[Federal Register Volume 75, Number 192 (Tuesday, October 5, 2010)]
[Notices]
[Pages 61521-61530]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-24815]


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NUCLEAR REGULATORY COMMISSION

[NRC-2010-0309]


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. 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

[[Page 61522]]

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 September 9, 2010, to September 22, 2010. 
The last biweekly notice was published on September 21, 2010 (75 
FR57521).

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.
    Written comments may be submitted by mail to the Chief, Rules, 
Announcements and Directives Branch (RADB), TWB-05-B01M, Division of 
Administrative Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be faxed to the RADB at 301-492-3446. 
Documents may be examined, and/or copied for a fee, at the NRC's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
    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 O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards

[[Page 61523]]

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 
<a href="/cdn-cgi/l/email-protection#fa929f9b8893949dd49e9599919f8eba948899d49d958c"><span class="__cf_email__" data-cfemail="731b1612011a1d145d171c10181607331d01105d141c05">[email protected]</span></a>, 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 NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in NRC's ``Guidance for Electronic 
Submission,'' which is available on the agency's public Web site at 
http://www.nrc.gov/site-help/e-submittals.html. Participants may 
attempt to use other software not listed on the Web site, but should 
note that the NRC's E-Filing system does not support unlisted software, 
and the NRC Meta System Help Desk will not be able to offer assistance 
in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through 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-submittals.html, by e-mail at 
<a href="/cdn-cgi/l/email-protection#e1acb2a9a5cfb384928e94938284a18f9382cf868e97"><span class="__cf_email__" data-cfemail="cc819f8488e29ea9bfa3b9beafa98ca2beafe2aba3ba">[email protected]</span></a>, 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 O1F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the ADAMS Public Electronic Reading

[[Page 61524]]

Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who encounter 
problems in accessing the documents located in ADAMS, should contact 
the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737, or by e-
mail to <a href="/cdn-cgi/l/email-protection#abdbcfd985d9ced8c4ded9c8ceebc5d9c885ccc4dd"><span class="__cf_email__" data-cfemail="750511075b0710061a00071610351b07165b121a03">[email protected]</span></a>.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: May 28, 2010.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) to allow manual operation of the 
containment spray system (CSS) and to change the setpoints for the 
refueling water storage tank (RWST).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1: Does the proposed amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The CSS and RWST are accident mitigation equipment. As such, 
changes in operation of these systems cannot have an impact on the 
probability of an accident.
    The RWST will continue to comply with all applicable regulatory 
requirements and design criteria following approval of the proposed 
changes (e.g., train separation, redundancy, and single failure). 
The water level on the containment floor will be higher at the start 
of transfer to the containment sump but will remain below the 
maximum design level analyzed for equipment submergence. The change 
in the sump pH will not result in a significant increase in 
radiological consequences of a LOCA [loss-of-coolant accident]. 
Therefore, the design functions performed by the equipment are not 
changed.
    The proposed change alters the method of controlling the safety 
system following a design basis event so that manual actions are 
substituted for automatic actions. Calculations and simulator 
exercises confirm these actions will be taken within the appropriate 
scenario sequence timing to provide containment cooling and source 
term reduction.
    The delay in CS [containment spray] operation will result in an 
increase in containment temperature, containment pressure, offsite 
dose, and control room dose during a LOCA or high energy line break 
inside containment. Containment analyses have been performed to 
demonstrate that containment pressure and temperature remain within 
the design limits and there is no significant impact on the 
environmental qualification for equipment inside containment. The 
reduction in fission product removal due to delayed CS operation 
does not result in exceeding the offsite dose and control room dose 
limits in 10 CFR 50.67. The analysis of the change in containment 
conditions due to a single failure of an operating spray pump and 
the suspension of CS determined that the pressure remained below the 
design limits.
    The proposed change to adopt [Technical Specification Task 
Force] TSTF-493, Rev. 4, on a limited basis clarifies requirements 
for instrumentation to ensure the instrumentation will actuate as 
assumed in the safety analysis. Instruments are not an assumed 
initiator of any accident previously evaluated. As a result, the 
proposed change will not increase the probability of an accident 
previously evaluated. The proposed change will ensure that the 
instruments actuate as assumed to mitigate the accidents previously 
evaluated. As a result, the proposed change will not increase the 
consequences of an accident previously evaluated.
    Based on this discussion, the proposed amendment does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    Criterion 2: Does the proposed amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The modification to the low level setpoint will not install any 
new plant equipment. The setpoint will continue to be included 
within the engineered safeguards features instrumentation and 
monitored according to the applicable surveillance requirements. The 
evaluation of the new level setpoint and the change in the 
switchover sequence concluded that the equipment aligned to the sump 
will continue to have sufficient suction pressure prior to 
containment sump suction switchover. The design of the RWST low 
level instrumentation complies with all applicable regulatory 
requirements and design criteria.
    The overall function of the CSS is not changed by this proposed 
amendment. The proposed change alters the method of controlling the 
safety system following a design basis event so that manual actions 
are substituted for automatic actions. Calculations confirm that 
these actions will be taken within the appropriate scenario sequence 
timing to provide containment cooling and source term reduction with 
no significant increase in radiological consequences and without 
exceeding containment design limits.
    The proposed change to adopt TSTF-493, Rev. 4 on a limited basis 
does not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation. The change does not alter 
assumptions made in the safety analysis but ensures that the 
instruments behave as assumed in the accident analysis. The proposed 
change is consistent with the safety analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    Criterion 3: Does the proposed amendment involve a significant 
reduction in a margin of safety?
    Response: No.
    The proposed change will increase the calculated radiological 
dose at the site boundary and in the control room. However, the 
calculations demonstrate that the dose consequences at the site 
boundary, low population zone, and control room remain within 
regulatory acceptance limits of 10 CFR 50.67.
    Additional analysis concluded:
    <bullet> Peak containment pressure for analyzed design basis 
accidents will not be significantly increased and containment design 
limits will not be exceeded.
    <bullet> Assumptions used in the environmental qualification of 
equipment exposed to the containment atmosphere remain bounding.
    <bullet> Pumps aligned to the RWST and to the containment sump 
will have adequate suction pressure.
    <bullet> The CSS will retain its ability to undergo all 
appropriate testing requirements following implementation of the 
proposed amendment. These testing requirements are conducted in 
accordance with the McGuire Inservice Testing Program and TS 3.6.6.
    It is estimated that the implementation of this license 
amendment request will result in an approximate 22% reduction in 
core damage frequency. This amendment request is based on the 
Nuclear Energy Institute (NEI) and the Pressurized Water Reactor 
(PWR) Owners Group initiative to extend the post-Loss of Coolant 
Accident (LOCA) injection phase and delay the onset of the 
containment sump recirculation phase.
    The proposed change to adopt TSTF-493, Rev. 4 on a limited basis 
clarifies the requirements for instrumentation to ensure the 
instrumentation will actuate as assumed in the accident analysis. No 
change is made to the accident analysis assumptions and no margin of 
safety is reduced as part of this change.
    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: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Gloria Kulesa.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana

    Date of amendment request: July 22, 2010.
    Description of amendment request: The proposed amendment would 
revise

[[Page 61525]]

