Biweekly Notice: Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 27882-27895 [2017-12732]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 82, Number 116 (Monday, June 19, 2017)]
[Notices]
[Pages 27882-27895]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-12732]


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NUCLEAR REGULATORY COMMISSION

[NRC-2017-0140]


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 May 23, 2017, to June 2, 2017. The last 
biweekly notice was published on June 6, 2017.

DATES: Comments must be filed by July 19, 2017. A request for a hearing 
must be filed by August 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-2017-0140. 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-2017-0140 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-2017-0140.
     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/

[[Page 27883]]

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-2017-0140 facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

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

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

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

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and petition for leave to intervene 
(petition) with respect to the action. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested persons should consult a 
current copy of 10 CFR 2.309. The NRC's regulations are accessible 
electronically from the NRC Library on the NRC's Web site at https://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of 
the regulations is available at the NRC's Public Document Room, located 
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. If a petition is filed, the 
Commission or a presiding officer will rule on the petition and, if 
appropriate, a notice of a hearing will be issued.
    As required by 10 CFR 2.309(d) the petition should specifically 
explain the reasons why intervention should be permitted with 
particular reference to the following general requirements for 
standing: (1) The name, address, and telephone number of the 
petitioner; (2) the nature of the petitioner's right under the Act to 
be made a party to the proceeding; (3) the nature and extent of the 
petitioner's property, financial, or other interest in the proceeding; 
and (4) the possible effect of any decision or order which may be 
entered in the proceeding on the petitioner's interest.
    In accordance with 10 CFR 2.309(f), the petition must also set 
forth the specific contentions which the petitioner seeks to have 
litigated in the proceeding. Each contention must consist of a specific 
statement of the issue of law or fact to be raised or controverted. In 
addition, the petitioner must provide a brief explanation of the bases 
for the contention and a concise statement of the alleged facts or 
expert opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to the specific sources and 
documents on which the petitioner intends to rely to support its 
position on the issue. The petition must include sufficient information 
to show that a genuine dispute exists with the applicant or licensee on 
a material issue of law or fact. Contentions must be limited to matters 
within the scope of the proceeding. The contention must be one which, 
if proven, would entitle the petitioner to relief. A petitioner who 
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene. 
Parties have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that party's admitted 
contentions, including the opportunity to present evidence, consistent 
with the NRC's regulations, policies, and procedures.
    Petitions must be filed no later than 60 days from the date of 
publication of this notice. Petitions and motions for leave to file new 
or amended contentions that are filed after the deadline will not be 
entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in 
accordance with the filing instructions in the ``Electronic Submissions 
(E-Filing)'' section of this document.

[[Page 27884]]

    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to establish when the hearing is held. If the final determination is 
that the amendment request involves no significant hazards 
consideration, the Commission may issue the amendment and make it 
immediately effective, notwithstanding the request for a hearing. Any 
hearing would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of the amendment unless the Commission finds an imminent 
danger to the health or safety of the public, in which case it will 
issue an appropriate order or rule under 10 CFR part 2.
    A State, local governmental body, Federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission by 
August 18, 2017. The petition must be filed in accordance with the 
filing instructions in the ``Electronic Submissions (E-Filing)'' 
section of this document, and should meet the requirements for 
petitions set forth in this section, except that under 10 CFR 
2.309(h)(2) a State, local governmental body, or federally recognized 
Indian Tribe, or agency thereof does not need to address the standing 
requirements in 10 CFR 2.309(d) if the facility is located within its 
boundaries. Alternatively, a State, local governmental body, Federally-
recognized Indian Tribe, or agency thereof may participate as a non-
party under 10 CFR 2.315(c).
    If a hearing is granted, any person who is not a party to the 
proceeding and is not affiliated with or represented by a party may, at 
the discretion of the presiding officer, be permitted to make a limited 
appearance pursuant to the provisions of 10 CFR 2.315(a). A person 
making a limited appearance may make an oral or written statement of 
his or her position on the issues but may not otherwise participate in 
the proceeding. A limited appearance may be made at any session of the 
hearing or at any prehearing conference, subject to the limits and 
conditions as may be imposed by the presiding officer. Details 
regarding the opportunity to make a limited appearance will be provided 
by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing and petition for leave to intervene (petition), any 
motion or other document filed in the proceeding prior to the 
submission of a request for hearing or petition to intervene, and 
documents filed by interested governmental entities that request to 
participate under 10 CFR 2.315(c), must be filed in accordance with the 
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 
46562, August 3, 2012). The E-Filing process requires participants to 
submit and serve all adjudicatory documents over the internet, or in 
some cases to mail copies on electronic storage media. Detailed 
guidance on making electronic submissions may be found in the Guidance 
for Electronic Submissions to the NRC and on the NRC Web site at https://www.nrc.gov/site-help/e-submittals.html. Participants may not submit 
paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to (1) request a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign submissions and access the E-Filing 
system for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a petition or 
other adjudicatory document (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a 
digital ID certificate and a docket has been created, the participant 
can then submit adjudicatory documents. Submissions must be in Portable 
Document Format (PDF). Additional guidance on PDF submissions is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the 
time the document is submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
document on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before adjudicatory documents are 
filed so that they can obtain access to the documents via the E-Filing 
system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic 
Filing Help Desk through the ``Contact Us'' link located on the NRC's 
public Web site at https://www.nrc.gov/site-help/e-submittals.html, by 
email to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m. 
and 6 p.m., Eastern Time, Monday through Friday, excluding government 
holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
stating why there is good cause for not filing electronically and 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff. Participants filing adjudicatory documents in this 
manner are responsible for serving the document on all other 
participants. Filing is considered complete by first-class mail as of 
the time of deposit in the mail, or

[[Page 27885]]

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.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3), Westchester 
County, New York

    Date of amendment request: December 14, 2016, as supplemented by 
letter dated April 19, 2017. Publicly available versions are in ADAMS 
under Package Accession No. ML16355A066 and Accession No. ML17114A467, 
respectively.
    Description of amendment request: The amendments would revise the 
Appendix C Technical Specifications (TS) Limiting Condition for 
Operation (LCO) 3.1.2 for IP2 and IP3 and Appendix A TS LCO 3.7.13 for 
IP2. These LCOs ensure that the fuel to be loaded into the Shielded 
Transfer Canister (STC) meets the design basis for the STC and has an 
acceptable rack location in the IP2 spent fuel pit before the STC is 
loaded with fuel. The proposed changes to these LCOs would increase the 
population of IP3 fuel eligible for transfer to the IP2 spent fuel pit 
and maintain full core offload capability for IP3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with NRC staff's edits in 
square brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would modify the IP2 and IP3 Technical 
Specifications (TS) to incorporate the results of revised 
criticality, thermal, and shielding and dose analyses and 
evaluations.
    [For IP2,] the proposed amendment was evaluated for impact on 
the following previously evaluated events and accidents: STC 
Criticality Accidents, SFP Criticality Accidents, Boron Dilution 
Accidents, Fuel Handling Accidents, Loss of Spent Fuel Pool [SFP] 
Cooling, and Natural Events.

