[Federal Register Volume 72, Number 232 (Tuesday, December 4, 2007)]
[Notices]
[Pages 68206-68224]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-23225]


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NUCLEAR REGULATORY COMMISSION


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 staff) 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 November 8, 2007 to November 21, 2007. The 
last biweekly notice was published on November 20, 2007 (72 FR 65360).

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 10 CFR 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. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license, 
and any person whose interest may be affected by this proceeding and 
who wishes to participate as a party in the proceeding must file a 
written request for a hearing and a petition for leave to intervene.
    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, 
Directives and Editing Branch, 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 delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area O1F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, 
person(s) may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license, and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
via electronic submission through the NRC E-Filing system for a hearing 
and a petition for leave to intervene. 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 01F21, 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 within 60 
days, 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

[[Page 68207]]

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 set forth 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 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, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, 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.
    A request for hearing or a petition for leave to intervene must be 
filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated on August 28, 2007, (72 FR 49139). The E-Filing process 
requires participants to submit and serve 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 a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
[email protected], 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 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(tm) to 
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ 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 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 technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, 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 petition 
and/or request 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). To be timely, filings must be submitted no later 
than

[[Page 68208]]

11:59 p.m. Eastern Time on the due date.
    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, 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. 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.
    For further details with respect to this amendment action, 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 01F21, 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. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County, 
North Carolina
    Date of amendments request: January 22, 2007, as supplemented by 
letters dated June 21, July 18, July 31, and October 15, 2007.
    Description of amendments request: The amendment would revise the 
Technical Specifications to support the transition to AREVA NP fuel and 
core design 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.
    The proposed amendments revise the list of NRC-approved 
analytical methods used to establish core operating limits. Core 
operating limits are established to ensure that fuel design limits 
are not exceeded during operating transients or accidents. The 
analytical methods used to determine core operating limits are those 
methods that have previously been found acceptable by the NRC and 
are required to be listed in the Technical Specification section 
governing the Core Operating Limits Report. The application of these 
NRC-approved analytical methods will continue to ensure that 
acceptable operating limits are established and applied to operation 
of the reactor core.
    The proposed amendments will add a new Technical Specification 
3.2.3, ``Linear Heat Generation Rate (LHGR),'' for fuel bundles, add 
a new definition to Technical Specification 1.1 for LHGR, and revise 
Technical Specifications 3.4.1 and 3.7.6 to incorporate restrictions 
on LHGR when in single recirculation loop operation or with an 
inoperable Turbine Bypass System. These LHGR limits will be 
established using NRC-approved analytical methods to ensure that 
fuel performance during normal, transient, and accident conditions 
is acceptable.
    Based on the above, the proposed amendments do not involve an 
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.
    As previously stated, the proposed amendments support transition 
from Global Nuclear Fuels Americas (GNF-A) fuel and core design and 
analysis services to AREVA NP fuel and core design and analysis 
services. The AREVA NP fuel assemblies which will be used in the 
BSEP Unit 1 and 2 cores will be similar in design to the GNF-A fuel 
that will be co-resident in the cores. The BSEP, Unit 1 and 2 cores 
in which this fuel will operate will be designed to meet all 
applicable design and licensing criteria. Adherence to these design 
and licensing criteria will not introduce any new modes of operation 
or introduce any new accident precursors, and thus will preclude the 
introduction 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 amendments will continue to require that core 
operating limits be determined using NRC-approved analytical 
methods. Acceptable fuel performance is obtained by ensuring that 
the peak cladding temperature (PCT) during a postulated design basis 
loss-of-coolant accident (LOCA) is maintained less than the limits 
specified in 10 CFR 50.46, and that the core remains in a coolable 
geometry following a postulated design basis LOCA. The proposed 
amendments ensure that adequate margin will continue to be 
maintained to the 2200 degree PCT limit of 10 CFR 50.46, and the use 
of NRC-approved analytical methods will continue to ensure 
acceptable fuel performance during normal operations, as well as 
during transient and accident conditions. Therefore, the proposed 
amendments do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Thomas H. Boyce.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina
    Date of amendments request: August 6, 2007.
    Description of amendments request: The amendment would revise the 
Technical Specifications (TSs) to implement Technical Specification 
Task Force (TSTF) Change TSTF-343, Revision 1, which allows the 
performance of visual examinations of the primary containment to be 
performed in accordance with the American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, 
Subsections IWE and IWL. The amendment would also make an 
administrative change to the TSs by eliminating a one-time requirement 
to perform containment leak rate testing that has already been 
completed.
    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 affects the frequency of visual examinations 
that will be performed for the concrete surfaces of the containment 
for the purpose of the Primary Containment Leakage Rate Testing 
Program. In addition, the proposed change allows those examinations 
to be performed during power operation as opposed to during a 
refueling outage. The frequency of visual examinations of the 
metallic and concrete surfaces of the containment and the mode of 
operation during which those examinations are performed has no 
relationship to or adverse impact on the probability of any of the 
initiating events assumed in the accident analyses. The proposed 
change would allow

