[Federal Register Volume 78, Number 83 (Tuesday, April 30, 2013)]
[Notices]
[Pages 25310-25320]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-10020]
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NUCLEAR REGULATORY COMMISSION
[NRC-2013-0074]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
Background
Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC) is publishing this regular biweekly notice. The Act
requires the Commission 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 April 4, 2013 to April 17, 2013. The last
biweekly notice was published on April 16, 2013 (78 FR 22563).
ADDRESSES: Please refer to Docket ID NRC-2013-0074 when contacting the
NRC about the availability of information regarding this document. You
may access information related to this document, which the NRC
possesses and is publicly available, using any of the following
methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0074. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected].
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
Fax comments to: RADB at 301-492-3446.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
SUPPLEMENTARY INFORMATION:
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2013-0074 when contacting the NRC
about the availability of information regarding this document. You may
access information related to this document, which the NRC possesses
and is publicly-available, by the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0074.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly-available documents online in the NRC
Library 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]. Documents may be viewed in
ADAMS by performing a search on the document date and docket number.
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-2013-0074 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.
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses and Combined Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Section 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
[[Page 25311]]
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.
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 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.
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.
All documents filed in the 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/apply-certificates.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
[[Page 25312]]
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 agency'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 Web site
at http://www.nrc.gov/site-help/e-submittals.html, by email at
[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 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 following three
factors in 10 CFR 2.309(c)(1): (i) The information upon which the
filing is based was not previously available; (ii) the information upon
which the filing is based is materially different from information
previously available; and (iii) the filing has been submitted in a
timely fashion based on the availability of the subsequent information.
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the NRC's PDR, located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to
ADAMS or who encounter problems in accessing the documents located in
ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209,
301-415-4737, or by email to [email protected].
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1 (RBS), West Feliciana
Parish, Louisiana
Date of amendment request: February 7, 2013.
Description of amendment request: Entergy Operations, Inc. (the
licensee), proposes to revise RBS Technical Specification (TS) 3.8.4,
``DC [Direct Current] Sources--Operating,'' Surveillance Requirements
(SRs) 3.8.4.2 and 3.8.4.5. The changes to the SRs will add new
acceptance criteria to address possible non-conservative conditions
when the battery connection resistances are at maximum values.
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 changes are to the surveillance requirements only.
The ability of the TS surveillance to ensure that the batteries have
the capacity to perform their specified safety functions with regard
to accident mitigation or meeting their licensing design basis
requirements is not reduced/diminished.
There are no design changes associated with this TS amendment.
The DC power system/batteries will retain adequate independency,
redundancy, capacity and testability to permit the functioning
required of the engineered safety features. The batteries will each
continue to independently
[[Page 25313]]
provide this capacity assuming a failure of a single active
component. The proposed change will not affect accident initiators
or precursors, or adversely alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated. The proposed change will not alter the
ability of structures, systems and components to perform their
intended functions to mitigate the consequences of an initiating
event. The proposed change does not physically alter safety related
systems nor affect the way in which safety related systems perform
their function.
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 involves only surveillance test acceptance
criteria. The ability of the TS surveillance to ensure that the
batteries have the capacity to perform their specified safety
functions with regard to accident mitigation or meeting their
licensing design basis requirements is not reduced/diminished.
There are no proposed design changes, nor are there any changes
in the method by which any safety related plant structure, system,
or component (SSC) performs its specified safety function. The
proposed change will not affect the normal method of plant operation
or change any operating parameters. Equipment performance necessary
to fulfill safety analysis missions will be unaffected. The proposed
change will not alter any assumptions required to meet the safety
analysis acceptance criteria. No new accident scenarios, transient
precursors, failure mechanisms, or limiting single failures will be
introduced because of this amendment. There will be no adverse
effect or challenges imposed on any safety related system because of
this amendment.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not reduce the ability of the TS
surveillance requirements to ensure that the station batteries have
adequate capacity to perform their engineered safety features
functions with regard to accident mitigation and meeting their
licensing design basis requirements. The lower battery inter-cell
connection resistance values are more restrictive, consistent with
design basis calculations and appropriately identified in
maintenance procedures. The proposed changes do not physically alter
safety related systems. There will be no effect on those plant
systems necessary to assure the accomplishment of protection
functions. The applicable radiological dose consequence acceptance
criteria will continue to be met.
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: Joseph A. Aluise, Associate General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
1 (ANO-1), Pope County, Arkansas
Date of amendment request: January 28, 2013.