[IP2] STC Criticality Accidents

    The STC criticality accident considered were: Abnormal 
temperature, dropped, mislocated, and misloaded fuel assemblies, and 
misalignment between the active fuel region and the neutron 
absorber.
    The probability of an STC criticality accident will not increase 
significantly due to the proposed changes because the individual 
fuel assemblies will be loaded into the STC in the same manner, 
using the same equipment, procedures, and other administrative 
controls (i.e. fuel move sheets) that are currently used.
    The consequences of an STC criticality accident are not changed 
because the reactivity analysis demonstrates that the same 
subcriticality criteria and requirements continue to be met for 
these accidents.

[IP2] SFP Criticality Accidents

    The SFP criticality accident of record considered the following 
accidents (1) a dropped fuel assembly or an assembly placed 
alongside a rack, (2) a misloaded fuel assembly, and (3) abnormal 
heat loads. Because the IP2 and IP3 fuel assemblies are identical 
[with] regards [to] those parameters that are utilized in the design 
basis criticality analysis (DBA) to qualify fresh fuel these 
accidents are bounding for IP3 fuel.
    The probability of an SFP criticality accident will not increase 
significantly due to the proposed changes because the individual 
fuel assemblies will be loaded into the SFP in the same manner, 
using the same equipment, procedures, and other administrative 
controls (i.e. fuel move sheets) that are currently used.
    The consequences of an SFP criticality accident are not changed 
because the reactivity analysis demonstrates that the same 
subcriticality criteria and requirements continue to be met for this 
accident.

[IP2] STC Thermal Accidents

    The thermal analyses demonstrate that the postulated accidents 
(rupture of the HI-TRAC water jacket, 50-gallon transported fuel 
tank rupture and fire, simultaneous loss of water from the water 
jacket and HI-TRAC annulus, fuel misload, hypothetical tipover, and 
crane malfunction) continue to meet their acceptance criteria.
    The probability of an STC thermal accident will not increase 
significantly because the individual fuel assemblies will be loaded 
into the SFP in the same manner, using the same equipment, 
procedures, and other administrative controls (i.e. fuel move 
sheets) that are currently used.
    The consequences of an STC thermal accident will not increase 
significantly because the thermal analysis demonstrates that the 
same thermal acceptance criteria and requirements continue to be met 
for this accident.

[IP2] Boron Dilution Accident

    The probability of a boron dilution event remains the same 
because the proposed change does not alter the manner in which the 
IP2 spent fuel cooling system or any other plant system is operated, 
or otherwise increase the likelihood of adding significant 
quantities of unborated water into the spent fuel pit.
    The consequences of the boron dilution event remains the same. 
The reactivity of the STC filled with the most reactive combination 
of approved fuel assemblies in unborated water results in a 
keff less than 0.95. Thus, even in the unlikely event of 
a complete dilution of the spent fuel pit water, the STC will remain 
safely subcritical.

[IP2] Fuel Handling Accident

    The probability of an FHA will not increase significantly due to 
the proposed changes because the individual fuel assemblies will be 
moved between the STC and the spent fuel pit racks and the STC and 
HI-TRAC will be moved in the same manner, using the same equipment, 
procedures, and other administrative controls (i.e. fuel move 
sheets) that are currently used.
    The consequences of the existing fuel handling accident remain 
bounding because the IP3 fuel assembly design is essentially the 
same as the IP2 design and the IP3 fuel assemblies to be transferred 
to IP2 will be cooled a minimum of 6 years. This compares with a 
cooling time of 84 hours used in the existing FHA radiological 
analysis. The 6-year cooling time results in a significant reduction 
in the radioactive source term available for release from a damaged 
fuel assembly compared to the source term considered in the design 
basis FHA radiological analysis. The consequences of the previously 
analyzed fuel assembly drop accident, therefore, continue to provide 
a

[[Page 27886]]

bounding estimate of offsite dose for this accident.

[IP2] Loss of Spent Fuel Pool Cooling

    The probability of a loss of spent fuel pit cooling remains the 
same because the proposed change does not alter the manner in which 
the IP2 spent fuel cooling loop is operated, designed or maintained.
    The consequences of a loss of spent fuel pit cooling remains the 
same because the thermal design basis for the spent fuel pit cooling 
loop provides for all fuel pit rack locations to be filled at the 
end of a full core discharge and therefore the design basis heat 
load effectively includes any heat load associated with the 
assemblies within the STC.

[IP2] Natural Events

    The natural events considered include the following accidents 
(1) a seismic event, (2) high winds, tornado and tornado missiles, 
(3) flooding and (4) a lightning strike.
    The probability of natural event will not increase due to the 
proposed changes because there are no elements of the proposed 
changes that influence the occurrence of any natural event.
    The consequences of a natural event will not increase due to the 
proposed changes because the structural analyses design limits 
continue to be met. A lightning strike may cause ignition of the VCT 
fuel but this event is addressed under STC thermal accidents.
    [For IP3,] the proposed amendment was evaluated for impact on 
the following previously evaluated events and accidents: STC 
Criticality Accidents, SFP Criticality Accidents, Boron Dilution 
Accidents, Fuel Handling Accidents, Loss of Spent Fuel Pool Cooling, 
and Natural Events.

[IP3] STC Criticality Accidents

    The STC criticality accident considered were: Abnormal 
temperature, dropped, mislocated, and misloaded fuel assemblies, and 
misalignment between the active fuel region and the neutron 
absorber.
    The probability of an STC criticality accident will not increase 
significantly due to the proposed changes because the individual 
fuel assemblies will be loaded into the STC in the same manner, 
using the same equipment, procedures, and other administrative 
controls (i.e. fuel move sheets) that are currently used.
    The consequences of an STC criticality accident are not changed 
because the reactivity analysis demonstrates that the same 
subcriticality criteria and requirements continue to be met for 
these accidents.

