[Federal Register Volume 76, Number 163 (Tuesday, August 23, 2011)]
[Notices]
[Pages 52699-52715]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2011-21212]
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NUCLEAR REGULATORY COMMISSION
[NRC-2011-0187]
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 July 28, 2011, to August 10, 2011. The last
biweekly notice was published on August 9, 2011 (76 FR 48908).
ADDRESSES: Please include Docket ID NRC-2011-0187 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-
2011-0187. Address questions about NRC dockets to Carol Gallagher 301-
492-3668; e-mail [email protected].
Mail comments to: 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.
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 e-mail
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-2011-0187.
[[Page 52700]]
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), Section 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; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example, in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
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 20852. 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, 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 e-mail 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
[[Page 52701]]
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 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 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 NRC guidance
available on the NRC 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 e-mail notice confirming receipt of the document. The
E-Filing system also distributes an e-mail 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 e-mail 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 PDR Reference staff at 1-800-397-4209,
301-415-4737, or by e-mail to [email protected].
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: April 13, 2011.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TSs) as a result of a revised Fuel
Handling Accident analysis. The new analysis determined that the
current TSs may not be conservative for all scenarios. The proposed
amendment would provide new applicability and/or action language in the
TSs that includes load movements over irradiated fuel assemblies.
Specifically, the amendment would modify the following TSs: TS 3.3.3.1
(Radiation Monitoring Instrumentation); TS 3.7.6.1 (Control Room
Emergency Air Filtration System); TS 3.7.6.3 (Control Room Air
Temperature--Operating); TS 3.7.6.4 (Control Room Air Temperature--
Shutdown); TS 3.8.1.2 (A.C.
[[Page 52702]]
[Alternating Current] Sources--Shutdown); TS 3.8.2.2 (DC Sources
[Direct Current]--Shutdown); TS 3.8.3.2 (On Site Power Distribution--
Shutdown); TS 3.9.3 (Decay Time); TS 3.9.4 (Containment Building
Penetrations); and TS 3.9.7 (Crane Travel--Fuel Handling Building).
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.
This proposed change revises Technical Specifications
applicability wording regarding the movement of fuel assemblies in
containment and the fuel storage pool to include load movements over
irradiated fuel assemblies. The proposed applicability is more
comprehensive than the current Applicability. This change was driven
by an analysis change and was not due to fuel handling equipment or
fuel movement methods. Expanding the applicability of the relevant
Technical Specifications is necessary to account for updated fuel
drop analyses which demonstrate that the impacted spent fuel
assemblies may be damaged.
Consequently, dropping of a non-irradiated fuel assembly, dummy
fuel assembly, or other load could result in a Fuel Handling
Accident that has radiological consequences. Changing the
applicability of the relevant Technical Specifications does not
affect the probability of a Fuel Handling Accident. The expanded
applicability provides assurance that equipment designed to mitigate
a Fuel Handling Accident is capable of performing its specified
safety function.
The dose consequences due to failure of two assemblies remain
within the Regulatory Guide 1.183 and 10 CFR 50.67 acceptance
criteria limits. The Exclusion Area Boundary (EAB), Low Population
Zone (LPZ), and Main Control Room (MCR) dose results and associated
regulatory limits are presented below.
----------------------------------------------------------------------------------------------------------------
Regulatory guide 1.183
New analysis limit 10 CFR 50.67 limit
----------------------------------------------------------------------------------------------------------------
EAB.................................. 4.56 rem TEDE.......... <6.3 rem TEDE.......... <25 rem TEDE.
LPZ.................................. 0.70 rem TEDE.......... <6.3 rem TEDE.......... <25 rem TEDE.
MCR.................................. 0.824 rem TEDE......... <5 rem TEDE............ <5 rem TEDE.
----------------------------------------------------------------------------------------------------------------
Consequently, 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 revised spent fuel handling analyses demonstrate that the
impacted fuel assemblies may be damaged as the result of a dropped
fuel assembly, dummy assembly, or load. The existing Technical
Specifications regarding movement of fuel assemblies are not
applicable for movement of non-irradiated fuel assemblies or other
loads. A drop of these loads could cause radiological consequences
during periods when the equipment required to mitigate those
consequences is not required to be OPERABLE in accordance with the
existing Technical Specifications.
The proposed changes to the Technical Specifications
applicability language regarding the movement of these loads in
containment and the fuel storage pool ensure that Limiting
Conditions of Operation and appropriate Required Actions for
required equipment are in effect during fuel movement. This provides
assurance that the Fuel Handling Accident will remain within the
initial assumptions of accident analyses.
Consequently, there is no possibility of a new or different kind
of accident due to this change.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed Technical Specifications change will not affect
protection criterion for plant equipment and will not reduce the
margin of safety. By extending the Applicability to the movement of
non-irradiated fuel assemblies, the current margin of safety is
maintained.
Consequently, there is no significant reduction in a margin of
safety due to this change.
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
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: June 2, 2011, as supplemented by letter
dated August 1, 2011.
Description of amendment request: The proposed amendment would
approve revision to the South Texas Project (STP), Units 1 and 2, Fire
Protection Program related to the alternate shutdown capability.
Specifically, STP Nuclear Operating Company (STPNOC) proposes to credit
the following manual operator actions in the control room prior to
evacuation due to a fire for meeting the alternate shutdown capability:
Main steam line isolation.
Closing the pressurizer power-operated relief valves block
valves.
Securing all reactor coolant pumps.
Feedwater isolation.
Securing the startup feedwater pump.
Letdown isolation.
Securing the charging pumps.
In addition, STPNOC proposes to credit the automatic trip of the
main turbine upon the initiation of a manual reactor trip for meeting
the alternate shutdown capability. A thermal-hydraulic analysis will
demonstrate that these operations will ensure that the reactor coolant
system (RCS) process variables remain within those values predicted for
a loss of normal alternating current (a-c) power, as required by
Section III.L.1 of Appendix R of Title 10 of the Code of Federal
Regulations (10 CFR) part 50.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design function of structures, systems and component are not
impacted by the proposed change. The proposed change involves
crediting operations in the control room prior to evacuation in the
event of a fire in order to meet safe shutdown performance criteria.
The proposed action will not initiate an event. The proposed actions
do not increase the probability of occurrence of a fire or any other
accident previously evaluated.
The proposed operations are feasible and reliable and
demonstrate that the unit can be safely shutdown in the event of a
fire. No significant consequences result from the performance of the
proposed operations.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The design function of structures, systems and component are not
impacted by the proposed amendment. The proposed change involves
operations in response to a fire. They do not involve new failure
mechanisms
[[Page 52703]]
or malfunctions that can initiate a new accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Thermal-hydraulic analysis demonstrates that the proposed
operations to be performed in the control room will ensure that the
RCS process variables remain within those values predicted for a
loss of normal a-c power, as required by 10 CFR 50, Appendix R,
Section III.L.1. The analysis demonstrates that a single spurious
operation before control of the plant is achieved through the
alternative or dedicated shutdown system will not adversely impact
the results of the analysis. After control of the plant is achieved
by the alternative or dedicated shutdown system, circuits subjected
to fire-induced circuit failures are isolated from the control
stations such that the safe shutdown operations will not be
compromised.
The need to perform the proposed operations can be readily
diagnosed and the operations can be performed in rapid succession by
control room operators at their normal control station. The actions
are straightforward and familiar to the operators. The actions have
been verified that they can be performed through demonstration. The
operations are backed up outside the control room such that
assurance exists they should not be negated by subsequent spurious
actuation signals from a postulated fire.
The automatic turbine trip action can reasonably be assumed to
occur with the credited manual reactor trip action that is part of
the current licensing basis.
Considerable defense-in-depth features exist in Fire Area 1
[control room is part of Fire Area 1] such that it is extremely
unlikely that a fire would result in evacuation of the control room.
The proposed operations are feasible and reliable and
demonstrate that the unit can be safely shutdown in the event of a
fire. The operations ensure that performance goals of Appendix R,
Section III.L.2 are met. The achievement of these goals provide
adequate margin from challenging any safety limits.
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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: Michael T. Markley.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: June 17, 2011 (TS-SQN-2011-07).