Description of amendment request: The ANO-1 Technical Specification
(TS) requirements are revised from requirements on battery cells to
requirements on the battery. This focuses the requirements on the
assumed safety function of the battery. The proposed amendment would
revise TS requirements related to direct current (DC) electrical
systems in TS Limiting Condition for Operation (LCO) 3.8.4, ``DC
Sources--Operating,'' LCO 3.8.5, ``DC Sources--Shutdown,'' and LCO
3.8.6, ``Battery Parameters.'' A new ``Battery Monitoring and
Maintenance Program'' is being proposed for Section 5.5,
``Administrative Controls--Programs and Manuals.''
These changes are consistent with the NRC-approved Technical
Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, ``DC
Electrical Rewrite--Update to TSTF-360.'' The availability of this TS
improvement was announced in the Federal Register on September 1, 2011
(76 FR 54510), as part of the consolidated line item improvement
process (CLIIP).
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 changes restructure the Technical Specifications
(TS) for the direct current (DC) electrical power system and are
consistent with TSTF-500, Revision 2. The proposed changes modify TS
Actions relating to battery and battery charger operability
requirements. The DC electrical power system, including associated
battery chargers, is not an initiator of any accident sequence
analyzed in the Safety Analysis Report (SAR). Rather, the DC
electrical power system supports equipment used to mitigate
accidents. The proposed changes to restructure TS and change
surveillances for batteries and chargers to incorporate the
applicable updates included in TSTF-500, Revision 2, will maintain
the same level of equipment performance required for mitigating
accidents assumed in the SAR. Operation in accordance with the
proposed TS would ensure that the DC electrical power system is
capable of performing its specified safety function as described in
the SAR. Therefore, the mitigating functions supported by the DC
electrical power system will continue to provide the protection
assumed by the analysis. A new licensee-controlled Battery
Monitoring and Maintenance Program will ensure appropriate
monitoring and maintenance that is consistent with industry
standards. In addition, the DC electrical power system is within the
scope of 10 CFR 50.65, ``Requirements for monitoring the
effectiveness of maintenance at nuclear power plants,'' which will
ensure the control of maintenance activities associated with the DC
electrical power system.
The integrity of fission product barriers, plant configuration,
and operating procedures as described in the SAR will not be
affected by the proposed changes. Therefore, the consequences of
previously analyzed accidents will not increase by implementing
these changes.
Therefore, the proposed changes do 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 changes involve restructuring the TS for the DC
electrical power system. The DC electrical power system, including
associated battery chargers, is not an initiator to any accident
sequence analyzed in the SAR. Rather, the DC electrical power system
supports equipment used to mitigate accidents. The proposed changes
to restructure the TS and change surveillances for batteries and
chargers to incorporate the applicable updates included in TSTF-500,
Revision 2, will maintain the same level of equipment performance
required for mitigating accidents assumed in the SAR. Administrative
and mechanical controls are in place to ensure the design and
operation of the DC systems continues to meet the plant design basis
described in the SAR.
Therefore, operation of the facility in accordance with this
proposed 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.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The equipment margins will be
[[Page 25314]]
maintained in accordance with the plant-specific design bases as a
result of the proposed changes. The proposed changes will not
adversely affect operation of plant equipment. These changes will
not result in a change to the setpoints at which protective actions
are initiated. Sufficient DC capacity to support operation of
mitigation equipment is ensured. The changes associated with the new
Battery Maintenance and Monitoring Program will ensure that the
station batteries are maintained in a highly reliable manner. The
equipment fed by the DC electrical sources will continue to provide
adequate power to safety-related loads in accordance with analysis
assumptions.
TS changes made in accordance with TSTF-500, Revision 2,
maintain the same level of equipment performance stated in the SAR
and the current TSs.
Therefore, the proposed changes do not involve a significant
reduction 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: Michael T. Markley.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
2 (ANO-2), Pope County, Arkansas
Date of amendment request: January 28, 2013.
Description of amendment request: The ANO-2 Technical Specification
(TS) requirements are revised from requirements on battery cells to
requirements on the battery. This focuses the requirements on the
assumed safety function of the battery. The proposed amendment would
revise the TS requirements related to direct current (DC) electrical
systems in TS Limiting Condition for Operation (LCO) 3.8.2.3, ``DC
Distribution--Operating,'' and LCO 3.8.2.4, ``DC Distribution--
Shutdown.'' Because ANO-2 is a custom TS plant, a new TS 3.8.3,
``Battery Parameters,'' would be created to capture the intent of
Standard TS (STS) LCO 3.8.6, ``Battery Parameters,'' as modified by
TSTF-500. A new ``Battery Monitoring and Maintenance Program'' is also
being proposed for Section 6.5, ``Administrative Controls--Programs and
Manuals.''
