[Federal Register Volume 81, Number 40 (Tuesday, March 1, 2016)]
[Notices]
[Pages 10675-10686]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-04346]


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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0040]


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 February 2, 2016, to February 12, 2016. The 
last biweekly notice was published on February 16, 2016.

DATES: Comments must be filed by March 31, 2016. A request for a 
hearing must be filed by May 2, 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 http://www.regulations.gov and search for Docket ID NRC-2016-0040. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected].
     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: Janet Burkhardt, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-1384, email: [email protected].

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2016-0040 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-2016-0040.
     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 the 
SUPPLEMENTARY INFORMATION section.
     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-2016-0040, facility name, unit 
number(s), 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 http://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

[[Page 10676]]

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 should 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. Should the 
Commission take 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. Should the Commission make 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 http://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 to rely to establish those facts or expert 
opinion. 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 NRC 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 May 
2, 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

[[Page 10677]]

section, except that under Sec.  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. Persons desiring to make a limited 
appearance are requested to inform the Secretary of the Commission by 
May 2, 2016.

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). 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 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 http://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,'' which is available on the agency's public Web 
site at http://www.nrc.gov/site-help/e-submittals.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 Meta System 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. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    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 http://www.nrc.gov/site-help/e-submittals.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 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 Meta System 
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 Meta System Help Desk is available between 8 a.m. and 8 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 
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

[[Page 10678]]

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 request 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.
    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).
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

Duke Energy Progress, Inc., Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2 (HBRSEP2), Darlington County, South Carolina

    Date of amendment request: November 2, 2015, as supplemented by 
letter dated December 22, 2015. Publicly-available versions are in 
ADAMS under Accession Nos. ML15307A069 and ML15356A481, respectively.
    Description of amendment request: The proposed amendment would 
revise the reactor coolant system (RCS) pressure and temperature (P/T) 
limits by replacing Technical Specification (TS) Section 3.4.3, ``RCS 
Pressure and Temperature (P/T) Limits,'' Figures 3.4.3-1 and 3.4.3-2, 
with figures that are applicable up to 50 effective full power years 
(EFPY).
    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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed RCS P/T limits are based on NRC-approved 
methodology and will continue to maintain appropriate limits for the 
HBRSEP2 RCS up to 50 EFPY. These changes provide appropriate limits 
for pressure and temperature during heatup and cooldown of the RCS, 
thus ensuring that the probability of RCS failure is maintained 
acceptably low. These limits are not directly related to the 
consequences of accidents.
    Therefore, the proposed amendment does not result in an increase 
in the probability or consequences of an accident previously 
evaluated.

2. Does the Proposed Change Create the Possibility of a New or 
Different Kind of Accident From Any Accident Previously Evaluated?

    Response: No.
    The proposed changes will continue to ensure that the RCS will 
be maintained within appropriate pressure and temperature limits 
during heatup and cooldown. No physical changes to the HBRSEP2 
systems, structures, or components are being implemented. There are 
no new or different accident initiators or sequences being created 
by the proposed Technical Specifications changes.
    Therefore, these changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

3. Does the Proposed Change Involve a Significant Reduction in a Margin 
of Safety?

    Response: No.
    The proposed changes ensure that the margin of safety for the 
fission product barriers protected by these functions will continue 
to be maintained. This conclusion is based on use of the applicable 
NRC-approved methodology for developing and establishing the 
proposed RCS P/T limits.
    Therefore, these changes do not involve a significant reduction 
in the margin of safety.

