[Federal Register Volume 79, Number 179 (Tuesday, September 16, 2014)]
[Notices]
[Pages 55507-55516]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-21833]
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NUCLEAR REGULATORY COMMISSION
[NRC-2014-0917]
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 21, 2014 to September 3, 2014. The
last
[[Page 55508]]
biweekly notice was published on September 2, 2014.
DATES: Comments must be filed by October 16, 2014. A request for a
hearing must be filed by November 17, 2014.
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-2014-0917. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected].
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: 3WFN-06-A44M, 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: Beverly A. Clayton, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-3475, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2014-0917 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-2014-0917.
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-2014-0917 in the subject line of your
comment submission, in order to ensure that the NRC is able to make
your comment submission available to the public in this docket.
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 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 by the above date, 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.
[[Page 55509]]
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 identify 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.
If a hearing is requested, 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.
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
ten 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-
[[Page 55510]]
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 security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, 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 Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of amendment request: June 30, 2014. A publicly-available
version is in ADAMS under Accession No. ML14184B384.
Description of amendment request: The amendment would revise the
Technical Specifications (TS) by reducing the allowed maximum rated
thermal power (RTP) at which the unit can operate when select High
Pressure Injection (HPI) System equipment is inoperable.
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 TS changes do not modify the reactor coolant system
pressure boundary, nor make any physical changes to the facility
design, material, or construction standards. The probability of any
design basis accident (DBA) is not affected by this change, nor are
the consequences of any DBA affected by this change. The new small
break loss-of-coolant accident (SBLOCA) partial-power analysis
demonstrates that all 10 CFR 50.46 acceptance criteria are
satisfied. Radiological consequences for loss-of-coolant accident
(LOCA) events are evaluated in ONS Updated Final Safety Analysis
Report Section 15.15 for the Maximum Hypothetical Accident. The
proposed changes will not impact assumptions and conditions
previously used in the radiological consequence evaluations for the
Maximum Hypothetical Accident. The proposed changes do not involve
changes to any structures, systems, or components (SSCs) that can
alter the probability for initiating a LOCA event.
Therefore, the proposed TS changes do not significantly increase
the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed TS changes reduce the allowed power level that the
unit may be operated at with select HPI equipment out-of-service.
The changes do not alter the plant configuration (no new or
different type of equipment will be installed) or make changes in
methods governing normal plant operation. No new failure modes are
identified, nor are any SSCs required to be operated outside the
design bases.
Therefore, the possibility of a new or different kind of
accident from any kind of accident previously evaluated is not
created.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed TS changes are supported by SBLOCA analyses which
demonstrate that the acceptance criteria of 10 CFR 50.46 are
satisfied. These analyses were performed in accordance with the
Evaluation Model described in AREVA Topical Report BAW-10192P-A. The
new SBLOCA analysis assumes a lower initial core power level (50% of
rated thermal power (RTP)) than what was previously analyzed in
support of TS 3.5.2 (i.e., 75% of RTP). The resulting peak cladding
temperature results for the new SBLOCA analysis are lower than the
existing analysis. In addition, a supplemental evaluation
demonstrated that failure to perform a desired operator action of
maintaining secondary-side pressure at 300 psig by throttling the
atmospheric dump valve during a SBLOCA did not result in adverse
affects to the new SBLOCA analysis results. Therefore, it is
concluded that the proposed amendment request will not result in a
significant decrease 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, 526 South Church Street--EC07H, Charlotte, NC
28202-1802.
NRC Branch Chief: Robert J. Pascarelli.
Duke Energy Progress Inc., Docket No. 50-261, H.B. Robinson Steam
Electric Plant, Unit 2 (HBRSEP2), Darlington County, South Carolina
Date of amendment request: June 20, 2014. A publicly-available
version is in ADAMS under Accession No. ML14188B015.
Description of amendment request: The amendment would revise
Technical
[[Page 55511]]
Specification (TS) 5.5.9.b.2 for the Steam Generator (SG) Program
accident-induced leakage performance criterion to correct an editorial
error in the accident-induced leakage rate value for any design-basis
accident other than a SG tube rupture.