These changes are consistent with the NRC-approved Technical
Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, ``DC
Electrical Rewrite--Update to TSTF-360.'' The availability of this TS
improvement was announced in the Federal Register on September 1, 2011
(76 FR 54510), as part of the consolidated line item improvement
process (CLIIP).
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 changes restructure the Technical Specifications
(TS) for the direct current (DC) electrical power system and are
consistent with TSTF-500, Revision 2. The proposed changes modify TS
Actions relating to battery and battery charger operability. The DC
electrical power system, including associated battery chargers, is
not an initiator of any accident sequence analyzed in the Safety
Analysis Report (SAR). Rather, the DC electrical power system
supports equipment used to mitigate accidents. The proposed changes
to restructure TS and change surveillances for batteries and
chargers to incorporate the applicable updates included in TSTF-500,
Revision 2, will maintain the same level of equipment performance
required for mitigating accidents assumed in the SAR. Operation in
accordance with the proposed TS would ensure that the DC electrical
power system is capable of performing its specified safety function
as described in the SAR. Therefore, the mitigating functions
supported by the DC electrical power system will continue to provide
the protection assumed by the analysis. A new licensee-controlled
Battery Monitoring and Maintenance Program will ensure appropriate
monitoring and maintenance that is consistent with industry
standards. In addition, the DC electrical power system is within the
scope of 10 CFR 50.65, ``Requirements for monitoring the
effectiveness of maintenance at nuclear power plants,'' which will
ensure the control of maintenance activities associated with the DC
electrical power system.
The integrity of fission product barriers, plant configuration,
and operating procedures as described in the SAR will not be
affected by the proposed changes. Therefore, the consequences of
previously analyzed accidents will not increase by implementing
these changes.
Therefore, the proposed changes do 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 changes involve restructuring the TS for the DC
electrical power system. The DC electrical power system, including
associated battery chargers, is not an initiator to any accident
sequence analyzed in the SAR. Rather, the DC electrical power system
supports equipment used to mitigate accidents. The proposed changes
to restructure the TS and change surveillances for batteries and
chargers to incorporate the applicable updates included in TSTF-500,
Revision 2, will maintain the same level of equipment performance
required for mitigating accidents assumed in the SAR. Administrative
and mechanical controls are in place to ensure the design and
operation of the DC systems continues to meet the plant design basis
described in the SAR.
Therefore, operation of the facility in accordance with this
proposed 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.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The equipment margins will be maintained in
accordance with the plant-specific design bases as a result of the
proposed changes. The proposed changes will not adversely affect
operation of plant equipment. These changes will not result in a
change to the setpoints at which protective actions are initiated.
Sufficient DC capacity to support operation of mitigation equipment
is ensured. The changes associated with the new Battery Maintenance
and Monitoring Program will ensure that the station batteries are
maintained in a highly reliable manner. The equipment fed by the DC
electrical sources will continue to provide adequate power to
safety-related loads in accordance with analysis assumptions.
TS changes made in accordance with TSTF-500, Revision 2,
maintain the same level of equipment performance stated in the SAR
and the current TSs.
Therefore, the proposed changes do not involve a significant
reduction 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: Michael T. Markley.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3, Nuclear Generating Plant (CR-3), Citrus County, Florida
Date of amendment request: March 20, 2013.
Description of amendment request: Due to the pending corporate name
[[Page 25315]]
change for the licensee of CR-3, the licensee is requesting that an
amendment be made to this license to reflect the change in the name of
the licensee from Florida Power Corporation to Duke Energy Florida,
Inc.
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?
The proposed amendment involves a change of the corporate name
from Florida Power Corporation to Duke Energy Florida, Inc. The
proposed amendment does not involve any change in the technical
qualifications of the licensee or the plant's design, configuration,
or operation. All Limiting Conditions for Operation, Limiting Safety
System Settings and Safety Limits specified in the CR-3 Improved
Technical Specifications remain unchanged. Also, the Physical
Security Plan and related plans, the Operator Training and
Requalification Program, the Quality Assurance Program, and the
Emergency Plan will not be materially changed by the proposed name
change. The corporate name change amendment will not affect the
executive oversight provided by the Chief Nuclear Officer and his
staff.
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?
The proposed amendment does not involve any change in the
plant's design, configuration, or operation. The current plant
design, design bases, and plant safety analysis will remain the
same.
The Limiting Conditions for Operations, Limiting Safety System
Settings and Safety Limits specified in the CR-3 Improved Technical
Specifications are not affected by the proposed corporate name
change. As such, the plant conditions for which the design basis
accident analysis was performed remain valid.