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

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

    Date of amendment request: January 15, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16015A456.
    Description of amendment request: The licensee has provided a 
formal notification to the NRC of the intention to permanently cease 
power operations of JAF at the end of the current operating cycle. Once 
certifications for permanent cessation of operation and permanent 
removal of fuel from the reactor are submitted to the NRC, certain 
staffing and training Technical Specifications (TSs) administrative 
controls will no longer be applicable or appropriate for the 
permanently defueled condition. Therefore, ENO is requesting approval 
of changes to the staffing and training requirements in Section 5.0, 
Administrative Controls, of the JAF TSs. Specifically, the amendment 
would revise and remove certain requirements in TS Sections 5.1, 
``Responsibility,'' 5.2, ``Organization,'' and 5.3, ``Plant Staff 
Qualifications.'' The proposed amendment would not be effective until 
the certification of permanent cessation of operation and certification 
of permanent removal of fuel from the reactor vessel are submitted to 
the NRC.
    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, with NRC staff revisions provided in [brackets], which 
is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would not take effect until JAF has 
permanently ceased operation and entered a permanently defueled 
condition. The proposed amendment would modify the JAF TS by 
deleting the portions of the TS that are no longer applicable to a 
permanently defueled facility, while modifying the other sections to 
correspond to the permanently defueled condition.
    The deletion and modification of provisions of the 
administrative controls do not directly affect the design of 
structures, systems, and components (SSCs) necessary for safe 
storage of irradiated fuel or the methods used for handling and 
storage of such fuel in the fuel pool. The changes to the 
administrative controls are administrative in nature and do not 
affect any accidents applicable to the safe management of irradiated 
fuel or the permanently shutdown and defueled condition of the 
reactor.
    In a permanently defueled condition, the only credible accident 
is the fuel handling accident [(FHA)].
    The probability of occurrence of previously evaluated accidents 
is not increased, since extended operation in a defueled condition 
will be the only operation allowed, and therefore bounded by the 
existing analyses. Additionally, the occurrence of postulated

[[Page 10679]]

accidents associated with reactor operation is no longer credible in 
a permanently defueled reactor. This significantly reduces the scope 
of applicable accidents.
    Therefore, the proposed amendment does not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated.

    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes have no impact on facility SSCs affecting 
the safe storage of irradiated fuel, or on the methods of operation 
of such SSCs, or on the handling and storage of irradiated fuel 
itself. The administrative removal of or modifications of the TS 
that are related only to administration of facility cannot result in 
different or more adverse failure modes or accidents than previously 
evaluated because the reactor will be permanently shutdown and 
defueled and JAF will no longer be authorized to operate the 
reactor.
    The proposed deletion of requirements of the JAF TS do not 
affect systems credited in the accident analysis for the [FHA] at 
JAF. The proposed TS will continue to require proper control and 
monitoring of safety significant parameters and activities.
    The proposed amendment does not result in any new mechanisms 
that could initiate damage to the remaining relevant safety barriers 
for defueled plants (fuel cladding and spent fuel cooling). Since 
extended operation in a defueled condition will be the only 
operation allowed, and therefore bounded by the existing analyses, 
such a condition does not create the possibility of a new or 
different kind 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Because the 10 CFR part 50 license for JAF will no longer 
authorize operation of the reactor or emplacement or retention of 
fuel into the reactor vessel once the certifications required by 10 
CFR 50.82(a)(1) are submitted, as specified in 10 CFR 50.82(a)(2), 
the occurrence of postulated accidents associated with reactor 
operation is no longer credible. The only remaining credible 
accident is a [FHA]. The proposed amendment does not adversely 
affect the inputs or assumptions of any of the design basis analyses 
that impact the FHA.
    The proposed changes are limited to those portions of the [TS] 
that are not related to the safe storage of irradiated fuel. The 
requirements that are proposed to be revised or deleted from the JAF 
[TS] are not credited in the existing accident analysis for the 
remaining applicable postulated accident; and as such, do not 
contribute to the margin of safety associated with the accident 
analysis. Postulated DBAs [Design Basis Accidents] involving the 
reactor are no longer possible because the reactor will be 
permanently shutdown and defueled and JAF will no longer be 
authorized to operate the reactor.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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 amendment request: September 15, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15259A042.
    Description of amendment request: The amendment would revise the 
GGNS Technical Specifications (TSs) to eliminate the ``Inservice 
Testing [IST] 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.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with NRC edits in [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. 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.'' Other requirements 
in the Section 5.5 IST Program are eliminated because the NRC has 
determined their inclusion in the TS is contrary to regulations. A 
new defined term, ``Inservice Testing Program,'' is added to the TS, 
which references the requirements of 10 CFR 50.55a(f). The proposed 
change also revises the SR Section 3.0, ``SR Applicability,'' Bases 
to explain the application of the usage rules to the Section 5.5 
testing requirements.
    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 periods under Code Case OMN-20 are 
equivalent to the current testing period allowed by the TS with the 
exception that testing periods 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 period 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.
    The proposed [changes to the] SR 3.0 Bases clarify the 
appropriate application of the existing TS requirements. Since the 
proposed change does not significantly affect system Operability, 
the proposed change will have no significant effect on the 
initiating events for accidents previously evaluated and will have 
no significant effect on the ability of the systems to mitigate 
accidents previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change 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. The 
proposed Bases change does not change the Operability requirements 
for plant systems or the actions taken when plant systems are not 
operable. The proposed Bases change clarifies the current 
application of the specifications.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.