Limiting Condition for Operation (LCO) 3.10.1, ``Inservice Leak and 
Hydrostatic Testing Operation,'' and the associated Bases, to expand 
its scope to include provisions for temperature excursions greater than 
200 degrees Fahrenheit ([deg]F) as a consequence of inservice leak and 
hydrostatic testing, and as a consequence of scram time testing 
initiated in conjunction with an inservice leak or hydrostatic test, 
while considering operational conditions to be in Mode 4. The proposed 
change is consistent with NRC-approved Technical Specification Task 
Force (TSTF) Improved Standard Technical Specification Traveller, TSTF-
484, ``Use of TS 3.10.1 for Scram Time Testing Activities,'' that was 
announced in the Federal Register on October 27, 2001 (71 FR 63050), as 
part of the consolidated Line Item Improvement Process (CCIIP).
    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.
    Technical Specifications currently allow for operation at > 200 
[deg]F while imposing MODE 4 requirements in addition to the 
secondary containment requirements required to be met. Extending the 
activities that can apply this allowance will not adversely impact 
the probability or consequences of an 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.
    Technical Specifications currently allow for operation at > 200 
[deg]F while imposing MODE 4 requirements in addition to the 
secondary containment requirements required to be met. No new 
operational conditions beyond those currently allowed by LCO 3.10.1 
are introduced. The extended allowances would result from operations 
that commence at reduced temperatures, but approach the normal MODE 
4 limit of 200 [deg]F prior to completion of the inspections or 
testing. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements or eliminate any existing requirements. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are consistent with the safety analysis assumptions 
and current plant operating practice.
    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.
    Technical Specifications currently allow for operation at > 200 
[deg]F while imposing MODE 4 requirements in addition to the 
secondary containment requirements required to be met. Extending the 
activities that can apply this allowance will not adversely impact 
any margin of safety. Allowing completion of inspections and testing 
and supporting completion of scram time testing initiated in 
conjunction with an inservice leak or hydrostatic test prior to 
power operation, results in enhanced safe operations by eliminating 
unnecessary maneuvers to control reactor temperature and pressure.
    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: Joseph A. Aluise, Assistant General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: August 19, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications to be consistent with Standard 
Technical Specifications 3.6.1.8 ``Suppression Chamber-to-Drywell 
Vacuum Breakers'' and 3.6.2.5 ``Drywell-to-Suppression Chamber 
Differential Pressure,'' along with the associated Bases, of NUREG-
1433, Revision 3, ``Standard Technical Specifications General Electric 
Plants, BWR/4,'' modified to account for plant specific design details.
    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 proposed amendment does not significantly increase the 
probability or consequences of an accident since it does not involve 
a modification to any plant equipment or affect how plant systems or 
components are operated. No design functions or design parameters 
are affected by the proposed amendment. The proposed amendment 
involves the operation and testing of Primary Containment systems 
but does not impact containment design or performance requirements. 
Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve any physical alteration of 
plant equipment and does not change the method by which any safety-
related system performs its function. No new or different types of 
equipment will be installed and the basic operation of installed 
equipment is unchanged. The methods governing plant operation and 
testing remain consistent with current safety analysis assumptions. 
The proposed amendment involves the operation and testing of Primary 
Containment systems but does not alter the way that the systems are 
operated or how the tests are performed. 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.
    The proposed change ensures that the safety functions of the 
pressure suppression chamber-drywell vacuum breakers and drywell-
suppression chamber differential pressure are fulfilled by 
incorporating the guidance of NUREG-1433. The proposed amendment 
does not involve a physical modification of the plant and does not 
change the design or function of any component or system. Therefore, 
the proposed amendment will not involve a significant reduction in 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Nancy Salgado.

[[Page 61526]]

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: August 10, 2010.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.9.3, ``Reactor Building 
Penetrations,'' to allow reactor building flow path(s) providing direct 
access from the reactor building atmosphere to the outside atmosphere 
to be unisolated under administrative control, during movement of 
irradiated fuel assemblies. The proposed change is consistent with 
Technical Specification Task Force (TSTF) Technical Change Traveler 
312, Revision 1, ``Administratively Control Containment Penetrations.''
    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 status of the penetration flow paths during fuel movement in 
the reactor building has no affect on the probability of the 
occurrence of any accident previously evaluated. The proposed change 
does not alter any plant equipment or operating practices in such a 
manner that the probability of an accident is increased. Since the 
consequences of a fuel handling accident (FHA) inside the reactor 
building with open penetrations flow paths is bounded by the current 
FHA analyses and the probability of an accident is not affected by 
the status of the penetration flow paths, 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 open reactor building penetration flow paths are not 
accident initiators. The proposed allowance to open the reactor 
building penetrations during fuel movement inside the reactor 
building will not adversely affect plant safety functions or 
equipment operating practices such that a new or different accident 
could be created. Therefore, the proposed change does not create the 
possibility of an accident of a different kind than previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Technical Specification (TS) 3.9.3 closure requirements for 
reactor building penetrations ensure that the consequences of a 
postulated FHA inside the reactor building during irradiated fuel 
handling activities are minimized. The Limiting Condition for 
Operation establishes reactor building closure requirements, which 
limit the potential escape paths for fission products by ensuring 
that there is at least one integral barrier to the release of 
radioactive material. The proposed change to allow the reactor 
building penetration flow paths to be open during refueling 
operations under administrative controls does not significantly 
affect the expected dose consequences of a FHA because the limiting 
FHA does not credit reactor building closure or filtration. The 
proposed administrative controls provide assurance that prompt 
closure of the penetration flow paths will be accomplished in the 
event of a[n] FHA inside the reactor building. The provisions to 
promptly isolate open penetration flow paths provide assurance that 
the offsite dose consequences of a[n] FHA inside containment will be 
minimized. Therefore, this 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, Assistant 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. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois; Docket Nos. 
STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle 
County, Illinois