The proposed amendment does not introduce a new mode of plant
operation or new accident precursors, does not involve any physical
alterations to plant configuration, or make changes to system set
points that could initiate 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 accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
The proposed amendment does not involve a change in the plant's
design, configuration, or operation. The proposed amendment does not
affect either the way in which the plant structures, systems, and
components perform their safety function or its design and licensing
bases.
Plant safety margins are established through Limiting Conditions
for Operation, Limiting Safety System Settings and Safety Limits
specified in the Technical Specifications. Because there is no
change to the physical design of the plant, there is no change to
any of these margins.
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: Lara S. Nichols, 550 South Tryon Street,
Charlotte NC 28202.
NRC Branch Chief: Jessie Quichocho.
NextEra Energy Seabrook, LLC., Docket No. 50-443, Seabrook Station,
Unit 1, Rockingham County, New Hampshire
Date of amendment request: March 13, 2013.
Description of amendment request: The proposed amendment will
revise the Seabrook Technical Specifications (TSs). Specifically, the
proposed amendment will modify the circuitry that initiates high-head
safety injection (SI) by adding a new permissive, cold leg injection
permissive. This permissive prevents opening of the high head SI valves
until the reactor coolant system pressure decreases to the cold leg
injection permissive set point.
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, along with NRC edits in square
brackets:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change adds an additional permissive before high
head safety injection is initiated to assist the operators in
mitigating the consequences of an inadvertent initiation of the
emergency core cooling system (ECCS). This change in the ECCS
actuation circuitry does not increase the probability of any
accident previously evaluated because:
There is no effect on any of the systems, structures,
or components that are used for normal operation of the plant,
There is no effect on any of the fission product
barriers,
This change will not affect the normal operating
procedures,
The revised circuitry will delay the initiation of high head SI
until reactor coolant pressure is below the CLIP [cold leg injection
permissive] setpoint; however, the proposed change does not
significantly increase the consequences of accidents previously
evaluated. The proposed change does not alter ECCS flow. The delayed
opening of the high head SI valves has been evaluated for the effect
on the consequences of the following:
Mass and energy release for steam line break accidents,
Steam line break--UFSAR section 15.1.5 (specifically
hot zero-power conditions)
Feedwater line break--UFSAR section 15.2.8
Inadvertent operation of emergency core cooling system
during power operation--UFSAR section 15.5.1
Chemical and volume control system malfunction that
increases reactor coolant inventory--UFSAR section 15.5.2
For all of the above evaluated accidents, the analysis results
continue to meet all the safety limits. For the inadvertent
initiation of ECCS event, the proposed change assists the operators
in mitigating the event by significantly extending the time for the
pressurizer to fill. Additional evaluations of small break LOCA
[loss-of-coolant accident], best estimate large break LOCA, long
term cooling, LOCA forces, cold overpressure mitigation/low
temperature over pressure protection, steam generator tube rupture,
and LOCA mass and energy release were performed and it was concluded
that they were not affected by this change.
In addition evaluations were performed for the centrifugal
charging pumps and reactor vessel internals; and for the NSSS
[nuclear steam supply system] design transients to determine if the
change in the timing of the high head injection would have an effect
and it was concluded that these components and transients are not
adversely affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change adds new components to the process
protection racks and solid state protection system similar to the
components and configurations that are already installed. The
sequence of operation of equipment used to mitigate the consequences
of an accident is changed; however, it does not add any different
types of equipment. The proposed change is a change to the
protection circuitry for the plant and not to the system or
equipment used for normal operation of the plant. It does not alter
any fluid flow paths or fission product barriers and does not change
the method of control of any plant systems used for normal
operations. The proposed change does not alter or prevent the
ability of the ECCS to perform its specified function to mitigate
the consequences of an initiating event within assumed acceptance
limits. The evaluation of the centrifugal charging pumps, reactor
internals, control systems and NSSS design transients confirmed that
new failure modes were not created.
[[Page 25316]]
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary, and containment structure) to limit the
level of radiation dose to the public. The proposed changes will not
relax any criteria used to establish safety limits and will not
relax any safety system settings. The safety analysis acceptance
criteria are not affected by this change. The proposed change will
not result in plant operation in a configuration outside the design
basis.
The proposed change does involve a change in the timing of the
mitigation of inadvertent ECCS actuation and steam line break.
This change provides additional time for mitigating the
inadvertent operation of emergency core cooling system during power
operation event prior to filling the pressurizer water solid, by
preventing the injection of high head safety injection when it is
not required.