[[Page 10680]]


    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 periods 
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 
period 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 
period, 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. The NRC has 
determined that statement to be incorrect. However, elimination of 
the statement will have no effect on plant operation or safety. The 
proposed changes to the SR 3.0 Bases clarify the application of the 
existing TS requirements and, as a result, have no significant 
effect on a margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Meena K. Khanna.

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

    Date of amendment request: December 23, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15357A250.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) Limiting Condition for Operation (LCO) 
3.10.1, to expand its scope to include provisions for temperature 
excursions greater than 212 degrees Fahrenheit ([deg]F) as a 
consequence of inservice leak and hydrostatic testing, and as a 
consequence of scram time testing initiated in conjunction with an 
inservice leak or hydrostatic test, while considering operational 
conditions to be in Mode 4. The proposed change is based on NRC-
approved Technical Specification Task Force (TSTF) Improved Standard 
Technical Specifications Change Traveler, TSTF-484, Revision 0, ``Use 
of TS 3.10.1 for Scram Time Testing Activities.''
    The NRC staff issued a Notice of Availability for TSTF-484 in the 
Federal Register on October 27, 2006 (71 FR 63050). The staff also 
issued a Federal Register notice on August 21, 2006 (71 FR 48561) that 
provided a model safety evaluation and a model no significant hazards 
consideration (NSHC) determination that licensees could reference in 
their plant-specific applications. In its application dated December 
23, 2015, the licensee affirmed the applicability of the model NSHC 
determination for PBAPS, Units 2 and 3.
    Basis for proposed NSHC determination: As required by 10 CFR 
50.91(a), the licensee has provided its analysis of the issue of NSHC, 
which is presented below:

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

    Technical Specifications currently allow for operation at 
greater than 212[emsp14][deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact the probability or consequences of an accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

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

    Technical Specifications currently allow for operation at 
greater than 212[emsp14][deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. No new operational conditions beyond those currently allowed by 
LCO 3.10.1 are introduced. The changes do not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements or eliminate any existing requirements. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are consistent with the safety analysis assumptions 
and current plant operating practice.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

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

    Technical Specifications currently allow for operation at 
greater than 212[emsp14][deg]F while imposing MODE 4 requirements in 
addition to the secondary containment requirements required to be 
met. Extending the activities that can apply this allowance will not 
adversely impact any margin of safety. Allowing completion of 
inspections and testing and supporting completion of scram time 
testing initiated in conjunction with an inservice leak or 
hydrostatic test prior to power operation results in enhanced safe 
operations by eliminating unnecessary maneuvers to control reactor 
temperature and pressure.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    Based on the above, 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 NSHC.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL 
60555.
    NRC Branch Chief: Douglas A. Broaddus.

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

    Date of amendment request: December 14, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15348A224.
    Description of amendment request: The amendment proposes to revise 
the technical specifications to increase the minimum required fuel oil 
in each standby diesel generator (DG) fuel oil day tank.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not increase the probability or the 
consequences of an accident previously evaluated. The DGs and

[[Page 10681]]

their associated emergency buses function to mitigate accidents. The 
proposed change does not involve a change in the operational limits 
or the design of the electrical power systems, change the function 
or operation of plant equipment, or affect the response of that 
equipment when called upon to operate.
    The proposed change to TS SR 3.8.1.4 confirms the minimum supply 
of fuel oil in each DG fuel oil day tank. The minimum value for the 
affected parameter is being increased in the conservative direction 
and assures the DGs' ability to fulfill their safety function.
    Therefore, based on the discussion above, 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 involve a change in the operational 
limits or the design capabilities of the electrical power systems. 
The proposed change does not alter the function or operation of 
plant equipment or introduce any new failure mechanisms. The 
evaluation that supports this request included a review of the DG 
fuel oil system to which this parameter applies.
    Therefore, this change 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?
    Response: No.
    Margins of safety are related to the confidence in 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 systems. 
Since the proposed change does not adversely affect the operation of 
any plant equipment, including equipment credited in protecting the 
fission product barriers, operation in the proposed manner will 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 Acting Branch Chief: Justin C. Poole.