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 is a correction to an editorial error in the
specified accident induced leakage performance criterion of TS
5.5.9.b.2. The error in TS 5.5.9.b.2 being addressed by this
proposed change was introduced at the time of the HBRSEP2 submittal
of the NRC-approved Technical Specification Task Force (TSTF)
traveler 449, Rev. 4, Steam Generator Tube Integrity. The accident-
induced leakage performance criterion will continue to be within the
limit assumed in the accident analysis. As a result, neither the
probability nor the consequences of any accident previously
evaluated will be affected.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new or different accidents result from the proposed changes.
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, it only corrects an editorial error in the
accident-induced leakage performance criterion specified in the SG
Program. The proposed changes are consistent with the safety
analysis assumptions and current plant operating practice.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change will have no effect on the margin of safety. This
proposed change corrects an editorial error in the accident-induced
leakage performance criterion specified in the SG Program.
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.
Acting NRC Branch Chief: Lisa M. Regner.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: November 14, 2013. A publicly-available
version is in ADAMS under Accession No. ML13323A516.
Description of amendment request: The proposed amendment would
eliminate operability requirements for secondary containment when
handling sufficiently decayed irradiated fuel or a fuel cask following
a minimum of 13 days after the permanent cessation of reactor
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:
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 modify the design or operation of
equipment used to move spent fuel or to perform core alterations.
The proposed changes cannot increase the probability of any
previously analyzed accident because they are based on changes in
Source Term, atmospheric dispersion and dose consequence analysis
methodology, not in procedures or equipment used for fuel handling.
The conservative re-analysis of the FHA [fuel-handling accident]
concludes that the radiological consequences are within the
regulatory limits established 10 CFR 50.67. This conclusion is based
on the Alternate Source Term and guidance provided in Appendix B of
Regulatory Guide 1.183 and analyses of fission product release and
transport path that does not take credit for dose mitigation
provided by engineered safeguards including secondary containment
and the SGT system. The results of the core alteration events, other
than the FHA, remain unchanged from the original design-basis that
showed these events do not result in fuel cladding damage or
radioactive release.
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 changes do not introduce any new modes of plant
operation and do not involve physical modifications to the plant.
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.
Regulation in 10 CFR 50.67 permits licensees to voluntarily
revise the accident source term used in design-basis radiological
consequence analyses. This license amendment application evaluates
the consequences of a design-basis fuel handling accident in
accordance with this regulation and Regulatory Guide 1.183. The
revised analysis concludes that the radiological consequences of the
fuel handling accident are less than the regulatory allowable
limits. Safety margins and analytical conservatisms are retained to
ensure the analysis adequately bounds all postulated event
scenarios. The selected assumptions and release models provide an
appropriate and prudent safety margin against unpredicted events in
the course of an accident and compensates for large uncertainties in
facility parameters, accident progression, radioactive material
transport and atmospheric dispersion. The proposed TS applicability
statements continue to ensure that the total effective dose
equivalent (TEDE) at the boundaries of the control room, the
exclusion area, and low population zone boundaries are below the
corresponding regulatory allowable limits in 10 CFR 50.67(b)(2).
Therefore, 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: Ms. Jeanne Cho, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company, LLC, Docket Nos. STN 50-456, STN 50-457 and
72-73, Braidwood Station, Units 1 and 2, Will County, Illinois
Exelon Generation Company, LLC, Docket Nos. STN 50-454, STN 50-455 and
72-68, Byron Station, Units 1 and 2, Ogle County, Illinois
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit 1, DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-10, 50-237, 50-249 and
72-37, Dresden Nuclear Power Station, Units 1, 2 and 3, Grundy County,
Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373, 50-374 and 72-70,
LaSalle County Station, Units 1 and 2, LaSalle County, Illinois
[[Page 55512]]
Exelon Generation Company, LLC, Docket Nos. 50-352, 50-353 and 72-65,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Exelon Generation Company, LLC, et al., Docket No. 50-219 and 72-15,
Oyster Creek Nuclear Generating Station, Ocean County, New Jersey
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
171, 50-277, 50-278 and 72-29, Peach Bottom Atomic Power Station, Units
1, 2 and 3, York and Lancaster Counties, Pennsylvania
Exelon Generation Company, LLC, Docket Nos. 50-254, 50-265 and 70-53,
Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Exelon Generation Company, LLC, Docket No. 50-320, Three Mile Island
Nuclear Station, Unit 2, Dauphin County, Pennsylvania
Date of amendment request: May 30, 2014. A publicly-available
version is in ADAMS under Accession No. ML14164A054.