[IP3] STC Thermal Accidents

    The thermal analyses demonstrate that the postulated accidents 
(rupture of the HI-TRAC water jacket, 50-gallon transported fuel 
tank rupture and fire, simultaneous loss of water from the water 
jacket and HI-TRAC annulus, fuel mislead, hypothetical tipover, and 
crane malfunction) continue to meet their acceptance criteria. The 
probability of an STC thermal accident will not increase 
significantly because the individual fuel assemblies will be loaded 
into the SFP in the same manner, using the same equipment, 
procedures, and other administrative controls (i.e. fuel move 
sheets) that are currently used.
    The consequences of an STC thermal accident will not increase 
significantly because the thermal analysis demonstrates that the 
same thermal acceptance criteria and requirements continue to be met 
for this accident.

[IP3] Boron Dilution Accident

    The probability of a boron dilution event remains the same 
because the proposed change does not alter the manner in which the 
IP3 spent fuel cooling system or any other plant system is operated, 
or otherwise increase the likelihood of adding significant 
quantities of unborated water into the spent fuel pit.
    The consequences of the boron dilution event remains the same. 
The reactivity of the STC filled with the most reactive combination 
of approved fuel assemblies in unborated water results in a 
keff less than 0.95. Thus, even in the unlikely event of 
a complete dilution of the spent fuel pit water, the STC will remain 
safely subcritical.

[IP3] Fuel Handling Accident

    The probability of an FHA will not increase significantly due to 
the proposed changes because the individual fuel assemblies will be 
moved between the STC and the spent fuel pit racks and the STC and 
HI-TRAC will be moved in the same manner, using the same equipment, 
procedures, and other administrative controls (i.e. fuel move 
sheets) that are currently used.
    The consequences of the existing fuel handling accident remain 
bounding because only IP3 fuel is moved in the IP3 spent fuel pit. 
The IP3 fuel assemblies to be transferred to IP2 will be cooled a 
minimum of 6 years. This compares with a cooling time of 84 hours 
used in the existing FHA radiological analysis. The 6-year cooling 
time results in a significant reduction in the radioactive source 
term available for release from a damaged fuel assembly compared to 
the source term considered in the design basis FHA radiological 
analysis. The consequences of the previously analyzed fuel assembly 
drop accident, therefore, continue to provide a bounding estimate of 
offsite dose for this accident.

[IP3] Loss of Spent Fuel Pool Cooling

    The probability of a loss of spent fuel pit cooling remains the 
same because the proposed change does not alter the manner in which 
the IP3 spent fuel cooling loop is operated, designed or maintained.
    The consequences of a loss of spent fuel pit cooling remains the 
same because the thermal design basis for the spent fuel pit cooling 
loop provides for all fuel pit rack locations to be filled at the 
end of a full core discharge and therefore the design basis heat 
load effectively includes any heat load associated with the 
assemblies within the STC.

[IP3] Natural Events

    The natural events considered include the following accidents 
(1) a seismic event, (2) high winds, tornado and tornado missiles, 
(3) flooding and (4) a lightning strike.
    The probability of natural event will not increase due to the 
proposed changes because there are no elements of the proposed 
changes that influence the occurrence of any natural event.
    The consequences of a natural event will not increase due to the 
proposed changes because the structural analyses design limits 
continue to be met. A lightning strike may cause ignition of the VCT 
fuel but this event is addressed under STC thermal accidents.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. No new modes of operation are introduced by the proposed 
changes. The proposed changes will not create any failure mode not 
bounded by previously evaluated accidents.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident, from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment would modify the TS to incorporate the 
results of revised criticality, thermal and shield and dose 
analyses. The margin of safety required by 10 CFR 50.58(b)(4) 
remains unchanged. New criticality evaluations for both the STC [and 
the IP2 SFP] confirm that operation in accordance with the proposed 
amendment continues to meet the required subcriticality margins. The 
thermal analyses demonstrate that the postulated accidents (rupture 
of the HI-TRAC water jacket, 50-gallon transported fuel tank rupture 
and fire, simultaneous loss of water from the water jacket and HI-
TRAC annulus, fuel misload, hypothetical tipover, and crane 
malfunction) continue to meet their acceptance criteria without a 
significant loss of safety margin. The shielding and dose analyses 
demonstrate that the shielding and radiation protection requirements 
continue to be met without a significant loss of safety margin.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

[[Page 27887]]

Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert 
Cliffs Nuclear Power Plant, Units 1 and 2, Calvert County, Maryland

    Date of amendment request: March 28, 2017. A publicly available 
version is in ADAMS under Accession No. ML17087A374.
    Description of amendment request: The amendments would revise the 
Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Technical 
Specifications (TSs) to change the low level of the refueling water 
tank (RWT) to reflect a needed increase in the required borated water 
volume and change the allowable value of the RWT level-low function.
    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 any accident previously 
evaluated?
    Response: No.
    The proposed amendment increases the required volume of water in 
the RWT to maintain the existing design requirements. The increase 
is necessary due to an increase in the RWT Level--Low RAS 
[recirculation actuation signal] setpoint, which allows more water 
to stay in the tank following a LOCA [loss-of-coolant accident]. The 
modification to the allowable value of the RWT level-low (function 
5a) resolves a non-conservative TS per the guidance of 
Administrative Letter 98-10 ``Dispositioning of Technical 
Specifications That Are Insufficient to Assure Plant Safety.''
    The RWT is not an accident initiator. The RWT is required to 
supply adequate borated water to perform its mitigation function as 
assumed in the accident analyses. With the proposed increase in the 
minimum required water volume, the RWT maintains its design margin 
for supplying the required amount of borated water to the reactor 
core and the containment sump.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    The proposed amendment increases the required volume of water in 
the RWT to maintain the existing design requirements. The increase 
is necessary due to an increase in the RWT Level--Low RAS setpoint, 
which allows more water to stay in the tank following a LOCA. The 
modification to the allowable value of the RWT level-low (function 
5a) resolves a non-conservative TS per the guidance of 
Administrative Letter 98-10 ``Dispositioning of Technical 
Specifications That Are Insufficient to Assure Plant Safety.''
    The proposed amendment does not impose any new or different 
requirements. The change does not alter assumptions made in the 
safety analyses. The proposed change is consistent with the safety 
analyses assumptions and current plant operating practice.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment increases the required volume of water in 
the RWT to maintain the existing design requirements. The increase 
is necessary due to an increase in the RWT Level--Low RAS setpoint, 
which allows more water to stay in the tank following a loss-of-
coolant accident. The modification to the allowable value of the RWT 
level-low (function 5a) resolves a non-conservative TS per the 
guidance of Administrative Letter 98-10 ``Dispositioning of 
Technical Specifications That Are Insufficient to Assure Plant 
Safety.''
    The proposed amendment does not affect the design, operation, 
and testing methods for systems, structures and components specified 
in applicable codes and standards (or alternatives approved for use 
by the NRC). With the proposed increase in the minimum required 
water volume, the RWT maintains its design margin for supplying the 
required amount of borated water to the reactor core and the 
containment sump. The RWT will continue to meet all of its 
requirements as described in the plant licensing basis (including 
the Updated Final Safety Analysis Report and the TS Bases). 
Similarly, there is no impact to Safety Analysis acceptance criteria 
as described in the plant licensing basis.
    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: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: James G. Danna.

Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: April 5, 2017. A publicly available 
version is in ADAMS under Accession No. ML17095A081.
    Description of amendment request: The amendment would revise the 
Nine Mile Point Nuclear Station, Unit 2, Technical Specifications to 
allow greater flexibility in performing surveillance testing in Modes 
1, 2, or 3 of emergency diesel generators and Class 1E batteries. The 
proposed changes are based on Technical Specifications Task Force 
(TSTF) Traveler TSTF-283-A, Revision 3, ``Modify Section 3.8 Mode 
Restriction Notes.''
    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 modify Mode restriction Notes to allow 
performance of the Surveillance in whole or in part to reestablish 
Emergency Diesel Generator (EDG) Operability, and to allow the 
crediting of unplanned events that satisfy the Surveillances. The 
EDGs and their associated emergency loads are accident mitigating 
features, and are not an initiator of any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not significantly increased. To manage any increase in 
risk, the proposed changes require an assessment to verify that 
plant safety will be maintained or enhanced by performance of the 
Surveillance in the current prohibited Modes. The radiological 
consequences of an accident previously evaluated during the period 
that the EDG is being tested to reestablish operability are no 
different from the radiological consequences of an accident 
previously evaluated while the EDG is inoperable. As a result, the 
consequences of any accident previously evaluated are not increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any 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 physical alteration to the 
plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The purpose of Surveillances is to verify that equipment is 
capable of performing its assumed safety function. The proposed 
changes will only allow the performance of the Surveillances to 
reestablish Operability,

[[Page 27888]]

and the proposed changes may not be used to remove an EDG from 
service. In addition, the proposed changes will potentially shorten 
the time that an EDG is unavailable because testing to reestablish 
Operability can be performed without a plant shutdown. The proposed 
changes also require an assessment to verify that plant safety will 
be maintained or enhanced by performance of the Surveillance in the 
normally prohibited Modes.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: April 27, 2017. A publicly available 
version is in ADAMS under Accession No. ML17121A449.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 5.5.12, ``Primary Containment Leakage 
Rate Testing Program,'' to allow for the permanent extension of the 
Type A integrated leak rate testing and Type C leak rate testing 
frequencies, and would also delete a one-time exception.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed activity involves revision of the Quad Cities 
Nuclear Power Station (QCNPS) Technical Specification (TS) 5.5.12, 
Primary Containment Leakage Rate Testing Program, to allow the 
extension of the QCNPS, Units 1 and 2, Type A containment integrated 
leakage rate test interval to 15 years, and the extension of the 
Type C local leakage rate test interval to 75 months. The current 
Type A test interval of 120 months (10 years) would be extended on a 
permanent basis to no longer than 15 years from the last Type A 
test. The existing Type C test interval of 60 months for selected 
components would be extended on a performance basis to no longer 
than 75 months. Extensions of up to nine months (total maximum 
interval of 84 months for Type C tests) are permissible only for 
non-routine emergent conditions.
    The proposed extension does not involve either a physical change 
to the plant or a change in the manner in which the plant is 
operated or controlled. The containment is designed to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. As such, 
the containment and the testing requirements invoked to periodically 
demonstrate the integrity of the containment exist to ensure the 
plant's ability to mitigate the consequences of an accident, and do 
not involve the prevention or identification of any precursors of an 
accident.
    The change in dose risk for changing the Type A Integrated Leak 
Rate Test (ILRT) interval from three-per-ten years to once-per-
fifteen-years, measured as an increase to the total integrated dose 
risk for all internal events accident sequences for QCNPS, is 1.0E-
02 person-rem/yr (0.31%) using the Electric Power Research Institute 
(EPRI) guidance with the base case corrosion included. The change in 
dose risk drops to 2.7E-03 person-rem/yr (0.08%) when using the EPRI 
Expert Elicitation methodology. The values calculated per the EPRI 
guidance are all lower than the acceptance criteria of less than or 
equal to 1.0 person-rem/yr or less than 1.0% person-rem/yr defined 
in Section 1.3 of Attachment 3 to this LAR. Therefore, this proposed 
extension does not involve a significant increase in the probability 
of an accident previously evaluated.
    As documented in NUREG-1493, ``Performance-Based Containment 
Leak-Test Program,'' dated January 1995, Types B and C tests have 
identified a very large percentage of containment leakage paths, and 
the percentage of containment leakage paths that are detected only 
by Type A testing is very small. The QCNPS, Units 1 and 2 Type A 
test history supports this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as: (1) Activity based, 
and, (2) time based. Activity based failure mechanisms are defined 
as degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment combined with the containment inspections performed in 
accordance with American Society of Mechanical Engineers (ASME) 
Section XI, and TS requirements serve to provide a high degree of 
assurance that the containment would not degrade in a manner that is 
detectable only by a Type A test. Based on the above, the proposed 
test interval extensions do not significantly increase the 
consequences of an accident previously evaluated.
    The proposed amendment also deletes an exception previously 
granted in amendments 220 and 214 to allow one-time extensions of 
the ILRT test frequency for QCNPS, Units 1 and 2, respectively. This 
exception was for an activity that has already taken place; 
therefore, this deletion is solely an administrative action that 
does not result in any change in how QCNPS, Units 1 and 2 are 
operated.
    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 amendment to TS 5.5.12, ``Primary Containment 
Leakage Rate Testing Program,'' involves the extension of the QCNPS, 
Units 1 and 2 Type A containment test interval to 15 years and the 
extension of the Type C test interval to 75 months. The containment 
and the testing requirements to periodically demonstrate the 
integrity of the containment exist to ensure the plant's ability to 
mitigate the consequences of an accident.
    The proposed change does not involve a physical modification to 
the plant (i.e., no new or different type of equipment will be 
installed), nor does it alter the design, configuration, or change 
the manner in which the plant is operated or controlled beyond the 
standard functional capabilities of the equipment.
    The proposed amendment also deletes an exception previously 
granted under TS Amendments 220 and 214 to allow the one-time 
extension of the ILRT test frequency for QCNPS, Units 1 and 2, 
respectively. This exception was for an activity that has already 
taken place; therefore, this deletion is solely an administrative 
action that does not result in any change in how the QCNPS units are 
operated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment to TS 5.5.12 involves the extension of 
the QCNPS, Units 1 and 2 Type A containment test interval to 15 
years and the extension of the Type C test interval to 75 months for 
selected components. This amendment does not alter the manner in 
which safety limits, limiting safety system set points, or limiting 
conditions for operation are determined. The specific requirements 
and conditions of the TS Containment Leak Rate Testing Program exist 
to ensure that the degree of containment structural integrity and 
leak-tightness that is considered in the plant safety analysis is 
maintained. The overall containment leak rate limit specified by TS 
is maintained.
    The proposed change involves the extension of the interval 
between Type A containment leak rate tests and Type C tests for 
QCNPS, Units 1 and 2. The proposed surveillance interval extension 
is bounded by the 15-year ILRT interval and the 75-month Type C test 
interval currently authorized

