[Federal Register Volume 77, Number 6 (Tuesday, January 10, 2012)]
[Notices]
[Pages 1514-1521]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2012-124]
=======================================================================
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[NRC-2011-0303]
Biweekly Notice; Applications and Amendments to Facility
Operating 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
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 December 15, 2011 to December 28, 2011. The
last biweekly notice was published on December 27, 2011 (76 FR 80972).
Addresses: Please include Docket ID NRC-2011-0303 in the subject
line of your comments. Comments submitted in writing or in electronic
form will be posted on the NRC Web site and on the Federal rulemaking
Web site http://www.regulations.gov. Because your comments will not be
edited to remove any identifying or contact information, the NRC
cautions you against including any information in your submission that
you do not want to be publicly disclosed.
The NRC requests that any party soliciting or aggregating comments
received from other persons for submission to the NRC inform those
persons that the NRC will not edit their comments to remove any
identifying or contact information, and therefore, they should not
include any information in their comments that they do not want
publicly disclosed.
You may submit comments by any one of the following methods.
Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for documents filed under Docket ID NRC-
2011-0303. Address questions about NRC dockets to Carol Gallagher (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, or by fax to RADB at (301) 492-3446.
You can access publicly available documents related to this notice
using the following methods:
NRC's Public Document Room (PDR): The public may examine
and have copied for a fee publicly available documents at the NRC's
PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike,
Rockville, Maryland 20852.
NRC's Agencywide Documents Access and Management System
(ADAMS): 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. From this page, the public
can gain entry into ADAMS, which provides text and image files of the
NRC's public documents. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC's PDR reference staff at 1-(800) 397-4209, (301) 415-4737, or by
email to [email protected]. From this page, the public can gain
entry into ADAMS, which provides text and image files of NRC's public
documents. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the NRC's PDR
reference staff at 1-(800) 397-4209, (301) 415-4737, or by email to
[email protected].
Federal Rulemaking Web Site: Public comments and
supporting materials related to this notice can be found at http://www.regulations.gov by searching on Docket ID: NRC-2011-0303.
Notice of Consideration of Issuance of Amendments to Facility Operating
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 Title 10 of the Code of Federal
Regulations (10 CFR) 50.92, 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; (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
[[Page 1515]]
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.
Requests for a hearing and a petition for leave to intervene shall be
filed in accordance with the Commission's ``Rules of Practice for
Domestic Licensing Proceedings'' 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 20874. The NRC
regulations are accessible electronically from the NRC Library on the
NRC 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 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 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 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
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 the NRC guidance
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is
[[Page 1516]]
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 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 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
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. 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. Non-timely filings
will not be entertained absent a determination by the presiding officer
that the petition or request should be granted or the contentions
should be admitted, based on a balancing of the factors specified in 10
CFR 2.309(c)(1)(i)-(viii).
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].
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: August 22, 2011.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 6.9.1.6, ``Core Operating Limits
Report,'' to add plant-specific methodology, ANP-3011 (P), ``Harris
Nuclear Plant Unit 1 Realistic Large Break LOCA [Loss-of-Coolant
Accident] Analysis,'' Revision 1, that implements AREVA's NRC-approved
topical report, EMF-2103(P)(A), ``Realistic Large Break LOCA
Methodology for Pressurized Water Reactors,'' and add EMF-2103(P)(A),
``Realistic Large Break LOCA Methodology for Pressurized Water
Reactors,'' Revision 2 or higher upon approval of the specific revision
by the NRC, to the TS 6.9.1.6.2 listing of analytical methods used to
determine the core operating limits, and eliminates extraneous detail
in TS 6.9.1.6 that cross references each method to the applicable TS
Section 3.0 specifications and parameters.
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 TR [topical report] underlying the proposed HNP [Shearon
Harris Nuclear Power Plant] methodology has been reviewed and
approved by the NRC for use in determining core operating limits and
for evaluation of LBLOCA [large break loss-of-coolant accident]. The
core operating limits to be developed using the new methodologies
for HNP will be established in accordance with the applicable
limitations as documented in the NRC SE [safety evaluation]. In the
April 9, 2003, NRC SE, the NRC concluded that the S-RELAP5 RLBLOCA
[realistic large break loss-of-coolant accident] methodology is
acceptable for referencing in licensing applications in accordance
with the stated limitations.