    Date of amendment request: June 29, 2010, as supplemented on August 
24, 2010.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TS) Section 3.4.12, ``Low Temperature 
Overpressure Protection (LTOP) System,'' to correct an inconsistency 
between the TS, and implementation of procedures and administrative 
controls for Safety Injection (SI) pumps required to mitigate a 
postulated loss of decay heat removal during mid-loop operation as 
discussed in NRC Generic Letter (GL) 88-17, ``Loss of Decay Heat 
Removal.''
    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 does not result in any physical changes to 
safety related structures, systems, or components. The proposed 
change revises TS 3.4.12 to correct an inconsistency between the TS, 
and implementation of procedures and administrative controls for SI 
pumps required to mitigate a postulated loss of decay heat removal 
during mid-loop operation as discussed in GL 88-17. Specifically, 
the proposed change adds a note to TS LCO [limiting condition for 
operation] 3.4.12 that states: ``For the purpose of protecting the 
decay heat removal function, one or more SI pumps may be capable of 
injecting into the RCS in MODE 5 and MODE 6 when the reactor vessel 
head is on provided pressurizer level is <= 5 percent.'' The 
proposed change corrects an oversight introduced during the 
conversion of the Braidwood Station and Byron Station TS to the ITS 
[Improved TS].
    The probability of occurrence of an accident is not increased 
since the proposed change will continue to require that no SI pumps 
are capable of injecting into the RCS in Modes 5 and 6 with 
pressurizer level greater than 5 percent.
    The NRC has previously evaluated the allowance for one or more 
SI pumps to be capable of injecting into the RCS in Mode 5 or Mode 6 
when the reactor vessel head is on provided pressurizer level is <= 
5 percent for the Braidwood Station and Byron Station. In a safety 
evaluation dated August 31, 1990, related to Braidwood Station, 
Units 1 and 2, Amendment 25, and Byron Station, Units 1 and 2, 
Amendment 38, the NRC concluded that allowing SI pump capability to 
inject into the RCS in Mode 5 or Mode 6 when the reactor vessel head 
is on provided pressurizer level is <= 5 percent was acceptable. The 
availability of SI pumps under these circumstances does not present 
a concern regarding cold overpressure protection since sufficient 
air volume exists which allows Operations personnel time to mitigate 
the transient. This is in contrast to the analyzed cold overpressure 
transients, in which the RCS is assumed to be water solid at the 
onset of the event.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises TS 3.4.12 to correct an 
inconsistency between the TS, and implementation of procedures and 
administrative controls for SI pumps required to mitigate a 
postulated loss of decay heat removal during mid-loop operation as 
discussed in GL 88-17. Specifically, the proposed change adds a note 
to TS LCO 3.4.12 that states: ``For the purpose of protecting the 
decay heat removal function, one or more SI pumps may be

[[Page 61527]]