[[Page 68209]]

visual examinations that are performed in accordance with NRC-
approved ASME Section XI Code requirements, except where relief has 
been granted by the NRC, to meet the intent of visual examinations 
specified by Regulatory Guide 1.163, without requiring additional 
visual examinations in accordance with the Regulatory Guide. The 
intent of early detection of deterioration will continue to be met 
by the more vigorous requirements of the Code-required visual 
examinations. As such, the safety function of the containment as a 
fission product barrier is maintained.
    The proposed change also includes the removal of an item in TS 
5.5.12 which was incorporated to establish deadlines for performing 
the performance-based Type A leakage tests in conjunction with 
changing, on a one-time basis, the Type A test frequency. The 
specified Unit 1 and Unit 2 Type A test have been completed. As 
such, removal of this item is an administrative change.
    The proposed change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. The proposed change does not involve the addition or removal 
of any equipment, or any design changes to the facility. Therefore, 
based on the above, the proposed change does not represent 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 the Primary Containment Leakage Rate 
Testing Program in TS 5.5.12 for consistency with the requirements 
of 10 CFR 50.55a(g)(4) for components classified as Code Class MC 
and CC. The proposed change affects the frequency of visual 
examinations that will be performed for the metallic and concrete 
surfaces of containment and allows those examinations to be 
performed during power operation as opposed to during a refueling 
outage.
    The proposed change does not involve a modification to the 
physical configuration of the plants (i.e., no new equipment will be 
installed), and does not revise the methods governing normal plant 
operation. Also, the proposed change will not impose any new or 
different requirements or introduce a new accident initiator, 
accident precursor, or malfunction mechanism.
    The proposed change also includes the removal of an item in TS 
5.5.12 which was incorporated to establish deadlines for performing 
the performance based Type A leakage tests in conjunction with 
changing, on a one-time basis, the Type A test frequency. The 
specified Unit 1 and Unit 2 Type A test have been completed. As 
such, removal of this item is an administrative change.
    As such, 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 the Primary Containment Leakage Rate 
Testing Program in TS 5.5.12 for consistency with the requirements 
of 10 CFR 50.55a(g)(4) for components classified as Code Class MC 
and CC. The proposed change allows some of those examinations to be 
performed during power operation as opposed to during a refueling 
outage. As previously stated, the proposed change does not involve a 
modification to the physical configuration of the plants and does 
not revise the methods governing normal plant operation. As such, 
the safety function of the containment as a fission product barrier, 
will be maintained and is not adversely impacted by the proposed 
change.
    The proposed change also includes the removal of an item in TS 
5.5.12 which was incorporated to establish deadlines for performing 
the performance-based Type A leakage tests in conjunction with 
changing, on a one-time basis, the Type A test frequency. The 
specified Unit 1 and Unit 2 Type A test have been completed. As 
such, removal of this item is an administrative 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336 Millstone Power 
Station, Unit No. 2, New London County, Connecticut
    Date of amendment request: February 20, 2007.
    Description of amendment request: The proposed amendment would 
revise the Millstone Power Station, Unit No. 2 (MPS2) Technical 
Specifications (TS) to eliminate Surveillance Requirement (SR) 4.5.2.e 
which requires flow rate verification for each charging pump. Charging 
pump flow is no longer relied upon for design basis mitigation at MPS2 
and the charging pumps have been classified as non-risk significant in 
the MPS2 Probabilistic Risk Assessment model. Therefore, the proposed 
amendment is requesting to remove the charging pump flow verification 
requirements currently located in the TS SR 4.5.2.e.
    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 FSAR [Final Safety Analysis Report] Chapter 14 accident 
analyses for MPS2 take no credit for the flow delivered by the 
charging pumps. Additionally, the proposed change does not modify 
any plant equipment or method of operation for any system, structure 
or component required for safe operation of the facility or 
mitigation of accidents assumed in the facility safety analyses. As 
such, the proposed amendment does not increase the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not modify any plant equipment or 
method of operation for any system, structure or component required 
for safe operation of the facility or mitigation of accidents 
assumed in the facility safety analyses. As such, no new failure 
modes are introduced by the proposed change. Consequently, the 
proposed amendment does not introduce any accident initiators or 
malfunctions that would cause a new or different kind of accident. 
Therefore, the proposed amendment 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 FSAR Chapter 14 accident analyses for MPS2 take no credit 
for the charging pumps. The TS change does not involve a significant 
reduction in a margin of safety because the proposed change does not 
affect equipment design or operation, and there are no changes being 
made to the technical specification required safety limits or safety 
system settings. The proposed change does not affect any of the 
assumptions used in the accident analysis, nor does it affect any 
method of operation for equipment important to plant safety. 
Therefore, the margin of safety is not impacted by the proposed 
amendment.

