[Federal Register Volume 81, Number 216 (Tuesday, November 8, 2016)]
[Notices]
[Pages 78643-78658]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-26824]


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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0226]


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

[[Page 78644]]

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 October 8, 2016, to October 24, 2016. The 
last biweekly notice was published on October 25, 2016.

DATES: Comments must be filed by December 8, 2016. A request for a 
hearing must be filed by January 9, 2017.

ADDRESSES: You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0226. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact 
the individual or individuals in the FOR FURTHER INFORMATION CONTACT 
section of this document.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Lynn Ronewicz, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-1927, email: [email protected].

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2016-0226, facility name, unit 
number(s), plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0226.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in this 
document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2016-0226, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov, as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

I. 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 a petition 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, which is available at the NRC's PDR, located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. The NRC's regulations are accessible electronically 
from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a petition is filed within 60 days, 
the Commission or a presiding officer designated by the Commission or 
by the Chief Administrative Judge of the Atomic

[[Page 78645]]

Safety and Licensing Board Panel, will rule on the petition; and the 
Secretary or the Chief Administrative Judge of the Atomic Safety and 
Licensing Board will issue a notice of a hearing or an appropriate 
order.
    As required by 10 CFR 2.309, a petition shall set forth with 
particularity the interest of the petitioner in the proceeding, and how 
that interest may be affected by the results of the proceeding. The 
petition should specifically explain the reasons why intervention 
should be permitted with particular reference to the following general 
requirements: (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. The petition 
must also set forth the specific contentions which the petitioner seeks 
to have litigated at 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 shall 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 those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion to support 
its position on the issue. The petition must include sufficient 
information to show that a genuine dispute exists with the applicant on 
a material issue of law or fact. Contentions shall 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 these requirements 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, 
and have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that person's admitted 
contentions consistent with the NRC's regulations, policies, and 
procedures.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day 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).
    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 decide 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 held 
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 any 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 January 9, 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. A State, local governmental body, Federally-recognized 
Indian Tribe, or agency thereof may also have the opportunity to 
participate under 10 CFR 2.315(c).
    If a hearing is granted, any person who does not wish, or is not 
qualified, to become a party to the proceeding may, in 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 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, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene (hereinafter 
``petition''), and documents filed by interested governmental entities 
participating 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. Participants may 
not submit paper copies of their filings unless they seek an exemption 
in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [email protected], or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a petition (even 
in instances in which the participant, or its counsel or 
representative, already holds an NRC-issued digital ID certificate). 
Based upon this information, the Secretary will establish an electronic 
docket for the hearing in this proceeding if the Secretary has not 
already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are available on the NRC's public Web site at 
http://www.nrc.gov/site-help/e-submittals/

[[Page 78646]]

adjudicatory-sub.html. Participants may attempt to use other software 
not listed on the Web site, but should note that the NRC's E-Filing 
system does not support unlisted software, and the NRC Electronic 
Filing Help Desk will not be able to offer assistance in using unlisted 
software.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a petition. 
Submissions should be in Portable Document Format (PDF). Additional 
guidance on PDF submissions is available on the NRC's public Web site 
at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing 
is considered complete at the time the documents are 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 documents on those 
participants separately. Therefore, applicants and other participants 
(or their counsel or representative) must apply for and receive a 
digital ID certificate before a hearing petition to intervene is filed 
so that they can obtain access to the document via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Electronic 
Filing Help Desk through the ``Contact Us'' link located on the NRC's 
public Web site at http://www.nrc.gov/site-help/e-submittals.html, by 
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m. 
and 7 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, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, in some instances, a petition will require including 
information on local residence in order to demonstrate a proximity 
assertion of interest in the proceeding. 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.
    The Commission will issue a notice or order granting or denying a 
hearing request or intervention petition, designating the issues for 
any hearing that will be held and designating the Presiding Officer. A 
notice granting a hearing will be published in the Federal Register and 
served on the parties to the hearing.
    For further details with to respect 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.

Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power 
Station (KPS), Carlton, Wisconsin

    Date of amendment request: September 14, 2015. A publicly available 
version is in ADAMS under Accession No. ML15261A236.
    Description of amendment request: The amendment would revise the 
Operating License and associated Technical Specifications to reflect 
removal of all KPS spent nuclear fuel from the spent fuel pool and its 
transfer to dry cask storage within an Independent Spent Fuel Storage 
Installation.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would modify the KPS renewed facility 
operating license and Technical Specification (TS) by deleting the 
portions of the license and TS that are no longer applicable to a 
facility with no spent nuclear fuel stored in the spent fuel pool, 
while modifying the remaining portions to correspond to all nuclear 
fuel stored within an Independent Spent Fuel Storage Installation 
(ISFSI). This amendment becomes effective upon removal of all spent 
nuclear fuel from the KPS spent fuel pool and its transfer to dry 
cask storage within an ISFSI.
    The definition of safety-related structures, systems, and 
components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are 
those relied on to remain functional during and following design 
basis events to assure:
    1. The integrity of the reactor coolant boundary;
    2. The capability to shutdown the reactor and maintain it in a 
safe shutdown condition; or
    3. The capability to prevent or mitigate the consequences of 
accidents which could result in potential offsite exposures 
comparable to the applicable guideline exposures set forth in 10 CFR 
50.43(a)(1) or 100.11.
    The first two criteria (integrity of the reactor coolant 
pressure boundary and safe shutdown of the reactor) are not 
applicable to a plant in a permanently defueled condition. The third 
criterion is related to preventing or mitigating the consequences of 
accidents that could result in potential offsite exposures exceeding 
limits. However, after all nuclear spent fuel assemblies have been 
transferred to dry cask storage within an ISFSI, none of the SSCs at 
KPS are required to be relied on for accident mitigation. Therefore, 
none of the SSCs at KPS meet the definition of a safety-related SSC 
stated in 10 CFR 50.2. The proposed deletion of