Description of amendment request: The proposed changes revise the
Emergency Plans for the affected facilities to adopt the Nuclear Energy
Institute's (NEl's) revised Emergency Action Level (EAL) schemes
described in NEI 99-01, Revision 6, ``Development of Emergency Action
Levels for Non-Passive Reactors,'' which has been endorsed by the NRC
in a letter dated March 28, 2013. A publicly-available version can be
found in ADAMS under Accession No. ML12346A463.
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:
The proposed changes have been reviewed considering the
applicable requirements of 10 CFR 50.47, 10 CFR 50, Appendix E, and
other applicable NRC documents. Exelon has evaluated the proposed
changes to the affected sites' Emergency Plans and determined that
the changes do not involve a Significant Hazards Consideration. In
support of this determination, an evaluation of each of the three
(3) standards, set forth in 10 CFR 50.92, ``Issuance of amendment,''
is provided below.
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to Exelon's EAL schemes to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, ``Development of
Emergency Action Levels for Non-Passive Reactors,'' do not reduce
the capability to meet the emergency planning requirements
established in 10 CFR 50.47 and 10 CFR Part 50, Appendix E. The
proposed changes do not reduce the functionality, performance, or
capability of Exelon's ERO [Emergency Response Organization] to
respond in mitigating the consequences of any design basis accident.
The probability of a reactor accident requiring implementation
of Emergency Plan EALs has no relevance in determining whether the
proposed changes to the EALs reduce the effectiveness of the
Emergency Plans. As discussed in Section D, ``Planning Basis,'' of
NUREG-0654, Revision 1, ``Criteria for Preparation and Evaluation of
Radiological Emergency Response Plans and Preparedness in Support of
Nuclear Power Plants'';
``. . . The overall objective of emergency response plans is to
provide dose savings (and in some cases immediate life saving) for a
spectrum of accidents that could produce offsite doses in excess of
Protective Action Guides (PAGs). No single specific accident
sequence should be isolated as the one for which to plan because
each accident could have different consequences, both in nature and
degree. Further, the range of possible selection for a planning
basis is very large, starting with a zero point of requiring no
planning at all because significant offsite radiological accident
consequences are unlikely to occur, to planning for the worst
possible accident, regardless of its extremely low likelihood . . .
.''
Therefore, Exelon did not consider the risk insights regarding
any specific accident initiation or progression in evaluating the
proposed changes.
The proposed changes do not involve any physical changes to
plant equipment or systems, nor do they alter the assumptions of any
accident analyses. The proposed changes do not adversely affect
accident initiators or precursors nor do they alter the design
assumptions, conditions, and configuration or the manner in which
the plants are operated and maintained. The proposed changes do not
adversely affect the ability of Structures, Systems, or Components
(SSCs) to perform their intended safety functions in mitigating the
consequences of an initiating event within the assumed acceptance
limits.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to Exelon's EAL schemes to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not involve any
physical changes to plant systems or equipment. The proposed changes
do not involve the addition of any new plant equipment. The proposed
changes will not alter the design configuration, or method of
operation of plant equipment beyond its normal functional
capabilities. All Exelon ERO functions will continue to be performed
as required. The proposed changes do not create any new credible
failure mechanisms, malfunctions, or accident initiators.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from those that have been
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes to Exelon's EAL schemes to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not alter or exceed a
design basis or safety limit. There is no change being made to
safety analysis assumptions, safety limits, or limiting safety
system settings that would adversely affect plant safety as a result
of the proposed changes. There are no changes to setpoints or
environmental conditions of any SSC or the manner in which any SSC
is operated. Margins of safety are unaffected by the proposed
changes to adopt the NEI 99-01, Revision 6 EAL scheme guidance. The
applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E
will continue to be met.
Therefore, the proposed changes do not involve any reduction in
a margin of safety.
In conclusion, and based on the considerations discussed above:
(1) There is reasonable assurance that the health and safety of
the public will not be endangered by the proposed changes to adopt
the EAL schemes established in NEI 99-01, Revision 6, as endorsed by
the U.S. Nuclear Regulatory Commission (NRC); (2) the changes will
be in compliance with the NRC's regulations; and (3) the issuance of
the amendments will not be inimical to the common defense and
security or to the health and safety of the public.
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: Bradley Fewell, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: Travis L. Tate.