    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, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Branch Chief: Harold K. Chernoff.

[[Page 68210]]

Dominion Nuclear Connecticut Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut.
    Date of amendment request: July 2, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specification (TS) 4.0.5 to reference the American 
Society of Mechanical Engineers (ASME) Code for Operation and 
Maintenance of Nuclear Power Plants (OM Code) instead of Section XI of 
the ASME Boiler and Pressure Vessel Code. Specifically, the proposed 
amendment would modify the inservice inspection (ISI) of ASME Code 
Class 1, 2, and 3 components and inservice testing of ASME Code Class 
1, 2, and 3 pumps and valves to reflect the requirements in the ASME OM 
Code. In addition, the redundant requirement in TS 4.0.5 to maintain an 
ISI program is being proposed for removal, based on duplicate 
regulatory requirements set forth in Title 10 of the Code of Federal 
Regulations (10 CFR) Section 50.55a.
    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 change does not modify any plant equipment and does 
not impact any failure modes that could lead to an accident. 
Additionally, the proposed change has no effect on the consequence 
of any analyzed accident since the change does not affect the 
function of any equipment credited for accident mitigation. The 
proposed change incorporates revisions to the ASME Code that result 
in a net improvement in the measures for testing pumps and valves. 
Removing from TS the duplicate requirement in the regulations to 
maintain an ISI program in accordance with ASME codes and standards 
does not impact any accident initiators or analyzed events or 
mitigation of events. No reduction in previous commitments to 10 CFR 
50.55a(g) are being proposed by this change.
    Based on the discussion above, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or adversely affect 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. The proposed change does not 
alter existing test criteria or frequencies. Additionally, there is 
no change in the types or increases in the amounts of any effluent 
that may be released off-site and there is no increase in individual 
or cumulative occupational exposure. The proposed changes 
incorporate revisions to the ASME Code that result in a net 
improvement in the measures for testing pumps and valves. Removal of 
the duplicate TS requirement to maintain an ISI program will not 
alter the commitment to the current ISI program requirements in 10 
CFR 50.55a or any other TS requirements related to inservice 
inspection.
    Based on the discussion above, 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 revises TS 4.0.5 regarding inservice testing 
of ASME Code Class 1, 2, and 3 pumps and valves, for consistency 
with the requirements of 10 CFR 50.55a(f)(4). The proposed change 
incorporates an administrative clarification to the frequencies for 
IST and incorporates revisions to the ASME Code that result in a net 
improvement in the measures for testing pumps and valves. No 
setpoints or safety limit settings are being revised. The safety 
function of the affected pumps and valves will continue to be 
confirmed through inspection and testing. Removal of the ISI program 
requirement from TS 4.0.5 does not remove the requirement from 
regulations, and therefore, will not diminish the current station 
approved programs and procedures that implement the regulatory 
criteria of 10 CFR 50.55a(g) to maintain an acceptable ISI program 
in accordance with the ASME Code.
    Based on the discussion above, 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: Lillian M. Cuoco, Esquire, Senior Nuclear 
Counsel, Dominion Nuclear Connecticut, Inc., Building 475, 5th Floor, 
Rope Ferry Road, Waterford, CT 06141-5127.
    NRC Branch Chief: Harold K. Chernoff.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina
Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Oconee Nuclear Station Independent Spent Fuel Storage Installation NRC 
License No. SNM-2503, Docket No. 72-4, Oconee County, South Carolina
    Date of amendment request: March 14, 2007.
    Description of amendment request: The amendments would revise the 
licenses to reflect the change in the name of the licensee from Duke 
Power Company LLC to Duke Energy Carolinas, LLC. The proposed 
amendments are a name change only. There is no change in the state of 
incorporation, registered agent, registered office, rights or 
liabilities of the company. Nor is there a change in the function of 
the licensee or the way in which it does business.
    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 amendments are for a name change only. The 
amendments do not involve any change in the technical qualifications 
of the licensee or the design, configuration, or operation of the 
nuclear units. All Limiting Conditions for Operation, Limiting 
Safety System Settings and Safety Limits specified in the Technical 
Specifications remain unchanged. Also, the Physical Security Plans 
and related plans, the Operator Training and Requalification 
Programs, the Quality Assurance Programs, and the Emergency Plans 
will not be materially changed by the proposed name change. The name 
change amendments will not affect the executive oversight provided 
by the Chief Nuclear Officer and his staff.
    Therefore, the proposed amendments do not involve any increase 
in the probability or consequences of an accident previously 
analyzed.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.