[[Page 78647]]

requirements in the TS does not affect systems credited in any 
accident analysis at KPS.
    Section 14 of the KPS Updated Safety Analysis Report (USAR) 
described the design basis accidents related to the spent fuel pool. 
These postulated accidents are predicated on spent fuel being stored 
in the spent fuel pool. With the removal of the spent fuel from the 
spent fuel pool, there are no remaining spent fuel assemblies to be 
monitored and there are no credible accidents that require the 
actions of a Certified Fuel Handler, Shift Manager, or a Non-
certified Operator to prevent occurrence or mitigate the 
consequences of an accident.
    The proposed changes do not have an adverse impact on the 
remaining decommissioning activities or any of their postulated 
consequences.
    The proposed changes related to the relocation of certain 
administrative requirements do not affect operating procedures or 
administrative controls that have the function of preventing or 
mitigating any accidents applicable to the safe management of 
irradiated fuel or decommissioning of the facility.
    Therefore, the proposed amendment does not involve a significant 
increase in the consequences of a previously evaluated accident.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes eliminate the operational requirements and 
certain design requirements associated with the storage of the spent 
fuel in the spent fuel pool, and relocate certain administrative 
controls to the Quality Assurance Program Description.
    After the removal of the spent fuel from the spent fuel pool and 
transfer to the ISFSI, there are no spent fuel assemblies that 
remain in the spent fuel pool. Coupled with a prohibition against 
storage of fuel in the spent fuel pool, the potential for fuel 
related accidents is removed. The proposed changes do not introduce 
any new failure modes.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    The removal of all spent nuclear fuel from the spent fuel pool 
into storage in casks within an ISFSI, coupled with a prohibition 
against future storage of fuel within the spent fuel pool, removes 
the potential for fuel related accidents.
    The design basis and accident assumptions within the KPS USAR 
and the TS relating to safe management and safety of spent fuel in 
the spent fuel pool are no longer applicable. The proposed changes 
do not affect remaining plant operations, systems, or components 
supporting decommissioning activities.
    The requirements for systems, structures, and components (SSCs) 
that have been deleted from the KPS TS are not credited in the 
existing accident analysis for any applicable postulated accident; 
and as such, do not contribute to the margin of safety associated 
with the accident analysis.
    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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
    NRC Branch Chief: Bruce A. Watson.

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

    Date of amendment request: July 25, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16207A532.
    Description of amendment request: The amendment would revise the 
Waterford Steam Electric Station, Unit 3 (Waterford 3), Technical 
Specifications (TSs) Section 6.5.8, ``Inservice Testing Program,'' to 
remove requirements duplicated in the American Society of Mechanical 
Engineers Code for Operation and Maintenance of Nuclear Power Plants 
Case OMN-20, ``Inservice Test Frequency.'' A new defined term, 
``Inservice Testing Program,'' will be added to the TS 1.0, 
``Definitions,'' section. The licensee states that the proposed change 
to the TS is consistent with 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'' (ADAMS 
Accession No. ML15294A555). However, the Waterford 3 TSs (NUREG-0973) 
are of an older standard version and have not been converted to the 
Improved Standard Technical Specifications (ISTSs). Therefore, Entergy 
has included in the application a table of TSs affected by the 
amendment, with variations and differences between the Waterford 3 TSs 
and the ISTSs listed in TSTF-545 discussed individually.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff edits in square 
brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS Chapter 6, ``Administrative 
Controls,'' Section 6.5, ``Programs'' by eliminating the ``Inservice 
Testing Program'' specification. Most requirements in the IST 
Program are removed, as they are duplicative of requirements in the 
ASME OM Code [American Society of Mechanical Engineers Code for 
Operation and Maintenance of Nuclear Power Plants], as clarified by 
Code Case OMN-20, ``Inservice Test Frequency.'' The remaining 
requirements in the Section 6.5.8, IST Program are eliminated [. . 
.]. A new defined term, ``Inservice Testing Program,'' is added to 
the TS, which references the requirements of 10 CFR 50.55a(f).
    Performance of inservice testing is not an initiator to any 
accident previously evaluated. As a result, the probability of 
occurrence of an accident is not significantly affected by the 
proposed change. Inservice test frequencies under Code Case OMN-20 
are equivalent to the current testing period allowed by the TS with 
the exception that testing frequencies greater than 2 years may be 
extended by up to 6 months to facilitate test scheduling and 
consideration of plant operating conditions that may not be suitable 
for performance of the required testing. The testing frequency 
extension will not affect the ability of the components to mitigate 
any accident previously evaluated as the components are required to 
be operable during the testing period extension. Performance of 
inservice tests utilizing the allowances in OMN-20 will not 
significantly affect the reliability of the tested components. As a 
result, the availability of the affected components, as well as 
their ability to mitigate the consequences of accidents previously 
evaluated, is not affected.
    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 alter the design or configuration 
of the plant. The proposed change does not involve a physical 
alteration of the plant; no new or different kind of equipment will 
be installed. The proposed change does not alter the types of 
inservice testing performed. In most cases, the frequency of 
inservice testing is unchanged. However, the frequency of testing 
would not result in a new or different kind of accident from any 
previously evaluated since the testing methods are not altered.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change eliminates some requirements from the TS in 
lieu of

[[Page 78648]]

requirements in the ASME Code, as modified by use of Code Case OMN-
20. Compliance with the ASME Code is required by 10 CFR 50.55a. The 
proposed change also allows inservice tests with frequencies greater 
than 2 years to be extended by 6 months to facilitate test 
scheduling and consideration of plant operating conditions that may 
not be suitable for performance of the required testing. The testing 
frequency extension will not affect the ability of the components to 
respond to an accident as the components are required to be operable 
during the testing period extension. The proposed change will 
eliminate the existing TS Surveillance Requirement (SR) 4.0.3 
(referenced as SR 3.0.3 in the ISTS [improved standard technical 
specification]) allowance to defer performance of missed inservice 
tests up to the duration of the specified testing frequency, and 
instead will require an assessment of the missed test on equipment 
operability. This assessment will consider the effect on a margin of 
safety (equipment operability). Should the component be inoperable, 
the Technical Specifications provide actions to ensure that the 
margin of safety is protected. The proposed change also eliminates a 
statement that nothing in the ASME Code should be construed to 
supersede the requirements of any TS. [. . .] However, elimination 
of the statement will have no effect on plant operation or safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William B. Glew, Jr., Associate General 
Counsel--Entergy Services, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Acting Branch Chief: Stephen S. Koenick.