Description of amendment request: The proposed amendment would
revise the licensing basis and the Technical Specifications to permit
the use of a more robust AREVA Advanced W17 high thermal performance
(HTP) fuel at Sequoyah Nuclear Plant (SQN), Units 1 and 2. This new
fuel has been selected to address fuel assembly distortion and its
resultant fuel handling issues. The proposed AREVA Advanced W17 HTP
fuel assembly design consists of standard uranium dioxide fuel pellets
with gadolinium oxide burnable poison and M5TM cladding. The
new fuel design ensures mechanical compatibility with the existing
fuel, reactor core, control rods, steam supply system, and fuel
handling system. The transition from the existing fuel (AREVA Mark-BW)
to new fuel (AREVA Advanced W17 HTP) is planned to occur over two
refueling cycles for each SQN unit.
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 reactor fuel and the analyses associated with it are not
accident initiators. The response of the fuel to an accident is
analyzed using conservative techniques and the results are compared
to approved acceptance criteria. These evaluation results will show
that the fuel response to an accident is within approved acceptance
criteria for cores loaded with the new AREVA Advanced W17 HTP fuel
and cores loaded with both AREVA Advanced W17 HTP and AREVA Mark-BW
fuel. Therefore, the change in fuel design does not affect accident
or transient initiation or consequences.
The addition of limits on DNBR [departure from nucleate boiling
ratio] and maximum local fuel pin centerline temperature to Safety
Limit Technical Specification 2.1.1 or the proposed change to the
Safety Limit Technical Specification Figure 2.1-1 does not require
any physical change to any plant system, structure, or component.
Specifying DNBR and maximum local fuel pin centerline temperature
and the change to the CSL [core safety limit] lines are consistent
with the Standard Review Plan (SRP) for ensuring that the fuel
design limits are met. Operations and analysis will continue to be
in compliance with Nuclear Regulatory Commission (NRC) regulations.
The new CSL limits will ensure DNBR and the peak fuel centerline
temperature is maintained for protecting the fuel. The addition of
DNBR limits or fuel pin centerline temperature limits, or changes to
the CSL lines do not impact the initiation or the mitigation of an
accident.
The proposed change Technical Specification Table 2.2-1 and
Figure 3.2-1 are revised to present a new loop flow and total core
flow design limit based on the new AREVA Advanced W17 HTP fuel and
the new steam generators (now installed for SQN Unit 1 and that will
be installed concurrently with the introduction of the new Advanced
W17 HTP fuel for SQN Unit 2). Core flow is not an accident initiator
and does not play a role in accident mitigation.
The core operating limits to be developed using the new
methodologies will be established in accordance with the applicable
limitations as documented in the appropriate NRC Safety Evaluation
reports. The proposed change to add and remove various topical
reports cited in Technical Specification 6.9.1.14.a (including
adding revision numbers and revision dates to current cited topical
reports) enables the use of appropriate methodologies to re-analyze
certain events. The proposed methodologies will ensure that the
plant continues to meet applicable design criteria and safety
analysis acceptance criteria. The proposed change to the list of
NRC-approved methodologies listed in Technical Specification
6.9.1.14.a is administrative in nature and has no impact on any
plant configuration or system performance relied upon to mitigate
the consequences of an accident. The proposed change will update the
listing of NRC-approved methodologies consistent with the transition
to AREVA Advanced W17 HTP fuel. Changes to the calculated core
operating limits may only be made using NRC-approved methods, must
be consistent with all applicable safety analysis limits and are
controlled by the 10 CFR 50.59 process. The list of methodologies in
the Technical Specifications does not impact either the initiation
of an accident or the mitigation of its consequences.
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.
Use of AREVA Advanced W17 HTP fuel in the SQN, Units 1 and 2,
reactor cores does not adversely affect any fission product barrier,
nor does it alter the safety function of safety systems, structures,
or components, or their roles in accident prevention or mitigation.
The operational characteristics of AREVA Advanced W17 HTP fuel are
bounded by the safety analyses. The AREVA Advanced W17 HTP fuel
design performs within fuel design limits and does not create the
possibility of a new or different type of accident.
The addition of limits on DNBR and maximum local fuel pin
centerline
[[Page 52704]]
temperature to Safety Limit Technical Specification 2.1.1 or the
proposed change to the Safety Limit Technical Specification Figure
2.1-1 does not require any physical change to any plant system,
structure, or component. Specifying DNBR and maximum local fuel pin
centerline temperature and the change to the CSL lines are
consistent with the SRP for ensuring that the fuel design limits are
met. Operations and analysis will continue to be in compliance with
NRC regulations. The new CSL limits will ensure DNBR and the peak
fuel centerline temperature is maintained for protecting the fuel.
The addition of DNBR limits or fuel pin centerline temperature
limits, or changes to the CSL lines do not affect any accident
initiators that would create a new accident.
The proposed change Technical Specification Table 2.2-1 and
Figure 3.2-1 are revised to present a new loop flow and total core
flow design limit based on the new AREVA Advanced W17 HTP fuel and
the new steam generators (now installed for SQN, Unit 1, and that
will be installed concurrently with the introduction of the new
Advanced W17 HTP fuel for SQN, Unit 2). Core flow is not an accident
initiator and does not play a role in accident mitigation and cannot
create the possibility of a new or different kind of accident.
The proposed change to the list of topical reports used to
determine the core operating limits is administrative in nature and
has no impact on any plant configuration or on system performance.
It updates the list of NRC-approved topical reports used to develop
the core operating limits. There is no change to the parameters
within which the plant is normally operated. The possibility of a
new or different accident is not created.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Use of AREVA Advanced W17 HTP fuel does not adversely affect any
fission product barrier, nor does it alter the safety function of
safety systems, structures, or components, or their roles in
accident prevention or mitigation. The operational characteristics
of AREVA Advanced W17 HTP fuel are bounded by the safety analyses.
The AREVA Advanced W17 HTP fuel design performs within fuel design
limits. The proposed changes do not result in exceeding design basis
limits. Therefore, the licensed safety margins are maintained.
The addition of limits on DNBR and maximum local fuel pin
centerline temperature to Safety Limit Technical Specification 2.1.1
or the proposed change to the Safety Limit Technical Specification
Figure 2.1-1 does not require any physical change to any plant
system, structure, or component. Specifying DNBR and maximum local
fuel pin centerline temperature and the change to the CSL lines are
consistent with the SRP for ensuring that the fuel design limits are
met. Operations and analysis will continue to be in compliance with
NRC regulations. The new CSL limits will ensure DNBR and the peak
fuel centerline temperature is maintained for protecting the fuel.
The addition of DNBR limits or fuel pin centerline temperature
limits, or changes to the CSL lines do not impact licensed safety
margins.
The proposed change Technical Specification Table 2.2-1 and
Figure 3.2-1 are revised to present a new loop flow and total core
flow design limit based on the new AREVA Advanced W17 HTP fuel and
the new steam generators (now installed for SQN Unit 1 and that will
be installed concurrently with the introduction of the new Advanced
W17 HTP fuel for SQN Unit 2). The proposed changes to core flow are
provided to ensure licensed safety margins are maintained.
The proposed change to the list of topical reports in Technical
Specification 6.9.1.14.a does not amend the cycle specific
parameters presently required by the Technical Specifications. The
individual Technical Specifications continue to require operation of
the plant within the bounds of the limits specified in the COLR
[core operating limits report]. The proposed change to the list of
analytical methods referenced in the COLR is administrative in
nature and does not impact the margin of safety.
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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, Tennessee 37902.
NRC Branch Chief: Douglas A. Broaddus.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: February 23, 2011.