FirstEnergy Nuclear Operating Company (FENOC), Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Perry, OH
Date of amendment request: March 25, 2014. A publicly-available
version is in ADAMS under Accession No. ML14084A165.
Description of amendment request: The proposed changes are
consistent with the NRC-approved Industry/Technical Specifications Task
Force (TSTF) Traveler, TSTF-425, Revision 3, ``Relocate Surveillance
Frequencies to
[[Page 55513]]
Licensee Control--RITSTF Initiative 5b.'' The proposed change relocates
surveillance frequencies to a licensee controlled program, the
Surveillance Frequency Control Program. This change is applicable to
licensees using probabilistic risk guidelines contained in NRC-approved
Nuclear Energy Institute (NEI) 04-10, ``Risk-Informed Technical
Specifications Initiative 5b, Risk-Informed Method for Control of
Surveillance Frequencies.''
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 any accident previously
evaluated?
Response: No.
The proposed change relocates the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program. Surveillance frequencies are
not an initiator to any accident previously evaluated. As a result,
the probability of any accident previously evaluated is not
significantly increased. The systems and components required by the
technical specifications for which the surveillance frequencies are
relocated are still required to be operable, meet the acceptance
criteria for the surveillance requirements, and be capable of
performing any mitigation function assumed in the accident analysis.
As a result, the consequences of any accident previously evaluated
are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical alteration of the
plant (that is, 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. 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.
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
the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the final safety analysis report and bases to the
TS [technical specification]), since these are not affected by
changes to the surveillance frequencies. Similarly, there is no
impact to safety analysis acceptance criteria as described in the
plant licensing basis. To evaluate a change in the relocated
surveillance frequency, FENOC will perform a probabilistic risk
evaluation using the guidance contained in NRC approved Nuclear
Energy Institute (NEI) 04-10, Revision 1, in accordance with the TS
Surveillance Frequency Control Program. NEI 04-10, Revision 1,
methodology provides reasonable acceptance guidelines and methods
for evaluating the risk increase of proposed changes to surveillance
frequencies consistent with Regulatory Guide 1.177, ``An Approach
for Plant-Specific, Risk-Informed Decision-making: Technical
Specifications.''
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based upon the reasoning presented above, FENOC concludes that
the requested change does not involve a significant hazards
consideration as set forth in 10 CFR 50.92(c), Issuance of
Amendment.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Travis L. Tate.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, San
Diego County, California
Date of amendment request: March 21, 2014. A publicly-available
version is in ADAMS under Accession No. ML14085A141.
Description of amendment request: The proposed amendment would
revise the Operating License and associated Technical Specifications
(TS) to reflect the permanent cessation of power operation. Because the
licenses for SONGS, Units 2 and 3 no longer authorize emplacement or
retention of fuel in the reactor vessel, the limiting conditions for
operation and associated surveillance requirements that do not apply in
the defueled condition are being proposed for deletion. The remaining
portions of the TS are being proposed for revision and incorporation as
the permanently defueled TS to provide a continuing acceptable level of
safety, which addresses the reduced scope of postulated design basis
accidents associated with a defueled plant, as described in the SONGS,
Units 2 and 3 safety analyses.
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.
SONGS Units 2 and 3 have permanently ceased operation. The
proposed amendment would modify the SONGS Units 2 and 3 facility
operating licenses and TS by deleting the portions of the licenses
and TS that are no longer applicable to a permanently defueled
facility, while modifying the remaining portions to correspond to
the permanently shutdown condition. This change is consistent with
the criteria set forth in 10 CFR 50.36 for the contents of TS.
Section 15 of the SONGS Updated Final Safety Analysis Report
(UFSAR) described the design basis accident (DBA) and transient
scenarios applicable to SONGS Units 2 and 3 during power operations.
With the reactors in a permanently defueled condition, the fuel
storage pools and their systems have been isolated and are dedicated
only to spent fuel storage. In this condition, the spectrum of
credible accidents is much smaller than for an operational plant. As
a result of the certifications submitted by SCE [Southern California
Edison] in accordance with 10 CFR 50.82(a)(1), and the consequent
removal of authorization to operate the reactors or to place or
retain fuel in the reactors in accordance with 10 CFR 50.82(a)(2),
most of the accident scenarios postulated in the UFSAR are no longer
possible.
