[Federal Register Volume 74, Number 133 (Tuesday, July 14, 2009)]
[Notices]
[Pages 34044-34052]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-16347]


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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0306]


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 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 18, 2009, to July 1, 2009. The last 
biweekly notice was published on June 30, 2009 (74 FR 31318).

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

[[Page 34045]]

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, Rulemaking 
and Directives Branch (RDB), 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 RDB 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 petitioner/requestor 
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 
petitioner/requestor 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 petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner/requestor is aware and on 
which the petitioner/requestor 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 petitioner/requestor to relief. 
A petitioner/requestor 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, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). 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 petitioner/requestor 
should contact the Office of the Secretary by e-mail at 
<a href="/cdn-cgi/l/email-protection#29414c485b40474e074d464a424c5d69475b4a074e465f"><span class="__cf_email__" data-cfemail="b1d9d4d0c3d8dfd69fd5ded2dad4c5f1dfc3d29fd6dec7">[email protected]</span></a>, or by calling (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/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the

[[Page 34046]]

petitioner/requestor (or its counsel or representative) already holds 
an NRC-issued digital ID certificate). Each petitioner/requestor will 
need to download the Workplace Forms Viewer<SUP>TM</SUP> to access the 
Electronic Information Exchange (EIE), a component of the E-Filing 
system. The Workplace Forms Viewer<SUP>TM</SUP> is free and is 
available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. 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.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it 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 filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE 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 
EIE 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 through the ``Contact Us'' link 
located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC Meta-System Help Desk, which is 
available between 8 a.m. and 8 p.m., Eastern Time, Monday through 
Friday, excluding government holidays. The Meta-System Help Desk can be 
contacted by telephone at 1-866-672-7640 or by e-mail at 
<a href="/cdn-cgi/l/email-protection#b8f5ebf0fc96eaddcbd7cdcadbddf8d6cadb96dfd7ce"><span class="__cf_email__" data-cfemail="3d706e7579136f584e52484f5e587d534f5e135a524b">[email protected]</span></a>.
    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.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the request and/
or petition should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii).
    Documents submitted in adjudicatory proceedings will appear in 
NBC'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, an Atomic Safety and Licensing Board, or 
a 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 submissions.
    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 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#730317015d0116001c06011016331d01105d141c05"><span class="__cf_email__" data-cfemail="95e5f1e7bbe7f0e6fae0e7f6f0d5fbe7f6bbf2fae3">[email protected]</span></a>.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendments request: May 13, 2009.
    Description of amendments request: The amendment would incorporate 
Technical Specification Task Force Traveler (TSTF) 479-A, Revision 0, 
``Changes to Reflect Revision of 10 CFR 50.55a,'' and TSTF 497-A, 
Revision 0, ``Limit Inservice Testing Program SR 3.0.2 Application to 
Frequencies of 2 Years or Less,'' approved by the Nuclear Regulatory 
Commission (NRC) on December 6, 2005 and October 4, 2006 respectively. 
The proposed changes revise the Improved Standard Technical 
Specification administrative controls of the Inservice Testing Program 
to be consistent with the requirements of Title 10 of the Code of 
Federal Regulations Section 50.55a(f)(4) for pumps and valves 
classified as American Society of Mechanical Engineers (ASME) Code 
Class 1, Class 2, and Class 3.
    The proposed change replaces references to ASME Section XI of the 
Boiler and Pressure Vessel Code with references to the ASME Code for 
Operation and Maintenance of Nuclear Power Plants within Technical 
Specification (TS) 5.5.8. In TS 5.5.8.b, the proposed change applies 
the extension of Surveillance Requirement 3.0.2 to other normal and 
accelerated inservice testing frequencies of 2 years or less that were 
not included in the frequencies of the table listed in TS 5.5.8.a.
    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. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change will replace, within TS 5.5.8, references to 
Section XI of ASME Boiler and Pressure Vessel Code, with references 
to the ASME Code for Operation and Maintenance of Nuclear Power 
Plants (OM Code). In addition the proposed change

