[Federal Register Volume 76, Number 124 (Tuesday, June 28, 2011)]
[Notices]
[Pages 37845-37851]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2011-16030]


-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[NRC-2011-0139]


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 to publish notice of any amendments issued, or 
proposed to be issued and grants the Commission the authority to issue 
and make immediately effective any amendment to an operating license 
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 June 2, 2011, to June 15, 2011. The last 
biweekly notice was published on June 14, 2011 (75 FR 34763).

ADDRESSES: Please include Docket ID NRC-2011-0139 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-
2011-0139. Address questions about NRC dockets to Carol Gallagher, 
telephone: 301-492-3668; e-mail: [email protected].
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
     Fax comments to: RADB at 301-492-3446.
    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, 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 available online 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-2011-0139.

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; (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 Commission's PDR, located at One White Flint North, Public File 
Area O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. The NRC regulations are accessible electronically from the NRC 
Library on the NRC Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or a presiding 
officer designated by the Commission or by the Chief Administrative 
Judge of the Atomic Safety and Licensing Board Panel, will rule on the 
request and/or

[[Page 37846]]

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 
ten (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 ID certificate, which allows the participant (or its counsel 
or representative) to digitally sign documents and access the E-
Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at http://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through EIE, 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-

[[Page 37847]]

submittals.html, by e-mail at [email protected], or by a toll-free 
call at 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://ehd.nrc.gov/EHD_Proceeding/home.asp, 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 Commission's PDR, located at One White Flint 
North, Public File Area O1 F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland.

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: April 11, 2011.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to define a new time limit 
for restoring inoperable Reactor Coolant System (RCS) leakage detection 
instrumentation to operable status; establish alternate methods of 
monitoring RCS leakage when one or more required monitors are 
inoperable; make a minor editorial change to correct a formatting issue 
to be consistent with the Technical Specifications Task Force (TSTF), 
``Writer's Guide for Plant-Specific Improved Technical 
Specifications,'' and the [Boiling-Water Reactor] BWR6 TS format and 
does not affect the intent of the TSTF or the NRC safety evaluation; 
and make TS Bases changes which reflect the proposed changes and more 
accurately reflect the contents of the facility design basis related to 
operability of the RCS leakage detection instrumentation. These changes 
are consistent with NRC-approved Revision 3 to TSTF Improved Standard 
Technical Specification (STS) Change Traveler TSTF-514, ``Revise BWR 
Operability Requirements and Actions for RCS Leakage Instrumentation.''
    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 clarifies the operability requirements for 
the RCS leakage detection instrumentation and reduces the time 
allowed for the plant to operate when the only TS-required operable 
RCS leakage detection instrumentation monitor is the drywell 
atmospheric gaseous radiation monitor. The monitoring of RCS leakage 
is not a precursor to any accident previously evaluated. The 
monitoring of RCS leakage is not used to mitigate the consequences 
of any accident previously evaluated.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change clarifies the operability requirements for 
the RCS leakage detection instrumentation and reduces the time 
allowed for the plant to operate when the only TS-required operable 
RCS leakage detection instrumentation monitor is the drywell 
atmospheric gaseous radiation monitor. The proposed change does not 
involve a physical alteration of the plant (no new or different type 
of equipment will be installed) or a change in the methods governing 
normal plant operation.
    Therefore, it is concluded that the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change clarifies the operability requirements for 
the RCS leakage detection instrumentation and reduces the time 
allowed for the plant to operate when the only TS-required operable 
RCS leakage detection instrumentation monitor is the drywell 
atmospheric gaseous radiation monitor. Reducing the amount of time 
the plant is allowed to operate with only the drywell atmospheric 
gaseous radiation monitor operable increase the margin of safety by 
increasing the likelihood that an increase in RCS leakage will be 
detected before it potentially results in gross failure.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: May 6, 2011.
    Description of amendment request: The proposed amendments would