This change delays the injection of high head safety injection
on a steam line break. The delay has no effect on the steamline
break mass and energy releases and the limiting analysis of record
hot zero power steam line break as discussed below.
An evaluation was performed to address the impact of the CLIP
modification on the steamline break (SLB) mass and energy release
stretch power uprate (SPU) analyses, the current analysis of record.
For the steamline break mass and energy analyses, the CLIP
modification has the potential to delay initiation of ECCS injection
by inhibiting auto-open of the cold leg injection valves until both
an S-signal and a CLIP signal are present. There are three parts to
the evaluation: Part 1 addresses the licensing-basis cases for
steamline break mass and energy release inside containment, part 2
addresses the licensing-basis cases for steamline break mass and
energy release outside containment, and part 3 addresses steamline
break s smaller than those analyzed for the updated final safety
analysis report (UFSAR) for which there may be an S-signal but no
signal associated with the CLIP.
Steamline break inside containment. For these breaks, two
different break types are analyzed: double-ended ruptures and split
breaks. All cases from the SPU analysis were reviewed with respect
to the timing of SI flow actuation from the analysis of record and
when SI flow delivery with CLIP occurs.
In the SPU steamline break mass and energy release analysis for
double-ended ruptures, the first signal is low steam pressure for
all cases. Using the SPU analysis output results, the assumed time
of SI flow delivery is compared to the time when SI flow delivery
with CLIP occurs. The results are that for all of the double-ended
ruptures, SI flow delivery with CLIP is not reached until after the
time assumed for SI flow delivery in the SPU analysis. Although an
increase in safety injection delay is considered nonconservative, a
sensitivity calculation was specifically performed to evaluate the
impact of safety injection and the results show that mass and energy
releases are not impacted by the increased delay time for safety
injection. These results were expected as the ECCS injection occurs
at relatively low flow rates due to high reactor coolant system
pressure, and boron injection occurs long after the return to power
has been mitigated by increasing reactor coolant system temperature.
Any delay in initiation of ECCS injection has a negligible effect on
core cooling throughout the event and core reactivity during the
initial return to power.
In the SPU steamline break mass and energy release analysis for
split breaks, the first signal is the time of the first high
containment pressure setpoint. Using the SPU analysis output
results, the assumed time of SI flow delivery is compared to the
time when SI flow delivery with CLIP occurs. The results are that
for all of the split breaks, SI flow delivery with CLIP is not
reached until after the time assumed for SI flow delivery in the SPU
analysis. Although an increase in safety injection delay is
considered non-conservative, a sensitivity calculation was
specifically performed to evaluate the impact of safety injection
and the results show that mass and energy releases are not impacted
by the increased delay time for Safety Injection. These results were
expected as the ECCS injection occurs at relatively low flow rates
due to high reactor coolant system pressure, and boron injection
occurs long after the return to power has been mitigated by
increasing reactor coolant system temperature. Any delay in
initiation of ECCS injection has a negligible effect on core cooling
throughout the event and core reactivity during the initial return
to power.
Steamline break outside containment. The SPU analysis for the
steamline break mass and energy release outside containment was also
evaluated for the CLIP modification. Each steamline break case
actuated ECCS flow on a low-low pressurizer pressure S-signal. The
CLIP modification requires an S-signal and a CLIP signal. The
results show that the credited S-signal is much later than the CLIP
signal. The results from the SPU analysis remain valid and bounding
for the CLIP modification.
Smaller Steamline breaks. For the condition involving an S-
signal actuation with pressurizer pressure above the CLIP setpoint,
sensitivity cases varying the start time for ECCS injection,
including no ECCS injection have concluded that the instantaneous
and integrated mass and energy releases are insensitive to the
injection start time. These results were expected as the ECCS
injection occurs at relatively low flow rates due to high reactor
coolant system pressure, and boron injection occurs long after the
return to power has been mitigated by increasing reactor coolant
system temperature. Any delay in initiation of ECCS injection has a
negligible effect on core cooling throughout the event and core
reactivity during the initial return to power.
The hot zero-power steamline break event remains bounding for
operation at the current uprate conditions. The CLIP modification
does not impact the limiting case for hot zero-power steamline break
results because the cold leg injection valves will be fully open
before the as-modeled high head safety injection flow starts. In
addition, sensitivity studies confirm that the maximum break size
remains bounding for the hot zero-power steamline break event with
the CLIP modification.
The above evaluation shows that the installation of a CLIP would
not impact the Seabrook steamline break mass and energy release
licensing basis or the hot zero-power steam line break results.
