Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 59659-59669 [2016-20391]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 81, Number 168 (Tuesday, August 30, 2016)]
[Notices]
[Pages 59659-59669]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-20391]


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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0180]


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses and Combined Licenses Involving No Significant 
Hazards Considerations

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, 
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is 
publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued, and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from August 2, 2016, to August 15, 2016. The 
last biweekly notice was published on August 16, 2016.

DATES: Comments must be filed by September 29, 2016. A request for a 
hearing must be filed by October 31, 2016.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2016-0180. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Shirley Rohrer, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-5411, email: Shirley.Rohrer@nrc.gov.

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2016-0180, facility name, unit 
number(s),

[[Page 59660]]

plant docket number, application date, and subject when contacting the 
NRC about the availability of information for this action. You may 
obtain publicly-available information related to this action by any of 
the following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2016-0180.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in this 
document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2016-0180, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at https://www.regulations.gov as well as entering the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

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

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

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

    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Agency 
Rules of Practice and Procedure'' in 10 CFR part 2. Interested 
person(s) 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 https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed within 60 days, the Commission or a presiding officer designated 
by the Commission or by the Chief Administrative Judge of the Atomic 
Safety and Licensing Board Panel, will rule on the request and/or 
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 for leave to intervene 
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 requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/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 requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the requestor/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 
requestor/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 requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends

[[Page 59661]]

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 amendment under consideration. The contention must be 
one which, if proven, would entitle the requestor/petitioner to relief. 
A requestor/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, including the opportunity to present evidence and to 
submit a cross-examination plan for cross-examination of witnesses, 
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)-(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 
October 31, 2016. 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 for leave to intervene 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, 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 hearing.docket@nrc.gov, 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 request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission to the NRC,'' which is available on the agency's 
public Web site at https://www.nrc.gov/site-help/electronic-sub-ref-mat.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.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the 
time the 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

[[Page 59662]]

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 request/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 https://www.nrc.gov/site-help/e-submittals.html, by 
email to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m. 
and 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 
https://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 hearing request and petition to intervene 
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.
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

Dominion Nuclear Connecticut, Inc. (DNC), Docket No. 50-336, Millstone 
Power Station, Unit No. 2 (MPS2), New London County, Connecticut

    Date of amendment request: May 25, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16153A026.
    Description of amendment request: The amendment would add the AREVA 
topical report, EMF-2103(P)(A), ``Realistic Large Break [loss of 
coolant accident] LOCA [RLBLOCA] Methodology for Pressurized Water 
Reactors,'' Revision 3, to MPS2 Technical Specification (TS) 6.9.1.8.b, 
``Core Operating Limits Report,'' which lists the analytical methods 
used to determine the core operating limits. The methodology in EMF-
2013(P)(A) for RLBLOCA has been used for the MPS2 LBLOCA analysis of 
the AREVA Standard CE-14 HTP fuel product with M5 cladding, which DNC 
plans to introduce beginning with the fresh fuel for MPS2 Cycle 25 in 
spring 2017.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with NRC staff revisions 
provided in [brackets]:

    1. Does the proposed [amendment] involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No.
    The proposed change to TS 6.9.1.8.b permits the use of the AREVA 
RLBLOCA methodology to analyze the MPS2 LBLOCA to ensure that the 
plant continues to meet the Emergency Core Cooling System (ECCS) 
performance acceptance criteria in 10 CFR 50.46. The RLBLOCA 
analysis demonstrates MPS2 continues to satisfy the 10 CFR 50.46 
ECCS performance acceptance criteria using an NRC-approved 
evaluation model. The proposed change to the list of NRC-approved 
methodologies listed in TS 6.9.1.8.b has no impact on how the plant 
is operated or configured. Addition of this methodology to the list 
of methodologies in TS 6.9.1.8.b does not impact either the 
probability or consequences of an accident currently evaluated in 
Chapter 14 of the [Updated Final Safety Analysis Report] UFSAR.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed [amendment] create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    The proposed change to TS 6.9.1.8.b adds topical report EMF-
2103(P)(A) to the list of approved methodologies for determining 
core operating limits at MPS2. The proposed amendment has no adverse 
effect on plant operation or accident mitigation equipment. The 
amendment does not create any new credible failure mechanisms, 
malfunctions, or accident initiators not considered in the current 
design basis accidents (DBAs). The response of the plant and 
operators following a DBA will not be changed. The proposed 
amendment does not create the possibility of a new failure mode 
associated with any equipment or human performance failures. Thus, 
the possibility of a new or different type of accident is not 
created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from those previously 
evaluated within the FSAR.
    3. Does the proposed [amendment] involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to TS 6.9.1.8.b adds topical report EMF-
2103(P)(A) to the list of approved methodologies for determining 
core operating limits at MPS2. Approved methodologies will be used 
to ensure that the plant continues to meet applicable design 
criteria and safety analysis acceptance criteria. The proposed 
amendment has no [e]ffect on the ability of the plant to mitigate 
DBAs and ensure consequences of the existing DBA remains bounding. 
The margin of safety to mitigate consequences of DBAs is not 
reduced. Structures, systems and components used to mitigate DBAs 
are not affected. No changes are being made to safety limits or 
safety system settings required by TS. 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

[[Page 59663]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: Travis L. Tate.

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

    Date of amendment request: July 11, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16193A005.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to eliminate TS Section 
5.5.7, ``Inservice Testing [IST] Program.'' A new defined term, 
``INSERVICE TESTING PROGRAM,'' is added to the TS Definitions section. 
This amendment request is consistent with Technical Specifications Task 
Force (TSTF)-545, Revision 3, ``TS Inservice Program Removal & Clarify 
SR [Surveillance Requirement] Usage Rule Application to Section 5.5 
Testing.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change would revise TS Chapter 5, Administrative 
Controls, Section 5.5, Programs and Manuals, by eliminating the TS 
5.5.7, Inservice Testing Program, specification. Most requirements 
in the IST Program would be removed, as they are duplicative of 
requirements in the ASME [American Society of Mechanical Engineers] 
OM Code [ASME 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 5.5 IST Program would be 
eliminated because the NRC has determined their inclusion in the TS 
is contrary to regulations. A new defined term, INSERVICE TESTING 
PROGRAM, would be added to the TS, which references the requirements 
of 10 CFR 50.55a(f),
    Performance of IST 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. 
IST 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 IST 
performed. In most cases, the frequency of IST would be 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 accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change would eliminate 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 would allow 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 would 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 would 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 would eliminate a statement 
that nothing in the ASME Code should be construed to supersede the 
requirements of any TS. The NRC has determined that statement to be 
incorrect. 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: Jeanne Cho, Senior Counsel, Entergy 
Services, Inc., 440 Hamilton Ave., White Plains, NY 10601.
    NRC Branch Chief: David J. Wrona.

LaCrosseSolutions, Inc., and Dairyland Power Cooperative, Docket Nos.: 
50-409 and 72-046, La Crosse Boiling Water Reactor (LACBWR), La Crosse 
County, Wisconsin

    Date of amendment request: June 27, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16200A083.
    Description of amendment request: The proposed amendment would 
amend the Possession Only License for the LACBWR to reflect the 
approval of the LACBWR License Termination Plan (LTP) when that review 
and approval process is completed by the NRC staff. The LTP will become 
a supplement to LACBWR's other decommissioning documents and will be 
implemented by the licensee to complete decommissioning activities at 
the LACBWR site. Once decommissioning is complete, a separate request 
will be made to the NRC by the licensee to terminate the LACBWR 
license.
    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?
    The only remaining accident following completion of fuel 
transfer to the [Independent Spent Fuel Storage Installation] 
(ISFSI) is a radioactive release accident where spontaneous release 
of the (non-ISFSI-related) radioactive source term remaining at the 
LACBWR site in a form and quantity is immediately released through 
an airborne or liquid release path.
    A radioactive release analysis was performed to establish the 
bounding event at the site considering the current stage of LACBWR 
decommissioning. 1.175 [Curies]

[[Page 59664]]