[[Page 37848]]

revise Technical Specification 3.7.3, ``Ultimate Heat Sink,'' to reduce 
the allowed sedimentation in the Core Standby Cooling System (CSCS) 
pond from <= 1.5 feet to <= 1.0 feet, which allows the temperature of 
the cooling water supplied to the plant to be increased from <= 101.25 
[deg]F to <= 101.95 [deg]F resulting in a higher volume of cooling 
water available in the CSCS pond. 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 will reduce the allowed sedimentation in the 
Core Standby Cooling System (CSCS) pond from <= 1.5 feet to <= 1.0 
feet, which allows the indicated temperature of the cooling water 
supplied to the plant from the CSCS pond to be increased from <= 
101.25 [deg]F to <= 101.95 [deg]F based on reduction in post-
accident heatup from 2.0 [deg]F to 1.3 [deg]F due to a resulting 
higher volume of cooling water available in the CSCS pond.
    Analyzed accidents are assumed to be initiated by the failure of 
plant structures, systems, or components. An inoperable ultimate 
heat sink (UHS) is not considered as an initiator of any analyzed 
events. As such, there is not a significant increase in the 
probability of a previously evaluated accident. Allowing the UHS to 
operate with a lower allowance for sedimentation at a higher 
allowable indicated temperature, will not affect the failure 
probability of any equipment. The current heat analysis calculations 
of record for LSCS, Units 1 and 2, assume a UHS post-accident peak 
inlet temperature of 104 [deg]F. The proposed temperature increase 
is based on an adjustment to post accident UHS heatup due to 
restricting the level of sedimentation allowed in the CSCS pond. The 
current analysis bounds the proposed change. This higher allowable 
indicated temperature does not impact the loss of coolant accident 
(LOCA) Peak Clad Temperature Analysis, LOCA Containment Analysis or 
the non-LOCA analyses; therefore, continued operation with a UHS 
temperature > 101.25 [deg]F but <= 101.95 [deg]F will not increase 
the consequences of an accident previously evaluated in the Updated 
Final Safety Analysis Report (UFSAR).
    Based on the information discussed above, the reduction in the 
allowable CSCS pond sedimentation depth to <= 1.0 feet in concert 
with an allowable UHS temperature of <= 101.95 [deg]F, has no effect 
on the results of the design basis event, and will continue to 
assure that each required heat exchanger can perform its safety 
function. The plant heat exchangers will continue to provide 
sufficient cooling for the heat loads during the most severe 30-day 
period. Since the proposed change has no impact on any analyzed 
accident, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change involves reducing the allowable 
sedimentation of the CSCS pond from <= 1.5 feet to <= 1.0 feet. This 
proposed action will not alter the manner in which equipment is 
operated, nor will the functional demands on credited equipment be 
changed. Reducing the CSCS pond sedimentation limit does not 
introduce any new or different modes of plant operation, nor does it 
affect the operational characteristics of any safety-related 
equipment or systems; as such, no new failure modes are being 
introduced. The proposed action does not alter assumptions made in 
the safety analysis. Increasing the allowable indicated temperature 
of the cooling water supplied to the plant from the CSCS pond from 
<= 101.25 [deg]F to <= 101.95 [deg]F has no impact on safety related 
systems. The plant is designed such that the residual heat removal 
(RHR) pumps on the unit undergoing the LOCH/loss of offsite power 
(LOOP) conditions would start upon the receipt of a signal, and 
would load onto their respective Emergency Diesel Generators' 
emergency bus during the LOOP event. The increase in the allowable 
indicated temperature of the cooling water supplied to the plant 
from the CSCS pond will not require operation of additional RHR 
pumps; therefore, system operation is unaffected by the proposed 
change.
    Based on the above information, 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 reduces the allowable sedimentation levels 
in the CSCS pond to <= 1.0 feet and consequently allows an increase 
in the allowable indicated temperature of the cooling water supplied 
to the plant from the CSCS pond to <= 101.95 [deg]F. The margin of 
safety is determined by the design and qualification of the plant 
equipment, the operation of the plant within analyzed limits, and 
the point at which protective or mitigative actions are initiated. 
The proposed action does not impact these factors as the analyzed 
peak post accident inlet temperature of the UHS is unaffected based 
on the reduced allowable sediment depth in the CSCS pond. This 
change is supported by an engineering analysis that determined that 
existing post-accident CSCS pond heatup rates calculations were 
overly conservative based on observed CSCS pond sedimentation being 
significantly less than predicted. No setpoints are affected, and no 
other change is being proposed in the plant operational limits as a 
result of this change. All accident analysis assumptions and 
conditions will continue to be met. Adequate design margin is 
available to ensure that the required margin of safety is not 
significantly reduced.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Jacob. I. Zimmerman.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant, Citrus County, Florida