The feedline break (FLB) has been reanalyzed with the additional
conservatism, with respect to the SPU FLB analysis, of assuming no
safety injection flow. The results of the analysis show that all the
safety limits continue to be met even with the additional
conservatism of no safety injection assumed. The assumption that
operator action is required to mitigate the consequences of a
chemical and volume control malfunction is not changed by this
modification. Before CLIP, the event was bounded by the inadvertent
ECCS actuation event and its associated operator action. With CLIP,
the event requires operator action to terminate charging and seal
injection flows. As discussed above, the consequences of the other
accidents evaluated remain bounded by the analyses of record. The
results of analyses and evaluations supporting the proposed change
demonstrate acceptance criteria continue to be met.
Therefore, these 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 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. James Petro, Managing Attorney, Florida
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Meena Khanna.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
1, Rockingham County, New Hampshire
Date of amendment request: March 27, 2013.
Description of amendment request: The proposed amendment will
revise the Seabrook Technical Specifications (TSs). Specifically, the
proposed amendment will modify TS requirements regarding steam
generator tube inspections and reporting as described in TS Task Force
(TSTF)-510, Revision 2, ``Revision to Steam Generator Program
Inspection Frequencies and Tube Sample Selection,'' using the
Consolidated Line Item Improvement Process (CLIIP). The
[[Page 25317]]
changes are consistent with Industry/TSTF Standard Technical
Specification Change Traveler, TSTF-510. The availability of this TS
improvement was announced in the Federal Register on October 27, 2011
(76 FR 66763), as part of the CLIIP.
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 revises the Steam Generator (SG) Program to
modify the frequency of verification of SG tube integrity and SG
tube sample selection. A steam generator tube rupture (SGTR) event
is one of the design basis accidents that is analyzed as part of a
plant's licensing basis. The proposed SG tube inspection frequency
and sample selection criteria will continue to ensure that the SG
tubes are inspected such that the probability of a SGTR is not
increased. The consequences of a SGTR are bounded by the
conservative assumptions in the design basis accident analysis. The
proposed change will not cause the consequences of a SGTR to exceed
those assumptions.
Therefore, it is concluded that this 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 changes to the Steam Generator Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
proposed change does not affect the design of the SGs or their
method of operation. In addition, the proposed change does not
impact any other plant system or component.
Therefore, it is concluded that this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
also isolate the radioactive fission products in the primary coolant
from the secondary system. In summary, the safety function of a SG
is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change will continue to require monitoring of the physical
condition of the SG tubes such that there will not be a reduction in
the margin of safety compared to the current requirements.
Therefore, it is concluded that 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James Petro, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Meena Khanna.
Virginia Electric and Power Company, Docket No. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of amendment request: February 22, 2013.
Description of amendment request: The proposed change will allow
the sequence and overlap limits to be exceeded and TS 3.1.6.C Action
entered if a failure is identified during the performance of
Surveillance Requirement (SR) 3.1.4.2, which verifies control rod
freedom of movement. This will align the sequence and overlap limit of
Condition A with the control bank insertion limit Condition B. The
control bank insertion limit of Condition B was modified with this same
change in Amendments 179 and 160. The subsequent change to Improved
Technical Specifications (ITS) added the Condition for sequence and
overlap limits but failed to include the exception if a failure is
identified during control rod freedom of movement 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:
Criterion 1--Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed amendment would modify the North Anna Power Station
current licensing basis by increasing the time that a single rod
bank may be permitted to be outside of sequence and overlap limits.
The new allowance only applies to minor sequence and overlap limit
differences. The proposed change will result in a small increase in
the probability that, at any given time, a control bank will be
inserted outside of sequence and overlap limits. However, the
probability of occurrence of previously evaluated accidents is not
affected, since the existing TS already permit a similar deviation
with respect to insertion limit. Only the allowed duration of the
sequence and overlap limits' exceedance is being changed.
The allowed misalignment is not a malfunction of equipment
important to safety; therefore, the probability of such a
malfunction is not increased. A single rod bank's position within 18
steps of its sequence and overlap limits does not significantly
increase the probability of a malfunction of a component important
to safety. This change does not impact the requirement that the rod
bank shall be operable (i.e., trippable); as such, it remains able
to fulfill its safety function. Therefore, the proposed amendment
does not involve a significant increase in the consequences of a
previously evaluated accident.
Therefore, neither the probability of occurrence nor the
consequences of an accident previously evaluated is significantly
increased.
Criterion 2--Does the change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed amendment does not create any new allowances for
operating the plant. Only the duration of an existing allowance is
being lengthened, with additional restrictions being applied during
the extended allowance. No physical changes are being made to any
portion of the plant, so no new accident causal mechanisms are being
introduced. The proposed change does not result in any new
mechanisms that could initiate damage to the reactor or its
principal safety barriers (i.e., fuel cladding, reactor coolant
system, or primary containment).