The definition of safety-related structures, systems, and
components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are
those relied on to remain functional during and following design
basis events to assure:
1. The integrity of the reactor coolant boundary;
2. The capability to shut down the reactor and maintain it in a
safe shutdown condition; or
3. The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures
comparable to the applicable guideline exposures set forth in 10 CFR
50.43(a)(1) or 100.11.
The first two criteria (integrity of the reactor coolant
pressure boundary and safe shut down of the reactor) are not
applicable to a plant in a permanently defueled condition. The third
criterion is related to preventing or mitigating the consequences of
accidents that could result in potential offsite
[[Page 55514]]
exposures exceeding limits. However, after the termination of
reactor operations at SONGS Units 2 and 3 and the permanent removal
of the fuel from the reactor vessels (following 17 months of decay
time after shut down) and purging of the contents of the waste gas
decay tanks, none of the SSCs at SONGS Units 2 and 3 are required to
be relied on for accident mitigation. Therefore, none of the SSCs at
SONGS Units 2 and 3 meet the definition of a safety-related SSC
stated in 10 CFR 50.2 (with the exception of the passive fuel
storage pool structure).
The deletion of TS definitions and rules of usage and
application, that are currently not applicable in a defueled
condition, has no impact on facility SSCs or the methods of
operation of such SSCs. The deletion of design features and safety
limits not applicable to the permanently shut down and defueled
status of SONGS Units 2 and 3 has no impact on the remaining DBA.
The removal of limiting conditions for operation (LCOs) or
surveillance requirements (SRs) that are related only to the
operation of the nuclear reactors or only to the prevention,
diagnosis, or mitigation of reactor-related transients or accidents
do not affect the applicable DBAs previously evaluated since these
DBAs are no longer applicable in the defueled mode. The safety
functions involving core reactivity control, reactor heat removal,
reactor coolant system inventory control, and containment integrity
are no longer applicable at SONGS Units 2 and 3 as a permanently
defueled plant. The analyzed accidents involving damage to the
reactor coolant system, main steam lines, reactor core, and the
subsequent release of radioactive material are no longer possible at
SONGS Units 2 and 3.
Since SONGS Units 2 and 3 has permanently ceased operation, the
future generation of fission products has ceased and the remaining
source term will decay. The radioactive decay of the irradiated fuel
since shut down of the reactor will have reduced the consequences of
the FHA [fuel handling accident] to levels well below those
previously analyzed. The relevant parameter (water level) associated
with the fuel pool provides an initial condition for the FHA
analysis and is included in the permanently defueled TS.
The fuel storage pool water level, fuel storage pool boron
concentration, and spent fuel assembly storage TS are retained to
preserve the current requirements for safe storage of irradiated
fuel.
Fuel pool cooling and makeup related equipment and support
equipment (e.g., electrical power systems) are not required to be
continuously available since there is sufficient time to effect
repairs, establish alternate sources of makeup flow, or establish
alternate sources of cooling in the event of a loss of cooling and
makeup flow to the fuel storage pool.
The deletion and modification of provisions of the
administrative controls does not directly affect the design of 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 shut down and defueled condition
of the reactors.
The probability of occurrence of previously evaluated accidents
is not increased, since extended operation in a defueled condition
is the only operation currently allowed, and therefore bounded by
the existing analyses. Additionally, the occurrence of postulated
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 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 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 removal of TS that are related only to the operation of
the nuclear reactor or only to the prevention, diagnosis, or
mitigation of reactor-related transients or accidents cannot result
in different or more adverse failure MODES or accidents than
previously evaluated because the reactor is permanently shut down
and defueled and SCE is no longer authorized to operate the
reactors.
The proposed deletion of requirements of the SONGS Unit 2 and
Unit 3 TS do not affect systems credited in the accident analysis.
The proposed permanently defueled TS (PDTS) continue to require
proper control and monitoring of safety significant parameters and
activities.
The proposed restriction on the fuel pool level is fulfilled by
normal operating conditions and preserves initial conditions assumed
in the analyses of the postulated DBA. The fuel storage pool water
level, fuel storage pool boron concentration, and spent fuel
assembly storage TS are retained to preserve the current
requirements for safe storage of irradiated fuel.
The proposed amendment does not result in any new mechanisms
that could initiate damage to the remaining relevant safety barriers
for defueled plants (i.e., fuel cladding and spent fuel cooling).