    Date of amendment request: March 24, 2011.
    Description of amendments request: The proposed amendment would 
adopt Technical Specification Task Force (TSTF), Improved Standard 
Technical Specifications Change Traveler, TSTF-248, Revision 0, 
``Revise Shutdown Margin Definition for Stuck Rod Exception,'' which 
modifies the definition of shutdown margin to include a provision 
allowing an exception to the highest reactivity worth stuck control rod 
penalty if there are two independent means of confirming that all 
control rods are fully inserted in the reactor core.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The revision to the Shutdown Margin (SDM) definition will result 
in analytical flexibility for determining SDM. Changes in the 
definition will not have an impact on the probability of an 
accident.
    The introduction of this definition change does not change 
continued compliance with all applicable regulatory requirements and 
design criteria (e.g., train separation, redundancy, and single 
failure). Therefore, since all plant systems will continue to 
function as designed, all plant parameters will remain within their 
design limits. As a result, the proposed change will not increase 
the consequences of an accident.
    Based on this discussion, the proposed LAR [license amendment 
request] does not significantly increase the probability or 
consequences of an accident previously evaluated.

[[Page 37849]]

    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Revising the definition of SDM in the Crystal River Unit 3 (CR-
3) Improved Technical Specifications (ITS) would not require core 
designers to revise any SDM calculation. Rather, it would afford the 
analytical flexibility for determining SDM for a particular 
circumstance.
    The proposed change does not involve any change in the design, 
configuration, or operation of the nuclear plant. The current plant 
safety analyses, therefore, remain complete and accurate in 
addressing the design basis events and in analyzing plant response 
and consequences.
    The Limiting Conditions for Operation, Limiting Safety System 
Settings and Safety Limits specified in the CR-3 ITS are not 
affected by the proposed change. As such, the plant conditions for 
which the design basis accident analysis were performed remain 
valid.
    The LAR does not introduce a new mode of plant operation or new 
accident precursors, does not involve any physical alterations to 
the plant configuration, or make changes to system setpoints that 
could initiate a new or different kind of accident.
    Therefore, the LAR does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does not involve a significant reduction in a margin of 
safety.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their accident mitigation 
functions. These barriers include the fuel and the fuel cladding, 
the reactor coolant system and the reactor containment building and 
containment related systems. The proposed change will not impact the 
reliability of these barriers to function. Radiological dose to 
plant operators or to the offsite public will not increase as a 
result of the proposed change. The change to the CR-3 ITS definition 
for SDM will not impact the safety barriers of the plant. Adequate 
SDM will continue to be assured for all operational conditions.
    Additionally, the current SDM calculation requires the 
consideration of the worth of the most reactive control rod to 
remain out of the core. This provides a margin of safety in that 
additional boron has to be injected to assure the reactor is shut 
down and remains shut down. This requirement will remain. However, 
once all control rods are verified to be fully inserted by two 
independent means, the conservatism of the additional boron 
concentration is balanced by the additional reactive worth of the 
inserted control rod and the additional boron will not be necessary 
to maintain the required SDM. The independent verification of all 
rods in will provide a very high confidence that adequate SDM will 
continue to be assured.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Douglas A. Broaddus.

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-220, Nine 
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York