[[Page 34047]]

adds words to TS 5.5.8.b which applies the extension allowance of 
the Surveillance Requirement 3.0.2 to other normal and accelerated 
inservice testing frequencies of two years or less that were not 
included in the frequencies of the table listed in TS 5.5.8.a.
    The proposed change is administrative, does not affect any 
accident initiators, does not affect the ability to successfully 
respond to previously evaluated accidents and does not affect 
radiological assumptions used in the evaluations. Thus, operation of 
the facility in accordance with the proposed change will not involve 
an increase in the probability or the 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. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed change will replace, within TS 5.5.8 references to 
Section XI of ASME Boiler and Pressure Vessel Code with references 
to the ASME OM Code. In addition the proposed change also adds words 
to TS 5.5.8.b which applies the extension allowance of Surveillance 
Requirement 3.0.2 to other normal and accelerated inservice testing 
frequencies of two years or less that were not included in the 
frequencies of the table listed in TS 5.5.8.a.
    The proposed change does not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or involve a change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. Additionally there is no change 
in the types or increase in the amounts of any effluent that may be 
released offsite and there is no increase in individual or 
cumulative occupational exposure.
    Therefore the proposed change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed change will replace, within TS 5.5.8 references to 
Section XI of ASME Boiler and Pressure Vessel Code with references 
to the ASME OM Code. In addition the proposed change also adds words 
to TS 5.5.8.b which applies the extension allowance of Surveillance 
Requirement 3.0.2 to other normal and accelerated inservice testing 
frequencies of two years or less that were not included in the 
frequencies of the table listed in TS 5.5.8.a.
    The proposed change does not involve a modification to the 
physical configuration of the operating units or change the methods 
governing normal plant operation. The proposed change incorporates 
revisions to the ASME Code that results in a net improvement in the 
measures for testing pumps and valves. The safety functions of the 
applicable pumps and valves will be maintained.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 
17th floor, Baltimore, MD 21202.
    NRC Acting Branch Chief: Richard Guzman.

DTE Energy, Inc., Docket No. 50-16, Enrico Fermi Atomic Power Plant, 
Unit 1, Monroe County, Michigan

    Date of amendment request: March 25, 2009.
    Description of amendment request: The proposed amendment would add 
a license condition incorporating a site license termination plan (LTP) 
for the Enrico Fermi Atomic Power Plant, Unit 1 (Fermi-1). The proposed 
license condition includes LTP change control criteria specifying when 
changes to the LTP require prior Nuclear Regulatory Commission 
approval. Since Fermi-1 is completely within the boundary of Unit 2, 
the Fermi-1 property would become part of Unit 2 site upon successful 
completion of license termination activities.
    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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. The change allows for the approval of the LTP and provides 
the criteria for when changes to the LTP require prior NRC approval. 
This change does not affect possible initiating events for the three 
postulated accidents previously evaluated in the Fermi-1 Safety 
Analysis Report (SAR), as updated, or alter the configuration or 
operation of the facility. Safety limits, limiting safety system 
settings, and limiting control systems are no longer applicable to 
Fermi-1 in the permanently defueled condition, and are therefore not 
considered further. The proposed change does not affect the 
boundaries used to evaluate compliance with liquid or gaseous 
effluent limits, and has no impact on plant operations.
    Therefore, the proposed license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the probability of a new or 
different accident from any accident previously evaluated?
    No. The safety analysis for the facility remains accurate as 
described in the Fermi-1 SAR, as updated, Section 8.4, Postulated 
Radiological Accidents. There are sections of the LTP that make 
reference to the decommissioning activities still remaining (e.g., 
removal of large components, decontamination, etc.), however these 
activities are performed in accordance with approved Fermi-1 work 
authorizations and undergo 10 CFR 50.59 screening prior to 
initiation. The plant conditions for which the postulated accidents 
have been evaluated are still valid and no new accident scenarios, 
failure mechanisms, or single failures are introduced by this 
amendment. The system operating procedures are not affected.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. There are no changes to the design or operation of the 
facility resulting from this amendment. The proposed change does not 
affect the boundaries used to evaluate compliance with liquid or 
gaseous effluent limits, and has no impact on plant shutdown 
operations. Accordingly, neither the postulated accident assumptions 
in the Fermi-1 SAR, as updated, nor the basis of the Technical 
Specifications are affected.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    Based upon the reasoning presented above 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 Pettinari.
    NRC Branch Chief: Andrew Persinko.