[[Page 27889]]

within NEI 94-01, Revision 3-A. Industry experience supports the 
conclusion that Types B and C testing detects a large percentage of 
containment leakage paths and that the percentage of containment 
leakage paths that are detected only by Type A testing is small. The 
containment inspections performed in accordance with ASME Section Xl 
and TS serve to provide a high degree of assurance that the 
containment would not degrade in a manner that is detectable only by 
Type A testing. The combination of these factors ensures that the 
margin of safety in the plant safety analysis is maintained. The 
design, operation, testing methods and acceptance criteria for Types 
A, B, and C containment leakage tests specified in applicable codes 
and standards would continue to be met, with the acceptance of this 
proposed change, since these are not affected by changes to the Type 
A and Type C test intervals.
    The proposed amendment also deletes exceptions previously 
granted to allow one-time extensions of the ILRT test frequency for 
QCNPS, Units 1 and 2. This exception was for an activity that has 
taken place; therefore, the deletion is solely an administrative 
action and does not change how QCNPS is operated and maintained. 
Thus, there is no reduction in any margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: April 9, 2017. A publicly available 
version is in ADAMS under Accession No. ML17101A637.
    Description of amendment request: The amendments would modify the 
Technical Specifications (TSs) to remove various reporting 
requirements. Specifically, the amendments would remove the 
requirements to prepare various special reports, the Startup Report, 
and the Annual Report. In addition, the amendments would revise the TSs 
to remove the completion time for restoring spent fuel pool water level 
to address inoperability of one of the two parallel flow paths in the 
residual heat removal or safety injection headers for the Emergency 
Core Cooling Systems and to make other administrative changes, 
including updating plant staff and responsibilities and correcting a 
misspelling.
    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 actions, surveillance requirements, and administrative 
controls associated with the proposed changes to the technical 
specifications (TS) are not initiators of any accidents previously 
evaluated, so the probability of accidents previously evaluated is 
unaffected by the proposed changes. The proposed changes do not 
alter the design, function, operation, or configuration of any plant 
structure, system, or component (SSC). The capability of any 
operable TS-required SSC to perform its specified safety function is 
not impacted by the proposed changes. As a result, the outcomes of 
accidents previously evaluated are unaffected. Therefore, the 
proposed changes do not result in a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not challenge the integrity or 
performance of any safety-related systems. No plant equipment is 
installed or removed, and the changes do not alter the design, 
physical configuration, or method of operation of any plant SSC. No 
physical changes are made to the plant, so no new causal mechanisms 
are introduced. Therefore, the proposed changes to the TS do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The ability of any operable SSC to perform its designated safety 
function is unaffected by the proposed changes. The proposed changes 
do not alter any safety analyses assumptions, safety limits, 
limiting safety system settings, or method of operating the plant. 
The changes do not adversely impact plant operating margins or the 
reliability of equipment credited in the safety analyses. Therefore, 
the proposed changes do not involve a significant reduction in the 
margin of safety.

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

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

    Date of amendment request: April 20, 2017. A publicly available 
version is in ADAMS under Accession No. ML17111A631.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) Section 3.1.2, ``Reactivity 
Anomalies,'' with a change to the method of calculating core reactivity 
for the purpose of performing the reactivity anomaly surveillance. The 
proposed change would allow performance of the reactivity anomaly 
surveillance on a comparison of monitored to predicted core reactivity. 
The reactivity anomaly verification is currently determined by a 
comparison of monitored versus predicted control rod density.
    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 does not affect any plant systems, 
structures, or components designed for the prevention or mitigation 
of previously evaluated accidents. The proposed change would only 
modify how the reactivity anomaly surveillance is performed. 
Verifying that the core reactivity is consistent with predicted 
values ensures that accident and transient safety analyses remain 
valid. This amendment changes the TS requirements such that, rather 
than performing the surveillance by comparing monitored to predicted 
control rod density, the surveillance is performed by a direct 
comparison of core keff. Present day on-line core 
monitoring systems, such as 3D MONICORE and ACUMEN, are capable of 
performing the direct measurement of reactivity.
    Therefore, since the reactivity anomaly surveillance will 
continue to be performed by a viable method, the proposed change 
does not involve a significant increase in the probability or 
consequence of a previously evaluated accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.

[[Page 27890]]

    The proposed change does not involve any changes to the 
operation, testing, or maintenance of any safety-related, or 
otherwise important to safety systems. All systems important to 
safety will continue to be operated and maintained within their 
design bases. The proposed changes to the Reactivity Anomalies TS 
will only provide a new, more efficient method of detecting an 
unexpected change in core reactivity.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change is to modify the method for performing the 
reactivity anomaly surveillance from a comparison of monitored to 
predicted control rod density to a comparison of monitored to 
predicted core keff. The direct comparison of 
keff provides a technically superior method of 
calculating any differences in the expected core reactivity. The 
reactivity anomaly surveillance will continue to be performed at the 
same frequency as is currently required by the TS, only the method 
of performing the surveillance will be changed. Consequently, core 
reactivity assumptions made in safety analyses will continue to be 
adequately verified. The proposed change has no impact to the margin 
of safety.