[[Page 68211]]

    The proposed amendments do not involve any change in the design, 
configuration, or operation of the nuclear plant. The current plant 
design, design bases, and plant safety analysis will remain the 
same.
    The Limiting Conditions for Operations, Limiting Safety System 
Settings and Safety Limits specified in the Technical Specifications 
are not affected by the proposed changes. As such, the plant 
conditions for which the design basis accident analyses were 
performed remain valid.
    The proposed amendments do not introduce a new mode of plant 
operation or new accident precursors, do not involve any physical 
alterations to plant configurations, or make changes to system 
setpoints that could initiate a new or different kind of accident.
    Therefore, the proposed amendments do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendments do not involve a change in the design, 
configuration, or operation of the nuclear plants. The change does 
not affect either the way in which the plant structures, systems, 
and components perform their safety function or their design and 
licensing bases.
    Plant safety margins are established through Limiting Conditions 
for Operation, Limiting Safety System Settings and Safety Limits 
specified in the Technical Specifications. Because there is no 
change to the physical design of the plant, there is no change to 
any of these margins.
    Therefore, the proposed amendments do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Evangelos C. Marinos.
Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington
    Date of amendment request: November 7, 2007.
    Description of amendment request: The proposed amendment would 
delete License Condition 2.F, which requires reporting of violations of 
certain other requirements contained in Section 2.C of the license.
    The NRC staff issued a ``Notice of Availability of Model 
Application Concerning Elimination of Typical License Condition 
Requiring Reporting of Violations of Section 2.C of Operating License 
Using the Consolidated Line Item Improvement Process'' in the Federal 
Register on November 4, 2005 (70 FR 67202). The notice referenced a 
model safety evaluation, a model no significant hazards consideration 
(NSHC) determination, and a model license amendment request published 
in the Federal Register on August 29, 2005 (70 FR 51098). In its 
application dated November 7, 2007, the licensee affirmed the 
applicability of the model NSHC determination which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change involves the deletion of a reporting 
requirement. The change does not affect plant equipment or operating 
practices and therefore does not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change is administrative in that it deletes a 
reporting requirement. The change does not add new plant equipment, 
change existing plant equipment, or affect the operating practices 
of the facility. Therefore, the 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 deletes a reporting requirement. The change 
does not affect plant equipment or operating practices and therefore 
does not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the analysis adopted by the licensee 
and, based upon this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendment involves NSHC.
    Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Thomas G. Hiltz.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2 (IP2), Westchester County, New York
    Date of amendment request: October 24, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) requirements related to the 
containment buffering agent used for pH control under post loss-of-
coolant accident (LOCA) conditions. Specifically, the proposal would 
approve the use of sodium tetraborate (STB) as the buffering agent 
instead of the currently approved compound, trisodium phosphate (TSP). 
The reason for this change in buffering agents is to minimize the 
potential for an adverse chemical interaction between the TSP and 
certain insulation materials in the containment that could degrade flow 
through the sump screens following certain design-basis accident 
scenarios such as a LOCA.
    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 amendment does not involve a significant increase 
in the probability of an accident previously evaluated because the 
containment buffering agent is not an initiator of any analyzed 
accident. The proposed change does not impact any failure modes that 
could lead to an accident.
    The proposed amendment does not involve a significant increase 
in the consequences of an accident previously evaluated. The 
buffering agent in containment is designed to buffer the acids 
expected to be produced after a LOCA and is credited in the 
radiological analysis for iodine retention. Utilizing STB as a 
buffering agent ensures the post LOCA containment sump mixture will 
have a pH >= 7.0. The proposed change of replacing TSP with STB 
results in the radiological consequences remaining within the limits 
of 10 CFR 50.67 as demonstrated by existing analyses of record.
    Therefore, 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.
    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 amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. STB is a passive component that is proposed to be used at 
IP2 as a buffering agent to increase the pH of the initially acidic 
post-LOCA containment water to a more neutral pH. Changing the 
proposed buffering agent from TSP to STB does not constitute an 
accident initiator or create a new or different

[[Page 68212]]