Therefore, the possibility for a new or different kind of
accident from any accident previously evaluated is not created.
Criterion 3--Does this change involve a significant reduction in a
margin of safety?
The proposed amendment does not affect the inputs or assumptions
of any of the design basis analyses that demonstrate the integrity
of the fuel cladding, reactor coolant system, or containment during
accident conditions. Operation within the proposed limits will not
cause unacceptable core radial peaking factors that could result in
exceeding departure from nucleate boiling (DNB) limits. Operation
within the sequence and overlap limit differences will not result in
shutdown margins lower than assumed in the accident analyses.
Control and Shutdown rods will remain fully operable (i.e.,
trippable) during the duration of the proposed extended allowance.
Therefore, it is concluded that this change does not involve a
significant reduction in the margin of safety.
[[Page 25318]]
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: Robert Pascarelli.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses and Combined Licenses, Proposed No
Significant Hazards Consideration Determination, and Opportunity for a
Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Southern California Edison, Docket No. 50-361, San Onofre Nuclear
Generating Station, Unit 2, San Diego County, California
Date of amendment request: April 5, 2013, as supplemented by letter
dated April 9, 2013.
Brief description of amendment request: The proposed amendment
makes a temporary change to the steam generator management program and
the license condition for maximum power. For the duration of Unit 2,
Cycle 17, the proposed amendment would change the terms ``full range of
normal operating conditions'' and ``normal steady state full power
operation'' and restricts operation to 70 percent of the maximum
authorized power level. ``Full range of normal operating conditions''
and ``normal steady state full power operation'' shall be based upon
the steam generators being operated under conditions associated with
reactor core power levels up to 70 percent Rated Thermal Power (2406.6
megawatts thermal).
Date of publication of individual notice in Federal Register: April
16, 2013 (78 FR 22576).
Expiration date of individual notice: May 16, 2013 (public
comments) and June 17, 2013 (hearing requests).
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 are available for public inspection at
the NRC's Public Document Room (PDR), located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible online through the Agencywide Documents Access and
Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to
[email protected].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of application for amendment: June 22, 2011, as supplemented
by letters dated December 9, 2011, January 27, 2012, and January 30,
2013.
Brief description of amendment: The amendments revised Technical
Specification (TS) 3.7.4, ``Atmospheric Dump Valves (ADVs).''
Specifically, the amendments revised the Limiting Condition for
Operation for TS 3.7.4, with corresponding revisions to the TS
Conditions, Required Actions, and Completion Times associated with one
or more inoperable ADV lines.
Date of issuance: April 11, 2013.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: Unit 1-191; Unit 2-191; Unit 3-191.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendment revised the Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: October 4, 2011 (76 FR
61394). The supplemental letters dated December 9, 2011, January 27,
2012, and January 30, 2013, 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 April 11, 2013.
No significant hazards consideration comments received: No.
Carolina Power and Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: October 22, 2012.
Brief Description of amendment: The amendment revised the Technical
Specification (TS) surveillance requirements for addressing missed
surveillances, and is consistent with the Nuclear Regulatory Commission
approved Revision 6 of Technical Specification Task Force (TSTF)
Standard TSs Change Traveler TSTF-358, ``Missed Surveillance
Requirements.''
Date of issuance: April 11, 2013.
Effective date: As of date of issuance and shall be implemented
within 90 days.
[[Page 25319]]
Amendment No.: 141.
Renewed Facility Operating License No. NPF-63: Amendment revised
the TSs.
Date of initial notice in Federal Register: November 27, 2012 (77
FR 70839).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 11, 2013.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit 3, Westchester County, New York
Date of application for amendment: May 23, 2012, as supplemented on
August 3, 2012.
Brief description of amendment: The amendment revises Technical
Specification 3.7.4, ``Atmospheric Dump Valves (ADVs),'' Limiting
Condition for Operation 3.7.4 to require four operable ADVs instead of
three.
Date of issuance: April 15, 2013.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 251.
Facility Operating License Nos. DPR-26 and DPR-64: The amendment
revised the License and the Technical Specifications.
Date of initial notice in Federal Register: September 14, 2012 (77
FR 56880).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 15, 2013.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station (VYNPS),
Vernon, Vermont
Date of amendment request: March 5, 2012, as supplemented on
November 20, 2012, March 26, March 29, and April 5, 2013.