capable of injecting into the RCS in MODE 5 and MODE 6 when the 
reactor vessel head is on provided pressurizer level is <= 5 
percent.'' The proposed change corrects an oversight introduced 
during the conversion of the Braidwood Station and Byron Station TS 
to the ITS.
    The proposed change is necessary for the purpose of mitigating 
the consequences of a loss of decay heat removal during mid-loop 
operations. Operation of at least one SI pump is required in some 
cases to prevent the core from uncovering. The only new 
configuration allowed by the proposed change is the potential of 
having one or more SI pumps available in Modes 5 and 6 with 
pressurizer level <= 5 percent. The potential overpressurization 
accident has been analyzed and accounted for by requiring 
pressurizer level to be <= 5 percent if one or more SI pumps are 
available.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises TS 3.4.12 to correct an 
inconsistency between the TS, and implementation of procedures and 
administrative controls for SI pumps required to mitigate a 
postulated loss of decay heat removal during mid-loop operation as 
discussed in GL 88-17. Specifically, the proposed change adds a note 
to TS LCO 3.4.12 that states: ``For the purpose of protecting the 
decay heat removal function, one or more SI pumps may be capable of 
injecting into the RCS in MODE 5 and MODE 6 when the reactor vessel 
head is on provided pressurizer level is <= 5 percent.'' The 
proposed change corrects an oversight introduced during the 
conversion of the Braidwood Station and Byron Station TS to the ITS.
    The proposed note allows one or more SI pumps to be capable of 
injecting into the RCS only when pressurizer level is <= 5 percent 
in Mode 5 and Mode 6 when the reactor vessel head is on. This 
provides protection to limit coolant input capacity during shutdown 
in which a pressure fluctuation due to coolant input from the SI 
pumps could occur more quickly than an operator could react, while 
providing an allowance for one or more SI pumps to be capable of 
injecting into the RCS during conditions in which a loss of decay 
heat removal could result in rapid core uncovery.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Robert D. Carlson.

Florida Power and Light Company (FPL), Docket Nos. 50-250 and 50-251, 
Turkey Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: August 5, 2010.
    Description of amendment request: The proposed amendments would 
revise technical specification (TS) 5.5.1 Fuel Storage--Criticality, to 
include new spent fuel storage patterns that account for both the 
increase in fuel maximum enrichment from 4.5 weight percentage (wt%) U-
235 to 5.0 wt% U-235 and the impact on the fuel of higher power 
operation proposed under the Extended Power Uprate (EPU) project. 
Although the fuel storage has been analyzed at the higher fuel 
enrichment in the new criticality analysis, the fuel enrichment limit 
of 4.5 wt% U-235 specified in TS 5.5.1 will not be changed under this 
license amendment request. The proposed TS changes and a new supporting 
criticality analysis are being submitted to revise the current 
licensing basis analysis for both new fuel and spent fuel pool storage.
    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?
    No. The proposed amendments do not change or modify the fuel, 
fuel handling processes, fuel storage racks, number of fuel 
assemblies that may be stored in the spent fuel pool (SFP), decay 
heat generation rate, or the spent fuel pool cooling and cleanup 
system. The proposed amendment was evaluated for impact on the 
following previously evaluated events and accidents:
    a. A fuel handling accident (FHA),
    b. A cask drop accident,
    c. A fuel mispositioning event,
    d. A spent fuel pool boron dilution event,
    e. A seismic event, and
    f. A loss of spent fuel pool cooling event.
    Although the proposed amendment will require increased handling 
of the fuel, the probability of a FHA is not significantly increased 
because the implementation of the proposed amendment will employ the 
same equipment and process to handle fuel assemblies that is 
currently used. Also, tests have confirmed that the Metamic inserts 
can be installed and removed without damaging the host fuel 
assemblies. The FHA radiological dose consequences associated with 
fuel enrichment at this level were addressed in LAR [license 
amendment request] 196 on Alternative Source Term implementation at 
EPU conditions and remain unchanged. Therefore, the proposed 
amendments do not significantly increase the probability or 
consequences of a FHA.
    The proposed amendments do not increase the probability of 
dropping a fuel transfer cask because they do not introduce any new 
heavy loads to the SFP and do not affect heavy load handling 
processes. Also, the insertion of Metamic rack inserts does not 
increase the consequences of the cask drop accident because the 
radiological source term of that accident is developed from a non-
mechanistically derived quantity of damaged fuel stored in the spent 
fuel pool. Therefore, the proposed amendments do not significantly 
increase the probability or consequences of a cask drop accident.
    Operation in accordance with the proposed amendment will not 
change the probability of a fuel mispositioning event because fuel 
movement will continue to be controlled by approved fuel handling 
procedures. These procedures continue to require identification of 
the initial and target locations for each fuel assembly that is 
moved. The consequences of a fuel mispositioning event are not 
changed because the reactivity analysis demonstrates that the same 
subcriticality criteria and requirements continue to be met for the 
worst-case fuel mispositioning event.
    Operation in accordance with the proposed amendment will not 
change the probability of a boron dilution event because the systems 
and events that could affect spent fuel pool soluble boron are 
unchanged. The consequences of a boron dilution event are unchanged 
because the proposed amendment reduces the soluble boron requirement 
below the currently required value and the maximum possible water 
volume displaced by the inserts is an insignificant fraction of the 
total spent fuel pool water volume.
    Operation in accordance with the proposed amendment will not 
change the probability of a seismic event. The consequences of a 
seismic event are not significantly increased because the forcing 
functions for seismic excitation are not increased and because the 
mass of storage racks with Metamic inserts is not appreciably 
increased. Seismic analyses demonstrate adequate stress levels in 
the storage racks when inserts are installed.
    Operation in accordance with the proposed amendment will not 
change the probability of a loss of SFP cooling event because the 
systems and events that could affect SFP cooling are unchanged. The 
consequences are not significantly increased because there are no 
changes in the SFP heat load or SFP cooling systems, structures or 
components. Furthermore, conservative analyses indicate that the 
current design requirements and criteria continue to be met with the 
Metamic inserts installed.
    Therefore, the proposed changes do 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?
    No. The proposed amendments do not change or modify the fuel, 
fuel handling processes, fuel racks, number of fuel assemblies that 
may be stored in the pool,