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

    Date of amendment request: August 31, 2015, as superseded by letter 
dated December 23, 2015. Publicly-available versions are in ADAMS under 
Accession Nos. ML15243A167 and ML15363A042, respectively.
    Description of amendment request: The licensee proposes to revise 
the FCS Updated Safety Analysis Report (USAR) to change the structural 
design methodology for Class I structures at FCS to use American 
Concrete Institute (ACI) ultimate strength requirements, with the 
exception of the containment structure (cylinder, dome, and base mat), 
the spent fuel pool, and the foundation mats. No change to the current 
licensing basis code of record is proposed for the excepted 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?
    Response: No.
    This LAR [license amendment request] revises the methodology 
used to design new or re-evaluate existing Class I structures other 
than the containment structure (cylinder, dome, and base mat), the 
spent fuel pool (SFP), and the foundation mats. These structures 
will continue to utilize the current license basis and thus are not 
affected by this change. The proposed change allows other Class I 
structures to apply the ultimate strength design (USD) method from 
the ACI 318-63 Code for normal operating/service load combinations.
    The ACI USD method is an accepted industry standard used for the 
design and analysis of reinforced concrete. A change in the 
methodology that an analysis uses to verify structure qualifications 
does not have any impact on the probability of accidents previously 
evaluated. Designs performed with the ACI USD method will continue 
to demonstrate that the Class I structures meet industry accepted 
ACI Code requirements. This LAR does not propose changes to the no 
loss-of-function loads, loading combinations, or required ultimate 
strength capacity.
    Calculations that apply the limit design method and use dynamic 
increase factors (DIF) of ACI 349-97, Appendix C will demonstrate 
that the concrete structures meet required design criteria. 
Therefore, these proposed changes will not pose a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The use of actual concrete strength based on original test data 
for the areas identified in Section 2.2 of this document and the use 
of 10% higher steel yield strength for the reactor cavity and 
compartment (RC&C) and containment internal structures (CIS) 
maintain adequate structural capacity. As such, these proposed 
changes do not pose a significant increase in the probability or 
consequences of an accident previously evaluated because the revised 
strength values are determined based on actual original test data 
using a high level of confidence.
    The controlled hydrostatic load is changed from live load to 
dead load for ultimate strength design in the definition. This is 
consistent with ACI-349-97 and therefore does not pose a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This LAR proposes no physical change to any plant system, 
structure, or component (SSC). Similarly, no changes to plant 
operating practices, operating procedures, computer firmware, or 
computer software are proposed. This LAR does not propose changes to 
the design loads used to design Class I structures. Application of 
the new methodology to the design or evaluation of Class I 
structures will continue to ensure that those structures will 
adequately house and protect equipment important to safety.
    Calculations that use the ACI USD method for normal operating/
service load combinations will continue to demonstrate that the 
concrete structures meet required design criteria. Calculations that 
apply the limit design method and use dynamic increase factors (DIF) 
of ACI 349-97, Appendix C will demonstrate that the concrete 
structures meet required design criteria. Use of the actual 
compressive strength of concrete based on 28-day test data (not age 
hardening) is permitted by the ACI 318-63 Code and ensures that the 
concrete structure is capable of performing its design function 
without alteration or compensatory actions of any kind. A 10% higher 
steel yield has minimal reduction on design margin for the RC&C or 
the CIS. The controlled hydrostatic load is changed from live load 
to dead load for ultimate strength design in the definition which is 
consistent with ACI-349-97.
    The use of these alternative methodologies for qualifying Class 
I structures does not have a negative impact on the ability of the 
structure or its components to house and protect equipment important 
to safety and thus, 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 is for the design of new or re-analysis of 
existing Class I structures with the exception of the containment 
structure, the spent fuel pool, and the foundation mats for which no 
change to the current licensing basis (CLB) is proposed.
    Utilization of the ACI 318-63 Code USD method applies only to 
the normal operating/service load cases and is already part of the 
CLB for no loss-of-function load cases. No changes to design basis 
loads are proposed;