The proposed change enables the use of new methodology to re-
analyze a LBLOCA. It does not, by itself, impact the current design
bases. Revised analysis may either result in continued conformance
with design bases or may change the design bases. If design basis
changes result from a revised analysis, the specific design changes
will be evaluated in accordance with HNP design change procedures
and 10 CFR 50.59.
The proposed change does not involve physical changes to any
plant structure, system, or component (SSC). Therefore, the
probability of occurrence for a previously analyzed accident is not
significantly increased.
The consequences of a previously analyzed accident are dependent
on the initial conditions assumed for the analysis, the behavior of
the fission product barriers during the analyzed accident, the
availability and successful functioning of the equipment assumed to
operate in response to the analyzed event, and the setpoints at
which these actions are initiated.
[[Page 1517]]
The proposed methodologies will ensure that the plant continues
to meet applicable design and safety analyses acceptance criteria.
The proposed change does not affect the performance of any equipment
used to mitigate the consequences of an analyzed accident. As a
result, no analysis assumptions are impacted and there are no
adverse effects on the factors that contribute to offsite or onsite
dose as a result of an accident. The proposed change does not affect
setpoints that initiate protective or mitigative actions. The
proposed change ensures that plant SSCs are maintained consistent
with the safety analysis and licensing bases.
Therefore, this amendment does not involve a significant
increase in the probability or consequences of a previously analyzed
accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
The proposed change does not involve any physical alteration of
plant SSCs. No new or different equipment is being installed and no
installed equipment is being operated in a different manner. There
is no change to the parameters within which the plant is normally
operated or in the setpoints that initiate protective or mitigative
actions. As a result, no new failure modes are being introduced.
Therefore, the 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.
There is no impact on any margin of safety resulting from the
incorporation of this new TR into the TS or deletion of cross-
reference information from the description of the COLR [core
operating limit report]. If design basis changes result from a
revised analysis that uses these new methodologies, the specific
design changes will be evaluated in accordance with HNP design
change procedures and 10 CFR 50.59. Any potential reduction in the
margin of safety would be evaluated for that specific design change.
Therefore, this amendment does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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 T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Douglas A. Broaddus.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: October 28, 2011.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Table 3.2.B to increase the
condensate storage tank low water level setpoint for the interlock to
high-pressure coolant injection (HPCI) pump suction valves. The
proposed amendment would also correct typographical errors in TS
numbering and referencing that were introduced in previous license
amendment requests.
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 increasing of the setpoint for the Condensate Storage Tank
(CST) low water level High Pressure Coolant Injection (HPCI) System
automatic suction transfer to the Suppression Pool is not a
precursor to any accident previously evaluated. The CST is not
utilized to mitigate the consequences of any accident previously
evaluated. The increase in the setpoint provides for HPCI pump
performance with the required flow to mitigate the accident
conditions. The proposed corrections to typographical errors
incurred in the prior License Amendments provide correct references
to the applicable existing Specifications, which is an
administrative change.
The proposed changes do not involve a change to the safety
function of the HPCI system operation. The proposed TS revision
involves no significant changes to the operation of any systems or
components in normal or accident operating conditions.
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 increasing of the setpoint for the Condensate Storage Tank
(CST) low water level High Pressure Coolant Injection (HPCI) System
automatic suction transfer to the Suppression Pool is not a
precursor to any accident previously evaluated. The CST is not
utilized to mitigate the consequences of any accident previously
evaluated. The increase in the setpoint provides for HPCI pump
performance with the required flow to mitigate the accident
conditions. The proposed corrections to typographical errors
incurred in the prior License Amendments provide correct references
to the applicable existing Specifications, which is an
administrative change.