Since extended operation in a defueled condition is the only
operation currently 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 amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Because the 10 CFR Part 50 licenses for SONGS Units 2 and 3 no
longer authorize operation of the reactors or emplacement or
retention of fuel into the reactor vessels, as specified in 10 CFR
50.82(a)(2), the occurrence of postulated accidents associated with
reactor operation is no longer credible. The remaining credible
accidents do not credit SSCs for mitigation. The proposed amendment
does not adversely affect the inputs or assumptions of any of the
design basis analyses that impact an accident.
The proposed changes are limited to those portions of TS and
license that are not related to the safe storage of irradiated fuel.
The requirements for SSCs that have been deleted from the SONGS TS
Units 2 and 3 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 involving the reactors are no longer
possible because the reactors are permanently shut down and defueled
and SCE is no longer authorized to operate the reactors.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety because the current design limits
continue to be met for the accidents of concern.
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 requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: Douglas A. Broaddus.
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: June 4, 2014. A publicly-available
version is in ADAMS under Accession No. ML14156A477.
Description of amendment request: The purpose of the proposed
license amendment request is to address proposed changes related to
departure from the plant-specific Design Control Document (DCD) Tier 1
(and corresponding Combined License Appendix C information) and Tier 2
material to reconcile differences in the various valve table entries.
Because this proposed change requires a departure from Tier 1
information in the Westinghouse Advanced Passive 1000 DCD, the licensee
also requested an exemption from the requirements of the Generic DCD
Tier 1 in accordance with 10 CFR 52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 55515]]
consideration, which is presented below:
1. Does the requested amendment involve a significant increase
in the probability or consequences of an accident previously
evaluated?
Response: No.
The proposed changes do not result in any physical changes to
the plant, and therefore do not change any safety-related design
requirement, qualification requirement or function. The proposed
changes do not involve any accident initiating event or component
failure, thus, the probabilities of the accidents previously
evaluated are not affected. The proposed changes do not affect the
radioactive material releases used in the accident analyses, thus,
the radiological releases in the accident analyses are not affected.
The proposed changes do not affect any postulated non-radioactive
accident scenario as evaluated in UFSAR [Updated Final Safety
Analysis Report] Chapter 15.
Therefore, the requested amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the requested amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not result in any physical changes to
the plant, and therefore do not adversely affect any structure,
system or component. No safety-related equipment qualification or
design function is affected. The proposed changes do not introduce a
new failure mode or create a new fault or sequence of events that
could result in a radioactive material release.
Therefore, the requested amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not result in any physical changes to
the plant, and therefore do not change valve performance, including
containment isolation. No safety acceptance criterion would be
exceeded or challenged. No safety related function would be
affected. Valve qualification would not be affected.
The proposed changes do not affect compliance with existing
design codes and regulatory criteria and do not affect any safety
analysis.
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: Mr. M. Stanford Blanton, Blach & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence Burkhart.
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 Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: September 12, 2013, as
supplemented by letters dated May 20 and July 22, 2014.
Brief description of amendments: The amendments modify Technical
Specification (TS) 3.3.2. Specifically, the change modifies setpoints
associated with the auxiliary feedwater pump suction transfer on low
suction pressure.
Date of issuance: August 27, 2014.
Effective date: This license amendment is effective as of its date
of issuance and shall be implemented within 60 days of issuance.
Amendment Nos.: 273 and 253. A publicly-available version is in
ADAMS under Accession No. ML14211A403; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and technical specifications.
Date of initial notice in Federal Register: December 10, 2013 (78
FR 74179). The supplemental letters dated May 20 and July 22, 2014,
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 27, 2014.
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 (PBAPS), Units 2 and
3, York and Lancaster Counties, Pennsylvania
Date of application for amendments: September 28, 2012, as
supplemented by letters dated February 15, 2013, May 7, 2013, May 24,
2013, June 4, 2013, June 27, 2013, July 30, 2013, July 31, 2013, August
5, 2013, August 22, 2013, August 29, 2013, September 13, 2013, October
11, 2013, October 15, 2013, October 31, 2013, December 6, 2013,
December 20, 2013, January 17, 2014, January 31, 2014 (2 letters),
February 20, 2014, February 28, 2014, March 10, 2014, March 17, 2014,
April 11, 2014, April 18, 2014, May 6, 2014, June 5, 2014, and June 20,
2014.