Description of amendment request: The proposed amendment would
revise the Wolf Creek Generating Station Technical Specifications (TSs)
3.3.7, ``Control Room Emergency Ventilation System (CREVS) Actuation
Instrumentation,'' 3.3.8, ``Emergency Exhaust System (EES) Actuation
Instrumentation,'' 3.7.10, ``Control Room Emergency Ventilation System
(CREVS),'' 3.7.11, ``Control Room Air Conditioning System (CRACS),''
3.7.13, ``Emergency Exhaust System (EES),'' 3.8.2, ``AC [Alternating
Current] Sources--Shutdown,'' 3.8.5, ``DC [Direct Current] Sources--
Shutdown,'' 3.8.8, ``Inverters--Shutdown,'' and 3.8.10, ``Distribution
Systems--Shutdown.'' Specifically, the proposed amendment would: (1)
Delete MODES 5 and 6 from the Limiting Condition for Operation (LCO)
Applicability for the CREVS and its actuation instrumentation (TS
3.7.10 and TS 3.3.7, respectively); (2) delete the Required Action from
TS 3.7.10 and TS 3.7.11 that requires verifying that the OPERABLE
CREVS/CRACS train is capable of being powered by an emergency power
source; (3) revise TS 3.7.13 by incorporating a 7-day Completion Time
for restoring an inoperable EES train to OPERABLE status during
shutdown conditions; (4) adopt NRC-approved Technical Specification
Task Force (TSTF) Change Traveler TSTF-36-A, Revision 4, ``Addition of
LCO 3.0.3 N/A [not applicable] to shutdown electrical power
specifications,'' for TSs 3.3.8, 3.7.13, 3.8.2, 3.8.5, 3.8.8, and
3.8.10; and (5) add a more restrictive change to the LCO Applicability
for TSs 3.8.2, 3.8.5, 3.8.8, and 3.8.10 such that these LCOs apply not
only during MODES 5 and 6, but also during the movement of irradiated
fuel assemblies regardless of the MODE in which the plant is operating.
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.
Deleting MODES 5 and 6 from the LCO Applicability of TSs 3.3.7
and 3.7.10 does not significantly increase the consequences of any
accident since it has been demonstrated that the radiological
consequences to control room occupants from a waste gas decay tank
rupture will remain much less than the regulatory limits with no
mitigation from the CREVS in MODES 5 and 6. The acceptance criteria
for this event will continue to be met.
Incorporation of a 7-day Completion Time for restoring an
inoperable EES train during shutdown conditions (i.e., during
movement of irradiated fuel assemblies in the fuel building) and the
deletion of Required Actions for verifying the availability of an
emergency power source when a CREVS/CRACS train is inoperable during
the same conditions, are operational provisions that have no impact
on the frequency of occurrence of the event for which the EES, CREVS
and CRACS are designed to mitigate. These systems have no bearing on
the occurrence of a fuel handling accident [(FHA)] as the systems
themselves are not associated with any of the potential initiating
sequences, mechanisms or occurrences--such as a failure of a lifting
device or crane, or an operator error--that could cause an FHA.
Since these systems are designed only to respond to an FHA as
accident mitigators
[[Page 52705]]
after the accident has occurred, and they have no bearing on the
occurrence of such an event themselves, the proposed changes to the
CREVS, CRACS, and EES Technical Specifications have no impact on the
probability of an accident previously evaluated.
With respect to deleting the noted Required Actions in TS 3.7.10
and TS 3.7.11 (for verifying that the OPERABLE CREVS/CRACS train is
capable of being powered from an emergency power source when one
CREVS/CRACS train is inoperable), such a change does not change the
LCO requirement for both CREVS/CRACS trains to be OPERABLE, nor to
the LCO requirements of the TS requirements pertaining to electrical
power sources/support for shutdown conditions. The change to the
Required Actions would thus not be expected to have a significant
impact on the availability of the CREVS and CRACS. That is, adequate
availability may be still assumed such that these systems would
continue to be available to provide their assumed function for
limiting the dose consequences of an FHA in accordance with the
accident analysis currently described in the [Updated Safety
Analysis Report].
With respect to the Completion Time for an inoperable EES train,
the consequences of a postulated accident are not affected by
equipment Completion Times as long as adequate equipment
availability is maintained. The proposed EES Completion Time is
based on the Completion Time specified in the Standard Technical
Specifications (STS) for which it may be presumed that the specified
Completion Time is acceptable and supports adequate EES
availability. As noted in the STS Bases, the 7-day Completion Time
for restoring an inoperable EES train takes into account the
availability of the other train. Since the STS-support Completion
Time supports adequate EES availability, it may be assumed that the
EES function would be available for mitigation of an FHA, thus
limiting offsite dose to within the currently calculated values
based on the current accident analysis. On this basis, the
consequences of applicable, analyzed accidents (i.e., the FHA) are
not increased by the proposed change.
The adoption of TSTF-36-A will not affect the equipment and LCOs
needed to mitigate the consequences of a[n] FHA in the fuel
building; however, this change will reduce the chances of an
unnecessary plant shutdown due to activities in the fuel building
that have no bearing on the operation of the rest of the plant and
the reactor core inside the containment building.
[redundant paragraph omitted]
The changes to the shutdown electrical specifications will add
an additional restriction that is consistent with the objective of
being able to mitigate a fuel handling accident during all
situations, including a full core offload, in which such an accident
could occur.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since there are
no design changes. All design, material, and construction standards
that were applicable prior to this amendment request will be
maintained. There will be no changes to any design or operating
limits.
The proposed changes will not adversely affect accident
initiators or precursors nor adversely alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated and maintained. The proposed changes will not
alter or prevent the ability of structures, systems, and components
(SSCs) from performing their intended functions to mitigate the
consequences of an initiating event within the assumed acceptance
limits.
The proposed changes do not physically alter safety-related
systems nor affect the way in which safety related systems perform
their functions. The proposed changes do not alter plant design or
operation; therefore, these changes will not increase the
probability of any accident.
All accident analysis acceptance criteria will continue to be
met with the proposed changes. The proposed changes will not affect
the source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of an
accident previously evaluated. After a postulated release from a
waste gas decay tank rupture no CREVS mitigation is required. The
applicable radiological dose criteria will continue to be met.
Therefore, the proposed changes will not increase the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
There are no proposed design changes nor are there any changes
in the method by which any safety related plant SSC performs its
specified safety function. The proposed changes 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 changes 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 as a
result of this amendment. There will be no adverse effect or
challenges imposed on any safety related system as a result of this
amendment.
The proposed amendment will not alter the design or performance
of the 7300 Process Protection System, Nuclear Instrumentation
System, or Solid State Protection System used in the plant
protection systems.
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
There will be no effect on those plant systems necessary to
assure the accomplishment of protection functions. There will be no
impact on the overpower limit, departure from nucleate boiling ratio
(DNBR) limits, heat flux hot channel factor [ ], nuclear enthalpy
rise hot channel factor [ ], loss of coolant accident peak cladding
temperature (LOCA PCT), peak local power density, or any other
margin of safety. The applicable radiological dose consequence
acceptance criteria will continue to be met. It has been
demonstrated that the CREVS and its actuation instrumentation are
not required to mitigate the control room radiological consequences
of a waste gas decay tank rupture.
The proposed changes do not eliminate any surveillances or alter
the frequency of surveillances required by the Technical
Specifications. None of the acceptance criteria for any accident
analysis will be changed.
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: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: April 22, 2011.
Description of amendment request: The proposed amendment would
revise the Wolf Creek Generating Station Technical Specification (TS)
5.3, ``Unit Staff Qualifications,'' by making two administrative
changes to TS 5.3.1.1. Specifically, these changes will remove the
operator license applicants' education and experience eligibility
requirements, and correct inadvertent omissions in previous amendments
relative to the Licensed Operators' and Senior Operators' qualification
requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change is an administrative change to reinstate the
qualification requirements for Licensed Operators and Senior
Licensed Operators that were inadvertently eliminated through the
issuance of Amendment No. 150 [issued
[[Page 52706]]
November 26, 2002] and Amendment No. 159 [issued January 31, 2005],
and to remove an unnecessary reference to a [National Academy for
Nuclear Training] NANT guideline. The proposed change does not
directly impact accidents previously evaluated. [Wolf Creek Nuclear
Operating Company's (WCNOC's)] licensed operator training program is
accredited by the NANT and is based on a systems approach to
training consistent with the requirements of 10 CFR Part 55.
Although licensed operator qualifications and training may have an
indirect impact on accidents previously evaluated, the NRC
considered this impact during the rulemaking process, and by
promulgation of the revised 10 CFR Part 55 rule, concluded that this
impact remains acceptable as long as the licensed operator training
program is certified to be accredited and is based on a systems
approach to training.