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

    Date of amendment request: June 3, 2009.
    Description of amendment request: The proposed change will modify 
Technical Specification (TS) 2.1.1.1 to account for the Combustion 
Engineering (CE) 16 x 16 Next Generation Fuel (NGF) and different U.S. 
Nuclear Regulatory Commission (NRC) reviewed and approved Departure 
from Nucleate Boiling (DNB) correlations. These new correlations will 
be implemented in the safety analyses for the next fuel cycle of

[[Page 34048]]

operation consistent with NRC-approved methodologies.
    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.
    No changes to plant equipment or operating procedures are 
required due to the change in the safety limit for DNBR [Departure 
from Nucleate Boiling Ratio]. This change does not impact any of the 
accident initiators. The analyses of the reload are performed using 
NRC-approved methodologies to ensure the Specified Acceptable Fuel 
Design Limits (SAFDLs), of which DNBR is one, are not violated. The 
current DNBR setpoint continues to ensure automatic protective 
action and is initiated to prevent exceeding the proposed DNBR 
safety limit.
    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 does not involve any plant modifications or 
change in the way the plant is designed to function. The proposed 
change is not associated with any accident precursor or initiator. 
The proposed change supports the loading and use of Next Generation 
Fuel (NGF) at Waterford 3 as previously approved by the NRC.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The use of the NRC-approved NGF WSSV-T correlation with the ABB-
NV correlation to establish a new bounding DNBR safety limit of 
1.24, preserves the DNBR margin of safety at a 95/95 level. The Core 
Protection Calculator (CPC) DNBR power adjustment addressable 
constant BERR1 is calculated based on the WSSV-T and ABB-NV CHF 
[Critical Heat Flux] correlations such that a CPC trip at a DNBR of 
1.26 using the CE-1 CHF correlation ensures that the bounding DNBR 
safety limit of 1.24 for the WSSV-T and ABB-NV CHF correlations will 
not be violated during normal operation and AOOs [anticipated 
operational occurrences] to at least a 95/95 probability/confidence 
level.
    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: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

FPL 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 amendment request: April 17, 2009.
    Description of amendment request: The proposed change would amend 
the Renewed Facility Operating Licenses DPR-24 and DPR-27 for Point 
Beach Nuclear Plant, Units 1 and 2, respectively, to reflect a change 
in the legal name of the Licensee from ``FPL Energy Point Beach, LLC'' 
to ``NextEra Energy Point Beach, LLC.''
    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.
    This request is for administrative changes only. No actual 
facility equipment or accident analyses will be affected by the 
proposed changes. Therefore, this request will have no impact on 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.
    This request is for administrative changes only. No actual 
facility equipment or accident analyses will be affected by the 
proposed changes and no failure modes not bounded by previously 
evaluated accidents will be created. Therefore, this request will 
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.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. This request is for 
administrative changes only. No actual plant equipment or accident 
analyses will be affected by the proposed changes. Additionally, the 
proposed changes will not relax any criteria used to establish 
safety limits, will not relax any safety system settings, and will 
not relax the bases for any limiting conditions of operation. 
Therefore, these proposed changes will 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: Antonio Fernandez, Esquire, Senior Attorney, 
FPL Energy Point Beach, LLC, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Lois M. James.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, 
Unit 2, Hamilton County, Tennessee

    Date of amendment request: May 21, 2009 (TSC 09-02).
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) Section 6.8.4.k, ``Steam 
Generator Program,'' for Unit 2 including associated TSs Bases 3/4.4.5, 
``Steam Generator (SG) Tube Integrity,'' to allow the implementation of 
SG tubing alternate repair criteria for axial indications in the 
Westinghouse Electric Company explosive tube expansion region below the 
top of the tubesheet and specify the W* distance for the SG cold legs.
    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.
    Of the various accidents previously evaluated, the proposed 
changes only affects the steam generator tube rupture (SGTR) event 
evaluation and the postulated steam line break (SLB) accident 
evaluation. Loss-of-coolant accident (LOCA) conditions cause a 
compressive axial load to act on the tube. Therefore, since the LOCA 
tends to force the tube into the tubesheet rather than pull it out, 
it is not a factor in this amendment request. Another faulted load 
consideration is a safe shutdown earthquake (SSE); however, the 
seismic analysis of Westinghouse 51 Series steam generators (SGs) 
has shown that axial loading of the tubes is negligible during an 
SSE.
    TVA's [Tennessee Valley Authority's] amendment request allows 
taking credit for