The proposed changes do not change the safety function of the
HPCI and RCIC [reactor core isolation cooling] systems. There is no
alteration to the parameters within which the plant is normally
operated. The increase in the setpoint is not a precursor to new or
different kinds of accidents and do not initiate new or different
kinds of accidents. The impact of these changes have been analyzed
and found to be acceptable within the design limits and plant
operating procedures. As a result, no new failure modes are being
introduced.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through the design of the
plant structures, systems, and components, the parameters within
which the plant is operated and the establishment of the setpoints
for the actuation of equipment relied upon to respond to an event
and design basis accidents. The proposed change increases the
setpoint at which protective actions are initiated, but does not
change the requirements governing operation or availability of
safety equipment assumed to operate to preserve the margin of
safety. The corrections to the typographical errors introduced in
prior License Amendments do not impact the safety margin.
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: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Nancy Salgado.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Unit 2, LaSalle County, Illinois
Date of amendment request: October 26, 2011.
Description of amendment request: The proposed amendment revises
license condition 2.C.(32) to require the installation of NETCO-SNAP-
IN[supreg] inserts to be completed no later than December 31, 2012, for
LaSalle County Station (LSCS) Unit 2. In addition, license condition
2.C.(31) is revised to apply until March 31, 2012, and a new license
condition 2.C.(34) is being
[[Page 1518]]
proposed to prohibit fuel storage after March 31, 2012, in spent fuel
pool (SFP) storage rack cells that have not been upgraded with the
NETCO-SNAP-IN[supreg] inserts.
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 LSCS Unit 2 Operating License to
accelerate the timeline for installation of the NETCO-SNAP-
IN[supreg] inserts in the LSCS Unit 2 SFP, and limit the time period
under which BORAFLEX\TM\ is credited as the neutron absorbing
material in the Unit 2 SFP. There are no changes to the SFP
criticality analysis associated with the proposed change. The SFP
criticality analysis was previously approved by the NRC and
continues to demonstrate that the effective neutron multiplication
factor, Keff, is less than or equal to 0.95 if the SFP is fully
flooded with unborated water. No physical changes to the plant are
proposed, no new plant equipment is being installed, and there are
no changes to the manner in which the plant is operated. Rather, the
proposed change is administrative because it involves accelerating
the timeline for installing the NETCO-SNAP-IN[supreg] inserts and
limiting the time period under which BORAFLEX\TM\ is credited as the
neutron absorbing material in the Unit 2 SFP.
The probability that a fuel assembly would be dropped is
unchanged by the proposed change. These events involve failures of
administrative controls, human performance, and equipment failures
that are unaffected by the proposed change. The proposed change does
not result in a significant increase in the consequence of an
accident previously analyzed. The criticality analysis that
demonstrates adequate margin to criticality for spent fuel storage
rack cells with rack inserts in the LSCS Unit 2 SFP, and adequate
criticality margin for assemblies accidentally dropped onto the
spent fuel storage racks, is not being changed. The consequences of
dropping a fuel assembly onto any other fuel assembly or other
structure are unaffected by the proposed change.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the LSCS Unit 2 Operating License to
accelerate the timeline for installation of the NETCO-SNAP-
IN[supreg] inserts in the LSCS Unit 2 SFP, and limit the time period
under which BORAFLEX\TM\ is credited as the neutron absorbing
material in the Unit 2 SFP. There are no changes to the SFP
criticality analysis associated with the proposed change. No
physical changes to the plant are proposed, and there are no changes
to the manner in which the plant is operated. Rather, the proposed
change is administrative because it involves accelerating the
timeline for installing the NETCO-SNAP-IN[supreg] inserts and
limiting the time period under which BORAFLEX\TM\ is credited as the
neutron absorbing material in the Unit 2 SFP.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the LSCS Unit 2 Operating License to
accelerate the timeline for installation of the NETCO-SNAP-
IN[supreg] inserts in the LSCS Unit 2 SFP, and limit the time period
under which BORAFLEX\TM\ is credited as the neutron absorbing
material in the Unit 2 SFP. Plant safety margins are established
through limiting conditions for operation, limiting safety system
settings, and safety limits specified in Technical Specifications.