Therefore, the proposed change will not increase 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 is an administrative change to reinstate the
qualification requirements for Licensed Operators and Senior
Licensed Operators that were inadvertently eliminated through the
issuance of Amendment No. 150 and Amendment No. 159, and to remove
an unnecessary reference to a NANT guideline. WCNOC's licensed
operator training program is accredited by the National Academy for
Nuclear Training and is based on a systems approach to training
consistent with the requirements of 10 CFR Part 55. Although
licensed operator qualifications and training may have an indirect
impact on accidents previously evaluated, the NRC considered this
impact during the rulemaking process, and by promulgation of the
revised 10 CFR Part 55 rule, concluded that this impact remains
acceptable as long as the licensed operator training program is
certified to be accredited and is based on a systems approach to
training.
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 is an administrative change to reinstate the
qualification requirements for Licensed Operators and Senior
Licensed Operators that were inadvertently eliminated through the
issuance of Amendment No. 150 and Amendment No. 159, and to remove
an unnecessary reference to a NANT guideline. As noted previously,
WCNOC's licensed operator training program is accredited and is
based on a systems approach to training consistent with the
requirements of 10 CFR Part 55. Licensed operator qualifications and
training can have an indirect impact on the margin of safety.
However, the NRC considered this impact during the rulemaking
process, and by promulgation of the revised 10 CFR Part 55 rule,
determined that this impact remains acceptable when licensees
maintain a licensed operator training program that is accredited and
based on a systems approach to training.
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: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
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' 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
PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail 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: July 22, 2010, as supplemented
by letter dated April 8, 2011.
Brief description of amendment: The amendment revised an element of
the methodology used in evaluating the radiological consequences of
design basis steam generator tube rupture (SGTR) accidents.
Specifically, the amendment revised the Palo Verde Nuclear Generating
Station (PVNGS) Updated Final Safety Analysis Report Section 15.6.6,
``Steam Generator Tube Rupture,'' to reflect a lower iodine spiking
factor assumed for the coincident event Generated Iodine Spike (GIS)
and the resulting reduction in the radiological consequences for the
Limiting SGTRLOPSF [Steam Generator Tube Rupture with Loss of Offsite
Power and Single Failure] Event.
Date of issuance: July 28, 2011.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: Unit 1--186; Unit 2--186; Unit 3--186.
Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74:
The amendment revised the Operating Licenses and the Updated Final
Safety Analysis Report.
Date of initial notice in Federal Register: December 28, 2010 (75
FR 81669).
The supplemental letter dated April 8, 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.
[[Page 52707]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 28, 2011.
No significant hazards consideration comments received: No.
Carolina Power and Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Unit 1 and 2, Brunswick County, North
Carolina
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit 2, Darlington County, South Carolina
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant Citrus County, Florida
Date of application for amendments: July 8, 2010, as supplemented
by letters dated September 23 and November 30, 2010; February 28 and
April 7, 2011.
Brief description of amendments: The amendments establish a fleet
Cyber Security Plan (CSP) in accordance with Title 10 of the Code of
Federal Regulations (10 CFR), Section 73.54, ``Protection of digital
computer and communication systems and networks,'' and in conformance
with the model CSP contained in Appendix A of Nuclear Energy Institute
(NEI) document NEI 08-09, ``Cyber Security Plan for Nuclear Power
Reactors,'' Revision 6, dated April 2010. The licensees' submittals
included the fleet CSP for Brunswick Steam Electric Plant, Units 1 and
2, H. B. Robinson Steam Electric Plant, Unit No. 2, Shearon Harris
Nuclear Power Plant, Unit 1, and Crystal River Unit 3 Nuclear
Generating Plant, the licensees' proposed changes to the facility
operating licenses, and a proposed CSP implementation schedule for each
facility.
The licensees' submittals dated November 30, 2010, and April 7,
2011, supplemented the licensees' CSP to address: (1) Scope of systems
in response to the October 21, 2010, the Nuclear Regulatory Commission
(NRC, Commission) decision; (2) records retention; and (3)
implementation schedule. The licensee provided, in its letter dated
April 7, 2011, a revised copy of the Carolina Power & Light Company and
Florida Power Corporation, Cyber Security Plan, Revision 0 that
incorporated all of the changes that the licensee had made to the
following sections of their CSP: Scope and purpose, defense-in-depth
protective strategies, document control and records retention and
handling, and deviations from NEI 08-09, Revision 6.
Date of issuance: July 29, 2011.
Effective date: The license amendments are effective as of the date
of their issuance. The implementation of the CSP, including the key
intermediate milestone dates and the full implementation date, shall be
in accordance with the implementation schedule submitted by the
licensees on April 7, 2011, and approved by the NRC staff with the
license amendments. All subsequent changes to the NRC-approved CSP
implementation schedule will require prior NRC approval pursuant to 10
CFR 50.90.
Amendment Nos.: Brunswick 1: 258, Brunswick 2: 286, Robinson 2:
226, Shearon Harris 1: 136, and Crystal River 3: 238.
Renewed Facility Operating License Nos. DPR-71, DPR-62, DPR-23, and
NPF-63; and Facility Operating License No. DPR-72.: Amendments changed
the facility operating licenses.
Date of initial notice in Federal Register: October 12, 2010 (75 FR
62595).
The supplements dated September 23 and November 30, 2010; February
28, 2011, and the Updated No Significant Hazards Consideration in
Enclosure 5 of the letter dated April 7, 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 July 29, 2011.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: July 27, 2010, as supplemented
by letters dated September 29, 2010, November 22, 2010, and March 30,
2011.
Brief description of amendment: The amendment approves the cyber
security plan and associated implementation schedule, and revises
Paragraph 2.E of Facility Operating License No. NPF-43 for Fermi 2, to
provide a license condition to require the licensee to fully implement
and maintain in effect all provisions of the NRC-approved Cyber
Security Plan. The proposed change is consistent with Nuclear Energy
Institute (NEI) 08-09, Revision 6, Cyber Security Plan for Nuclear
Power Reactors.
Date of issuance: July 28, 2011.
Effective date: This license amendment is effective as of the date
of its issuance. The implementation of the CSP, including the key
intermediate milestone dates and the full implementation date, shall be
in accordance with the implementation schedule submitted by the
licensee on July 27, 2010, as supplemented by letters dated September
29, 2010, November 22, 2010, and March 30, 2011, and approved by the
NRC staff with this license amendment. All subsequent changes to the
NRC-approved CSP implementation schedule will require prior NRC
approval pursuant to 10 CFR 50.90.
Amendment No.: 185.
Facility Operating License No. NPF-43: Amendment revised the
License.
Date of initial notice in Federal Register: December 7, 2010 (75 FR
76043).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original
application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 27, 2011.
No significant hazards consideration comments received: No.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of application for amendment: June 1, 2010, as supplemented by
letters dated January 18, 2011, March 14, 2011, and June 27, 2011.
Brief description of amendment: The amendment revised the Kewaunee
licensing basis, approving the licensee to operate the load tap
changers (LTCs) on two new transformers in the automatic mode. The LTCs
are designed to compensate for potential offsite power voltage
variations and will provide added assurance that acceptable voltage is
maintained for safety-related equipment.
Date of issuance: July 29, 2011.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 209.
Renewed Facility Operating License No. DPR-43: Amendment did not
revise the Technical Specifications.
Date of initial notice in Federal Register: August 10, 2010 (75 FR
48374).
The supplements dated January 18, 2011, March 14, 2011, and June
27, 2011, provided additional information that clarified the
application, did not expand the scope of the application, and
[[Page 52708]]
did not change the Commission's proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 29, 2011.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone
Power Station, Unit 3, New London County, Connecticut
Date of amendment request: July 21, 2010.
Description of amendment request: The amendment relocates Millstone
Power Station, Unit No. 3 (MPS3) Technical Specification (TS) 3/4.7.14,
``Area Temperature Monitoring,'' and the associated Table 3.7-6, ``Area
Temperature Monitoring,'' to the MPS3 Technical Requirements Manual.
Date of issuance: July 27, 2011.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 250.