[[Page 10682]]

therefore, new designs or re-evaluations of existing Class I 
structures shall still prove capable of coping with design basis 
loads.
    Use of the actual compressive strength of concrete based on 28-
day test data (not age hardening) is justified and further 
constrained by limiting its application to areas where the concrete 
is not exposed to harsh conditions. ACI 349-97, Appendix C is an 
accepted design code used in the nuclear industry. Calculations 
using DIFs per ACI 349-97, Appendix C must demonstrate that the 
Class I structures continue to meet an appropriate design code 
widely used in the nuclear industry. The use of a 10% higher steel 
yield was conservatively derived from original test data and has 
minimal reduction on design margin for the RC&C or the CIS. The 
controlled hydrostatic load is changed from live load to dead load 
for ultimate strength design in the definition which is consistent 
with ACI-349-97.
    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 Branch Chief: Robert J. Pascarelli.

Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant 
(WBN), Unit 2, Rhea County, Tennessee

    Date of amendment request: December 31, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15365A595.
    Description of amendment request: The amendment would revise 
License Condition 2.C(4) to permit the use of the Fuel Rod Performance 
and Design 4 Thermal Conductivity Degradation (PAD4TCD) computer 
program for the second cycle of plant operation.
    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 consequence of an accident previously evaluated?
    Response: No.
    The Emergency Core Cooling System (ECCS) response to a large 
break Loss-of-Coolant Accident (LOCA) as described in the WBN Unit 2 
Final Safety Analysis Report (FSAR) Chapter 15 incorporated an 
explicit evaluation of the effects of Thermal Conductivity 
Degradation (TCD). The FSAR evaluation considered fuel burn-up 
values that represent multi-cycle cores where the effects of TCD 
would be more evident. These analyses showed that the calculated 
peak clad temperature was 1776[emsp14][deg]F [degrees Fahrenheit] 
which provides a large margin to the regulatory limit specified in 
10 CFR 50.46 of 2200[emsp14][deg]F.
    The change to License Condition 2.C(4) does not change the 
safety analysis or any plant feature or design. Thus it is concluded 
that a significant increase in the consequences of an accident 
previously evaluated will not occur as a result of the proposed 
change.
    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 create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed change to [L]icense [C]ondition 2.C(4) does not change 
or modify the plant design, introduce any new modes of plant 
operation, change or modify the design of the ECCS, or change or 
modify the accident analyses presented in the WBN Unit 2 FSAR.
    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 safety analyses for WBN Unit 2 described in the FSAR have 
explicitly accounted for the potential effects of TCD where 
applicable. The results of these analyses have established that WBN 
Unit 2 can operate safely and in the unlikely event that a design 
basis event occurs, there are large margins to the regulatory limits 
explicitly accounting for TCD. This proposed change to License 
Condition 2.C(4) does not change these analyses or conclusions.
    Thus, the proposed change does not result in 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, WT 6A-K, Knoxville, Tennessee 37902.
    NRC Branch Chief: Benjamin G. Beasley.

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

    Date of amendment request: December 10, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15352A108.
    Description of amendment request: The proposed license amendment 
would revise Technical Specification (TS) 3.2.1, ``Heat Flux Hot 
Channel Factor FQ(Z)).'' Specifically, by relocating 
required operating space reductions (Power and Axial Flux Difference) 
to the Core Operating Limits Report, accompanied by verification for 
each reload cycle; and by defining 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 documents NSAL-09-5, Rev. 1 and NSAL-15-
1, Rev. 0 respectively.
    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 for resolution of Westinghouse notification 
documents NSAL-09-05, Rev. 1 and NSAL-15-1, Rev. 0 is intended to 
address deficiencies identified within the existing NAPS Technical 
Specifications and to return them to their as-designed function. 
Operation in accordance with the revised TS ensures that the 
assumptions for initial conditions of key parameter values in the 
safety analyses remain valid and does not result in actions that 
would increase the probability or consequences of any accident 
previously evaluated.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or the 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?
    Response: No.
    Operation in accordance with the revised TS and its limits 
precludes new challenges to [structures, systems and components 
(SSCs)] that might introduce a new type of accident. All design and 
performance criteria will continue to be met and no new single 
failure mechanisms will be created. The proposed change for 
resolution of Westinghouse notification documents NSAL-09-5, Rev. 1 
and NSAL-15-1, Rev. 0 does not involve the alteration of plant 
equipment or