[[Page 34049]]

how the tubesheet enhances the tube integrity in the Westinghouse 
Electric Company explosive tube expansion (WEXTEX) region by 
precluding tube deformation beyond its initial expanded outside 
diameter. For the SGTR and SLB events, the required structural 
margins of the SG tubes will be maintained due to the presence of 
the tubesheet. Tube rupture is precluded for axial cracks in the 
WEXTEX region due to the constraint provided by the tubesheet. 
Therefore, the normal operating 3[Delta]P margin and the postulated 
accident 1.43[Delta]P margin against burst are maintained.
    The W* length supplies the necessary resistive force to preclude 
pullout loads under both normal operating and accident conditions. 
The contact pressure results from the WEXTEX process, thermal 
expansion mismatch between the tube and tubesheet and from the 
differential pressure between the primary and secondary side. 
Therefore, the proposed change results in no significant increase in 
the probability or the occurrence of an SGTR or SLB accident.
    The proposed changes do not affect other systems, structures, 
components or operational features. Therefore, based on the above 
evaluation, the proposed changes do not involve a significant 
increase in the probability of an accident previously evaluated.
    The consequences of an SGTR event are primarily affected by the 
primary-to-secondary flow rate and the time duration of the primary-
to-secondary flow during the event. Primary-to-secondary flow rate 
through a postulated ruptured tube (i.e., complete severance of a 
single SG tube) is not affected by the proposed change since the 
flow rate is based on the inside diameter of a SG tube and the 
pressure differential. TVA's amendment request does not change 
either of these. The duration of primary-to-secondary leakage is 
based on the time required for an operator to determine that an SGTR 
has occurred, the time to identify and isolate the faulty SG, and 
ensure termination of radioactive release to the atmosphere from the 
faulty SG. TVA's amendment request does not affect the duration of 
the primary-to-secondary leakage because it does not change the 
control room indicators with which an operator would determine that 
an SGTR has occurred. The consequences of an SGTR are secondarily 
affected by primary-to-secondary leakage, which could occur due to 
axial cracks remaining in service in the WEXTEX region in a non-
faulted SG. During a SGTR, the primary-to-secondary differential 
pressure is less than or equal to the normal operating differential 
pressure; therefore, the primary-to-secondary leakage due to axial 
cracks in the WEXTEX region of a non-faulted SG during a SGTR would 
be less than or equal to the primary-to-secondary leakage 
experienced during normal operation. Primary-to-secondary leakage is 
considered in the calculation determining the consequences of a SGTR 
and the value is bounding.
    The postulated SLB has the greatest primary-to-secondary 
pressure differential, and therefore could experience the greatest 
primary-to-secondary leakage. TVA's amendment request allows axial 
cracks to remain in service in the WEXTEX region, which have the 
possibility of primary-to-secondary leakage during a postulated SLB 
accident. However, the primary-to-secondary leakage would be limited 
by the amount the crack face can open (compared to a similar free-
span axial crack) and by the restriction resulting from the tube to 
tubesheet contact pressure which create a restricted leakage path 
from the upper tip of the crack to the top of the WEXTEX expansion. 
TVA's amendment request requires the aggregate leakage, (i.e., the 
combined leakage for the tubes with axial cracks in the WEXTEX 
region) plus the combined leakage developed by other alternate 
repair criteria (ARC), to remain below the maximum allowable SLB 
primary-to-secondary leakage rate limit such that the doses are 
maintained to less than a fraction of the 10 CFR 100 limits and also 
less than the General Design Criteria (GDC)-19 limits.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    TVA's amendment request does not introduce any physical changes 
to the Sequoyah Unit 2 SGs. TVA's amendment request takes credit for 
how the tubesheet enhances the SG tube integrity in the WEXTEX 
region by precluding tube deformation beyond its initial expanded 
outside diameter and allows axial cracks in the WEXTEX region to 
remain in service if prescribed criteria are met. Removal of the 
existing primary water stress corrosion cracking (PWSCC) axial at 
dented tube support plate ARC incorporates the more conservative TS 
limit for SG tube plugging. A failure to meet SG tube integrity 
results in an SGTR. Because the circumferential cracks detected 
within the WEXTEX region are required to be plugged, it is highly 
unlikely that a W* tube would fail as a result of a circumferential 
defect. Therefore, a tube severance which would strike neighboring 
tubes and create a multiple tube rupture is not credible.
    The proposed change does not introduce any new equipment or any 
change to existing equipment. No new effects on existing equipment 
are created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The amendment request maintains the structural margins of the SG 
tubes for both normal and accident conditions that are required by 
Regulatory Guide 1.121.
    For primarily axially oriented cracking located within the 
tubesheet, tube burst is precluded because of the presence of the 
tubesheet. WCAP-14797 defines a length, W*, of degradation free 
expanded tubing that provides the necessary resistance to tube 
pullout due to the pressure induced forces (with applicable safety 
factor applied). Application of the W* criteria will preclude 
unacceptable primary-to-secondary leakage during all plant 
conditions. The methodology for determining leakage provides for 
large margins between calculated and actual leakage values in the W* 
criteria.
    Plugging of the SG tubes reduces the reactor coolant flow margin 
for core cooling. Implementation of the proposed changes is expected 
to result in plugging of fewer tubes than with the current criteria. 
Thus, implementation of the proposed changes will maintain the 
margin of flow that may have otherwise been reduced by tube 
plugging.
    It is concluded that the proposed changes do not result in a 
significant reduction of margin with respect to plant safety as 
defined in the Updated Final Safety Analysis Report or TSs.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Thomas H. Boyce.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: April 13, 2009.
    Description of amendment request: The proposed change revises 
Technical Specifications (TS) Table 3.7-1 Operator Action 3.b and 
provides direction if the less-than-required-minimum-operable-channels-
condition for the nuclear flux intermediate range occurs between 7% and 
11% of rated power.
    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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed change revises TS Table 3.7-1 Operator Action 
3.b and provides direction if the less-than-required-minimum-
operable-channels-condition for the nuclear flux intermediate range 
occurs between 7% and 11% of rated power. Required action between 7% 
and 11% of rated power is currently not addressed in the Operator 
Action 3.b. The proposed TS change does not involve a physical 
change to any structures, systems, or components (SSCs) at Surry 
Power Station; nor does it change any of the previously evaluated 
accidents in the