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

    Date of amendment request: August 22, 2016. A publicly available 
version is in ADAMS under Accession No. ML16236A300.
    Description of amendment request: The amendment would (1) revise 
Technical Specification (TS) 4.2.1, ``Reactor Core, Fuel Assemblies,'' 
to add Optimized ZIRLO\TM\ as an approved fuel rod cladding material, 
(2) revise TS 5.6.5.b to add the Westinghouse topical reports for 
Optimized ZIRLO\TM\ and ZIRLO[supreg], and (3) revise TS 5.6.5.b with a 
non-technical change to the Reference 11 title (replace a semicolon 
with a period).
    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 would allow the use of Optimized 
ZIRLOTM clad nuclear fuel in the reactors. The NRC 
approved topical report WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-
A, ``Optimized ZIRLOTM,'' prepared by Westinghouse 
Electric Company LLC (Westinghouse), addresses Optimized ZIRLO and 
demonstrates that Optimized ZIRLOTM has essentially the 
same properties as currently licensed ZIRLO[supreg]. The fuel 
cladding itself is not an accident initiator and does not affect 
accident probability. With the approved exemption, use of Optimized 
ZIRLOTM fuel cladding will continue to meet all 10 CFR 
50.46 acceptance criteria and, therefore, will not increase the 
consequences of an accident.
    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.
    Use of Optimized ZIRLO\TM\ clad fuel will not result in changes 
in the operation or configuration of the facility. Topical Report 
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, demonstrated that the 
material properties of Optimized ZIRLOTM are similar to 
those of standard ZIRLO[supreg]. Therefore, Optimized 
ZIRLOTM fuel rod cladding will perform similarly to those 
fabricated from standard ZIRLO[supreg], thus precluding the 
possibility of the fuel cladding becoming an accident initiator and 
causing a new or different type of accident. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will not involve a significant reduction in 
the margin of safety. Topical Report WCAP-12610-P-A & CENPD-404-P-A, 
Addendum 1-A, demonstrated that the material properties of the 
Optimized ZIRLO\TM\ are not significantly different from those of 
standard ZIRLO[supreg]. Optimized ZIRLO\TM\ is expected to perform 
similarly to standard ZIRLO[supreg] for all normal operating and 
accident scenarios, including both loss of coolant accident (LOCA) 
and non-LOCA scenarios. For LOCA scenarios, where the slight 
difference is Optimized ZIRLO\TM\ material properties relative to 
standard ZIRLO[supreg] could have some impact on the overall 
accident scenario, plant-specific LOCA analyses using Optimized 
ZIRLOTM properties will demonstrate that the acceptance 
criteria of 10 CFR 50.46 have been satisfied.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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, Unit Nos. 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. Publicly-available versions are in ADAMS under

[[Page 78649]]

Accession Nos. ML16209A218 and ML16280A402, respectively.
    Description of amendment request: The amendments would revise the 
Inservice Testing Program requirements in each plant's technical 
specifications (TSs). For each plant, the changes include deleting the 
current TS for the Inservice Testing Program, adding a new defined 
term, ``INSERVICE TESTING PROGAM,'' to the TSs, and revising other TSs 
to reference this new defined term instead of the deleted TS. The 
licensee stated that the proposed changes are based on 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'' (ADAMS Accession No. ML15294A555), with some 
variations.
    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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS Chapter 5, ``Administrative 
Controls,'' Section 5.5, ``Programs and Manuals,'' or equivalent, by 
deleting the ``lnservice Testing Program'' specification. A new 
defined term, ``INSERVICE TESTING PROGRAM,'' is added to the TS, 
which references the requirements of 10 CFR 50.55a(f), ``Inservice 
testing requirements.'' The regulations in 10 CFR 50.55a(f) require 
that specified pumps and valves meet the inservice test requirements 
in the American Society of Mechanical Engineers (ASME) Code for 
Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and 
addenda. Most requirements currently in the TS lnservice Testing 
Program are duplicative of requirements in the ASME OM Code and 
addenda, as modified by NRC-approved alternatives or reliefs. The 
proposed change primarily affects the required frequency for 
performing ASME OM Code required tests for pumps and valves which 
are covered by the Inservice Testing Program. The proposed change 
would allow a longer interval between some tests and require a 
shorter interval between other tests; the effect of the change to 
specific test intervals depends on the plant-specific licensing 
basis.
    Performance of inservice testing is not an initiator to any 
accident previously evaluated. As a result, the probability of 
occurrence of an accident is not significantly affected by the 
proposed change. Changing the required test frequency of pumps and 
valves will not affect the ability of the components to mitigate any 
accident previously evaluated, as the components are required to be 
operable. If components required by the TSs are found to be 
inoperable, the TSs specify the actions required to ensure safe 
operation of the facility, and these actions are not altered by the 
proposed change. Performance of inservice tests in accordance with 
the ASME OM Code, as modified by NRC-approved alternatives or 
reliefs, will not significantly affect the reliability of the tested 
components. As a result, the availability of the affected 
components, as well as their ability to mitigate the consequences of 
accidents previously evaluated, is not significantly affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not alter the design or configuration 
of the plant. The proposed change does not involve a physical 
alteration of the plant; no new or different kind of equipment will 
be installed. The proposed change does not alter the types of 
inservice testing performed. Changes to the frequency of testing 
would not result in a new or different kind of accident from any 
previously evaluated since the testing methods are not altered. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change eliminates some requirements from the TSs in 
lieu of requirements in the ASME OM Code, as modified by NRC-
approved alternatives or reliefs. Compliance with the ASME OM Code 
is required by 10 CFR 50.55a. Changes to the required test frequency 
will not affect the ability of the components to respond to an 
accident, as the components are required to be operable. The 
proposed change also eliminates a provision which allowed, under 
certain circumstances, the licensee to delay declaring equipment 
inoperable due to a missed surveillance. This change will not have a 
significant effect on plant operation or safety, as the licensee 
will still be required by TSs to assess component operability. If 
components required by the TSs are found to be inoperable, the TSs 
specify the actions required to ensure safe operation of the 
facility, and these actions are not altered by the proposed change. 
The proposed change also eliminates a statement that nothing in the 
ASME OM Code should be construed to supersede the requirements of 
any TS. Elimination of the statement will not have a significant 
effect on plant operation or safety. Therefore, the proposed change 
does not involve a significant reduction in a margin of safety.

    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 Acting Branch Chief: G. Edward Miller.