[[Page 61528]]

decay heat generation rate, or the spent fuel pool cooling and 
cleanup system. The effects of operating with the proposed amendment 
are listed below. The proposed amendments were evaluated for the 
potential of each effect to create the possibility of a new or 
different kind of accident:
    a. Addition of inserts to the fuel storage racks,
    b. New storage patterns,
    c. Additional weight from the inserts,
    d. Insert movement above fuel, and
    e. Displacement of fuel pool water by the inserts.
    Each insert will be placed between a fuel assembly and the 
storage cell wall, taking up some of the space available on two 
sides of the fuel assembly. Tests confirm that the insert can be 
installed and removed without damaging the fuel assembly. Analyses 
demonstrate that the presence of the inserts does not adversely 
affect spent fuel cooling, seismic capability, or subcriticality. 
The aluminum (alloy 6061) and boron carbide materials of 
construction have been shown to be compatible with nuclear fuel, 
storage racks and spent fuel pool environments, and generate no 
adverse material interactions. Therefore, placing the inserts into 
the spent fuel pool storage racks cannot cause a new or different 
kind of accident.
    Operation with the proposed fuel storage patterns will not 
create a new or different kind of accident because fuel movement 
will continue to be controlled by approved fuel handling procedures. 
These procedures continue to require identification of the initial 
and target locations for each fuel assembly that is moved. There are 
no changes in the criteria or design requirements pertaining to fuel 
storage safety, including subcriticality requirements, and analyses 
demonstrate that the proposed storage patterns meet these 
requirements and criteria with adequate margins. Therefore, the 
proposed storage patterns cannot cause a new or different kind of 
accident.
    Operation with the added weight of the Metamic inserts will not 
create a new or different accident. The net effect of the adding the 
maximum number of inserts is to add less than one percent to the 
weight of the loaded racks. Furthermore, the analyses of the racks 
with Metamic inserts installed demonstrate that the stress levels in 
the rack modules continue to be considerably less than allowable 
stress limits. Therefore, the added weight from the inserts cannot 
cause a new or different kind of accident.
    Operation with insert movement above stored fuel will not create 
a new or different kind of accident. The insert with its handling 
tool weighs considerably less than the weight of a single fuel 
assembly. Single fuel assemblies are routinely moved safely over 
fuel assemblies and the same level of safety in design and operation 
will be maintained when moving the inserts. Furthermore, the effect 
of a dropped insert to block the top of a storage cell has been 
evaluated in thermal-hydraulic analyses. Therefore, the movement of 
inserts cannot cause a new or different kind of accident.
    Whereas the installed rack inserts will displace a very small 
fraction of the fuel pool water volume and impose a very small 
reduction in operator response time to previously-evaluated SFP 
accidents, the reduction will not promote a new or different kind of 
accident. Also, displacement of water along two sides of a stored 
fuel assembly may have some local reduction in the peripheral 
cooling flow; however, this effect would be small compared to the 
flow induced through the fuel assembly and would in no way promote a 
new or different kind of accident.
    The accidents and events previously analyzed and presented in 
the Boraflex Remedy and Alternative Source Term LARs remain 
bounding.
    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 the margin of safety?
    No. The proposed change was evaluated for its effect on current 
margins of safety as they relate to criticality, structural 
integrity, and spent fuel heat removal capability.
    The margin of safety for subcriticality required by 10 CFR 
50.68(b)(4) is unchanged. New criticality analysis confirms that 
operation in accordance with the proposed amendment continues to 
meet the required subcriticality margins.
    The structural evaluations for the racks and spent fuel pool 
with Metamic inserts installed show that the rack and spent fuel 
pool are unimpaired by loading combinations during seismic motion, 
and there is no adverse seismic-induced interaction between the rack 
and Metamic inserts.
    The proposed change does not affect spent fuel heat generation 
or the spent fuel pool cooling systems. A conservative analysis 
indicates that the design basis requirements and criteria for spent 
fuel cooling continue to be met with the Metamic inserts in place, 
and displacing coolant. Thermal hydraulic analysis of the local 
effects of an installed rack insert blocking peripheral flow show a 
small increase in local water and fuel clad temperatures, but will 
remain within acceptable limits including no departure from nucleate 
boiling.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    Based on the above discussion, FPL has determined that the 
proposed change does not involve a significant hazards 
consideration.