(Ci) of radioactive material is conservatively estimated in the 
analysis to be present on plant surfaces, and as such represents the 
assumed total non-ISFSI radioactive source term remaining at the 
LACBWR site. The LACBWR analysis of postulated release events 
separately considers the portion of this remaining radioactive 
contamination that is immediately releasable as airborne 
contamination and that is immediately releasable as contaminated 
liquid.
    A conservative fraction of 30 percent of the total remaining 
source term is assumed in the analysis to be immediately available 
for airborne release. The analysis results demonstrate that the 
consequences of releasing 30 percent of the non-ISFSI radioactive 
source term remaining at the LACBWR site to the atmosphere are well 
within the applicable 10 CFR 100.11 and [U.S. Environmental 
Protection Agency] (EPA) [Protective Action Guides] (PAG) limits.
    The portion of the total remaining source term conservatively 
assumed in the analysis to be available for liquid release at any 
one time is 80 percent of the radioactively contaminated liquid 
stored in the site retention tank. In the unlikely event that 80 
percent of the retention tank volume at a total radionuclide 
concentration of 3.9E-03 [mu]Ci/cc were to be released from the 
retention tank at a flow rate of 20 [gallons per minute] (gpm), the 
normal effluent concentration limits of 10 CFR 20, Appendix B, Table 
2, would not be exceeded. Thus, the liquid release analysis 
demonstrates that there is no reasonable likelihood that a 
postulated radioactive liquid release event could result in 
exceeding the normal effluent concentration limits of 10 CFR 20, 
Appendix B.
    With consideration for the current stage of LACBWR 
decommissioning and with spent nuclear fuel now stored in the ISFSI, 
the bounding radioactive release analysis, for both airborne and 
liquid releases, confirms that the minimal radioactive material 
resulting from LACBWR operation and remaining on the LACBWR site is 
insufficient for any potential event to result in exceeding dose 
limits or otherwise involving a significant adverse effect on public 
health and safety.
    The proposed change does not affect the boundaries used to 
evaluate compliance with liquid or gaseous effluent limits, and has 
no impact on plant operations. The proposed changes do not have an 
adverse impact on the remaining decommissioning activities or any 
decommissioning related postulated accident consequences.
    The proposed changes related to the approval of the LTP do not 
affect operating procedures or administrative controls that have the 
function of preventing or mitigating the remaining decommissioning 
design basis accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The accident analysis for the facility related to 
decommissioning activities is described in the [Decommissioning 
Plan/Post-Shutdown Decommissioning Activities Report] (D-Plan/
PSDAR). The requested license amendment is consistent with the plant 
activities described in the D-Plan/PSDAR. Thus, the proposed changes 
do not affect the remaining plant systems, structures, or components 
in a way not previously evaluated.
    There are sections of the LTP that refer to the decommissioning 
activities still remaining. These activities are performed in 
accordance with approved site processes and undergo a 10 CFR 50.59 
review as required prior to initiation. The proposed amendment 
merely makes mention of these processes and does not bring about 
physical changes to the facility.
    Therefore, the facility conditions for which the remaining 
postulated accident has been evaluated is still valid and no new 
accident scenarios, failure mechanisms, or single failures are 
introduced by this amendment. The system operating procedures are 
not affected.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The LTP is a plan for demonstrating compliance with the 
radiological criteria for license termination as provided in 10 CFR 
20.1402. The margin of safety defined in the statements of 
consideration for the final rule on the Radiological Criteria for 
License Termination is described as the margin between the 100 
[millirem per year] (mrem/yr) public dose limit established in 10 
CFR 20.1301 for licensed operation and the 25 mrem/yr dose limit to 
the average member of the critical group at a site considered 
acceptable for unrestricted use (one of the criteria of 10 CFR 
20.1402). This margin of safety accounts for the potential effect of 
multiple sources of radiation exposure to the critical group. Since 
the License Termination Plan is designed to comply with the 
radiological criteria for license termination for unrestricted use, 
the LTP supports this margin of safety.
    In addition, the LTP provides the methodologies and criteria 
that will be used to perform remediation activities of residual 
radioactivity to demonstrate compliance with the [As Low As 
Reasonably Achievable] (ALARA) criterion of 10 CFR 20.1402.
    Additionally, the LTP is designed with recognition that (a) the 
methods in MARSSIM (Multi-Agency Radiation Survey and Site 
Investigation Manual) and (b) the building surface contamination 
levels are not directly applicable to use with complex nonstructural 
components. Therefore, the LTP states that nonstructural components 
remaining in buildings (e.g., pumps, heat exchangers, etc.) will be 
evaluated against the criteria of Regulatory Guide 1.86, 
``Termination of Operating Licenses for Nuclear Reactors,'' to 
determine if the components can be released for unrestricted use. 
The LTP also states that materials, surveyed and evaluated as a-part 
of normal decommissioning activities and prior to implementation of 
the final radiation surveys, will be surveyed for release using 
current site procedures to demonstrate compliance with the ``no 
detectable'' criteria. Such materials that do not pass these 
criteria will be controlled as contaminated.
    Also, as previously discussed, the bounding radioactive release 
accident analysis for decommissioning is based on a conservative 
estimate of the radioactive material remaining onsite. Since the 
bounding accident results in a release of more airborne and liquid 
radioactivity than can be released from planned LTP decommissioning 
events, the margin of safety associated with the consequences of 
decommissioning accidents is not reduced by this activity.
    Thus, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Russ Workman, General Counsel, Energy 
Solutions, 299 South Main Street, Suite 1700, Salt Lake City, Utah 
84111.
    NRC Branch Chief: Bruce Watson.