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

    Date of amendment request: August 26, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16245A288.
    Description of amendment request: The amendment would revise the 
CNS Technical Specifications (TSs) to eliminate Section 5.5.6, 
``Inservice Testing [IST] Program,'' to remove requirements duplicated 
in the American Society of Mechanical Engineers Code for Operation and 
Maintenance of Nuclear Power Plants (ASME OM Code) Case OMN-20, 
``Inservice Test Frequency.'' A new defined term, ``Inservice Testing 
Program,'' will be added to TS Section 1.1, ``Definitions.'' The 
licensee stated that the proposed change to the TSs is consistent with 
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'' (ADAMS Accession No. ML15294A555), 
with no proposed technical variations or deviations. However, in some 
cases, the CNS TSs use different section titles or numbering for 
surveillance requirements than the Standard Technical Specifications on 
which TSTF-545 was based, so the licensee changed the TSTF-545 
numbering to be consistent with the CNS TS numbering.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff edits in [square 
brackets]:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS Chapter 5, ``Administrative 
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating 
the ``Inservice Testing Program'' specification. Most requirements 
in the Inservice Testing Program are removed, as they are 
duplicative of requirements in the ASME OM Code, as clarified by 
Code Case OMN-20, ``Inservice Test Frequency.'' The remaining 
requirements in the Section 5.5 IST Program

[[Page 78650]]

are eliminated [. . .]. A new defined term, ``Inservice Testing 
Program,'' is added to the TS, which references the requirements of 
10 CFR 50.55a(f).
    Performance of inservice testing is not an initiator to any 
accident previously evaluated. As a result, the probability of 
occurrence of an accident is not significantly affected by the 
proposed change. Inservice test frequencies under Code Case OMN-20 
are equivalent to the current testing period allowed by the TS with 
the exception that testing frequencies greater than 2 years may be 
extended by up to 6 months to facilitate test scheduling and 
consideration of plant operating conditions that may not be suitable 
for performance of the required testing. The testing frequency 
extension will not affect the ability of the components to mitigate 
any accident previously evaluated as the components are required to 
be operable during the testing period extension. Performance of 
inservice tests utilizing the allowances in OMN-20 will not 
significantly affect the reliability of the tested components. As a 
result, the availability of the affected components, as well as 
their ability to mitigate the consequences of accidents previously 
evaluated, is not affected.
    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 alter the design or configuration 
of the plant. The proposed change does not involve a physical 
alteration of the plant; no new or different kind of equipment will 
be installed. The proposed change does not alter the types of 
inservice testing performed. In most cases, the frequency of 
inservice testing is unchanged. However, the frequency of testing 
would not result in a new or different kind of accident from any 
previously evaluated since the testing methods are not altered.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change eliminates some requirements from the TS in 
lieu of requirements in the ASME Code, as modified by use of Code 
Case OMN-20. Compliance with the ASME Code is required by 10 CFR 
50.55a. The proposed change also allows inservice tests with 
frequencies greater than 2 years to be extended by 6 months to 
facilitate test scheduling and consideration of plant operating 
conditions that may not be suitable for performance of the required 
testing. The testing frequency extension will not affect the ability 
of the components to respond to an accident as the components are 
required to be operable during the testing period extension. The 
proposed change will eliminate the existing TS SR 3.0.3 allowance to 
defer performance of missed inservice tests up to the duration of 
the specified testing frequency, and instead will require an 
assessment of the missed test on equipment operability. This 
assessment will consider the effect on a margin of safety (equipment 
operability). Should the component be inoperable, the Technical 
Specifications provide actions to ensure that the margin of safety 
is protected. The proposed change also eliminates a statement that 
nothing in the ASME Code should be construed to supersede the 
requirements of any TS. [. . .] However, elimination of the 
statement will have no effect on plant operation or 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: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Acting Branch Chief: Stephen S. Koenick.

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

    Date of amendment request: September 2, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16246A321.
    Description of amendment request: The amendment would revise the 
Nuclear Radiological Emergency Response Plan (RERP) for FCS for the 
plant condition following permanent cessation of power operations and 
defueling. The proposed FCS RERP changes would revise the shift 
staffing and Emergency Response Organization (ERO) staffing.
    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 the FCS RERP do not impact the function 
of plant structures, systems, or components (SSCs). The proposed 
changes do not affect accident initiators or precursors, nor does it 
alter design assumptions. The proposed changes do not prevent the 
ability of the on-shift staff and ERO to perform their intended 
functions to mitigate the consequences of any accident or event that 
will be credible in the permanently defueled condition. The proposed 
changes only remove positions that will no longer be credited in the 
FCS RERP.
    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 reduce the number of on-shift and ERO 
positions commensurate with the hazards associated with a 
permanently shut down and defueled facility. The proposed changes do 
not involve installation of new equipment or modification of 
existing equipment, so that no new equipment failure modes are 
introduced. Also, the proposed changes do not result in a change to 
the way that the equipment or facility is operated so that no new 
accident initiators are created.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed changes are 
associated with the FCS RERP staffing and do not impact operation of 
the plant or its response to transients or accidents. The change 
does not affect the Technical Specifications. The proposed changes 
do not involve a change in the method of plant operation, and no 
accident analyses will be affected by the proposed changes. Safety 
analysis acceptance criteria are not affected by the proposed 
changes. The revised FCS RRP will continue to provide the necessary 
response staff with the proposed changes.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 
K Street NW., Washington, DC 20006-3817.
    NRC Acting Branch Chief: Stephen S. Koenick.

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

    Date of amendment request: September 28, 2016. A publicly-

[[Page 78651]]

available version is in ADAMS under Accession No. ML16273A502.
    Description of amendment request: The amendment would modify the 
Technical Specifications to make administrative changes to align 
staffing for decommissioning Fort Calhoun Station, Unit No. 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 only impact administrative requirements 
associated with staff qualification, staff titles, personnel 
staffing levels, and clarification of systems used during 
decommissioning. The proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated because: (1) The proposed amendment does not 
represent a change to any system design, (2) the proposed amendment 
does not alter, degrade, or prevent action described or assumed in 
any accident in the USAR [updated safety analysis report] from being 
performed, (3) the proposed amendment does not alter any assumptions 
previously made in evaluating radiological consequences, and [(4)] 
the proposed amendment does not affect the integrity of any fission 
product barrier. No safety related equipment is affected by the 
proposed change.
    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 do not alter the physical design, safety 
limits, or safety analysis assumptions associated with the operation 
of the plant. Hence, the proposed changes do not introduce any new 
accident initiators, nor do these changes reduce or adversely affect 
the capabilities of any plant structure or system in the performance 
of their safety function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits or limiting safety system settings are determined. The safety 
analysis acceptance criteria are not affected by these proposed 
changes. Further, the proposed changes do not change the design 
function of any equipment assumed to operate in the event of an 
accident.
    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: David A. Repka, Esq., Winston & Strawn, 1700 
K Street NW., Washington, DC 20006-3817.
    NRC Acting Branch Chief: Stephen S. Koenick.