[[Page 10683]]

introduce unique operational modes or accident precursors. It thus 
does not create the potential for a different kind of accident.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Operation in accordance with the revised TS and its limits 
preserves the margins assumed in the initial conditions for key 
parameters assumed in the safety analysis. This ensures that all 
design and performance criteria associated with the safety analysis 
will continue to be met and that the margin of safety is not 
affected.
    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 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: Michael T. Markley.

ZionSolutions, LLC. (ZS), Docket Nos. 50-295 and 50-304, Zion Nuclear 
Power Station (ZNPS), Units 1 and 2, Lake County, Illinois

    Date of amendment request: January 7, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16008B080.
    Description of amendment request: The amendment would approve a 
revision to the ZNPS Defueled Station Emergency Plan (DSEP) to 
implement an Independent Spent Fuel Storage Installation (ISFSI)-Only 
emergency plan. The major proposed changes to the DSEP include the 
removal of non-ISFSI related emergency event types; transfer of 
responsibility for implementing the emergency plan to ISFSI Management, 
and a revised emergency plan organization.
    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.
    ZS has, in effect, an NRC-approved emergency plan. The credible 
accidents involving the ISFSI and [Modular Advanced Generation 
Nuclear All-Purpose Storage System (MAGNASTOR)] system have been 
analyzed and determined that none result in doses to the public 
beyond the owner-controlled boundary (Figure 2-2 of the emergency 
plan) that would exceed the [U.S. Environmental Protection Agency 
Protective Action Guides (EPA PAGs)]. These analyses have not 
changed. With decommissioning completed, the ZNPS site-related 
accidents previously analyzed are no longer credible.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident from 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?
    Response: No.
    ZS has, in effect, an NRC-approved emergency plan. The credible 
accidents involving the ISFSI and MAGNASTOR system have been 
analyzed and determined that none result in doses to the public 
beyond the owner-controlled boundary that would exceed the EPA PAGs. 
With decommissioning substantially completed (Safe Transition to an 
ISFSI only [emergency plan] is contingent on reducing plant side 
curie content to a level where a credible scenario no longer exists 
which could trigger a plant side Emergency Action Level (EAL) 
Threshold Value. Safe Transition will be a bounding number based on 
a calculated value of plant side curie inventory and will occur 
prior to the completion of decommissioning sometime in late 2016 or 
early 2017); the ZNPS site accidents previously analyzed are no 
longer credible. Accidents associated with the ISFSI are addressed 
in the MAGNASTOR [Final Safety Analysis Report (FSAR)].
    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.
    Margin of safety is related to the ability of the fission 
product barriers (fuel cladding, primary containment) to perform 
their design functions during and following postulated accidents. ZS 
has, in effect, an NRC-approved emergency plan. The credible 
accidents involving the ISFSI and MAGNASTOR system have been 
analyzed and determined that none result in doses to the public 
beyond the owner-controlled boundary that would exceed the EPA PAGs. 
With spent fuel located at the ISFSI and decommissioning 
substantially completed, the ZNPS plant-related accidents previously 
analyzed are no longer credible.
    Therefore, the proposed amendment does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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, Deputy General Counsel, 
EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT 
84101.
    NRC Branch Chief: Bruce A. Watson, CHP.