South Carolina Electric and 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: June 16, 2016, as revised August 8, 
2016. Publicly-available versions are in ADAMS under Accession Nos. 
ML16168A257 and ML16221A649, respectively.
    Description of amendment request: The requested amendment proposes 
to depart from approved AP1000 Design Control Document Tier 2* and 
associated Tier 2 information in the Updated Final Safety Analysis 
Report (UFSAR). Specifically, the requested amendment proposes to 
depart from UFSAR text and figures that describe the connections 
between floor modules and structural wall modules in the containment 
internal structures.
    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?

[[Page 59665]]

    Response: No.
    The design functions of the nuclear island structures are to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the nuclear island. 
The nuclear island structures are structurally designed to meet 
seismic Category I requirements as defined in Regulatory Guide 1.29.
    The change of the design details for the floor modules and the 
connections between floor modules and the structural wall modules, 
and the change to more clearly state the design requirement that 
these connections meet criteria and requirements of American 
Concrete Institute (ACI) 349 and American Institute of Steel 
Construction (AISC) N690, do not have an adverse impact on the 
response of the nuclear island structures to safe shutdown 
earthquake ground motions or loads due to anticipated transients or 
postulated accident conditions. The change of the design details for 
the connections between floor modules and the structural wall 
modules, and the clarification of design requirements for these 
connections, do not impact the support, design, or operation of 
mechanical and fluid systems. There is no change to plant systems or 
the response of systems to postulated accident conditions. There is 
no change to the predicted radioactive releases due to normal 
operation or postulated accident conditions. The plant response to 
previously evaluated accidents or external events is not adversely 
affected, nor does the change described create any new accident 
precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change is to revise design details for the floor 
modules and the connections between floor modules and the structural 
wall modules, and more clearly state the design requirement that 
these connections meet criteria and requirements of ACI 349 and AISC 
N690. The clarification and changes to the design details for the 
floor modules and the connections between floor modules and the 
structural wall modules do not change the design requirements of the 
nuclear island structures. The clarification and changes of the 
design details for the floor modules and the connections between 
floor modules and the structural wall modules do not result in a new 
failure mechanism for the nuclear island structures or new accident 
precursors. As a result, the design function of the nuclear island 
structures is not adversely affected by the proposed change.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes, thus, no margin of 
safety is reduced. The acceptance limits for the design of seismic 
Category I structures are included in the codes and standards used 
for the design, analysis, and construction of the structures. The 
two primary codes for the seismic Category I structures are American 
Institute of Steel Construction (AISC) N690 and American Concrete 
Institute (ACI) 349. The changes to the design of the connection of 
the floor module to the structural wall modules in the containment 
internal structures satisfy applicable provisions of AISC N690 and 
ACI 349 and supplemental requirements included in the UFSAR.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety previously evaluated.