The proposed change does not alter these established safety margins.
For SFP criticality, the required safety margin is 5% including a
conservative margin to account for engineering and manufacturing
uncertainties. The proposed change does not alter the criticality
analysis for the SFP and does not affect the SFP criticality safety
margin.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Jacob I. Zimmerman.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Units 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: May 27, 2011.
Description of amendment request: The proposed amendment would
modify Technical Specifications (TSs) to allow the BVPS-1 containment
spray additive, sodium hydroxide (NaOH), to be replaced by sodium
tetraborate (NaTB). Also, an administrative change to the BVPS-2
license is required.
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.
Use of NaTB in lieu of NaOH would not involve a significant
increase in probability of a previously evaluated accident because
the containment spray additive is not an initiator of any analyzed
accident. The NaTB would be stored and delivered by a passive method
that does not have potential to affect plant operations. Any
existing NaOH delivery system equipment which remains in place but
is removed from service would meet existing seismic and electrical
requirements. Therefore the change in additive, including removal of
NaOH equipment from service, would not result in any failure modes
that could initiate an accident.
The spray additive is used to mitigate the consequences of a
LOCA [loss-of-coolant accident]. Use of NaTB as an additive in lieu
of NaOH would not involve a significant increase in the consequences
of a previously evaluated accident because the amount of NaTB
specified in the proposed TS would achieve a pH of 7 or greater,
consistent with the current licensing basis. This pH is sufficient
to achieve long-term retention of iodine by the containment sump
fluid for the purpose of reducing accident related radiation dose
following a LOCA.
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.
Regarding the proposed use of NaTB in lieu of NaOH, the NaTB
would be stored and delivered by a passive method that does not have
potential to affect plant operations. Any existing NaOH delivery
system equipment that is removed from service would meet existing
seismic and electrical requirements. Hydrogen generation would not
be significantly impacted by the change.
Therefore, no new failure mechanisms, malfunctions, or accident
initiators would be introduced by the proposed change, and it would
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.
Since the quantity of NaTB specified in the amended TS would
reduce the potential for undesirable chemical effects while
achieving radiation dose reductions, corrosion control and hydrogen
generation effects that are comparable to NaOH, the proposed change
does not involve a significant reduction in a margin of safety. The
primary function of an additive is to reduce LOCA consequences by
[[Page 1519]]
controlling the amount of iodine fission products released to
containment atmosphere from reactor coolant accumulating in the sump
during a LOCA. Because the amended [TS] would achieve a pH of 7 or
greater using NaTB, dose related safety margins would not be
significantly reduced. Use of NaTB reduces the potential for
undesirable chemical effects that could interfere with recirculation
flow through the sump strainers. Any existing NaOH delivery system
equipment that remains in place but is removed from service would
meet existing seismic and electrical requirements and would not
interfere with operation of the existing containment or containment
spray system.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, FirstEnergy Nuclear
Operating Company, FirstEnergy Corporation, 76 South Main Street,
Akron, OH 44308.
NRC Branch Chief: Nancy L. Salgado.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
1, Rockingham County, New Hampshire
Date of amendment request: November 17, 2011.
Description of amendment request: The proposed change would revise
the applicability of the figures in the Technical Specifications for
the reactor coolant system (RCS) pressure-temperature limits and the
cold overpressure protection setpoints. The proposed change revises the
applicability of the figures from 20 effective full-power years (EFPY)
to 23.7 EFPY.
Basis for proposed no significant hazards consideration (NSHC)
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of NSHC, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change does not impact the physical function of
plant structures, systems, or components (SSCs) or the manner in
which SSCs perform their design function. The proposed change
neither adversely affects accident initiators or precursors, nor
alters design assumptions. The proposed change does not alter or
prevent the ability of operable SSCs to perform their intended
function to mitigate the consequences of an initiating event within
assumed acceptance limits. The change does not affect the integrity
of the RCS pressure boundary. The proposed change to the
applicability of the RCS pressure-temperature limits and the cold
overpressure protection setpoints continues to protect the integrity
of the RCS pressure boundary.