Renewed Facility Operating License No. NPF-49: The amendment
revised the License and Technical Specifications.
Date of initial notice in Federal Register: March 22, 2011 (76 FR
16007).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 27, 2011.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of application of amendments: July 14, 2010.
Brief description of amendments: The amendments revised the
Technical Specifications related to the adoption of technical
specification task force technical change Traveler 52, Revision 3, to
implement option B of Appendix J to Title 10 of the Code of Federal
Regulations (10 CFR), part 50.
Date of Issuance: July 28, 2011.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 375, 377, and 376.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: December 14, 2010 (75
FR 77909).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 28, 2011.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of application of amendments: June 10, 2009, as supplemented
by letters dated December 18, 2009, and August 25, 2010.
Brief description of amendments: The amendments change the
Technical Specifications (TSs) and authorize changes to the ``Updated
Final Safety Analysis Report'' (UFSAR) to allow the use of CASMO-4/
SIMULATE-3 methodology for application to reactor core designs
containing low enrichment uranium fuel bearing lumped burnable and/or
gadolinia integral absorbers.
Date of Issuance: August 2, 2011.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 377, 379, and 378.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the licenses and the TSs and authorized UFSAR
changes.
Date of initial notice in Federal Register: March 19, 2010 (75 FR
13314).
The supplements dated December 15, 2009, and August 25, 2010,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 2, 2011.
No significant hazards consideration comments received: No.
Duke Power Company, LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Duke Power Company, LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Duke Power Company, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: July 28, 2010, as supplemented
March 3, 2011.
Brief description of amendments: The amendments approve changes to
each station emergency plans to allow changes to the minimum staffing
requirement during emergencies.
Date of Issuance: July 29, 2011.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: Catawba 1 and 2-265/261.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the licenses and emergency plan.
Amendment Nos.: McGuire 1 and 2--263/243
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and emergency plan.
Amendment Nos. Oconee 1, 2 and 3--376/378/377
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the licenses and emergency plan.
Date of initial notice in Federal Register: September 7, 2010 (75
FR 54393).
The supplement dated March 3, 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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 29, 2011.
No significant hazards consideration comments received: No.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: July 22, 2010, as supplemented by
letters dated September 23 and November 30, 2010, and February 15 and
April 4, 2011.
Brief description of amendment: The amendment approved the cyber
security plan (CSP) and associated implementation schedule, and added
new Paragraph 2.E to Facility Operating License No. NPF-47 to provide a
license condition to require the licensee to fully implement and
maintain in effect all provisions of the NRC-approved Cyber Security
Plan. The proposed change is generally consistent with Nuclear Energy
Institute (NEI) 08-09, Revision 6, ``Cyber Security Plan for Nuclear
Power Reactors.''
Date of issuance: July 29, 2011.
Effective date: This license amendment is effective as of the date
of its issuance. The implementation of the CSP, including the key
intermediate milestone dates and the full implementation date, shall be
in accordance with the implementation schedule submitted by the
licensee on
[[Page 52709]]
April 4, 2011, and approved by the NRC staff with this license
amendment. All subsequent changes to the NRC-approved CSP
implementation schedule will require prior NRC approval pursuant to 10
CFR 50.90.
Amendment No.: 171.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: October 12, 2010 (75 FR
62596).
The supplemental letters dated September 23 and November 30, 2010,
and February 15 and April 4, 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 amendment is contained
in a Safety Evaluation dated July 29, 2011.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247, and 50-
286, Indian Point Nuclear Generating Unit 1, 2 and 3, (IP1, IP2, and
IP3), Westchester County, New York
Date of application for amendment: July 8, 2010, as supplemented by
letters dated February 18, April 1, and June 29, 2011.
Brief description of amendment: The licensee's application for the
proposed amendments to the Facility Operating Licenses (FOLs) includes:
(1) The proposed Cyber Security Plan (CSP), (2) an implementation
schedule, and (3) a proposed statement to be added to the existing FOL
Physical Protection license conditions requiring Entergy to fully
implement and maintain in effect all provisions of the Commission-
approved CSP as required by 10 CFR 73.54, ``Protection of digital
computer and communication systems and networks.'' A Federal Register
notice dated March 27, 2009, issued the final rule that amended 10 CFR
Part 73. The regulations in 10 CFR 73.54, establish the requirements
for a CSP. This regulation specifically requires each licensee
currently licensed to operate a nuclear power plant under Part 50 of
this chapter to submit a CSP that satisfies the requirements of the
Rule. Each submittal must include a proposed implementation schedule,
and implementation of the licensee's CSP must be consistent with the
approved schedule. The background for this application is addressed by
the NRC Notice of Availability, Federal Register Notice, Final Rule 10
CFR Part 73, Power Reactor Security Requirements, published on March
27, 2009 (74 FR 13926).
Date of issuance: August 2, 2011.
Effective date: These license amendments are effective as of the
date of their issuance. The implementation of the CSP, including the
key intermediate milestone dates and the full implementation date,
shall be in accordance with the implementation schedule submitted by
the licensee on July 8, 2010, as supplemented by letters dated February
18, April 1, and June 29, 2011, and approved by the NRC staff with
these license amendments. All subsequent changes to the NRC-approved
CSP implementation schedule will require prior NRC approval pursuant to
10 CFR 50.90.
Amendment Nos.: 55 for IP1, 266 for IP2, and 243 for IP3,
respectively.
Facility Operating License Nos. DPR-5, DPR-26, and DPR-64: The
amendment revised the Licenses.
Date of initial notice in Federal Register: October 12, 2010 (75 FR
62596).
The supplements dated February 18, April 1, and June 29, 2011,
provided additional information that clarified the application, did not
expand 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 August 2, 2011.
No significant hazards consideration comments received: Yes. The
Safety Evaluation dated August 2, 2011, provides the discussion of the
comments received from New York State.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of application for amendment: July 26, 2011, supplemented by
letters dated September 27, 2010, November 30, 2010, February 15, 2011,
and April 4, 2011.
Brief description of amendment: The amendment approves the cyber
security plan and associated implementation schedule, and revises
Paragraph 2.E of Facility Operating License No. DPR-20 for Palisades
Nuclear Plant, to provide a license condition to require the licensee
to fully implement and maintain in effect all provisions of the NRC-
approved Cyber Security Plan. The proposed change is generally
consistent with Nuclear Energy Institute (NEI) 08-09, Revision 6, Cyber
Security Plan for Nuclear Power Reactors.
Date of issuance: July 27, 2011.
Effective date: This license amendment is effective as of the date
of its issuance. The implementation of the CSP, including the key
intermediate milestone dates and the full implementation date, shall be
in accordance with the implementation schedule submitted by the
licensee on July 26, 2010, as supplemented by letters dated September
27, 2010, November 30, 2010, February 15, 2011, and April 4, 2011, and
approved by the NRC staff with this license amendment. All subsequent
changes to the NRC-approved CSP implementation schedule will require
prior NRC approval pursuant to 10 CFR 50.90.
Amendment No.: 243.
Facility Operating License No. DPR-20: Amendment revised the
Renewed Facility Operating License.
Date of initial notice in Federal Register: December 7, 2010 (75 FR
76044).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original
application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 27, 2011.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
1, Pope County, Arkansas
Date of amendment request: August 10, 2010, as supplemented by
letter dated June 10, 2011.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.9.3, ``Reactor Building Penetrations,'' to allow
reactor building flow path(s) providing direct access from the reactor
building atmosphere to the outside atmosphere to be unisolated under
administrative control, during movement of irradiated fuel assemblies.
The proposed change is consistent with Technical Specification Task
Force (TSTF) Technical Change Traveler TSTF-312, Revision 1,
``Administratively Control Containment Penetrations.''
Date of issuance: August 10, 2011.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 245.
Renewed Facility Operating License No. DPR-51: Amendment revised
the Technical Specifications/license.
Date of initial notice in Federal Register: October 5, 2010 (75 FR
61526).
[[Page 52710]]
The supplemental letter dated June 10, 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 amendment is contained
in a Safety Evaluation dated August 10, 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: March 19, 2010, as supplemented
by letters dated June 16, 2010, October 29, 2010, December 3, 2010,
January 14, 2011, and March 23, 2011.