    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: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Douglas A. Broaddus.

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Virginia Electric and Power Company: Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Unit Nos. 1 and 2, Located in Louisa County, 
Virginia; and 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 
2, Located in Surry County, Virginia

    Date of amendment request: May 6 and February 10, 2010.
    Brief description of amendment request: The proposed amendments 
will add Optimized ZIRLO as an acceptable fuel rod cladding material 
and in addition, propose adding the Westinghouse topical report for 
Optimized ZIRLO to the analytical methods used to determine the core 
operating limits listed in the Technical Specifications.
    Date of publication of individual notice in Federal Register: 
August 27, 2010 (75 FR 52781).
    Expiration date of individual notice: Comments, September 27, 2010; 
Hearing, October 26, 2010.

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

[[Page 61529]]

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 Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management System (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1-(800) 397-4209, (301) 415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#750511075b0710061a00071610351b07165b121a03"><span class="__cf_email__" data-cfemail="aadaced884d8cfd9c5dfd8c9cfeac4d8c984cdc5dc">[email protected]</span></a>.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of application for amendment: October 30, 2009, as 
supplemented by letters dated April 29 and August 24, 2010.
    Brief description of amendment: The amendments consisted of 
administrative changes to update the licenses and the technical 
specifications as a result of changes that were approved in previously 
issued amendments. The amendments removed requirements that are no 
longer applicable due to the completion of power uprates, the 
replacement of steam generators, the removal of part-length control 
element assemblies, the completion of the core protection calculator 
upgrade, and made a minor administrative change to the nomenclature of 
the containment sump trash racks and screens.
    Date of issuance: September 10, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: Unit 1-179; Unit 2-179; Unit 3-179.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendment revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2010 (75 FR 
4113). The supplemental letters dated April 29 and August 24, 2010, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 10, 2010.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: November 19, 2009.
    Brief description of amendment: The amendment revises the charcoal 
testing criteria in Technical Specification 5.5.9, ``Ventilation Filter 
Testing Program.''
    Date of issuance: September 13, 2010.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 265.
    Facility Operating License No. DPR-26: The amendment revised the 
License and the Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2010 (75 FR 
4115).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 13, 2010.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: September 9, 2009.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3/4 .9.7, ``Crane Travel--Fuel Handling Building,'' 
to permit certain operations needed for dry cask storage of spent 
nuclear fuel. Specifically, the proposed change to this TS, while 
continuing to prohibit travel of a heavy load over irradiated fuel 
assemblies in the spent fuel pool, would permit travel of loads in 
excess of 2,000 pounds over a transfer cask containing irradiated fuel 
assemblies, provided a single-failure-proof handling system is used.
    Date of issuance: September 13, 2010.
    Effective date: As of the date of issuance and shall be implemented 
prior to the start of the dry cask storage operations.
    Amendment No.: 227.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: November 17, 2009 (74 
FR 59261). The supplemental letters dated June 8 and July 22, 2010, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 13, 2010.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: October 22, 2009.
    Brief description of amendment: The amendment modified the 
Technical Specifications (TS) Table 2.2-1 and Table 3.3-1. 
Specifically, the TS changes clarify TS Table 2.2-1 Notes (1) and (5), 
TS Table 3.3-1 Notes (a) and (c), and TS Table 3.3-1 Actions 2 and 3, 
which have resulted in Plant Protection System redundancy issues with 
respect to verbatim compliance. While the changes modified the table 
notations for the 10<SUP>-4</SUP> percent Bistable in the Tables, they 
still maintain the safety function associated with the Core Protection 
Calculators and High Logarithmic Power trip functions, and with the 
small hysteresis for the 10<SUP>-4</SUP> percent Bistable, there is a 
negligible impact on the Control Element Assembly withdrawal analysis. 
Additionally, the calculated peak power and heat flux are not 
significantly changed.
    Date of issuance: September 13, 2010.