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

    Date of amendment request: August 30, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16243A233.
    Description of amendment request: The amendments would revise the 
Salem Generating Station, Unit Nos. 1 and 2 (Salem), Technical 
Specifications (TSs), Section 6.8.4.j, ``Inservice Testing Program,'' 
to remove requirements duplicated in the American Society of Mechanical 
Engineers (ASME) Code for Operation and Maintenance of Nuclear Power 
Plants (OM Code) Case OMN-20, ``Inservice Test Frequency.'' A new 
defined term, ``Inservice Testing Program,'' will be added to the TS 
1.0, ``Definitions,'' section. The licensee stated that the proposed 
change to the TS is consistent with 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'' 
(ADAMS Accession No. ML15294A555). However, the Salem TSs use different 
numbering than the Standard Technical Specifications on which TSTF-545 
was based, so the licensee changed the TSTF-545 numbering to be 
consistent with the Salem TS numbering.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff edits in square 
brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS Chapter 6, ``Administrative 
Controls,'' Section 6.8, ``Procedures and Programs,'' by eliminating 
the ``Inservice Testing Program'' specification. Most requirements 
in the Inservice Testing Program are removed, as they are 
duplicative of requirements in the ASME OM Code, as clarified by 
Code Case OMN-20, ``Inservice Test Frequency.'' The remaining 
requirements in the Section 6.8 IST [Inservice Testing] Program are 
eliminated [. . .]. A new defined term, ``Inservice Testing 
Program,'' is added to the TS, which references the requirements of 
10 CFR 50.55a(f).
    Performance of inservice testing is not an initiator to any 
accident previously evaluated. As a result, the probability of 
occurrence of an accident is not significantly affected by the 
proposed change. Inservice test frequencies under Code Case OMN-20 
are equivalent to the current testing period allowed by the TS with 
the exception that testing frequencies greater than 2 years may be 
extended by up to 6 months to facilitate test scheduling and 
consideration of plant operating conditions that may not be suitable 
for performance of the required testing. The testing frequency 
extension will not affect the ability of the components to mitigate 
any accident previously evaluated as the components are required to 
be operable during the testing period extension. Performance of 
inservice tests utilizing the allowances in OMN-20 will not 
significantly affect the reliability of the tested components. As a 
result, the availability of the affected components, as well as 
their ability to mitigate the consequences of accidents previously 
evaluated, is not affected.
    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 alter the design or configuration 
of the plant. The proposed change does not involve a physical 
alteration of the plant; no new or different kind of equipment will 
be installed. The proposed change does not alter the types of 
inservice testing performed. In most cases, the frequency of 
inservice testing is unchanged. However, the frequency of testing 
would not result in a new or different kind of accident from any 
previously evaluated since the testing methods are not altered.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change eliminates some requirements from the TS in 
lieu of requirements in the ASME Code, as modified by use of Code 
Case OMN-20. Compliance with the ASME Code is required by 10 CFR 
50.55a. The proposed change also allows inservice tests with 
frequencies greater than 2 years to be extended by 6 months to 
facilitate test scheduling and consideration of plant operating 
conditions that may not be

[[Page 78652]]

suitable for performance of the required testing. The testing 
frequency extension will not affect the ability of the components to 
respond to an accident as the components are required to be operable 
during the testing period extension. The proposed change will 
eliminate the existing TS 4.0.3 allowance to defer performance of 
missed inservice tests up to the duration of the specified testing 
frequency, and instead will require an assessment of the missed test 
on equipment operability. This assessment will consider the effect 
on a margin of safety (equipment operability). Should the component 
be inoperable, the TS provide actions to ensure that the margin of 
safety is protected. The proposed change also eliminates a statement 
that nothing in the ASME Code should be construed to supersede the 
requirements of any TS. [. . .] However, elimination of the 
statement will have no effect on plant operation or 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: Jeffrie J. Keenan, PSEG Nuclear LLC-N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Douglas A. Broaddus.

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

    Date of amendment request: September 22, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16270A582.
    Description of amendment request: The changes would amend Combined 
License Nos. NPF-93 and NPF-94 for the Virgil C. Summer Nuclear 
Station, Units 2 and 3, respectively. The amendments propose changes to 
the Updated Final Safety Analysis Report (UFSAR) in the form of 
departures from the incorporated plant-specific Design Control Document 
Tier 2 information and involve related changes to the Combined 
Operating License Appendix C (and corresponding plant-specific design 
control document Tier 1) information. Specifically, the proposed 
departures consist of changes to the design reliability assurance 
program (D-RAP) to identify the covers for the in-containment refueling 
water storage tank vents and overflow weirs as the risk-significant 
components included in the D-RAP and to differentiate between the rod 
drive motor-generator (MG) sets field control relays and the rod drive 
power supply control cabinets in which the relays are located.
    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 in-containment refueling water storage tank (IRWST) provides 
flooding of the refueling cavity for normal refueling. The tank also 
serves as a heat sink during Passive Residual Heat Removal (PRHR) 
Heat Exchanger (HX) operation and in the event of a loss-of-coolant-
accident (LOCA) provides injection in support of long-term RCS 
cooling. This activity adds normally closed covers to the IRWST 
vents and overflow weirs to prevent debris from entering the tank, 
prevent over-pressurization and accommodate volume and mass 
increases in the tank. The vent and overflow weir covers open upon 
differential pressures between the IRWST and containment.
    The rod drive MG sets provide the power to the control rod drive 
mechanisms through the reactor trip switchgear. This activity 
revises the equipment description and equipment tag associated with 
the risk-significant control relays which open to de-energize the 
rod drive MG sets and permit rods to drop.
    The proposed changes to add the IRWST vent and overflow weir 
covers and to change the description of the equipment and equipment 
tag related to the rod drive MG sets does not inhibit the SSCs from 
performing their safety-related function. The design bases of the 
IRWST vents and overflow weirs are not modified as a result of the 
addition of the covers to the vents and overflow weirs and the 
change to the control cabinet relay description and equipment tag. 
This proposed amendment does not have an adverse impact on the 
response to anticipated transients or postulated accident conditions 
because the functions of the SSCs are not changed. Required IRWST 
venting is not affected for any accident conditions. Required DAS 
functions are not affected for any accident conditions. Safety-
related structure, system, component (SSC) or function is not 
adversely affected by this change. The changes to include the IRWST 
covers and to change the control cabinet relay description and tag 
number do not involve an interface with any SSC accident initiator 
or initiating sequence of events, and thus, the probabilities of the 
accidents evaluated in the 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. Probabilistic Risk Assessment (PRA) modeling and analyses 
associated with the SSCs are not impacted by this change.
    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 the design of the IRWST vent and 
overflow weir covers do not adversely affect any safety-related 
equipment, and do not add any new interfaces to safety-related SSCs. 
No system or design function or equipment qualification is affected 
by these changes. The changes do not introduce a new failure mode, 
malfunction or sequence of events that could affect plant safety or 
safety-related equipment as the simplistic design of the cover 
louvers and hinged flappers are not considered unique designs. No 
new credible failure modes are introduced by the addition of the 
covers.
    The proposed changes to the description and equipment tag 
associated with the risk-significant control relays for the rod 
drive MG sets do not adversely affect any safety-related equipment, 
and do not add any new interfaces to safety-related SSCs. No system 
or design function or equipment qualification is affected by these 
changes. The changes do not introduce a new failure mode, 
malfunction or sequence of events that could affect plant safety or 
safety-related equipment because the design function of the control 
relays, control cabinets, or rod drive MG sets is not changed.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes maintain compliance with the applicable 
Codes and Standards, thereby maintaining the margin of safety 
associated with these SSCs. The proposed changes do not alter any 
applicable design codes, code compliance, design function, or safety 
analysis. Consequently, no safety analysis or design basis 
acceptance limit/criterion is challenged or exceeded by the proposed 
change, thus the margin of safety is not reduced. Because no safety 
analysis or design basis acceptance limit/criterion is challenged or 
exceeded by these changes, no margin of safety is reduced.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