[[Page 34050]]

Updated Final Safety Analysis Report (UFSAR). Thus, this change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    No. The proposed change revises TS Table 3.7-1 Operator Action 
3.b, and provides required action between 7% and 11% of rated power, 
which is currently not addressed in the Operator Action 3.b. The 
proposed change does not involve a physical change to any SSCs, and 
there is no impact on their design function. The proposed change 
does not affect initiators of analyzed events. Therefore, the 
proposed change does not introduce any new failures that could 
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?
    No. The proposed change provides required action for 
inoperability of nuclear flux intermediate range instrumentation 
between 7% and 11% of rated power in TS Table 3.7-1 Operator Action 
3.b. Margin of safety is established through the design of plant 
SSCs, the parameters within which the plant is operated, and the 
establishment of the setpoints for the actuation of equipment relied 
upon to respond to an event. The proposed change does not impact the 
condition or performance of SSCs relied upon for accident mitigation 
or any safety analysis assumptions. Therefore, the proposed 
amendment 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: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 
23219.
    NRC Branch Chief: Melanie C. Wong.

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.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of amendment request: May 31, 2009.
    Description of amendment request: Revise Technical Specification 
3.1.3.4, related to requirements for Control Element Assembly (CEA) 
drop time to increase the available margin for CEA drop time testing.
    Date of publication of individual notice in the Federal Register: 
June 1, 2009 (74 FR 26261).
    Expiration date of individual notice: July 1, 2009 (Comments) and 
July 31, 2009 (Hearing).