[[Page 61530]]

    Effective date: As of the date of issuance and shall be implemented 
90 days from the date of issuance.
    Amendment No.: 228.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 15, 2009 (74 
FR 66384).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 13, 2010.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2 (Braidwood), Will County, Illinois; 
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2 
(Byron), Ogle County, Illinois

    Date of application for amendment: March 29, 2010.
    Brief description of amendment: The amendments revise Technical 
Specification (TS) 5.5.7, ``Reactor Coolant Pump Flywheel Inspection 
Program,'' to extend the reactor coolant pump (RCP) motor flywheel 
examination frequency from the currently-approved 10-year inspection 
interval to an interval not to exceed 20 years for certain Braidwood 
and Byron RCPs. These changes are consistent with TS Task Force (TSTF) 
traveler TSTF-421, ``Revision to RCP Flywheel Inspection Program (WCAP-
15666),'' Revision 0, that has been approved generically for the 
Westinghouse Standard Technical Specifications, NUREG-1431. A notice 
announcing the availability of this proposed TS change using the 
Consolidated Line Item Improvement Process was published in the Federal 
Register on October 22, 2003 (68 FR 60422).
    Date of issuance: September 16, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: Braidwood Unit 1-163; Braidwood Unit 2-163; Byron 
Unit No. 1-169; and Byron Unit No. 2-169.
    Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: 
The amendments revise the TSs and Licenses.
    Date of initial notice in Federal Register: May 18, 2010 (75 FR 
27827).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 16, 2010.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois

    Date of application for amendment: April 2, 2010, as supplemented 
by letters dated June 19, 2009, and March 31, 2010.
    Brief description of amendment: The amendment revises the Exelon 
Nuclear Radiological Emergency Plan Annex for Clinton Station, Table B-
1, ``Minimum Staffing Requirements for the On-Shift Clinton Station 
ERO,'' to increase the Non-Licensed Operator staffing from two to four, 
allow in-plant protective actions to be performed by personnel assigned 
to other functions, and replace a Mechanical Maintenance person with a 
Non-Licensed Operator.
    Date of issuance: September 21, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 191.
    Facility Operating License No. NPF-62: The amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: June 1, 2010 (75 FR 
30445).
    The June 19, 2009, and March 31, 2010, supplement, 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 amendment is contained 
in a Safety Evaluation dated September 21, 2010.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 27, 2010, as 
supplemented by letters dated May 12, and May 13, 2010.
    Brief description of amendments: The amendments would revise the 
Operating License and technical Specifications to implement an increase 
of approximately 1.65 percent in rated thermal power from the current 
licensed thermal power of 3489 megawatts thermal (MWt) to 3546 MWt.
    Date of issuance: September 16, 2010.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days for Unit 1 and within 90 days of completion of refueling 
outage L2R13, which is currently scheduled for March 2011, for Unit 2.
    Amendment Nos.: 198/185.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: May 11, 2010 (75 FR 
26289).
    The May 12, and May 13, 2010, 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 September 16, 2010.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 23rd day of September 2010.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2010-24815 Filed 10-4-10; 8:45 am]
BILLING CODE 7590-01-P


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