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

NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant (Point Beach), Units 1 and 2, Town of Two Creeks, 
Manitowoc County, Wisconsin

    Date of amendment request: March 31, 2017. A publicly available 
version is in ADAMS under Accession No. ML17090A511.
    Description of amendment request: The amendments would document a 
risk-informed resolution strategy to resolve low risk, legacy design 
code non-conformances associated with construction trusses in the 
containment buildings of Point Beach, Units 1 and 2. The proposed 
license amendment request (LAR) is a risk-informed licensing basis 
change. The proposed change is acceptance of the final configuration of 
the construction trusses, including the attached containment spray 
piping and ventilation ductwork, and the containment liners/walls 
adjacent to the trusses, using a risk-informed resolution. Accordingly, 
the proposed change meets the criteria set forth in Regulatory Guide 
(RG) 1.174, ``An Approach for Using Probabilistic Risk Assessment [PRA] 
in Risk-Informed Decisions on Plant-Specific Changes to the Licensing 
Basis,'' and the generic guidance in RG 1.200, ``An Approach for 
Determining the Technical Adequacy of Probabilistic Risk Assessment 
Results for Risk-Informed Activities.''
    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 probability of an accident previously evaluated is not 
changed. The containment structures and the containment spray piping 
and ventilation ducts attached to the construction trusses are 
accident mitigation equipment. They are not accident initiators.
    The acceptance of the final configuration of Point Beach Units 1 
and 2 results in a change in core damage frequency and large early 
release frequency that is within acceptance guidelines and does not 
involve a significant reduction in the margin of safety. Although 
failures are postulated in the PRA analysis, the engineering 
calculations in support of the LAR conclude that the construction 
trusses and the associated structures/components remain structurally 
sound in the event of a design basis seismic or thermal event and 
there is no adverse impact or change to any station SSC's 
[structure, system, and components] design function and there is no 
change to accident mitigation response.
    This change has no impact on station fire risk caused by a 
seismic event.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not install any new or different type 
of equipment in the plant. The proposed change does not create any 
new failure modes for existing equipment or any new limiting single 
failures. Engineering calculations conclude the construction 
trusses, equipment supported by the trusses, and containment liners 
remain capable of withstanding design basis seismic and thermal 
events and remain capable of performing their designated design 
functions. Additionally, the proposed change does not involve a 
change in the methods governing normal plant operation, and all 
safety functions will continue to perform as previously assumed in 
the accident analyses. Thus, the proposed change does not adversely 
affect the design function or operation of any structures, systems 
and components important to safety.
    There are no new accidents identified associated with acceptance 
of the final modified configuration of Unit 1 and the current 
configuration of Unit 2.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The effects of the change, [Delta]CDF [core damage frequency] 
and [Delta]LERF, [large early release frequency] are within the 
acceptance guidelines shown in Figures 4 and 5 of Regulatory Guide 
1.174. Consequently, the change does not result in a significant 
reduction in the margin of safety.
    The containment structures and liners, construction trusses, and 
equipment supported by the trusses remain fully capable of 
performing their specified design functions as concluded by 
supporting engineering calculations.
    Modifications associated with implementation of NFPA [National 
Fire Protection Association] 805 are planned that will provide 
protection of the reactor coolant system feed and bleed capability 
and result in additional safety margin.
    The proposed change does not affect the margin of safety 
associated with confidence in the ability of the fission product 
barriers (i.e., fuel cladding, reactor coolant system pressure 
boundary, and containment structure) to limit the level of radiation 
dose to the public. The proposed change does not alter any safety 
analyses assumptions, safety limits, limiting safety system 
settings, or methods of operating the plant. The changes do not 
adversely impact the reliability of equipment credited in the safety 
analyses. The proposed change does not adversely affect systems that 
respond to safely shutdown the plant and to maintain the plant in a 
safe shutdown condition.
    The station will implement new seismic and thermal event limits 
to ensure the construction trusses and associated equipment are 
inspected and/or analyzed for any event exceeding elastic stress 
limits to determine their capability to withstand a subsequent 
design basis event prior to Unit restart.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William Blair, Managing Attorney--Nuclear, 
Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard, 
Juno Beach, FL 33408-0420.
    NRC Branch Chief: David J. Wrona.

[[Page 27891]]

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: April 27, 2017. A publicly available 
version is in ADAMS under Accession No. ML17118A049.
    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 
(Tier 1) changes require additional changes to corresponding Tier 2 
component data information in Updated Final Safety Analysis Report 
(UFSAR) Chapter 9. Because this proposed change requires a departure 
from Tier 1 information in the Westinghouse Electric Company's AP1000 
Design Control Document, the licensee also requested an exemption from 
the requirements of the Generic Design Control Document 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 (VAS) unit coolers and reduced chilled 
water system (VWS) cooling coil flow rates do not adversely impact 
the 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 VWS safety-related 
function of containment isolation is not 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 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.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, 50-296, and 72-
052, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, and 
Independent Spent Fuel Storage Installation (ISFSI), Limestone County, 
Alabama

Tennessee Valley Authority, Docket Nos. 50-327, 50-328, and 72-034, 
Sequoyah Nuclear Plant (SQN), Units 1 and 2, and ISFSI, Hamilton 
County, Tennessee

Tennessee Valley Authority (TVA), Docket Nos. 50-390, 50-391, and 72-
1048, Watts Bar Nuclear Plant (WBN), Units 1 and 2, and ISFSI, Rhea 
County, Tennessee

    Date of amendment request: January 4, 2017. A publicly available 
version is in ADAMS under Accession No. ML17004A340.
    Description of amendment request: The amendments would modify the 
Emergency Plans for BFN, Units 1, 2, and 3, and its ISFSI; SQN, Units 1 
and 2, and its ISFSI; and WBN, Units 1 and 2, and its ISFSI, to adopt 
the Emergency Action Level (EAL) schemes based on Nuclear Energy 
Institute (NEI) 99-01, Revision 6, which has been endorsed by the NRC 
as documented in a letter dated March 28, 2013 (ADAMS Accession No. 
ML12346A463). The proposed changes to TVA's EAL schemes to adopt the 
guidance in NEI 99-01, Revision 6, do not reduce the capability to meet 
the emergency planning requirements established in 10 CFR 50.47 and 10 
CFR part 50, Appendix E. The proposed changes do not reduce the 
functionality, performance, or capability of TVA's Emergency Response 
Organization (ERO) to respond in mitigating the consequences of 
accidents. The TVA ERO functions will continue to be performed as 
required.
    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 consequence of an accident previously evaluated?
    Response: No.
    The proposed changes to TVA's EAL schemes to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, ``Development of 
Emergency Action Levels