kind of accident previously analyzed. The proposed amendment does 
not involve operation of any required systems, structures or 
components in a manner or configuration different from those 
previously recognized or evaluated. No new failure mechanisms will 
be introduced by the changes being requested.
    Therefore, the proposed amendment 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 amendment does not involve a significant reduction 
in a margin of safety. The proposed amendment of changing the 
buffering agent from TSP to STB results in equivalent control of 
maintaining sump pH at 7.0 or greater, thereby controlling 
containment atmosphere iodine and ensuring the radiological 
consequences of a LOCA are within regulatory limits. The use of STB 
also reduces the potential for exacerbating sump screen blockage due 
to a chemical interaction between TSP and certain calcium sources 
used in containment. This proposed amendment eliminates the 
formation of calcium phosphate precipitate thereby reducing the 
overall amount of precipitate that may be formed in a postulated 
LOCA. The buffer change would minimize the potential chemical 
effects and should enhance the ability of the emergency core cooling 
system to perform the post-accident mitigating functions.
    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: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3 (IP3), Westchester County, New York
    Date of amendment request: October 24, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) requirements regarding the setpoint 
and definition of the low-low level alarm on the Refueling Water 
Storage Tank (RWST). Specifically, the proposal would revise the 
setpoint of the low-low level alarm from a range of greater than or 
equal to 10.5 ft and less than or equal to 12.5 ft to a range of 
greater than or equal to 9.0 ft and less than or equal to 11.0 ft, and 
revise the definition of the RWST ``low level alarm'' to ``low-low 
level alarm.'' The reason for these changes is to ensure that adequate 
water is supplied to the containment floor to eliminate the risk of 
vortexing and/or draw down at the sump strainer modules following a 
small-break loss-of-coolant accident (LOCA). The proposed changes are 
being requested to support resolution of the pressurized-water reactor 
sump performance issue involving debris accumulation, Generic Safety 
Issue (GSI)-191.
    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 changes to the Technical Specifications are 
consistent with the assumptions of all design basis accidents, as 
they exist currently and as affected subsequent to the 
implementation of the proposed amendment. The change in the RWST 
low-low level alarm setpoint range has been demonstrated to be 
within the safety margins for post-accident parameters and, in most 
cases, actually beneficial to plant post-accident response 
capability. The RWST is designed to respond to a variety of 
accidents, and, for operation in Modes 1 through 4, it serves no 
other purpose. Therefore, any adjustment of an intermediate level 
setpoint cannot increase the probability of a design basis accident. 
The change in the definition of the RWST ``low level alarm'' to 
``low-low level alarm'' is editorial and therefore does not affect 
the function of the alarm. Therefore, the proposed changes do 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 changes represent a minor adjustment to an existing 
setpoint range. The effect of the changes will be to assure 
recirculation flow following a LOCA with consideration for sump 
strainer installation, in response to GSI-191. However, the RWST 
will continue to perform its function in essentially the same manner 
that it has since original plant design. No changes in equipment 
operation or procedural control will result from this amendment that 
could possibly degrade the performance of the RWST or cause it to be 
operated in a manner inconsistent with existing design basis 
assumptions. The change in the definition of the RWST ``low level 
alarm'' to ``low-low level alarm'' is editorial and therefore does 
not affect the function of the alarm. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes improve the margin to safety, especially 
with respect to post-accident temperature/pressure and dose 