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 Carolinas, LLC, Docket No. 50-369, McGuire Nuclear Station, 
Unit 1, Mecklenburg County, North Carolina

    Date of amendment request: August 28, 2015, as supplemented by 
letter dated November 13, 2015.
    Brief description of amendment: The amendment provides a temporary

[[Page 10684]]

extension to the Completion Time for Technical Specification 3.5.2, 
``ECCS [Emergency Core Cooling Systems]--Operating,'' Condition A. The 
temporary extension will be used to allow the licensee to effect an on-
line repair of the Residual Heat Removal (RHR) pump motor air handling 
unit.
    Date of issuance: February 3, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 281. A publicly-available version is in ADAMS under 
Accession No. ML16004A352; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Operating License No. NPF-9: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 27, 2015 (80 FR 
65810). The supplemental letter dated November 13, 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 amendment is contained 
in a Safety Evaluation dated February 3, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 19, 2015, as supplemented by 
letter dated November 5, 2015.
    Description of amendment request: The amendments revised (1) 
technical specifications (TSs) by replacing AREVA Topical Report ANP-
10298PA, ``ACE/ATRIUM 10XM Critical Power Correlation,'' Revision 0, 
March 2010, with Revision 1, March 2014, of the same topical report; 
and (2) Appendix B, ``Additional Conditions,'' by removing the license 
condition issued by Amendment Nos. 262 and 290 for Units 1 and Unit 2, 
respectively.
    Date of issuance: February 9, 2016.
    Effective date: Once approved, the Unit 1 amendment shall be 
implemented prior to start-up from the 2016 Unit 1 refueling outage, 
and the Unit 2 amendment shall be implemented prior to start-up from 
the 2017 Unit 2 refueling outage.
    Amendment Nos.: 269 and 297. A publicly-available version is in 
ADAMS under Accession No. ML16019A029; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Facility Operating License Nos. DPR-71, and DPR-62: Amendments 
revised the renewed facility operating licenses and TSs.
    Date of initial notice in Federal Register: April 28, 2015 (80 FR 
23603). The supplemental letter dated November 5, 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 amendment is contained 
in a Safety Evaluation dated February 9, 2016.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 12, 2014, as supplemented 
by letters dated September 4, 2014, and April 3 and August 11, 2015.
    Brief description of amendment: The amendment revised the CGS 
Technical Specifications (TSs) to risk-inform requirements regarding 
selected Required Actions end states by incorporating Technical 
Specification Task Force (TSTF) Change Traveler TSTF-423, Revision 1, 
``Technical Specification End States, NEDC-32988-A.'' The Notice of 
Availability for TSTF-423, Revision 1, was published in the Federal 
Register on February 18, 2011 (76 FR 9164).
    Date of issuance: February 3, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 236. A publicly-available version is in ADAMS under 
Accession No. ML15216A266; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-21: The amendment 
revised the Facility Operating License and TSs.
    Date of initial notice in Federal Register: November 12, 2014 (79 
FR 67200). The supplemental letters dated April 3 and August 11, 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 amendment is contained 
in a Safety Evaluation dated February 3, 2016.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket No. 50-277, 
Peach Bottom Atomic Power Station (PBAPS), Unit 2, York and Lancaster 
Counties, Pennsylvania

    Date amendment request: December 5, 2014, as supplemented by letter 
dated April 30, 2015.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) related to the Safety Limit Minimum Critical Power 
Ratios. The changes resulted from a cycle-specific analysis performed 
to support the operation of PBAPS, Unit 2, in the current Cycle 21. The 
re-analysis was performed to accommodate operation in the Maximum 
Extended Load Line Limit Analysis Plus (MELLLA+) operating domain based 
on a separate license amendment request dated September 4, 2014.
    Date of issuance: February 8, 2016.
    Effective date: As of the date of issuance, and shall be 
implemented prior to operation in the MELLLA+ operating domain.
    Amendment No.: 304. A publicly-available version is in ADAMS under 
Accession No. ML15343A165; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-44: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: March 3, 2015 (80 FR 
11495). The supplemental letter dated April 30, 2015, provided 
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 February 8, 2016.
    No significant hazards consideration comments received: No.