    Date of amendment request: May 25, 2011.
    Description of amendment request: The proposed license amendment 
would delete an outdated reference to a specific date delineated in 
License Condition 2.B.(2) to be consistent with the wording found in 
the corresponding license condition at multiple stations including Nine 
Mile Point Unit 2 and Calvert Cliffs Units 1 and 2. This license 
condition authorizes NMPNS to ``* * * receive, possess and use at any 
time special nuclear material as reactor fuel, in accordance with the 
limitations for storage and amounts required for reactor operation, as 
described in the Final Safety Analysis Report as supplemented and 
amended as of February 4, 1976.'' The proposed change will remove the 
words ``as of February 4, 1976.''
    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 NMP1 Technical Specifications (TS) and Updated Final Safety 
Analysis Report (UFSAR) provide the specific limitations on the 
number of fuel assemblies in the NMP1 spent fuel pool, fresh fuel 
storage vault, and the reactor core. Removing the outdated reference 
to the February 4, 1976 UFSAR from License Condition 2.B.(2) has no 
effect on these limitations or on the supporting evaluations. The 
proposed change does not affect a precursor to any accident 
previously evaluated nor does it affect the ability of any system to 
mitigate the consequences of any accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The NMP1 TS and UFSAR provide the specific limitations on the 
number of fuel assemblies in the NMP1 spent fuel pool, fresh fuel 
storage vault, and the reactor core. Removing the outdated reference 
to the February 4, 1976 UFSAR from License Condition 2.B.(2) has no 
effect on these limitations or on the supporting evaluations. The 
proposed change does not introduce a new mode of plant operation and 
does not involve a physical modification to the plant. The change 
will not introduce new accident initiators or impact the assumptions 
made in a safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is related to confidence in the ability of the 
fission product barriers to perform their design functions during 
and following postulated accidents. The NMP1 TS and UFSAR provide 
the specific limitations on the number of fuel assemblies in the 
NMP1 spent fuel pool, fresh fuel storage vault, and the reactor 
core. Removing the outdated reference to the February 4, 1976, UFSAR 
from License Condition 2.B.(2) has no effect on these limitations or 
on the supporting evaluations. Accordingly, no margin of safety is 
affected.
    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: Carey W. Fleming, Senior Counsel, 
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 
200C, Baltimore, MD 21202.
    NRC Acting Branch Chief: John P. Boska

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-410, Nine 
Mile Point Nuclear Station Unit No. 2 (NMP 2), Oswego County, New York

    Date of amendment request: March 30, 2011.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3.4.7, ``RCS [Reactor 
Coolant System] Leakage Detection Instrumentation,'' to define a new 
time limit for restoring inoperable RCS leakage detection 
instrumentation to operable status and establish alternate methods of 
monitoring RCS leakage when required

[[Page 37850]]

monitors are inoperable. The proposed changes would be consistent with 
the NRC-approved Revision 3 to Technical Specification Task Force 
(TSTF), Improved Standard Technical Specification (STS) Change Traveler 
TSTF-514, ``Revise BWR [boiling-water reactor] Operability Requirements 
and Actions for RCS Leakage Instrumentation.'' The NRC staff issued a 
Notice of Availability of the models for referencing in license 
amendment applications in the Federal Register on December 17, 2010 (75 
FR 79048) as part of the consolidated line item improvement process.
    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 clarifies the operability requirements for 
the RCS leakage detection instrumentation and reduces the time 
allowed for the plant to operate when the only TS-required operable 
Reactor Coolant System (RCS) leakage detection instrumentation 
monitor is the drywell atmospheric gaseous radioactivity monitor. 
The monitoring of RCS leakage is not a precursor to any accident 
previously evaluated. The monitoring of RCS leakage is not used to 
mitigate the consequences of any accident previously evaluated.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident previously evaluated?
    Response: No.
    The proposed change clarifies the operability requirements for 
the RCS leakage detection instrumentation and reduces the time 
allowed for the plant to operate when the only TS-required operable 
RCS leakage detection instrumentation monitor is the drywell 
atmospheric gaseous radioactivity monitor. The proposed change does 
not involve a physical alteration of the plant (no new or different 
type of equipment will be installed) or a change in the methods 
governing normal plant operation.
    Therefore, it is concluded that the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    4. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change clarifies the operability requirements for 
the RCS leakage detection instrumentation and reduces the time 
allowed for the plant to operate when the only TS-required operable 
RCS leakage detection instrumentation monitor is the drywell 
atmospheric gaseous radioactivity monitor. Reducing the amount of 
time the plant is allowed to operate with only the drywell 
atmospheric gaseous radioactivity monitor operable increases the 
margin of safety by increasing the likelihood that an increase in 
RCS leakage will be detected before it potentially results in gross 
failure.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Carey W. Fleming, Senior Counsel, 
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite 
200C, Baltimore, MD 21202.
    NRC Acting Branch Chief: Douglas V. Pickett

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. (See ADDRESSES section.)

Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: September 16, 2010, as 
supplemented by letter dated March 31, 2011.
    Brief description of amendments: The amendments revised Technical 
Specification 3.3.2, ``Engineered Safety Feature Actuation System 
(ESFAS) Instrumentation,'' to replace the references to the outdated 
logic per train per doghouse with updated references which reflect 
License Amendment Nos. 249 and 243 granted by the U.S. Nuclear 
Regulatory Commission (NRC) staff on April 2, 2009.
    Date of issuance: June 13, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 264 and 260.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: January 25, 2011 (76 FR 
4384). The supplement dated March 31, 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 amendments is 
contained in a Safety Evaluation dated June 13, 2011.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Date of application for amendment: June 30, 2009, as supplemented 
by letters dated. January 25, July 1, November 8, 2010, and January 31, 
March 16 and May 4, 2011.
    Brief description of amendment: The proposed amendments revised 
Technical Specification (TS) 3.7.9, ``Ultimate Heat Sink (UHS),'' to 
add additional essential service water (SX) cooling tower fan 
requirements as a function of SX pump discharge temperature reflective 
of a revised analysis for the UHS.
    Date of issuance: June 14, 2011.

[[Page 37851]]

    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 173/173.
    Facility Operating License Nos. NPF-37 and NPF-66: The amendment 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: September 8, 2009 (74 
FR 46241). The January 25, July 1, November 8, 2010, and January 31, 
March 16 and May 4, 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 June 14, 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: June 22, 2010, supplemented on 
January 13, 2011.
    Brief description of amendment: The amendments revised the 
containment spray nozzles obstruction surveillance frequency specified 
in Surveillance Requirement 3.6.6.5 from a fixed ``10 years'' to 
``Following maintenance that could result in nozzle blockage.''
    Date of issuance: June 1, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 314 (for Unit 1) and 298 (for Unit 2).
    Facility Operating License Nos. DPR-58 and DPR-74: Amendment 
revised the Renewed Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 24, 2010 (75 FR 
52042).
    The supplemental information dated January 13, 2011, contained 
clarifying information, did not change the scope of the original 
application or the initial no significant hazards consideration 
determination, and does 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 June 1, 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: June 1, 2010, as supplemented 
by letters dated July 9 and November 22, 2010.
    Brief description of amendments: The amendments consist of revising 
the current license basis regarding a postulated reactor vessel head 
drop (RVHD) event to conform to the NRC-endorsed guidance of Nuclear 
Energy Institute (NEI) 08-05, ``Industry Initiative on Control of Heavy 
Loads,'' Revision 0. The proposed change to the license basis will 
revise Chapter 14.3.6, ``Reactor Vessel Head Drop Event,'' of the Final 
Safety Analysis Report.
    Date of issuance: June 1, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 242, 246.
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revise the Final Safety Analysis Report, Chapter 14.3.6, 
Reactor Vessel Head Drop Event.
    Date of initial notice in Federal Register: September 21, 2010 (75 
FR 57526). The supplemental letters contained clarifying information 
and did not change the 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 June 1, 2011.
    No significant hazards consideration comments received: No.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: July 23, 2009, as supplemented 
by letter dated May 3, 2011.
    Brief description of amendment: The amendment revises technical 
specification actions requiring suspension of operations involving 
positive reactivity addition and revises various notes precluding 
reduction in boron concentration. The amendment is consistent with 
TSTF-286, Revision 2, Define ``Operations Involving Positive Reactivity 
Additions.''
    Date of issuance: June 8, 2011.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 112.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2011 (76 FR 
12765). The letter dated May 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 as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 8, 2011.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Unit Nos. 1 and 2, Appling County Georgia and Southern 
Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle 
Electric Generating Plant, Units 1 and 2, Burke County, Georgia

    Date of application for amendments: December 16, 2010.
    Brief description of amendments: The amendments revised the 
Technical Specifications Section 2.0 ``Safety Limits,'' removing the 
requirement to report a Safety Limit Violation, that is redundant to 
existing regulations, Title 10 of the Code of Federal Regulations (10 
CFR) Section 50.36(c)(8) ``Written Reports.''
    Date of issuance: June 13, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 264, 208 (Hatch) and 161, 143 (Vogtle).
    Facility Operating License Nos. NPF-68 and NPF-81 for Vogtle Units 
1 and 2 respectively and DPR-57 and NPF-5 for Hatch Units 1 and 2 
respectively: Amendments revised the licenses and the technical 
specifications.
    Date of initial notice in Federal Register: February 22, 2011 (76 
FR 9828).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 13, 2011.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland this 16th day of June 2011.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2011-16030 Filed 6-27-11; 8:45 am]
BILLING CODE 7590-01-P