consequences during injection and, most importantly, pump 
performance under postulated sump debris conditions during 
recirculation. Significant margin is available to preclude air 
ingestion in the ECCS [emergency core cooling system] pumps, and 
sufficient time is available for the operators to perform the 
switchover to recirculation. The change in the definition of the 
RWST ``low level alarm'' to ``low-low level alarm'' is editorial and 
therefore does not affect the function of the alarm. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

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

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

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

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

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania


[[Page 68213]]


Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of amendment request: July 19, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Sections 5.3.1/6.3.1, ``Unit Staff 
Qualifications,'' for operator license applicants in accordance with 
current industry standards for education and experience eligibility 
requirements. The proposed amendment would permit changes to the unit 
staff qualification education and experience eligibility requirements 
for licensed operators. The proposal will bring Exelon Generation 
Company, LLC (EGC) and AmerGen Energy Company, LLC (AmerGen) sites in 
alignment with current industry practices and facilitate the 
development of a pre-initial licensed operator training program.
    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 the 
proposed amendment involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    Response: No.
    Licensed operator qualification and training can have an 
indirect impact on accidents previously evaluated. However, the NRC 
considered this impact during the rulemaking process, and by 
promulgation of the revised 10 CFR 55 rule, determined that this 
impact remains acceptable when licensees have an accredited licensed 
operator training program which is based on a systems approach to 
training (SAT). The NRC has concluded in RIS [Regulatory Issue 
Summary] 2001-01 and NUREG-1021 that standards and guidelines 
applied by INPO [the Institute of Nuclear Power Operations] in their 
accredited training programs are equivalent to those put forth by or 
endorsed by the NRC. Therefore, maintaining an INPO accredited SAT 
licensed operator training program is equivalent to maintaining an 
NRC approved licensed operator training program which conforms with 
applicable NRC Regulatory Guidelines or NRC endorsed industry 
standards. The proposed changes conform to ACAD [air containment 
atmosphere distribution] 00-003, Revision 1 licensed operator 
education and experience eligibility requirements.
    Based on the above, EGC and AmerGen conclude that the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Will operation of the facility in accordance with the 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment involves changes to the licensed operator 
training programs, which are administrative in nature. The EGC and 
AmerGen licensed operator training programs have been accredited by 
INPO and are based on SAT.
    Based on the above discussion, EGC and AmerGen conclude that the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed TS changes are administrative in nature. The 
proposed TS changes do not affect plant design, hardware, system 
operation, or procedures for accident mitigation systems. The 
proposed changes do not impact the performance or proficiency 
requirements for licensed operators. As a result, the ability of the 
plant to respond to and mitigate accidents is unchanged by the 
proposed TS changes. Therefore, these changes do not involve a 
significant reduction in a margin of safety.
    Based on the above, EGC and AmerGen conclude that the proposed 
changes do not involve a significant reduction in a margin of 
safety.
    Based on the above evaluation of the three criteria, EGC and 
AmerGen conclude that the proposed amendment presents no significant 
hazards consideration under the standards set forth in 10 CFR 
50.92(c), and, accordingly, a finding of ``no significant hazards 
consideration'' is justified.