Brief description of amendments: The amendments authorize an
increase in the maximum licensed thermal power level for PBAPS, Units 2
and 3, from 3514 megawatts thermal (MWt) to 3951 MWt, which is an
increase of approximately 12.4 percent.
Date of issuance: August 25, 2014.
Effective date: For PBAPS, Unit 2, the amendment is effective as of
its date of issuance and shall be implemented prior to startup from
refueling outage P2R20. For PBAPS, Unit 3, the
[[Page 55516]]
amendment is effective as of its date of issuance and shall be
implemented prior to startup from refueling outage P3R20.
Amendments Nos.: 293 and 296. A publicly-available version is in
ADAMS under Accession No. ML14133A046; 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 Facility Operating Licenses and the Technical
Specifications.
Date of initial notice in Federal Register: April 9, 2013 (78 FR
21168). The letters dated February 15, 2013, May 7, 2013, May 24, 2013,
June 4, 2013, June 27, 2013, July 30, 2013, July 31, 2013, August 5,
2013, August 22, 2013, August 29, 2013, September 13, 2013, October 11,
2013, October 15, 2013, October 31, 2013, December 6, 2013, December
20, 2013, January 17, 2014, January 31, 2014 (2 letters), February 20,
2014, February 28, 2014, March 10, 2014, March 17, 2014, April 11,
2014, April 18, 2014, May 6, 2014, June 5, 2014, and June 20, 2014,
provided clarifying information that did not change the initial
proposed no significant hazards consideration determination or expand
the application beyond the scope of the original Federal Register
notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 25, 2014.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear
Power Plant, Wayne County, New York
Date of amendment request: February 28, 2013, as supplemented by
letters dated June 19, and November 11, 2013 and January 22, March 14,
March 26, and June 6, 2014.
Brief description of amendment: The amendment revised the Ginna
Nuclear Power Plant Technical Specifications (TSs) to revise the
allowable containment average air temperature from ``<=
120[emsp14][deg]F'' to ``<= 125[emsp14][deg]F'' for TS 3.6.5
``Containment Air Temperature.''
Date of issuance: August 12, 2014.
Effective date: As of the date of issuance to be implemented within
45 days.
Amendment No.: 116. A publicly-available version is in ADAMS under
Accession No. ML14232A125; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-18: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: November 26, 2013 (78
FR 70594). The supplemental letters dated June 19, and November 11,
2013, and January 22, March 14, March 26, and June 6, 2014, 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 12, 2014.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: July 23, 2013, as supplemented
August 5, 2014.
Brief description of amendments: The amendments revise the
Technical Specification (TS) requirements and add license conditions
related to control room envelope habitability in accordance with the
Nuclear Regulatory Commission approved Revision 3 of Technical
Specification Task Force (TSTF) Standard Technical Specifications
Change Traveler TSTF-448, ``Control Room Habitability.''
Date of issuance: August 29, 2014.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: Unit 1-268 and Unit 2-212. A publicly-available
version is in ADAMS under Accession No. ML14147A410; documents related
to this amendment are listed in the Safety Evaluation enclosed with the
amendment.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Renewed Facility Operating licenses and the
Technical Specifications.
Date of initial notice in Federal Register: September 3, 2013 (78
FR 54290). The supplement dated August 5, 2014, 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 August 29, 2014.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County
Date of application for amendment: February 22, 2013.
Brief description of amendment: The amendment revised Technical
Specification 3.1.6, ``Control Bank Insertion Limits,'' to include
text, in Condition A, stating, ``for reasons other than Condition C.''
This text addition modifies Condition A, for control bank sequence or
overlap limits, to include language currently in Condition B, for
control bank insertion limits, this change would point to Condition C,
which, if applicable, would allow the specified completion time to
restore the control bank to within the insertion limit to be increased
from 2 hours to 72 hours. This would align the description of the
sequence and overlap limit of Condition A with the description of
control bank insertion limit Condition B.
Date of issuance: August 27, 2014.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 272 and 254. A publicly-available version is in
ADAMS under Accession No. ML14188C453; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
revised the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 30, 2013 (78 FR
25317).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 27, 2014.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 5th day of September 2014.
For the Nuclear Regulatory Commission.
A. Louise Lund,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2014-21833 Filed 9-15-14; 8:45 am]
BILLING CODE 7590-01-P