    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.
    Acting NRC Branch Chief: Jennifer Dixon-Herrity.

South Carolina Electric and 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: July 19, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16202A035.
    Description of amendment request: The amendment request proposes 
changes to the Technical Specifications and Updated Final Safety 
Analysis Report (UFSAR) Tier 2 information to update the Protection and 
Safety Monitoring System (PMS) to align with the requirements in 
Institute of Electrical and Electronics Engineers (IEEE) 603-1991, 
``IEEE Standard Criteria for Safety Systems for Nuclear Power 
Generating Stations.'' IEEE 603-1991, Clause 6.6, ``Operating 
Bypasses,'' imposes requirements on the operating bypasses (i.e., 
``blocks'' and ``resets'') used for the AP1000 PMS. The PMS functional 
logic for blocking the source range neutron flux doubling signal shown 
in UFSAR Figure 7.2-1 (Sheet 3) requires revision to fully comply with 
this requirement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change modifies the PMS logic used to terminate an 
inadvertent boron dilution accident which results in a source range 
flux doubling signal. An inadvertent boron dilution is caused by the 
failure of the demineralized water transfer and storage system or 
chemical and volume control system, either by controller, operator 
or mechanical failure. The proposed changes to PMS and Technical 
Specification requirements do not adversely affect any of these 
accident initiators or introduce any component failures that could 
lead to a boron dilution event; thus the probabilities of accidents 
previously evaluated are not affected. The proposed changes do not 
adversely interface with or adversely affect any system containing 
radioactivity or affect any radiological material release source 
term; thus the radiological releases in an accident are not 
affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The accident analysis evaluates events involving a decrease in 
reactor coolant system boron concentration due to a malfunction of 
the chemical and volume control system in Modes 1 through 6. The 
Technical Specifications currently provide administrative controls 
to prevent a boron dilution event in Mode 6. The proposed change 
would provide additional PMS interlocks and administrative controls 
for prevention of a boron dilution event applicable in Modes 2, 3, 
4, and 5. The proposed changes to the PMS design do not adversely 
affect the design or operation of safety related equipment or 
equipment whose failure could initiate an accident from what is 
already described in the licensing basis. These changes do not 
adversely affect fission product barriers. No safety analysis or 
design basis acceptance limit/criterion is challenged or exceeded by 
the requested change.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change would add additional restrictions on the 
source range flux doubling signal operational bypass to align it 
with the requirements in IEEE 603 and provide assurance that the 
protection logic is enabled whenever the plant is in a condition 
where protection might be required. These changes to the PMS design 
do not adversely impact nor affect the design, construction, or 
operation of any plant [structure, system, and components (SSCs)], 
including any equipment whose failure could initiate an accident or 
a failure of a fission product barrier. No analysis is adversely

[[Page 59666]]

affected by the proposed changes. Furthermore, no system function, 
design function, or equipment qualification will be adversely 
affected by the changes.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety previously evaluated.

    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.
    Acting NRC Branch Chief: Jennifer Dixon-Herrity.