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

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

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

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Exelon Generation Company, LLC, Docket No. 50-352 and No. 50-353, 
Limerick Generating Station, Unit 1 and 2, Montgomery County, 
Pennsylvania

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3,York 
and Lancaster Counties, Pennsylvania

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of amendment request: August 8, 2007.
    Description of amendment request: The proposed amendment replaces 
references to Section XI of the American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code with references to the 
ASME Code for Operation and Maintenance of Nuclear Power Plants (OM 
Code) in the applicable technical specification (TS) section for the 
Inservice Testing Program (IST) for the Exelon Generation Company, LLC, 
and AmerGen Energy Company, LLC, (the licensees) plants that have 
implemented industry Improved Technical Specifications. The proposed 
changes are based on TS Task Force (TSTF) 479-A, Revision 0, ``Changes 
to Reflect Revision of 10 CFR 50.55a,'' as modified by TSTF-497, 
Revision 0, ``Limit Inservice Testing Program SR [Surveillance 
Requirement] 3.0.2 Application to Frequencies of 2 Years or Less.'' In 
addition, the proposed amendment adds a provision in the applicable TS 
section to only apply the extension allowance of SR 3.0.2 to the 
frequency table listed in the TS as part of the IST and to normal and 
accelerated inservice testing frequencies of two years or less, as 
applicable.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise the applicable TS Section to conform 
to the requirements of 10 CFR 50.55a, ``Codes and

[[Page 68214]]