[[Page 78653]]

    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: September 23, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16271A378.
    Description of amendment request: The amendments would revise the 
completion date for License Condition 2.C(9)b for Unit 1, and License 
Condition 2.C(3) for Unit 2, regarding the date for completion of 
permanent modifications to the Fort Loudoun Dam to prevent overtopping 
due to the probable maximum flood. The change is needed to accommodate 
the current Tennessee Department of Transportation schedule for 
completion of highway construction that will facilitate access to 
complete the modifications to the Fort Loudoun Dam.
    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 revise the completion date for License 
Condition 2.C(9)b for WBN Unit 1 and License Condition 2.C(3) for 
WBN Unit 2 regarding the completion of permanent modifications to 
the Fort Loudoun Dam from February 1, 2017, to June 30, 2018, do not 
affect the structures, systems, or components (SSCs) of the plant, 
affect plant operations, or any design function or an analysis that 
verifies the capability of an SSC to perform a design function. No 
change is being made to any of the previously evaluated accidents in 
the WBN Updated Final Safety Analysis Report (UFSAR).
    The proposed changes do not (1) require physical changes to 
plant SSCs; (2) prevent the safety function of any safety-related 
system, structure, or component during a design basis event; (3) 
alter, degrade, or prevent action described or assumed in any 
accident described in the WBN UFSAR from being performed because the 
safety-related SSCs are not modified; (4) alter any assumptions 
previously made in evaluating radiological consequences; or (5) 
affect the integrity of any fission product barrier.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not introduce any new accident causal 
mechanisms, because no physical changes are being made to the plant, 
nor do they affect any plant systems that are potential accident 
initiators.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The margin of safety associated with the acceptance criteria of 
any accident is unchanged. The proposed changes will have no effect 
on the availability, operability, or performance of safety-related 
systems and components.
    The proposed change will not adversely affect the operation of 
plant equipment or the function of equipment assumed in the accident 
analysis.
    The proposed amendment does not involve changes to any safety 
analyses assumptions, safety limits, or limiting safety system 
settings. The changes do not adversely affect plant-operating 
margins or the reliability of equipment credited in the safety 
analyses.
    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: Sherry A. Quirk, Executive Vice President 
and General Counsel, Tennessee Valley Authority, 400 West Summit Hill 
Drive, 6A Tower West, Knoxville, TN 37902.
    NRC Acting Branch Chief: Jeanne A. Dion.

II. Previously Published Notices of Consideration of Issuance of 
Amendments to Facility Operating Licenses and Combined Licenses, 
Proposed No Significant Hazards Consideration Determination, and 
Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

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

    Date of amendment request: August 3, 2016, as supplemented by 
letter dated October 4, 2016. Publicly available versions are in ADAMS 
under Accession Nos. ML16230A003 and ML16291A495, respectively.
    Description of amendment request: The amendments would revise the 
Technical Specification (TS) requirements for the Control Room 
Emergency Ventilation System (CREVS). The licensee proposed the changes 
to align the CREVS TSs more closely with the applicable Standard 
Technical Specifications. Consequently, the requirements to immediately 
suspend irradiated fuel movement would be relocated, in most cases, to 
coincide with the commencement of unit shutdown in the event the 
allowable outage time (AOT) cannot be met for an inoperable CREVS 
component or control room envelope (CRE) boundary. The proposed 
amendments would also eliminate the TS Limiting Condition for Operation 
Actions and Surveillance Requirements associated with the CREVS kitchen 
and lavatory ventilation exhaust duct isolation dampers.
    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.
    Relocating the requirement to immediately suspend irradiated 
fuel movement from the determination of inoperability to the 
expiration of the AOT is consistent with the Westinghouse Standard 
Technical Specifications (STS) for an inoperable CREVS train and 
thereby establishes a commensurate level of safety. This change does 
not impact the functioning of the fuel handling system and so does 
not significantly increase the probability of a fuel handling 
accident. The removal of the kitchen and lavatory area exhaust 
damper requirements aligns the licensing basis with the current 
design and enhances the reliability of the CRE. The CREVS is not an 
initiator of an accident. Hence, neither of the proposed changes 
increase the probability of an accident previously evaluated.
    The proposed changes do not impair the CREVS' capability to 
provide a protected environment from which operators can control the 
Units for all postulated events in the presence of a single failure. 
For an inoperable CRE boundary in any plant MODE, the suspension of 
fuel movement for

[[Page 78654]]

the first 24 hours, during which the effectiveness of the mitigating 
actions are verified, ensures no increase in the consequences of a 
fuel handling accident. The proposed change aligns the licensing 
bases for the kitchen and lavatory ventilation exhaust pathways with 
a more reliable physical barrier design.
    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.
    Relocating the requirement to immediately suspend irradiated 
fuel movement until expiration of the AOT is consistent with the 
Westinghouse STS and hence does not introduce a new type of accident 
than previously evaluated or change the methods governing normal 
plant operation. Aligning the Control Room kitchen and lavatory 
ventilation exhaust pathway licensing bases with their current 
design does not introduce new failure modes for existing equipment 
or result in any new limiting single failure modes. The proposed 
changes do not challenge the performance or integrity of any safety-
related system.
    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 changes align the current CREVS TS ACTION(s) with 
the Westinghouse STS and the licensing bases for the Control Room 
kitchen and lavatory ventilation exhaust pathways with their current 
design. As such, the proposed changes do not involve changes to any 
safety analyses assumptions, safety limits, or limiting safety 
system settings nor do they adversely impact plant operating margins 
or the reliability of equipment credited in the safety analyses.
    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 S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB, 
Juno Beach, FL 33408-0420.
    NRC Acting Branch Chief: Jeanne A. Dion.