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the 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 Systems (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 at1 (800) 397-4209, (301) 415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#5a2a3e2874283f29352f28393f1a342839743d352c"><span class="__cf_email__" data-cfemail="651501174b1700160a10170600250b17064b020a13">[email protected]</span></a>.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of application for amendment: September 12, 2008.
    Brief description of amendment: The amendment revises Fermi 2 Plant 
Operating License, Appendix A, Technical Specifications (TS) to remove 
statements relating to Nuclear Power Plant Staff Working Hours in 
Section 5.2.2, ``Unit Staff,'' specifically deleting subsection 
5.2.2.e. The requested amendment removes references to and limits 
imposed by Nuclear Regulatory Commission Generic Letter (GL) 82-12, 
``Nuclear Power Plant Staff Working Hours'' from the administrative 
controls section of the Fermi 2 TS. The regulations in 10 CFR Part 26, 
Subpart I, supersede the guidelines in GL 82-12.
    Date of issuance: June 19, 2009.
    Effective date: As of the date of issuance and shall be implemented 
no later than October 1, 2009.
    Amendment No.: 181.
    Facility Operating License No. NPF-43: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: December 2, 2008 (73 FR 
73353).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 19, 2009.
    No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: January 14, 2009.
    Brief description of amendment: The amendment: (1) Deleted 
Technical Specification (TS) surveillance requirement (SR) 3.1.3.2 and 
revised SR 3.1.3.3, (2) removed the reference to SR 3.1.3.2 from 
Required Action A.2 of TS

[[Page 34051]]

3.1.3, ``Control Rod OPERABILITY,'' (3) renumbered SRs 3.1.3.3 through 
3.1.3.5 to reflect the deletion of SR 3.1.3.2, and (4) revised Example 
1.4-3 in Section 1.4, ``Frequency,'' to clarify the applicability of 
the 1.25 surveillance test interval extension. The changes are in 
accordance with NRC-approved TS Task Force (TSTF) change traveler TSTF-
475, Revision 1, ``Control Rod Notch Testing Frequency and SRM Insert 
Control Rod Action.''
    Date of issuance: June 30, 2009.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 212.
    Facility Operating License No. NPF-21: The amendment revised the 
Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: February 10, 2009 (74 
FR 6665).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 30, 2009.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 25, 2008, as supplemented by 
letters dated April 14 and 29, 2009.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.7.5, ``Main Turbine Bypass System.'' The change 
provides an alternative to the existing Limiting Condition for 
Operation for the Main Turbine Bypass System (MTBS). The revised TS 
will require that the MTBS be operable or that the Average Planar 
Linear Heat Generation Rate, the Minimum Critical Power Ratio, and the 
Linear Heat Generation Rate limits for the inoperable MTBS be placed in 
effect as specified in the Core Operating Limits Report.
    Date of issuance: June 24, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 163.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 11, 2008 (73 FR 
13023). The supplemental letters dated April 14 and 29, 2009, 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 24, 2009.
    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: June 21, 2007, as supplemented 
by letter dated January 30, 2009.
    Brief description of amendment: A change is proposed to the 
technical specifications (TSs) of Clinton Power Station, Unit 1, 
consistent with TS Task Force (TSTF) change TSTF-423 to the standard 
technical specifications (STSs) for boiling-water reactor plants to 
allow, for some systems, entry into hot shutdown rather than cold 
shutdown to repair equipment, if risk is assessed and managed 
consistent with the program in place for complying with the 
requirements of Title 10 of the Code of Federal Regulations (10 CFR) 
Section 50.65(a)(4). The proposed amendment would modify the TS to 
risk-informed requirements regarding selected required action end 
states provided in TSTF-423, Revision 0, ``Technical Specification End 
States, NEDC-32988-A.''
    Date of issuance: June 26, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 187.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: February 12, 2008 (73 
FR 8070). The January 30, 2009, 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 June 26, 2009.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: June 30, 2008, as supplemented 
on August 13, 2008.
    Brief description of amendments: The proposed amendment modified 
Technical Specification (TS) requirements related to Refueling Water 
Tank (RWT) minimum contained volume of borated water. The proposed 
changes would make permanent the current administrative RWT minimum 
level of 32.5 feet for both units.
    Date of Issuance: June 30, 2009.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 209 and 157.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 10, 2009 (74 
FR 6665). The supplement dated August 13, 2008, 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 June 30, 2009.
    No significant hazards consideration comments received: No.

Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota

    Date of application for amendments: June 26, 2008, as supplemented 
by letters dated March 16 and May 1, 2009.
    Brief description of amendments: The amendments revise the Facility 
Operating Licenses by revising the licensing basis loss-of-coolant 
accident and main steam line break accident radiological dose 
consequences as currently described in the Prairie Island Nuclear 
Generating Plant Units 1 and 2 Updated Safety Analysis Report. The 
amendments also make concomitant amendments to Technical Specifications 
(TSs) 3.3.5, 3.4.17, and 3.6.3, which are necessary to implement the 
revised analyses.
    Date of issuance: June 19, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 191, 180.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 9, 2008 (73 
FR 52420). The supplemental letters contained clarifying information 
and did not change the initial no significant hazards consideration 
determination, and did not expand the scope of the original Federal 
Register notice.

[[Page 34052]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 19, 2009.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of application for amendments: June 27, 2008, as supplemented 
letters dated August 13, 2008, and February 5, 2009.
    Brief description of amendments: The changes consist of revisions 
to the Technical Specifications in support of the replacement of the 
steam generators (SGs) at San Onofre Nuclear Generating Station 
(SONGS), Units 2 and 3. The changes reflect revised SG inspection and 
repair requirements, and revised peak containment post-accident 
pressure resulting from installation of the replacement SGs.
    Date of issuance: June 25, 2009.
    Effective date: Upon issuance; to be implemented for Unit 2 prior 
to entry into Mode 4 during the Unit 2 Cycle 16 refueling outage 
return-to-service, and to be implemented for Unit 3 prior to entry into 
Mode 4 during the Unit 3 Cycle 16 refueling outage return-to-service.
    Amendment Nos.: Unit 2--220; Unit 3-213.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: September 23, 2008 (73 
FR 54867). The supplemental letters dated August 13, 2008, and February 
5, 2009, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 25, 2009.
    No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: June 3, 2008, as supplemented by 
letters dated January 9 and 23, 2009.
    Brief description of amendment: The amendment revised Technical 
Specifications (TSs) 3.3.7, 3.3.8, 3.7.10, 3.7.13, 3.8.2, 3.8.5, 3.8.8, 
and 3.8.10. The amendment (1) deleted MODES 5 and 6 from the Control 
Room Emergency Ventilation System and its actuation instrumentation in 
TS 3.7.10 and TS 3.3.7; (2) adopted U.S. Nuclear Regulatory Commission 
(NRC)-approved Technical Specification Task Force (TSTF) change 
traveler TSTF-36-A for TSs 3.3.8, 3.7.13, 3.8.2, 3.8.5, 3.8.8, and 
3.8.10; and (3) added a more restrictive change to the Limiting 
Condition for Operation (LCO) applicability for TSs 3.8.2, 3.8.5, 
3.8.8, and 3.8.10 such that these LCOs apply not only during MODES 5 
and 6, but also during the movement of irradiated fuel assemblies 
regardless of the MODE in which the plant is operating.
    Date of issuance: June 30, 2009.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 192.
    Facility Operating License No. NPF-30: The amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 7, 2008 (73 FR 
58678). The supplemental letters dated January 9 and 23, 2009, 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 30, 2009.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of application for amendments: June 9, 2008.
    Brief description of amendments: These amendments revise the 
Technical Specifications to revise action statements in TS 3.12 for 
insertion limit and shutdown margin requirements, revise the 
applicability for the operability of the rod position indication and 
bank demand position indication systems, revise/add action statements 
for rod position indication, and add action statements for group step 
demand counters.
    Date of issuance: June 25, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 265, 264.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments change the licenses and the technical specifications.
    Date of initial notice in Federal Register: July 29, 2009 (73 FR 
43957).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 25, 2009.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 6th day of July 2009.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
 [FR Doc. E9-16347 Filed 7-13-09; 8:45 am]
BILLING CODE 7590-01-P


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