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

    Date of amendment request: July 25, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16207A340.
    Description of amendment request: The amendment request proposes 
changes to a plant-specific Tier 1 (and combined license Appendix C) 
table and the Updated Final Safety Analysis Report (UFSAR) tables to 
clarify the flow area for the Automatic Depressurization System (ADS) 
fourth stage squib valves and to reduce the minimum effective flow area 
for the second and third stage ADS control valves. Pursuant to the 
provisions of 10 CFR 52.63(b)(1), an exemption from elements of the 
design as certified in the 10 CFR part 52, Appendix D, design 
certification rule is also requested for the plant-specific Design 
Control Document Tier 1 material departures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not adversely affect the operation of 
any systems or equipment that initiate an analyzed accident or alter 
any structures, systems, and components (SSC) accident initiator or 
initiating sequence of events. The proposed changes do not adversely 
affect the physical design and operation of the second and third 
stage ADS control valves and fourth stage ADS squib valves, 
including as-installed inspections, testing, and maintenance 
requirements, as described in the UFSAR. Therefore, the operation of 
the second and third stage ADS control valves and fourth stage ADS 
squib valves is not adversely affected.
    The proposed changes do not adversely affect the ability of the 
second and third stage ADS control valves and fourth stage ADS squib 
valves to perform their design functions. The designs of the second 
and third stage ADS control valves and fourth stage ADS squib valves 
continue to meet the same regulatory acceptance criteria, codes, and 
standards as required by the UFSAR. In addition, the proposed 
changes maintain the capabilities of the second and third stage ADS 
control valves and fourth stage ADS squib valves to mitigate the 
consequences of an accident and to meet the applicable regulatory 
acceptance criteria. The proposed changes do not adversely affect 
the prevention and mitigation of other abnormal events, e.g., 
anticipated operational occurrences, earthquakes, floods and turbine 
missiles, or their safety or design analyses. Therefore, the 
consequences of the accidents evaluated in the UFSAR are not 
affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that may initiate a new or different kind of accident, 
or alter any SSC such that a new accident initiator or initiating 
sequence of events is created. The proposed changes do not adversely 
affect the physical design and operation of the second and third 
stage ADS control valves and fourth stage ADS squib valves, 
including as-installed inspections, testing, and maintenance 
requirements, as described in the UFSAR. Therefore, the operation of 
the second and third stage ADS control valves and fourth stage ADS 
squib valves is not adversely affected. These proposed changes do 
not adversely affect any other SSC design functions or methods of 
operation in a manner that results in a new failure mode, 
malfunction or sequence of events that affect safety-related or 
nonsafety-related equipment. Therefore, this activity does not allow 
for a new fission product release path, result in a new fission 
product barrier failure mode, or create a new sequence of events 
that results in significant fuel cladding failures.
    Therefore, the 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 existing safety margins. The 
proposed changes maintain the capabilities of the second and third 
stage ADS control valves and fourth stage ADS squib valves to 
perform their design functions. The proposed changes maintain 
existing safety margin through continued application of the existing 
requirements of the UFSAR, while updating the acceptance criteria 
for verifying the design features necessary to confirm the second 
and third stage ADS control valves and fourth stage ADS squib valves 
perform the design functions required to meet the existing safety 
margins in the safety analyses. Therefore, the proposed changes 
satisfy the same design functions in accordance with the same codes 
and standards as stated in the UFSAR. These changes do not adversely 
affect any design code, function, design analysis, safety analysis 
input or result, or design/safety margin.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes, and no margin of 
safety is reduced. Therefore, the requested amendment does not 
involve a significant reduction in a margin of safety.

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

Southern Nuclear Operating Company, Inc., Docket No. 50-425, Vogtle 
Electric Generating Plant, Unit 2, Burke County, Georgia

    Date of amendment request: August 12, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16225A619.
    Description of amendment request: The licensee proposes to modify 
the Vogtle Electric Generating Plant, Unit 2, Technical Specifications 
(TSs) Limiting Condition for Operation 3.7.9, ``Ultimate Heat Sink 
(UHS),'' such that with the 2B Nuclear Service Cooling Water (NSCW) 
transfer pump inoperable for refurbishment, the Completion Time of 
Condition 3.7.9.D.2.2 would be 46 days as opposed to 31 days. This TS 
change would be a one-time change and in effect only for the 2B NSCW 
transfer pump for the remainder of Cycle 19.
    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.