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

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

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

    Date of amendment request: September 22, 2015.
    Brief description of amendment: The amendment approved the proposed 
name change from Duke Energy Florida, Inc. to Duke Energy Florida, LLC.
    Date of issuance: October 12, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 250.
    Facility Operating License No. DPR-72: The amendment revised the 
facility operating license.
    Date of initial notice in Federal Register: August 16, 2016 (81 FR 
54614).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 12, 2016.
    No significant hazards consideration comments received: No.

Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Hartsville, South Carolina

    Date of amendment request: November 19, 2015, as supplemented by 
letter dated August 18, 2016.
    Brief description of amendment: The amendment revised the technical 
specifications (TSs) to allow the extension of the Type A containment 
test interval to 15 years and the extension of the Type B and Type C 
test intervals for selected components to 120 months and 75 months, 
respectively. The amendment also deleted from the TSs an already 
implemented one-time extension of the Type A test frequency.
    Date of issuance: October 11, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 247. A publicly-available version is in ADAMS under 
Accession No. ML16201A195; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-23: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: March 15, 2016 (81 FR 
13841). The supplemental letter dated August 18, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 11, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 4, 2016.

[[Page 78655]]

    Brief description of amendments: The amendments revised the CCNPP 1 
and 2 Technical Specifications (TSs) to include Surveillance 
Requirement (SR) 3.5.2.10 in the list of applicable surveillances of SR 
3.5.3.1 as part of the implementation of Technical Specifications Task 
Force (TSTF) Improved Standard Technical Specifications Change Traveler 
TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas 
Accumulation.''
    Date of issuance: October 7, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 319 (Unit 1) and 297 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16263A001; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: May 24, 2016 (81 FR 
32806).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 7, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 20, 2016, as supplemented by letter 
dated August 11, 2016.
    Brief description of amendments: The amendments revised Technical 
Specification 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting Air,'' 
to replace the required stored inventory of lube oil for the diesel 
generators (specified in number of gallons) with inventory requirements 
based on diesel generator operating time (specified in number of days). 
The changes are based on Revision 1 to Technical Specifications Task 
Force (TSTF) Improved Standard Technical Specifications Change Traveler 
TSTF-501, ``Relocate Stored Fuel Oil and Lube Oil Volume Values to 
Licensee Control.''
    Date of issuance: October 14, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 310 (Unit 2) and 314 (Unit 3). A publicly-available 
version is in ADAMS under Accession No. ML16235A405; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 2016 (81 FR 
46962). The supplemental letter dated August 11, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 14, 2016.
    No significant hazards consideration comments received: No.

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

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

    Date of amendment request: February 6, 2015, as supplemented by 
letters dated September 1, 2015, and January 20, January 28, April 26, 
June 22, and September 28, 2016.
    Brief description of amendments: The amendments revised the 
technical specifications (TSs) for both DNPS, Units Nos. 2 and 3, and 
QCNPS, Unit Nos. 1 and 2, to support the use of AREVA nuclear fuel; 
both facilities currently operate using a Westinghouse nuclear fuel 
design. Specifically, the TSs for the core operating limits report (TS 
5.6.5.b) are revised to include NRC-approved AREVA methodologies and to 
delete methodologies no longer in use. The transient analyses take 
credit for conservatism in the scram speed performance; therefore, a 
new surveillance requirement (SR) associated with linear heat 
generation rate (LHGR) is added to the TSs (SR 3.2.3.2). This 
demonstrates scram speed distribution is consistent with that used in 
the transient analyses. The TSs associated with the limiting condition 
for operation (LCO 3.7.7) for the main turbine bypass system is revised 
to include requirements to use the minimum critical power ratio limits 
(LCO 3.2.2) and LHGR limits (LCO 3.2.3) during operations when at 
greater than or equal to (=) 25 percent of rated thermal 
power and the main turbine bypass system is inoperable.
    To increase the margin to the maximum reactor pressure vessel (RPV) 
acceptance criteria for certain anticipated transient without scram 
(ATWS) transients, the SRs for the allowable value (AV) for the ATWS 
recirculation pump trip (ATWS-RPT) on high RPV steam dome pressure are 
modified (SR 3.3.4.1.4.b). The ATWS-RPT AV for DNPS, Unit Nos. 2 and 3, 
is lowered to less than or equal to 1,198 pounds per square inch gauge 
(psig). The ATWS-RPT AV for QCNPS, Unit Nos. 1 and 2, is lowered to 
less than or equal to 1,195 psig.
    Date of issuance: October 20, 2016.
    Effective date: As of the date of issuance and shall be implemented 
prior to entering into MODE 2 on the first plant startup following the 
next refueling outage for each unit.
    Amendment Nos.: 251 and 244 (DNPS, Unit Nos. 2 and 3) and 264 and 
259 (GCNPS, Unit Nos. 1 and 2). A publicly-available version is in 
ADAMS under Accession No. ML16221A061; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-19, DPR-25, DPR-29, and 
DPR-30: Amendments revised the Renewed Facility Operating Licenses and 
TSs.
    Date of initial notice in Federal Register: November 3, 2015 (80 FR 
67800). The supplemental letters dated January 20, January 28, April 
26, June 22, and September 28, 2016, provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety evaluation dated October 20, 2016.
    No significant hazards consideration comments received: No.

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: December 14, 2015, as supplemented by 
letters dated March 9, 2016, and June 1, 2016. Publicly-available 
versions are in ADAMS under Accession Nos. ML15348A396, ML16069A217, 
and ML16153A084, respectively.
    Brief description of amendments: The amendments revised the design 
bases in