Brief description of amendment: The changes implement an extension
of the Technical Specification (TS) allowed outage time (AOT) for the
Unit 1 and Unit 2 Suppression Pool Cooling (SPC) mode of the Residual
Heat Removal (RHR) system, the Residual Heat Removal Service Water
(RHRSW) system, the Emergency Service Water (ESW) system, and the A.C.
Sources--Operating (Emergency Diesel Generators) from 72 hours to seven
(7) days in order to allow for repairs of the RHRSW system piping. The
AOT extension would only be allowed once every other calendar year, for
each unit, with the opposite unit shutdown, reactor vessel head
removed, reactor cavity flooded, and certain other specific
compensatory measures, in effect.
Date of issuance: July 29, 2011.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 203 and 165.
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: May 18, 2010 (75 FR
27828).
The supplements dated June 16, 2010, October 29, 2010, December 3,
2010, January 14, 2011, and March 23, 2011, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed and did not change the NRC
staff's original proposed no significant hazards determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 29, 2011.
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
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit 1 and 2, Ogle County, Illinois
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit 1, DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket No. 50-352 and No. 50-353,
Limerick Generating Station, Unit 1 and 2, Montgomery County,
Pennsylvania
Exelon Generation Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendments: November 23, 2009, as
supplemented by letters dated July 23, September 24, November 18,
December 21, 2010, March 31, May 19, and July 11, 2011.
Brief description of amendments: The amendments were submitted in
accordance with the provisions of Title 10 of the Code of Federal
Regulations (10 CFR) 50.4 and 10 CFR 50.90 and requests NRC approval of
the Exelon Generation Company, LLC (Exelon) Cyber Security Plan (CSP),
provides an Implementation Schedule, and adds a sentence to the
existing Physical Protection license condition to require Exelon to
fully implement and maintain in effect all provisions of the Commission
approved CSP.
Date of issuance: August 10, 2011.
Effective date: These license amendments are effective as of the
date of their issuance. The implementation of the CSP, including the
key intermediate milestone dates and the full implementation date,
shall be in accordance with the implementation schedule submitted by
the licensee on November 23, 2009 as supplemented by letters dated July
23, September 24, November 18, December 21, 2010, March 31, May 19, and
July 11, 2011, and approved by the NRC staff with these license
amendments. All subsequent changes to the NRC-approved CSP
implementation schedule will require prior NRC approval pursuant to 10
CFR 50.90.
Amendment Nos.: 168, 168, 175, 175, 194, 238, 231, 203, 190, 204,
166, 280, 281, 283, 249, 244, 275.
Facility Operating License Nos. NPF-72, NPF-77, NPF-37, NPF-66,
NPF-62, DPR-19, DPR-25, NPF-11, NPF-18, NPF-39, NPF-85, DPR-16, DPR-44,
DPR-56, DPR-29, DPR-30, DPR-50: The amendments revised the Licenses.
Date of initial notice in Federal Register: April 12, 2011 (75 FR
20379).
The July 23, September 24, November 18, December 21, 2010, March
31, May 19, and July 11, 2011, supplements contained clarifying
information 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 August 10, 2011.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of application for amendments: July 22, 2010, as supplemented
by letters dated September 28, 2010, November 29, 2010, February 3,
2011, and April 6, 2011.
Brief description of amendments: The amendments to the Renewed
Facility Operating Licenses (FOL) include: (1) The proposed BVPS-1 and
2 Cyber Security Plan (CSP), (2) an implementation schedule, and (3) a
proposed sentence to be added to the existing renewed FOL Physical
Protection license condition for BVPS-1 and 2 requiring FirstEnergy
Nuclear Operating Company to fully implement and maintain in effect all
provisions of the Commission-approved BVPS-1 and 2 CSP as required by
Title 10 of the Code of Federal Regulations (10 CFR) 73.54,
``Protection of digital computer
[[Page 52711]]
and communication systems and networks.'' A Federal Register notice
dated March 27, 2009, issued the final rule that amended 10 CFR Part
73. The regulations in 10 CFR 73.54, establish the requirements for a
CSP. This regulation specifically requires each licensee currently
licensed to operate a nuclear power plant under part 50 of this chapter
to submit a CSP that satisfies the requirements of the Rule. Each
submittal must include a proposed implementation schedule and
implementation of the licensee's CSP must be consistent with the
approved schedule. The background for this application is addressed by
the NRC Notice of Availability, Federal Register Notice, Final Rule, 10
CFR Part 73, Power Reactor Security Requirements, published on March
27, 2009 (74 FR 13926).
Date of issuance: July 28, 2011.
Effective date: These license amendments are effective as of the
date of its issuance. The implementation of the CSP, including the key
intermediate milestone dates and the full implementation date, shall be
in accordance with the implementation schedule submitted by the
licensee on July 22, 2010, as supplemented by letters dated September
28, 2010, November 29, 2010, February 3, 2011, and April 6, 2011, and
approved by the Nuclear Regulatory Commission (NRC) staff with this
license amendment. All subsequent changes to the NRC-approved CSP
implementation schedule will require prior NRC approval pursuant to 10
CFR 50.90.
Amendment Nos.: 287 for BVPS-1 and 174 for BVPS-2.
Facility Operating License Nos. DPR-66 and NPF-73: The amendments
revised the License.
Date of initial notice in Federal Register: October 12, 2010, 75 FR
62599.
The supplements dated September 28, 2010, November 29, 2010,
February 3, 2011, and April 6, 2011, provided additional information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the NRC 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 July 28, 2011.
No significant hazards consideration comments received: No.
Florida Power and Light Company (FPL), Docket Nos. 50-250 and 50-251,
Turkey Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: July 28, 2010, as supplemented
by letters dated September 27 and November 19, 2010, and April 5 and
June 30, 2011.
Brief description of amendments: The amendment includes three
parts: The proposed plan, an implementation schedule, and a sentence
added to the existing Physical Protection license condition to require
FPL to fully implement and maintain in effect all provisions of the
Commission approved cyber security plan (CSP) as required by amended
Title 10 of the Code of Federal Regulations (10 CFR) part 73. The
proposed CSP was submitted in accordance with 10 CFR 73.54,
``Protection of digital computer and communication systems and
networks.''
Date of issuance: July 29, 2011.
Effective date: These license amendments are effective as of the
date of their issuance. The implementation of the CSP, including the
key intermediate milestone dates and the full implementation date,
shall be in accordance with the implementation schedule submitted by
the licensee on July 28, 2010, as supplemented by letters dated
September 27 and November 19, 2010, and April 5 and June 30, 2011, and
approved by the NRC staff with these license amendments. All subsequent
changes to the NRC-approved CSP implementation schedule will require
prior NRC approval pursuant to 10 CFR 50.90.
Amendment Nos: Unit 3--245 and Unit 4--241.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the licenses.
Date of initial notice in Federal Register: December 7, 2010 (75 FR
76045).
The supplements dated September 27 and November 19, 2010, and April
5 and June 30, 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 July 29, 2011.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company (IandM), Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendment: July 19, 2010, as supplemented
by letters dated September 28, 2010, November 30, 2010, and April 8,
2011.
Brief description of amendment: The amendments approve the Cyber
Security Plan and associated implementation schedule, and revises
License Condition 2.D of the Renewed Facility Operating Licenses for
Units 1 and 2. The amendments specify that the licensee fully implement
and maintain in effect all provisions of the Commission approved CSP as
required by 10 CFR 73.54.
Date of issuance: July 28, 2011.
Effective date: These license amendments are effective as of the
date of issuance. The implementation of the CSP, including the key
intermediate milestone dates and the full implementation date, shall be
in accordance with the implementation schedule submitted by the
licensee on April 8, 2011, and approved by the NRC staff with these
license amendments. All subsequent changes to the NRC-approved CSP
implementation schedule will require prior NRC approval pursuant to 10
CFR 50.90.
Amendment Nos.: 315 (for Unit 1) and 299 (for Unit 2).
Facility Operating License No. DPR-74: Amendments revised the
Renewed Facility Operating Licenses.
Date of initial notice in Federal Register: October 12, 2010 (75 FR
62600).