[[Page 59667]]

    The proposed change does not alter any plant equipment or 
operating practices in such a manner that the probability of an 
accident is increased. The proposed changes will not alter 
assumptions relative to the mitigation of an accident or transient 
event. Furthermore, the UHS will remain capable of adequately 
responding to a design basis event during the period of the extended 
CT [Completion Time]. 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 accident from any accident previously evaluated?
    Response: No.
    The proposed change does not introduce any new or unanalyzed 
modes of operation. The refurbishment of the pump does not involve 
any unanalyzed modifications to the design or operational limits of 
the NSCW system. The redundant pump and compensatory measures 
allowed by the Technical Specifications will remain unaffected. 
Therefore, no new failure modes or accident precursors are created 
due to the pump refurbishment during the extended Completion Time. 
For the reasons noted above, the proposed change will not create the 
possibility of a new or different accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is related to the ability of the fission 
product barriers to perform their design functions during and 
following an accident. These barriers include the fuel cladding, the 
reactor coolant system, and the containment. The performance of 
these fission product barriers will not be affected by the proposed 
change; therefore, the margin to the onsite and offsite radiological 
dose limits are not significantly reduced.
    During the extended Completion Time for the 2B NSCW transfer 
pump, the NSCW system and the UHS will remain capable of mitigating 
the consequences of a design basis event such as a LOCA [loss-of-
coolant accident]. Technical Specifications Action 3.7.9.D.2.1 will 
be taken to provide an alternate method of basin transfer.
    For the reasons noted above, there is no significant reduction 
in a margin of safety.

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

Attorney for licensee: Jennifer M. Buettner, Associate General Counsel, 
Southern Nuclear Operating Company, Inc., 40 Inverness Center Parkway, 
Birmingham, AL 35242.

    NRC Branch Chief: Michael T. Markley.

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

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

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power 
Station, Unit No. 3 (MPS3), New London County, Connecticut

    Date of amendment request: August 31, 2015.
    Brief description of amendment: The amendment revised the MPS3 
Design Features--Fuel Storage Technical Specification 5.6.3, 
``Capacity,'' to specify the spent fuel pool storage capacity limit in 
terms of the total number of fuel assemblies.
    Date of issuance: August 4, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 270. A publicly-available version is in ADAMS under 
Accession No. ML16206A001; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-49: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: November 24, 2015 (80 
FR 73235).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 4, 2016.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne 
County, Mississippi

    Date of application for amendment: September 15, 2015.
    Brief description of amendment: The amendment revised the GGNS 
Technical Specifications (TSs) to eliminate the ``Inservice Testing 
Program,'' specification in Section 5.5, ``Programs and Manuals,'' 
which is superseded by Code Case OMN-20. A new defined term, 
``INSERVICE TESTING PROGRAM,'' would be added to TS Section 1.1, 
``Definitions.'' This request is consistent with TS Task Force (TSTF)-
545, Revision 1, ``TS Inservice Testing Program Removal & Clarify SR 
[Surveillance Requirement] Usage Rule Application to Section 5.5 
Testing.''
    Date of issuance: August 4, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No: 211. A publicly-available version is in ADAMS under 
Accession No. ML16140A133; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-29: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 1, 2016 (81 FR 
10679).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 4, 2016.
    No significant hazards consideration comments received: No.

[[Page 59668]]

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station (Braidwood), Units 1 and 2, Will County, Illinois and 
Docket Nos. STN 50-454 and STN 50-455, Byron Station (Byron), Unit Nos. 
1 and 2, Ogle County, Illinois