[[Page 27892]]

for Non-Passive Reactors,'' do not reduce the capability to meet the 
emergency planning requirements established in 10 CFR 50.47 and 10 
CFR [Part] 50, Appendix E. The proposed changes do not reduce the 
functionality, performance, or capability of TVA's ERO to respond in 
mitigating the consequences of any design basis accident.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facilities or the manner in which the plants 
are operated and maintained. The proposed change does not adversely 
affect the ability of structures, systems, and components (SSC) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptable limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated. 
Further, the proposed changes do not increase the types and amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposure.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to TVA's EAL schemes to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not involve any 
physical changes to plant systems or equipment. The proposed changes 
do not involve the addition of any new plant equipment. The proposed 
changes will not alter the design configuration, or method of 
operation of plant equipment beyond its normal functional 
capabilities. All TVA ERO functions will continue to be performed as 
required. The proposed changes do not create any new credible 
failure mechanisms, malfunctions, or accident initiators.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from those that have been 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to TVA's EAL schemes to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not alter or exceed a 
design basis or safety limit. There is no change being made to 
safety analysis assumptions, safety limits, or limiting safety 
system settings that would adversely affect plant safety as a result 
of the proposed changes. There are no changes to setpoints or 
environmental conditions of any SSC or the manner in which any SSC 
is operated. Margins of safety are unaffected by the proposed 
changes to adopt the NEI 99-01, Revision 6, EAL scheme guidance. The 
applicable requirements of 10 CFR 50.47 and 10 CFR [Part] 50, 
Appendix E will continue to be met.
    Therefore, the proposed changes do not involve any reduction in 
a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Branch Chief: Benjamin G. Beasley.

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

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

    Date of amendment request: August 19, 2015, as supplemented by 
letters dated May 4, October 3, and November 17, 2016.
    Brief description of amendments: The amendments revised the 
Robinson Technical Specification (TS) 5.6.5.b and the Harris TS 
6.9.1.6.2 to adopt the methodology reports DPC-NE-1008-P, Revision 0, 
``Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse 
Reactors''; DPC-NF-2010, Revision 3, ``Nuclear Physics Methodology for 
Reload Design''; and DPC-NE-2011-P, Revision 2, ``Nuclear Design 
Methodology Report for Core Operating Limits of Westinghouse 
Reactors,'' for application specific to Robinson and Harris.
    Date of issuance: May 18, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: 253 (Robinson) and 157 (Harris). A publicly 
available version is in ADAMS under Accession No. ML17102A923; 
documents related to these amendments are listed in the Safety 
Evaluations enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-23 and NPF-63: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: February 2, 2016 (81 FR 
5492). The supplemental letter dated May 4, 2016, provided additional 
information that expanded the scope of the application as originally 
noticed, and changed the NRC staff's original proposed no significant 
hazards consideration determination as published in the Federal 
Register. Accordingly, the NRC published a second proposed no 
significant hazards consideration determination in the Federal Register 
on August 2, 2016 (81 FR 50746). This notice superseded the original 
notice in its entirety. The supplemental letters dated October 3 and 
November 17, 2016, provided additional information that clarified the 
application, did not expand the scope beyond the second notice, and did 
not change the NRC staff's proposed no significant hazards 
consideration determination as published in the Federal Register.

[[Page 27893]]

    The Commission's related evaluations of the amendments are 
contained in the Safety Evaluations dated May 18, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 21, 2015, as supplemented by 
letters dated June 29, July 13, August 15, November 1, November 17, 
2016, and February 27, 2017.
    Brief description of amendments: The amendments adopted the 
approved changes to Standard Technical Specifications for General 
Electric (BWR/4) [Boiling Water Reactor] Plants, NUREG-1433, Revision 
4, to allow relocation of specific technical specification surveillance 
frequencies to a licensee-controlled program. The changes are described 
in Technical Specification Task Force (TSTF) Traveler, TSTF-425, 
Revision 3, ``Relocate Surveillance Frequencies to Licensee Control--
RITSTF Initiative 5b'' (ADAMS Package Accession No. ML090850642), and 
are described in the Notice of Availability published in the Federal 
Register on July 6, 2009 (74 FR 31996).
    Date of issuance: May 24, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment Nos.: 276 (Unit 1) and 304 (Unit 2). A publicly available 
version is in ADAMS under Accession No. ML17096A129; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: March 29, 2016 (81 FR 
17504). The supplemental letters dated June 29, July 13, August 15, 
November 1, November 17, 2016, and February 27, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 24, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 26, 2016, as supplemented by letter 
dated December 19, 2016.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) by adding a new Administrative Controls section to 
establish, implement, and maintain a Diesel Fuel Oil Testing Program. 
It also relocated to this program the current TS surveillance 
requirements (SRs) for evaluating diesel fuel oil, along with the SRs 
for draining, sediment removal, and cleaning of each main fuel oil 
storage tank at least once every 10 years. In addition, the licensee 
took an exception to NRC Regulatory Guide 1.137, Revision 1, ``Fuel-Oil 
Systems for Standby Diesel Generators,'' to allow for the ability to 
perform sampling of new fuel oil offsite prior to its addition to the 
fuel oil storage tanks.
    Date of issuance: May 24, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 158. A publicly available version is in ADAMS under 
Accession No. ML17048A184; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-63: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: October 11, 2016 (81 FR 
70178). The supplemental letter dated December 19, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 24, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 29, 2016, as supplemented by letter 
dated November 4, 2016.
    Brief description of amendment: The amendment revised the Shearon 
Harris Nuclear Power Plant, Unit 1, Technical Specification (TS) 3/
4.11.1.4, ``Liquid Holdup Tanks''; TS 3/4.11.2.5, ``Explosive Gas 
Mixture''; and TS 6.8.4.j, ``Gas Storage Tank Radioactivity Monitoring 
Program.'' The amendment deleted TS Definition 1.16, ``GASEOUS RADWASTE 
TREATMENT SYSTEM''; TS 3/4.11.1.4, ``Liquid Holdup Tanks''; and TS 3/
4.11.2.5, ``Explosive Gas Mixture.'' The amendment relocated the 
deleted requirements for these TSs to licensee control under TS 
6.8.4.j, ``Gas Storage Tank Radioactivity Monitoring Program.'' The 
description for TS 6.8.4.j, ``Gas Storage Tank Radioactivity Monitoring 
Program,'' was modified to include the controls for potentially 
explosive gas mixtures contained in the Gaseous Waste Processing System 
and the quantity of radioactivity contained in unprotected outdoor 
liquid storage tanks. The amendment relocated requirements associated 
with TS 3/4.11.1.4 and TS 3/4.11.2.5 to the licensee-controlled Plant 
Programs Procedure PLP-114, ``Relocated Technical Specifications and 
Design Basis Requirements.''
    Date of issuance: May 25, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 159. A publicly available version is in ADAMS under 
Accession No. ML17074A672; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-63: The amendment 
revised the Facility Operating License and TSs.
    Date of initial notice in Federal Register: October 25, 2016 (81 FR 
73433). The supplemental letter dated November 4, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 25, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 11, 2016.
    Brief description of amendment: The amendment approved adoption of 
NRC-approved Technical Specifications Task Force (TSTF) Standard 
Technical Specifications Change Traveler TSTF-545, Revision 3, ``TS 
[Technical Specification] Inservice Testing Program