The supplemental letters contain clarifying information, did not
change the scope of the license amendment request, did not change the
NRC staff's initial proposed finding of no significant hazards
consideration determination, and did not expand the scope of the
original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 28, 2011.
No significant hazards consideration comments received: No.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of application for amendment: July 14, 2010, as supplemented
by letters dated September 27, 2010, November 17, 2010, April 5, 2011,
and June 22, 2011.
Brief description of amendment: The amendment approves the Cyber
Security Plan and associated implementation schedule, and revises
License Condition
[[Page 52712]]
2.C.(5) of the Renewed Facility Operating License. The amendment
specifies that the licensee fully implement and maintain in effect all
provisions of the Commission approved CSP, as required by 10 CFR 73.54.
Date of issuance: July 29, 2011.
Effective date: This license amendment is effective as of the date
of its issuance. The implementation of the CSP, including the key
intermediate milestone dates and the full implementation date, shall be
in accordance with the implementation schedule submitted by the
licensee on April 5, 2011, and approved by the NRC staff with this
license amendment. All subsequent changes to the NRC-approved CSP
implementation schedule will require prior NRC approval pursuant to 10
CFR 50.90.
Amendment No.: 278.
Renewed Facility Operating License No. DPR-49: The amendment
revised the Renewed Facility Operating License.
Date of initial notice in Federal Register: November 9, 2010 (75 FR
68836).
The supplemental letters contain clarifying information, did not
change the scope of the license amendment request, did not change the
NRC staff's initial proposed finding of no significant hazards
consideration determination, and did not expand the scope of the
original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 29, 2011.
No significant hazards consideration comments received: No.
NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: January 27, 2010, as
supplemented by letters dated August 30, 2010, and May 3, 2011.
Brief description of amendments: The amendments revise Technical
Specification 3.8.3, ``Diesel Fuel Oil and Starting Air,'' to specify
an increased minimum diesel fuel oil storage volume and associated
surveillance requirement for the Emergency Diesel Generators.
Date of issuance: August 4, 2011.
Effective date: This license amendment is effective as of the date
of issuance and shall be implemented within 60 days of the date of
issuance.
Amendment Nos.: 244 (for Unit 1) and 248 (for Unit 2).
Renewed Facility Operating License Nos. DPR-24 and DPR-27:
Amendments revised the Technical Specifications and Renewed Facility
Operating License.
Date of initial notice in Federal Register: November 30, 2010 (75
FR 74096).
The August 30, 2010, and May 3, 2011, supplements 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 August 4, 2011.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota (NSPM), Docket No. 50-263,
Monticello Nuclear Generating Plant, Wright County, Minnesota
Date of application for amendment: July 20, 2010, and supplemented
by letters dated September 24, 2010, November 30, 2010, February 21,
2010, April 1, 2011, and May 26, 2011.
Brief description of amendment: The amendment approves the Cyber
Security Plan (CSP) and associated implementation schedule, and revises
License Condition 2.C.3 of the Renewed Facility Operating License DPR-
22 for Monticello Nuclear Generating Plant. The amendment specifies
that the licensee fully implement and maintain in effect all provisions
of the Commission approved CSP as required by 10 CFR 73.54.
Date of issuance: August 2, 2011.
Effective date: This license amendment is effective as of the date
of its issuance. The implementation of the CSP, including the key
intermediate milestone dates and the full implementation date, shall be
in accordance with the implementation schedule submitted by the
licensee on April 1, 2011, and approved by the NRC staff with this
license amendment. All subsequent changes to the NRC-approved CSP
implementation schedule will require prior NRC approval pursuant to 10
CFR 50.90.
Amendment No.: 166.
Facility Operating License No. DPR-22. Amendment revised the
Renewed Facility Operating License.
Date of initial notice in Federal Register: October 12, 2010 (75 FR
62604).
The licensee's supplemental letters contained clarifying
information, did not change the scope of the original license amendment
request, did not change the NRC staff's initial proposed finding of no
significant hazards consideration determination, and did not expand the
scope of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 2, 2011.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue
County, Minnesota
Date of application for amendment: July 20, 2010, and supplemented
by letters dated September 24, 2010, November 30, 2010, February 21,
2011, April 1, 2011, and May 26, 2011.
Brief description of amendment: The amendments approve the Cyber
Security Plan (CSP) and associated implementation schedule, and revise
License Condition 2.C.(3) of the Facility Operating Licenses for each
unit at Prairie Island Nuclear Generating Plant. The amendments specify
that the licensee fully implement and maintain in effect all provisions
of the Commission-approved CSP as required by 10 CFR 73.54.
Date of issuance: July 29, 2011.
Effective date: These license amendments are effective as of the
date of their issuance. The implementation of the CSP, including the
key intermediate milestone dates and the full implementation date,
shall be in accordance with the implementation schedule submitted by
the licensee on April 1, 2011, and approved by the NRC staff with these
license amendments. All subsequent changes to the NRC-approved CSP
implementation schedule will require prior NRC approval pursuant to 10
CFR 50.90.
Amendment Nos.: 202 (for Unit 1) and 189 (for Unit 2).
Facility Operating License Nos. DPR-42 and DPR-60. Amendments
revised the Facility Operating Licenses
Date of initial notice in Federal Register: October 12, 2010 (75 FR
62604).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 29, 2011.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-354, 50-272, and 50-311, Hope Creek
Generating Station and Salem Nuclear Generating Station, Unit Nos. 1
and 2, Salem County, New Jersey
Date of application for amendments: July 14, 2010, as supplemented
by letters dated September 28, 2010, April 1, 2011, June 6, 2011, and
July 6, 2011.
Brief description of amendments: The amendments approve the Cyber
Security Plan (CSP) and associated implementation schedule for Hope
Creek Generating Station and Salem Nuclear Generating Station, Unit
Nos. 1
[[Page 52713]]
and 2. In addition, the amendments revise the existing license
condition regarding physical protection in the each of the three
facility operating licenses (FOLs) to require the licensee to fully
implement and maintain in effect all provisions of the Nuclear
Regulatory Commission (NRC)-approved CSP. The amendment was submitted
pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 73.54,
which requires licensees currently licensed to operate a nuclear power
plant under 10 CFR part 50 to submit a CSP for NRC review and approval.
Date of issuance: July 28, 2011.
Effective date: The license amendments are effective as of the date
of issuance. The implementation of the CSP, including the key
intermediate milestone dates and the full implementation date, shall be
in accordance with the implementation schedule submitted by the
licensee by letter dated June 6, 2011, and approved by the NRC staff
with these license amendments. All subsequent changes to the NRC-
approved CSP implementation schedule will require prior NRC approval
pursuant to 10 CFR 50.90.
Amendment Nos.: 189, 300 and 283.
Facility Operating License Nos. NPF-57, DPR-70 and DPR-75: The
amendments revised the FOLs.
Date of initial notice in Federal Register: October 12, 2010 (75 FR
62606).
The letters dated September 28, 2010, April 1, 2011, June 6, 2011,
and July 6, 2011, provided clarifying information that did not change
the initial proposed no significant hazards consideration determination
or expand the application beyond the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 28, 2011.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: July 22, 2010, as supplemented
by letters dated September 29 and November 30, 2010, and March 31 and
June 16, 2011.
Brief description of amendments: The amendments approved the cyber
security plan (CSP) and associated implementation schedule, and revised
Paragraph 2.E of Facility Operating License Nos. NPF-10 and NPF-15,
respectively, for San Onofre Nuclear Generating Station, Units 2 and 3,
to provide a license condition to require the licensee to fully
implement and maintain in effect all provisions of the NRC-approved
Cyber Security Plan. The proposed change is consistent with Nuclear
Energy Institute (NEI) 08-09, Revision 6, ``Cyber Security Plan for
Nuclear Power Reactors.''
Date of issuance: July 28, 2011.
Effective date: These license amendments are effective as of the
date of issuance. The implementation of the CSP, including the key
intermediate milestone dates and the full implementation date, shall be
in accordance with the implementation schedule submitted by the
licensee on March 31 and June 16, 2011, and approved by the NRC staff
with these license amendments. All subsequent changes to the NRC-
approved CSP implementation schedule will require prior NRC approval
pursuant to 10 CFR 50.90.