Brief description of amendment: The amendment revised the VYNPS
License Condition 3.P and 3.Q to clarify that the information in the
updated final safety analysis report (UFSAR) supplement submitted
pursuant to Section 54.21(d) of Title 10 of the Code of Federal
Regulations (10 CFR), as revised during the license renewal application
review process, and as supplemented by commitments of Appendix A of
Supplement 2 of NUREG-1907, can be incorporated as part of the UFSAR
and may be changed without prior NRC approval provided the requirements
of 10 CFR 50.59 have been previously satisfied.
Date of Issuance: April 17, 2013.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 256.
Facility Operating License No. DPR-28: The amendment revised the
License.
Date of initial notice in Federal Register: April 3, 2012 (77 FR
20074). The supplemental correspondence dated November 20, 2012, March
26, March 29, and April 5, 2013, 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 this amendment is contained
in a Safety Evaluation dated April 17, 2013.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Date of application for amendment: March 22, 2012, as supplemented
by letter dated. December 3, 2012.
Brief description of amendment: The amendment modifies technical
specification (TS) requirements regarding steam generator tube
inspections and reporting as described in Technical Specifications Task
Force (TSTF)-510, ``Revision to Steam Generator Program Inspection
Frequencies and Tube Sample Selection,'' with proposed variations and
deviations.
Date of issuance: March 21, 2013.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 172 and 172, respectively.
Facility Operating License Nos. NPF-72 and NPF-77: The amendment
revised the Technical Specifications and License.
Date of initial notice in Federal Register: (77 FR 31660; May 29,
2012).
The December 3, 2012, supplement did not increase the scope of the
application and did not change the NRC staff's initial proposed finding
of no significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 21, 2013.
No significant hazards consideration comments received: No.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
1, Rockingham County, New Hampshire
Date of amendment request: December 20, 2012.
Description of amendment request: The amendment revised the
Seabrook Technical Specifications (TS) TS 6.7.6.m, ``Reactor Coolant
Pump Flywheel Inspection Program.'' The amendment extends the reactor
coolant pump (RCP) motor flywheel examination frequency from the
currently approved 10-year inspection interval, to an interval not to
exceed 20 years. The changes are consistent with Industry/Technical
Specification Task Force (TSTF) Standard Technical Specification Change
Traveler, TSTF-421, ``Revision to RCP Flywheel Inspection Program
(WCAP-15666).'' The availability of this TS improvement was announced
in the Federal Register on October 22, 2003, as part of the
consolidated line item improvement process (CLIIP).
Date of issuance: April 4, 2013.
Effective date: As of its date of issuance and shall be implemented
within 60 days.
Amendment No.: 134.
Facility Operating License No. NPF-86: The amendment revised the
License and TS.
Date of initial notice in Federal Register: January 22, 2013 (78 FR
4473).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 4, 2013.
No significant hazards consideration comments received: No.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
1, Rockingham County, New Hampshire
Date of amendment request: November 17, 2011, as supplemented by
letters dated December 3, 2012, and January 9, 2013.
Description of amendment request: The change revised the
applicability of the figures in the Technical Specifications for the
reactor coolant system pressure-temperature limits and the cold
overpressure protection setpoints. The change revised the applicability
of the figures from 20 effective full-power years (EFPY) to 23.7 EFPY.
Date of issuance: April 15, 2013.
Effective date: As of its date of issuance and shall be implemented
within 60 days.
Amendment No.: 135.
Facility Operating License No. NPF-86: The amendment revised the TS
and the License.
Date of initial notice in Federal Register: January 10, 2012 (77 FR
1519). The supplements dated December 3, 2012, and January 9, 2013,
provided additional information that clarified the application, did not
expand
[[Page 25320]]
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 15, 2013.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: November 30, 2011, as supplemented by
letters dated August 16 and December 7, 2012, and March 3, 2013.
Brief description of amendment: The amendment revised the Technical
Specification 3.8.1, ``AC [Alternating Current] Sources--Operating,''
Surveillance Requirements related to Diesel Generator test loads,
voltage, and frequency. The changes correct non-conservative Diesel
Generator load values that are currently under administrative controls.
Date of issuance: April 11, 2013.
Effective date: As of its date of issuance and shall be implemented
within 90 days of the date of issuance.
Amendment No.: 204.
Renewed Facility Operating License No. NPF-42. The amendment
revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: June 12, 2012 (77 FR
35078). The supplemental letters dated August 16 and December 7, 2012,
and March 3, 2013, 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 April 11, 2013.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 22nd day of April 2013.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2013-10020 Filed 4-29-13; 8:45 am]
BILLING CODE 7590-01-P