[[Page 27894]]

Removal & Clarify SR [Surveillance Requirement] Usage Rule Application 
to Section 5.5 Testing,'' dated October 21, 2015. Specifically, the 
amendment deleted Palisades Nuclear Plant TS 5.5.7, ``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 Palisades Nuclear Plant 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: May 30, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 262. A publicly available version is in ADAMS under 
Accession No. ML17082A465; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-20: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: August 30, 2016 (81 FR 
59663).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 30, 2017.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert 
Cliffs Nuclear Power Plant, Units 1 and 2, Calvert County, Maryland

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

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

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

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

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

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

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

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

    Date of amendment request: July 26, 2016, as supplemented by letter 
dated October 6, 2016.
    Brief description of amendments: The amendments revised the 
Inservice Testing Program requirements in each plant's technical 
specifications (TSs). The changes included deleting the current TS 
requirements for the Inservice Testing Program, adding a new defined 
term, ``INSERVICE TESTING PROGRAM,'' to the TSs, and revising other TSs 
to reference this new defined term instead of the deleted program.
    Date of issuance: May 26, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 191, 192, 197, 197, 320, 298, 212, 254, 247, 223, 
209, 227, 161, 313, 317, 266, 261, 124, and 290. A publicly available 
version is in ADAMS under Accession No. ML17073A067. Documents related 
to these amendments are listed in the Safety Evaluations enclosed with 
the amendments.
    Facility Operating License Nos.: NPF-72, NPF-77, NPF-37, NPF-66, 
DPR-53, DPR-69, NPF-62, DPR-19, DPR-25, NPF-11, NPF-18, DPR-63, NPF-69, 
DPR-44, DPR-56, DPR-29, DPR-30, DPR-18, and DPR-50. Amendments revised 
the Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: November 8, 2016 (81 FR 
78648).
    The Commission's related evaluations of the amendments are 
contained in Safety Evaluations dated May 26, 2017.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania

    Date of amendment request: June 24, 2016, as supplemented by 
letters dated September 13, 2016; December 15, 2016; and March 16, 
2017.
    Brief description of amendment: The amendment modified the Renewed 
Facility Operating License to reflect the direct transfer of Toledo 
Edison Company's 18.26 percent leased interest in Beaver Valley Power 
Station, Unit 2, and Ohio Edison Company's 21.66 percent leased 
interest in Beaver Valley Power Station, Unit 2, from FirstEnergy 
Nuclear Operating Company to FirstEnergy Nuclear Generation, LLC.
    Date of issuance: May 30, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 187. A publicly available version is in ADAMS under 
Accession No. ML17115A123.
    Renewed Facility Operating License No. NPF-73: Amendment revised 
the Renewed Facility Operating License.
    Date of initial notice in Federal Register: January 23, 2017 (82 FR 
7880). The supplemental letter dated March 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 NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 14, 2017.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment request: July 21, 2016, as supplemented by letter 
dated September 26, 2016.
    Brief description of amendments: The amendments revised the Donald 
C. Cook Nuclear Plant, Units 1 and 2, Technical Specification (TS) 
Surveillance Requirements (SRs), consistent with the NRC-approved 
Technical Specifications Task Force (TSTF) Traveler, TSTF-545, Revision 
3, ``TS Inservice Testing Program Removal & Clarify SR Usage Rule 
Application to Section 5.5 Testing.'' Specifically, the change revised 
the TSs to eliminate Section 5.5.6, ``Inservice Testing Program.'' A 
new defined term, ``INSERVICE TESTING PROGRAM,'' was added to the TS 
Definitions section. TS SRs that previously referred to the Inservice 
Testing Program from Section 5.5.6 were revised to refer to the new 
defined term, ``INSERVICE TESTING PROGRAM.''
    Date of issuance: May 24, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: 335 (Unit 1) and 317 (Unit 2). A publicly available 
version is in ADAMS under Accession No. ML17103A106; documents related

[[Page 27895]]

to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-58 and DPR-74: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: September 27, 2016 (81 
FR 66307). The supplemental letter dated September 26, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 24, 2017.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Units No. 1 and No. 2, Surry County, Virginia

    Date of amendment request: May 18, 2016, as supplemented by letters 
dated February 10, 2017; March 1, 2017; and March 10, 2017.
    Brief description of amendments: The amendments revised Technical 
Specification 3.14 ``Circulating and Service Water Systems,'' to extend 
the Allowed Outage Time for only one operable Service Water flow path 
to the Changing Pump Service Water subsystem and to the Main Control 
Room/Emergency Switchgear Room air conditioning subsystem.
    Date of issuance: May 31, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 289 (Unit 1) and 289 (Unit 2). A publicly available 
version is in ADAMS under Accession No. ML17100A253; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: October 25, 2016 (81 FR 
73443). The supplemental letters dated February 10, 2017; March 1, 
2017; and March 10, 2017, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 31, 2017.
    No significant hazards consideration comments received: No.

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

    For the Nuclear Regulatory Commission.
Eric J. Benner,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2017-12732 Filed 6-16-17; 8:45 am]
BILLING CODE 7590-01-P