Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 52699-52715 [2011-21212]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 76, Number 163 (Tuesday, August 23, 2011)]
[Notices]
[Pages 52699-52715]
From the Federal Register Online via the Government Printing 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 https://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 https://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 Carol.Gallagher@nrc.gov.
     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 
https://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 pdr.resource@nrc.gov.
     Federal Rulemaking Web Site: Public comments and 
supporting materials related to this notice can be found at https://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 https://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 hearing.docket@nrc.gov, 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 https://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 
https://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 https://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 https://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 https://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
MSHD.Resource@nrc.gov, 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 
https://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 https://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 pdr.resource@nrc.gov.

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 https://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 
pdr.resource@nrc.gov.

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 notice