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, which revises the applicability of the RCS
pressure-temperature limits and the cold overpressure protection
setpoints, will not impact the accident analysis. The change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed), a significant change
in the method of plant operation, or new operator actions. The
proposed change will not introduce failure modes that could result
in a new accident. The RCS pressure-temperature limits and the cold
overpressure protection setpoints are not accident initiators. The
change does not alter assumptions made in the safety analysis.
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 change does 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 change does not
involve a significant change in the method of plant operation, and
no accident analyses will be affected by the proposed changes.
Additionally, 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 not adversely affect systems that respond to safely
shutdown the plant and to maintain the plant in a safe shutdown
condition. The proposed change to the applicability of the RCS
pressure-temperature limits and the cold overpressure protection
setpoints continues to protect the integrity of the RCS pressure
boundary.
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 NSHC.
Attorney for licensee: M.S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Harold K. Chernoff.
Notice of Issuance of Amendments to Facility Operating 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.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, 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 electronically 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
NRC's PDR Reference staff at 1 (800) 397-4209, (301) 415-4737, or by
email to [email protected].
[[Page 1520]]
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: March 31, 2011, as supplemented
by letter dated August 12, 2011.
Brief description of amendment: The amendments relocated certain
surveillance frequencies to a licensee-controlled program (the
Surveillance Frequency Control Program) in accordance with Technical
Specification Task Force (TSTF) Improved Standard Technical
Specifications Change Traveler TSTF-425, Revision 3, ``Relocate
Surveillance Frequencies to Licensee Control--RITSTF (Risk Informed
Technical Specification Task Force) Initiative 5b.'' The amendments
also approved two deviations from TSTF-425, Revision 3: an
administrative change which would allow it to retain a definition that
also appears in a portion of the plants' technical specifications that
are not subject to TSTF-425, and TS Bases changes recommended by the
NRC to the TSTF in a letter dated April 14, 2010.
Date of issuance: December 15, 2011.
Effective date: As of the date of issuance and shall be implemented
within 180 days from the date of issuance.
Amendment No.: Unit 1--188; Unit 2--188; Unit 3--188.
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: June 14, 2011 (76 FR
34765). The supplemental letter dated August 12, 2011, 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 December 15, 2011.
No significant hazards consideration comments received: No.
Carolina Power and Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: July 12, 2011.
Brief Description of amendments: The license amendments revised
Brunswick Steam and Electric Plant, Units 1 and 2 Technical
Specification (TS) 3.4.5, ``RCS Leakage Detection Instrumentation,''
consistent with the NRC-approved Technical Specification Task Force
(TSTF) Standard Technical Specification Change Traveler, TSTF-514,
``Revise BWR [Boiling Water Reactor] Operability Requirements and
Actions for RCS [Reactor Coolant System] Leakage Instrumentation,''
Revision 3. The availability of this TS improvement was announced in
the Federal Register on December 17, 2010 (75 FR 79048) as part of the
consolidated line item improvement process.
Date of issuance: December 21, 2011.
Effective date: Date of issuance, shall be implemented within 60
days of the effective date.
Amendment Nos.: Unit 1--260 and Unit 2--288.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: September 6, 2011 (76
FR 55127).
The Commission's related evaluation of the amendments is contained
in a safety evaluation dated December 21, 2011.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendment: April 6, 2011.
Brief description of amendment: The amendments modify the actions
to be taken when the containment atmosphere gaseous radioactivity
monitoring system and the primary containment pressure and temperature
monitoring system are the only operable reactor coolant leakage
detection monitoring systems. The modified actions require additional,
more frequent monitoring of other indications of Reactor Coolant System
(RCS) leakage and provide appropriate time to restore another
monitoring system to operable status. This change is consistent with
the U.S. Nuclear Regulatory Commission-approved safety evaluation on
Technical Specification Task Force (TSTF) Traveler, TSTF-514-A,
Revision 3, ``Revised [Boiling Water Reactor] BWR Operability
Requirements and Actions for RCS Leakage Instrumentation,'' dated
November 24, 2010.