[[Page 10685]]

Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket No. 50-278, 
Peach Bottom Atomic Power Station (PBAPS), Unit 3, York and Lancaster 
Counties, Pennsylvania

    Date amendment request: April 30, 2015, as supplemented by letter 
dated August 6, 2015.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) related to the Safety Limit Minimum Critical Power 
Ratios. The changes resulted from a cycle-specific analysis performed 
to support the operation of PBAPS, Unit 3, in the current Cycle 21. The 
re-analysis was performed to accommodate operation in the Maximum 
Extended Load Line Limit Analysis Plus (MELLLA+) operating domain based 
on a separate license amendment request dated September 4, 2014.
    Date of issuance: February 8, 2016.
    Effective date: As of the date of issuance, and shall be 
implemented prior to operation in the MELLLA+ operating domain.
    Amendment No.: 308. A publicly-available version is in ADAMS under 
Accession No. ML15343A177; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-56: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: July 7, 2015 (80 FR 
38773). The supplemental letter dated August 6, 2015, provided 
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 February 8, 2016.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 30, 2014, and supplemented by 
letters dated December 12, 2014, and July 20, 2015.
    Description of amendment: The amendment authorizes changes to the 
VEGP Units 3 and 4 Updated Final Safety Analysis Report (USFAR) in the 
form of departures from the incorporated plant-specific Design Control 
Document Tier 2* information. The proposed amendment would allow 
changes to correct editorial errors and promote consistency with the 
UFSAR Tier 1 and 2 information.
    Date of issuance: February 1, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 45. A publicly-available version is in ADAMS under 
Accession No. ML15335A060; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: September 30, 2014 (79 
FR 58812). The supplemental letters dated December 12, 2014, and July 
20, 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 amendment is contained 
in the Safety Evaluation dated February 1, 2016.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear 
Plant, Unit 3, Limestone County, Alabama

    Date of amendment request: March 6, 2015, as supplemented by letter 
dated July 7, 2015.
    Brief description of amendment: The amendment revised the Technical 
Specification (TS) Safety Limit Minimum Critical Power Ratio (SLMCPR) 
numeric values. The change decreased the numeric values of SLMCPR in TS 
Section 2.1.1.2 for single and two reactor recirculation loop operation 
based on the Cycle 18 SLMCPR evaluation.
    Date of issuance: February 9, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 279. A publicly-available version is in ADAMS under 
Accession No. ML15317A478; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-68: Amendment revised 
the Facility Operating License and TS.
    Date of initial notice in Federal Register: July 7, 2015 (80 FR 
38777). The supplemental letter dated July 7, 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 February 9, 2016.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 9, 2015, as supplemented 
by letters dated April 8, August 12, and December 10, 2015.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) requirements regarding steam generator tube 
inspections and reporting as described in TS Task Force (TSTF) traveler 
TSTF-510, Revision 2, ``Revision to Steam Generator Program Inspection 
Frequencies and Tube Sample Selection,'' with some minor administrative 
differences.
    Date of issuance: February 2, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 215. A publicly-available version is in ADAMS under 
Accession No. ML15324A114; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-30: The amendment 
revised the Operating License and TSs.
    Date of initial notice in Federal Register: June 9, 2015 (80 FR 
32630). The supplemental letters dated August 12 and December 10, 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 amendment is contained 
in a Safety Evaluation dated February 2, 2016.
    No significant hazards consideration comments received: No.

[[Page 10686]]

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

    Date of amendment request: May 4, 2015, as supplemented by letter 
dated August 5, 2015.
    Description of amendment request: The proposed amendments authorize 
modification of the Emergency Action Level (EAL) Technical Basis 
Document, EAL RA2.1, to revise the instrumentation used to classify an 
event under this EAL. Specifically, this would correct the equipment 
identification number from the ``GW-RI-178-1 Process Vent Normal 
Range'' monitor to the ``VG-RI-180-1 Vent Stack `B' Normal Range'' 
monitor for Initiating Condition RA2, EAL RA2.1.
    Date of issuance: January 21, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 277 and 259. A publicly-available version is in 
ADAMS under Accession No. ML15307A300; documents related to these 
amendments are listed in the Safety Evaluation enclosed with these 
amendments.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
changed the licenses.
    Date of initial notice in Federal Register: July 7, 2015 (80 FR 
38764). The supplemental letter dated August 5, 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.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 21, 2016.
    No significant hazards consideration comments received: Yes.

    Dated at Rockville, Maryland, this 22nd day of February 2016.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2016-04346 Filed 2-29-16; 8:45 am]
 BILLING CODE 7590-01-P