    Date of application for amendments: February 23, 2016.
    Brief description of amendment: The amendments revise technical 
specifications (TSs) 4.2.1, ``Fuel Assemblies,'' and 5.6.5, ``Core 
Operating Limits Report (COLR),'' to allow the use of Optimized 
ZIRLO\TM\ fuel cladding material in Braidwood, Units 1 and 2, and 
Byron, Unit Nos. 1 and 2 and to add WCAP-12610-P-A, ``VANTAGE+ Fuel 
Assembly Reference Core Report,'' and Addendum 1-A to Topical Report 
WCAP-12610-P-A and CENPD-404-P-A, ``Optimized ZIRLO'' to the list of 
documents previously reviewed and approved by the NRC.
    Date of issuance: August 1, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos: 190/196. A publicly-available version is in ADAMS 
under Accession No. ML16180A251; documents related to these amendments 
are listed in the Safety Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and 
NPF-66: The amendments revised the Technical Specifications and 
Licenses.
    Date of initial notice in Federal Register: May 10, 2016 (81 FR 
28897).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 1, 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: March 24, 2016, as supplemented by 
letter dated May 11, 2016.
    Brief description of amendments: The amendments revised the 
frequency for cycling of the recirculation pump discharge valves as 
specified in Technical Specification (TS) Surveillance Requirement (SR) 
3.5.1.5. Specifically, the amendments changed the frequency for the SR 
such that it is performed in accordance with the Inservice Testing 
Program.
    Date of issuance: August 10, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendments Nos.: 309 (Unit 2) and 313 (Unit 3). A publicly-
available version is in ADAMS under Accession No. ML16165A002; 
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 TSs.
    Date of initial notice in Federal Register: June 7, 2016 (81 FR 
36619).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 10, 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: January 29, 2016.
    Brief description of amendments: The amendments revised the CNP, 
Units 1 and 2, technical specification (TS) requirements to address 
Generic Letter 2008-01, ``Managing Gas Accumulation in Emergency Core 
Cooling, Decay Heat Removal, and Containment Spray Systems,'' as 
described in the Technical Specifications Task Force (TSTF) Traveler, 
TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas 
Accumulation.''
    Date of issuance: August 4, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment Nos.: 331--Unit 1 and 312--Unit 2. A publicly-available 
version is in ADAMS under Accession No. ML16195A004; documents related 
to this amendment are listed in the Safety Evaluation enclosed with the 
amendment.
    Renewed Facility Operating License Nos. DPR-58 and DPR-74: The 
amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: March 15, 2016 (81 FR 
13843).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 4, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 24, 2014, as supplemented by 
letter dated September 28, 2015.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) by adopting 21 previously NRC-approved 
Technical Specifications Task Force (TSTF) Travelers and one request 
not associated with TSTF Travelers. SNC stated that these TSTF 
Travelers are generic changes chosen to increase the consistency 
between the Joseph M. Farley Nuclear Plant, Units 1 and 2; the Improved 
Standard Technical Specifications for Westinghouse plants (NUREG-1431); 
and the TSs of the other plants in the SNC fleet.
    Date of issuance: August 3, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: 203 (Unit 1) and 199 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML15233A448; 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 and TSs.
    Date of initial notice in Federal Register: February 3, 2015 (80 FR 
5804). The supplemental letter dated September 28, 2015, 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 August 3, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 16, 2016.
    Brief description of amendments: The amendments revised the 
Technical Specifications to allow the use of Optimized 
ZIRLOTM as an approved fuel rod cladding.
    Date of issuance: August 4, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 182 (Unit 1) and 163 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16179A386; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.

[[Page 59669]]

    Renewed Facility Operating License Nos. NPF-68 and NPF-81: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: May 24, 2016 (81 FR 
32809).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 4, 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: March 16, 2016.
    Brief description of amendments: The amendments revised the 
Technical Specifications to allow the use of Optimized 
ZIRLOTM as an approved fuel rod cladding.
    Date of issuance: August 4, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 204 (Unit 1) and 200 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16179A386; 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: Amendments 
revised the Renewed Facility Operating Licenses and Technical 
Specifications.
    Date of initial notice in Federal Register: May 24, 2016 (81 FR 
32808).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 4, 2016.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: January 27, 2016, as supplemented by 
letter dated May 19, 2016.
    Brief description of amendment: The amendment revised the Technical 
Specifications to allow the use of Optimized ZIRLO\TM\ as an approved 
fuel rod cladding.
    Date of issuance: The amendment is effective upon issuance and 
shall be implemented within 90 days of the date of issuance.
    Effective date: August 3, 2016.
    Amendment No.: 216. A publicly-available version is in ADAMS under 
Accession No. ML16179A293; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-42. The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 12, 2016 (81 FR 
21603). The supplemental letter dated May 19, 2016, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the 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 3, 2016.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 18th day of August 2016.
    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2016-20391 Filed 8-29-16; 8:45 am]
 BILLING CODE 7590-01-P