Date of issuance: December 19, 2011.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment Nos.: 205 and 167.
Facility Operating License Nos. NPF-39 and NPF-85. These amendments
revised the license and the technical specifications.
Date of initial notice in Federal Register: August 9, 2011 (76 FR
48911).
The Commission's related evaluation of the amendment is contained
in Safety Evaluation dated December 19, 2011.
No significant hazards consideration comments received: No.
Attorney for licensee: J. Bradley Fewell, Esquire, Associate
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: December 20, 2010, as
supplemented by letters dated July 20, September 1, and October 5,
2011. The July 20, 2011, submittal entirely replaced the licensee's
submittal dated December 20, 2010.
Brief description of amendment: Florida Power Corporation (the
licensee) will be constructing and operating an onsite independent
spent fuel storage installation, under its general license, in order to
maintain full-core offload capacity in the spent fuel pools located in
the CR-3 auxiliary building (AB). In support of future dry shielded
canister/transfer cask loading operation, the licensee is replacing the
AB overhead crane. This amendment approved departure from a method for
evaluating the replaced AB overhead crane, revisions to the CR-3 Final
Safety Analysis Report (FSAR), and changes to the associated
commitments in the FSAR.
Date of issuance: December 27, 2011.
Effective date: Date of issuance, to be implemented within 180
days. The FSAR changes shall be implemented in the next periodic update
made in accordance with 10 CFR 50.71(e).
Amendment No.: 239.
Facility Operating License No. DPR-72: Amendment approved revisions
to the FSAR Sections 5.1.1.1.h, 9.6.1.5.a.5, and 9.6.3.1 as indicated
in the NRC's safety evaluation dated December 27, 2011.
Date of initial notice in Federal Register: September 6, 2011 (76
FR 55129). The supplements dated September 1 and October 5, 2011,
[[Page 1521]]
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.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York
Date of application for amendment: May 27, 2009, as supplemented on
August 28, 2009, December 23, 2009, February 19, 2010, April 16, 2010,
May 7, 2010, June 3, 2010, June 30, 2010, July 9, 2010, July 30, 2010,
September 16, 2010, October 8, 2010, October 28, 2010, November 5,
2010, December 10, 2010, December 13, 2010, January 19, 2011, January
31, 2011, February 4, 2011, March 23, 2011, May 9, 2011, June 13, 2011,
July 15, 2011, August 5, 2011, August 19, 2011, September 23, 2011,
October 27, 2011, and November 1, 2011.
Brief description of amendment: The amendment changes the NMP2
Technical Specifications to increase the maximum steady-state reactor
core power level from 3,467 megawatts thermal (MWt) to 3,988 MWt, which
is an increase from the current license of approximately 15 percent.
The proposed increase in power level is considered an extended power
uprate.
Date of issuance: December 22, 2011.
Effective date: As of the date of issuance to be implemented within
90 days.
Amendment No.: 140.
Renewed Facility Operating License No. NPF-69: The amendment
revises the License and TSs.
Date of initial notice in Federal Register: October 10, 2009 (74 FR
53778). The supplemental letters dated August 28, 2009, December 23,
2009, February 19, 2010, April 16, 2010, May 7, 2010, June 3, 2010,
June 30, 2010, July 9, 2010, July 30, 2010, September 16, 2010, October
8, 2010, October 28, 2010, November 5, 2010, December 10, 2010,
December 13, 2010, January 19, 2011, January 31, 2011, February 4,
2011, March 23, 2011, May 9, 2011, June 13, 2011, July 15, 2011, August
5, 2011, August 19, 2011, September 23, 2011, October 27, 2011, and
November 1, 2011, provided additional information that clarified the
application and did not expand the scope of the application as
originally noticed, and did not change the Nuclear Regulatory
Commission staff's initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 22, 2011.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 29th day of December 2011.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2012-124 Filed 1-9-12; 8:45 am]
BILLING CODE 7590-01-P