Amendment Nos.: Unit 2--225; Unit 3--218.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: November 9, 2010 (75 FR
68836).
The supplemental letters dated September 29 and November 30, 2010,
and March 31 and June 16, 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 July 28, 2011.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Joseph M. Farley Nuclear
Plant, Units 1 and 2, Docket Nos. 50-348 and 50-364, Houston County,
Alabama; Edwin I. Hatch Nuclear Plant, Units 1 and 2, Docket Nos. 50-
321 and 50-366, Appling County, Georgia; Vogtle Electric Generating
Plant, Units 1 and 2, Docket Nos. 50-424 and 50-425, Burke County,
Georgia
Date of amendment request: July 16, 2010, as supplemented March 28
and April 11, 2011.
Brief description of amendment request: The amendments approve the
licensee's Cyber Security Plan and Implementation Schedule.
Date of issuance: July 28, 2011.
Effective date: These license amendments are effective as of the
date of their issuance. The implementation of the cyber security plan
(CSP), including key intermediate milestone dates and the full
implementation date, shall be in accordance with the implementation
schedule submitted by the licensee by letter dated April 11, 2011, and
approved by the NRC staff with these license amendments. All subsequent
changes to the NRC-approved CSP implementation schedule will require
prior NRC approval pursuant to 10 CFR 50.90.
Amendment Nos: Farley 1 and 2--186/181; Hatch 1 and 2--265/209;
Vogtle 1 and 2--162/144.
Facility Operating License (Farley) NPF-2 and NPF-8; (Hatch) DPR-57
and NPF-5; (Vogtle) NPF-68 and NPF-81: The amendments changed the
licenses and the technical specifications.
Date of initial notice in Federal Register: April 12, 2011 (76 FR
20381)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 28, 2011.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: November 23, 2009, as
supplemented on December 18, 2009; July 23 and October 1, 2010; April 7
and July 15, 2011 (TS-470).
Description of amendment request: On March 27, 2009, the Federal
Register Notice (74 FR 13926) published the final rule that amended
Title 10 of the Code of Federal Regulations (10 CFR) Part 73,
``Physical Protection of Plants and Materials.'' Specifically, the
regulations in 10 CFR 73.54 ``Protection of Digital Computer and
Communication Systems and Networks,'' establish the requirements for a
cyber security program to protect digital computer and communication
systems and networks against cyber attacks. The proposed amendment
included the proposed Cyber Security Plan, its implementation schedule,
and a revised Physical Protection license condition for Browns Ferry
Nuclear Plant, Units 1, 2, and 3 to fully implement and maintain in
effect all provisions of the Nuclear Regulatory Commission approved
Cyber Security Plan as required by 10 CFR 73.54.
Date of issuance: July 29, 2011.
Effective date: This license amendment is effective as of the date
of issuance. The implementation of the cyber security plan (CSP),
including the key intermediate milestone dates and the full
implementation date, shall be in accordance with the implementation
schedule submitted by the licensee on
[[Page 52714]]
April 7, 2011, and approved by the NRC staff with this license
amendment. All subsequent changes to the NRC-approved CSP
implementation schedule will require prior NRC approval pursuant to 10
CFR 50.90.
Amendment Nos.: Unit 1--279, Unit 2--306, and Unit 3--265.
Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the licenses.
Date of initial notice in Federal Register: December 7, 2010 (75 FR
76046).
The above Federal Register notice was based on the supplement dated
December 18, 2009. The supplements dated July 23 and October 1, 2010;
April 7 and July 15, 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.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated July 29, 2011.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendment: November 23, 2009, as
supplemented on December 11 and December 18, 2009; July 23 and October
1, 2010; April 7 and July 15, 2011 (TS 09-06).
Brief description of amendment: On March 27, 2009, the Federal
Register Notice (74 FR 13926) published the final rule that amended
Title 10 of the Code of Federal Regulations (10 CFR) part 73,
``Physical Protection of Plants and Materials.'' Specifically, the
regulations in 10 CFR 73.54 ``Protection of Digital Computer and
Communication Systems and Networks,'' establish the requirements for a
cyber security program to protect digital computer and communication
systems and networks against cyber attacks. The proposed amendment
included the proposed Cyber Security Plan, its implementation schedule,
and a revised physical protection license condition for Sequoyah
Nuclear Plant, Units 1 and 2 to fully implement and maintain in effect
all provisions of the Nuclear Regulatory Commission approved Cyber
Security Plan as required by 10 CFR 73.54.
Date of issuance: July 29, 2011.
Effective date: This license amendment is effective as of the date
of issuance. The implementation of the Cyber Security Plan (CSP),
including the key intermediate milestone dates and the full
implementation date, shall be in accordance with the implementation
schedule submitted by the licensee on April 7, 2011, and approved by
the NRC staff with this license amendment. All subsequent changes to
the NRC-approved CSP implementation schedule will require prior NRC
approval pursuant to 10 CFR 50.90.
Amendment Nos.: Unit 1--329 and Unit 2--322.
Facility Operating License DPR-77 and DPR-79: Amendments revised
the licenses.
Date of initial notice in Federal Register: December 7, 2010 (75 FR
76046).
The above Federal Register notice was based on the supplement dated
December 18, 2009. The supplements dated July 23 and October 1, 2010;
April 7 and July 15, 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.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated July 29, 2011.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: November 25, 2009, as
supplemented by letters dated April 22, May 14, August 24, September
29, and November 4, 2010, and February 23, 2011.
Brief description of amendment: The amendment revised Technical
Specification 3.3.2, ``Engineered Safety Feature Actuation System
(ESFAS) Instrumentation,'' to provide a 24-hour Completion Time (CT)
for restoration of an inoperable Balance of Plant (BOP) ESFAS train and
extends the CTs associated with individual instrument channels in the
BOP ESFAS train to maintain overall consistency of related TS actions.
Date of issuance: July 28, 2011.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 201.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: May 18, 2010 (75 FR
27833).
The supplemental letters dated April 22, May 14, August 24,
September 29, and November 4, 2010, and February 23, 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 amendment is contained
in a Safety Evaluation dated July 28, 2011.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: August 5, 2010, as supplemented
by letters dated March 23, May 3, and July 25, 2011.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) by relocating specific surveillance frequencies to
a licensee-controlled program with the guidance of Nuclear Energy
Institute (NEI) 04-10, ``Risk-Informed Technical Specifications
Initiative 5b, Risk-Informed Method for Control of Surveillance
Frequencies.'' The amendment adopted NRC-approved Technical
Specification Task Force (TSTF)-425, Revision 3, ``Relocate
Surveillance Frequencies to Licensee Control--RITSTF [Risk-Informed
TSTF] Initiative 5b.'' When implemented, TSTF-425 relocates most
periodic frequencies of TS surveillances to a licensee-controlled
program, the Surveillance Frequency Control Program (SFCP), and
provides requirements for the new program in the Administrative
Controls section of the TSs.
Date of issuance: July 29, 2011.
Effective date: As of its date of issuance and shall be implemented
within 180 days from the date of issuance.
Amendment No.: 202.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 11, 2011 (76 FR
1649).
The supplemental letters dated March 23, May 3, and July 25, 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.
[[Page 52715]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 29, 2011.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of application for amendments: July 12, 2010.
Brief Description of amendments: These amendments revise the
Technical Specifications (TSs) to: (1) Correct an error in TS 3.12.E.5,
(2) delete duplicative requirements in TS 3.12.E.2 and TS 3.12.E.4, (3)
relocate the shutdown margin value in TS 3.12 and the TS 3.12 Basis to
the Core Operating Limits Report (COLR), and 4) expand the TS 6.2 list
of parameters defined in the COLR.
Date of issuance: July 28, 2011.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 275 and 275.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments change the licenses and the technical specifications.
Date of initial notice in Federal Register: May 17, 2011 (76 FR
28477).
The Commission's related evaluation of the amendments is contained
in a safety evaluation dated July 28, 2011.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 11th day of August 2011.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2011-21212 Filed 8-22-11; 8:45 am]
BILLING CODE 7590-01-P