[[Page 78656]]

the updated final safety analysis report to reflect the use of a new 
criticality safety assessment for fuel channel bow/bulge methodology to 
support the performance of criticality safety evaluation for ATRIUM-
10XM fuel design in the spent fuel pool.
    Date of issuance: October 17, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 263 (Unit 1) and 258 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16231A131; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-29 and DPR-30: The 
amendments revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: May 3, 2016 (81 FR 
26586). The March 9, 2016, supplement corrected a deficiency in the 
Holtec affidavit in the original submittal and did not change the NRC 
staff's initial proposed finding of no significant hazards 
consideration. The June 1, 2016, supplement contained clarifying 
information and did not change the NRC staff's initial proposed finding 
of no significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 17, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 14, 2014, as supplemented by 
letters dated February 12, July 17, August 24, August 28, November 16, 
and December 17, 2015, and February 19, May 6, July 12, and September 
15, 2016.
    Brief description of amendments: The amendments revised the CNP, 
Units 1 and 2, Technical Specifications (TSs) by replacing the limit on 
reactor coolant system (RCS) gross specific activity with a new limit 
on RCS noble gas specific activity. The noble gas specific activity 
limit is based on a new DOSE EQUIVALENT Xenon (Xe)-133 definition that 
replaces the E Bar average disintegration energy definition. In 
addition, the DOSE EQUIVALENT Iodine (I)-131 definition is revised to 
allow the use of additional thyroid dose conversion factors. The 
changes are consistent with NRC-approved industry Technical 
Specifications Task Force (TSTF) Standard Technical Specification 
change traveler, TSTF-490, Revision 0, ``Deletion of E-Bar Definition 
and Revision to Reactor Coolant System Specific Activity Technical 
Specification,'' with approved deviations. Additionally, the amendments 
revised the CNP, Units 1 and 2, licensing basis and TSs to adopt the 
alternative source term as allowed in 10 CFR 50.67.
    Date of issuance: October 20, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment Nos.: 332 for Unit 1 and 314 for Unit 2. A publicly-
available version is in ADAMS under Accession No. ML16242A111; 
documents related 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: March 31, 2015 (80 FR 
17091). The supplemental letters dated July 17, August 24, August 28, 
November 16, and December 17, 2015, and February 19, May 6, July 12, 
and September 15, 2016, provided additional information that clarified 
the application, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 20, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 21, 2016, as supplemented by 
letter dated August 29, 2016.
    Brief description of amendment: The amendment revised Section 2.0, 
``Safety Limits (SLs),'' of the CNS Technical Specifications by 
revising the two recirculation loop and single recirculation loop 
safety limit minimum critical power ratio values to reflect the results 
of a cycle-specific calculation.
    Date of issuance: October 17, 2016.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from Refuel Outage 29.
    Amendment No.: 257. A publicly-available version is in ADAMS under 
Accession No. ML16272A137; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-46: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: July 5, 2016 (81 FR 
43664). The supplemental letter dated August 29, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 17, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 14, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15289A233.
    Brief description of amendment: The amendment revised the DAEC 
Technical Specifications Section 5.5.6, ``Inservice Testing Program,'' 
to provide consistency with the requirements of 10 CFR 50.55a(f)(4) for 
inservice testing of pumps and valves and remove requirements that are 
redundant to the requirements of 10 CFR 50.55a.
    Date of issuance: October 17, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 298. A publicly-available version is in ADAMS under 
Accession No. ML16263A245; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-49: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: December 22, 2015 (80 
FR 79621).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 17, 2016.
    No significant hazards consideration comments received: No.

[[Page 78657]]

Northern States Power Company--Minnesota, Docket No. 50-282, Prairie 
Island Nuclear Generating Plant, Unit 1, Goodhue County, Minnesota

    Date of amendment request: April 7, 2016.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Surveillance Requirement (SR) 3.8.4.3 to allow a 
one-time extension of 1 month for the TS SR frequency.
    Date of issuance: October 13, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 7 days of issuance.
    Amendment No.: 218. A publicly-available version is in ADAMS under 
Accession No. ML16256A514; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. DPR-42: Amendment revised the 
Facility Operating License and TSs.
    Date of initial notice in Federal Register: June 21, 2016 (81 FR 
40360).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 13, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 8, 2016.
    Brief description of amendment: The amendment revised the HCGS 
Technical Specifications. Specifically, the safety limit minimum 
critical power ratio for single recirculation loop operation is 
revised. The change results from a cycle-specific analysis performed to 
support the operation of HCGS in upcoming Cycle 21.
    Date of issuance: October 13, 2016.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from the fall 2016 refueling outage.
    Amendment No.: 200. A publicly-available version is in ADAMS under 
Accession No. ML16270A038; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-57: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: August 2, 2016 (81 FR 
50748).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 13, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 25, 2016.
    Brief description of amendments: The amendments updated Attachment 
M, ``License Condition Changes''; Attachment S, ``Modification and 
Implementation Items''; and Attachment W, ``Fire Probabilistic Risk 
Analysis Insights,'' of the previously approved National Fire 
Protection Association (NFPA) 805 amendment.
    Date of issuance: October 17, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 205 (Unit 1) and 201 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16232A000; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-2 and NPF-8: The 
amendments revised the Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: June 7, 2016 (81 FR 
36623).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 17, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 19, 2016, as supplemented by 
letter dated March 1, 2016.
    Description of amendment: The amendments authorized changes to the 
VCSNS, Units 2 and 3, Updated Final Safety Analysis Report in the form 
of departures from the incorporated plant-specific Design Control 
Document Tier 2* information. The changes are related to changes to 
construction methods and construction sequence used for the composite 
floors and roof of the auxiliary building.
    Date of issuance: August 25, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 51 (for Units 2 and 3). A publicly-available 
version is in ADAMS under Package Accession No. ML16202A279; documents 
related to these amendments are listed in the Safety Evaluation 
enclosed with the amendments.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendments 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: March 15, 2016 (81 FR 
13837). The supplemental letter dated March 1, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 25, 2016.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Unit Nos. 1 and 2, Louisa County, Virginia

    Date of amendment request: December 10, 2016, as supplemented by 
letter dated June 15, 2016.
    Brief description of amendments: The amendments modified Technical 
Specification (TS) 3.2.1, ``Heat Flux Hot Channel Factor 
(FQ(Z)).'' The amendments relocate required operating space 
reductions to the Core Operating Limits Report, accompanied by 
verification for each reload cycle, and define TS surveillance 
requirements for steady-state and transient FQ(Z) and 
corresponding actions with which to apply an appropriate penalty factor 
to measured results, as identified in Westinghouse Nuclear Safety 
Advisory Letter (NSAL)-09-5, Revision 1, ``Relaxed Axial Offset Control 
FQ Technical Specification Actions,'' and NSAL-15-1, Revision 0, ``Heat 
Flux Hot Channel Factor Surveillance Requirements,'' respectively.
    Date of issuance: October 17, 2016.
    Effective date: As of the date of issuance and shall be implemented 
before September 30, 2017.
    Amendment Nos.: 278 (Unit No. 1) and 261 (Unit No. 2). A publicly 
available version is in ADAMS under Accession No. ML16252A478; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: March 1, 2016 (81 FR 
10682).

[[Page 78658]]

The supplemental letter dated June 15, 2016, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 17, 2016.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 27th day of October, 2016.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2016-26824 Filed 11-7-16; 8:45 am]
BILLING CODE 7590-01-P