of Steel Construction N690-1994, ``Specification for the 
Design, Fabrication, and Erection of Steel Safety-Related Structures 
for Nuclear Facilities.'' Since some of these coupler welds are already 
installed and embedded in concrete, the licensee proposed to 
demonstrate the adequacy of these inaccessible coupler welds through 
previously-performed visual testing examinations of the couplers and 
static tension testing of a representative sample of accessible, 
uninstalled couplers produced concurrently with those already 
installed.
    Date of issuance: June 27, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 80 (Unit 3) and 79 (Unit 4). A publicly-available 
version is in ADAMS under Package Accession No. ML17107A275; documents 
related to these amendments are listed in the Safety Evaluation 
enclosed with the amendments.
    Facility Operating License No. NPF-91 and NPF-92: Amendments 
revised the UFSAR in the form of departures from the incorporated 
plant-specific DCD Tier 2* information.
    Date of initial notice in Federal Register: November 8, 2017 (81 FR 
78666). The supplement, dated February 13, 2017, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 27, 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: October 17, 2016, as supplemented by 
letter dated March 6, 2017.
    Brief description of amendments: The amendments revised selected 
Technical Specification (TS) Surveillance Requirements (SRs) for 
alternating current electrical sources because of delays in the startup 
of Watts Bar Nuclear Plant, Unit 2. Specifically, the amendments 
revised the TSs to permit a one-time extension of the specified 18-
month interval for performing the required SRs.
    Date of issuance: June 28, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 114 (Unit 1) and 12 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17138A100; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-90 and NPF-96: Amendments 
revised the Facility Operating Licenses and TSs.
    Date of initial notice in the Federal Register: February 28, 2017 
(82 FR 12138). The supplemental letter dated March 6, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 28, 2017.
    No significant hazards consideration comments received: No.

TEX Operations Company LLC, Docket Nos. 50-445 and 50-446, Comanche 
Peak Nuclear Power Plant (CPNPP), Unit Nos. 1 and 2, Somervell County, 
Texas

    Date of amendment request: December 14, 2016.
    Brief description of amendments: The amendments revised the 
licensee name from ``TEX Operations Company LLC'' to ``Vistra 
Operations Company LLC'' in the CPNPP, Unit No. 1, Facility Operating 
License (FOL) NPF-87; CPNPP, Unit No. 2, FOL (NPF-89); and the title 
page of the Environmental Protection Plan.
    Date of issuance: June 29, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 169 (Unit 1) and 169 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17129A024; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: February 28, 2017 (82 
FR 12139).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 29, 2017.
    No significant hazards consideration comments received: No.

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

    For the Nuclear Regulatory Commission.
Eric J. Benner,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2017-14743 Filed 7-17-17; 8:45 am]
 BILLING CODE 7590-01-P