standards,'' paragraph (f) regarding the inservice testing of pumps 
and valves. The current TS reference the ASME Boiler and Pressure 
Vessel Code, Section XI, requirements for the inservice testing of 
ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes 
would reference the ASME OM Code as applicable, which is consistent 
with 10 CFR 50 .55a, paragraph (f), ``Inservice testing 
requirements.'' In addition, the proposed changes clarify that the 
extension allowance of SR 3.0.2 only applies to the frequency table 
listed in the TS, if applicable, as part of the Inservice Testing 
Program and to normal and accelerated inservice testing frequencies 
of two years or less. The definitions of the frequencies are not 
changed by this license amendment request.
    The proposed changes are administrative in nature, do not affect 
any accident initiators, do not affect the ability to successfully 
respond to previously evaluated accidents and do not affect 
radiological assumptions used in the evaluations. Thus, the 
probability or radiological consequences of any accident previously 
evaluated are not increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes revise the applicable TS Section to conform 
to the requirements of 10 CFR 50.55a(f) regarding the inservice 
testing of pumps and valves. The current TS Section references the 
ASME Boiler and Pressure Vessel Code, Section XI, requirements for 
the inservice testing of ASME Code Class 1, 2, and 3 pumps and 
valves. The proposed changes would reference the ASME OM Code as 
applicable, which is consistent with 10 CFR 50.55a(f). In addition, 
the proposed changes clarify that the extension allowance of SR 
3.0.2 only applies to the frequency table listed in the TS, if 
applicable, as part of the Inservice Testing Program and to normal 
and accelerated inservice testing frequencies of two years or less. 
The definitions of the frequencies are not changed by this license 
amendment request.
    The proposed changes to the applicable TS Section do not affect 
the performance of any structure, system, or component credited with 
mitigating any accident previously evaluated and do not introduce 
any new modes of system operation or failure mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed changes revise the applicable TS Section for 
Braidwood Station Units 1 and 2, Byron Station Units 1 and 2, 
Dresden Nuclear Power Station Units 2 and 3, Limerick Generating 
Station Units 1 and 2, Oyster Creek Generating Station, Peach Bottom 
Atomic Power Station Units 2 and 3, Quad Cities Nuclear Power 
Station Units 1 and 2, and Three Mile Island Unit 1 to conform to 
the requirements of 10 CFR 50.55a(f) regarding the inservice testing 
of pumps and valves.
    The current TS Section references the ASME Boiler and Pressure 
Vessel Code, Section XI, requirements for the inservice testing of 
ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes 
would reference the ASME OM Code as applicable, which is consistent 
with the 10 CFR 50.55a(f). In addition, the proposed changes clarify 
that the extension allowance of SR 3.0.2 only applies to the 
frequency table listed in the TS, if applicable, as part of the 
Inservice Testing Program and to normal and accelerated inservice 
testing frequencies of two years or less. The definitions of the 
frequencies are not changed by this license amendment request.
    The proposed changes do not modify the safety limits or 
setpoints at which protective actions are initiated and do not 
change the requirements governing operation or availability of 
safety equipment assumed to operate to preserve the margin of 
safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois 
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station 
(QCNPS), Units 1 and 2, Rock Island County, Illinois
    Date of application for amendment request: August 1, 2007.
    Description of amendment request: The proposed amendment would 
revise the technical specification (TS) allowable value (AV) for the 
Reactor Protection System (RPS) Instrumentation Function 10, ``Turbine 
Condenser Vacuum--Low,'' specified in TS Table 3.3.1.1-1, ``Reactor 
Protection System Instrumentation.'' The proposed amendment also 
revises the Channel Functional Test (CFT) and Channel Calibration (CC) 
Surveillance Test Interval (STI) for DNPS TS Table 3.3.1.1-1, Function 
10. As part of the DNPS STI revision, surveillance requirement (SR) 
3.3.1.10, ``Channel Calibration,'' which is specific to the Turbine 
Condenser Vacuum--Low instrument function, is deleted since it is no 
longer applicable.
    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. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Revision of Allowable Value
    The proposed license amendment implements a revised AV for the 
Turbine Condenser Vacuum--Low scram instrument function at DNPS, 
Units 2 and 3 and QCNPS, Units 1 and 2.
    The proposed changes to the DNPS and QCNPS Turbine Condenser 
Vacuum--Low scram AV do not require modification [of] any system 
interface or affect the probability of any event initiators at the 
facilities. Overall RPS performance will remain within the bounds of 
the previously performed accident analyses, since no hardware 
changes are proposed.
    There will be no degradation in the performance of, or an 
increase in the number of challenges imposed on safety-related 
equipment that are assumed to function during an accident situation. 
The proposed changes will not alter any assumptions or change any 
mitigation actions in the radiological consequence evaluations in 
the Updated Final Safety Analysis Report. The proposed changes are 
consistent with safety analysis assumptions and resultant 
consequences.
    For these reasons, the proposed DNPS and QCNPS AV changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    Relaxation of STIs (DNPS only)
    The proposed license amendment implements a revised CFT and CC 
STI for the Turbine Condenser Vacuum--Low scram instrument function 
at DNPS Units 2 and 3. The proposed DNPS TS change to increase the 
CFT STI for the Turbine Condenser Vacuum--Low scram instrument 
function is based on an analytical method that has been reviewed and 
approved by the NRC [Nuclear Regulatory Commission].
    The proposed change to relax the CFT STI implements 
recommendations from a generic evaluation that was developed by 
General Electric (GE) and the Boiling Water Reactor Owners' Group 
(BWROG), and subsequently approved by the NRC. This licensing 
topical report (LTR) assessed the reliability of TS actuation 
instrumentation and concluded that extending AOTS [allowed outage 
times] and CFT STIs for test and repair activities would enhance 
operational safety.
    The proposed DNPS TS change to increase the CC STI for the 
Turbine Condenser Vacuum--Low scram instrument function is based 
upon a revised setpoint error analysis that provides revised AVs, 
trip setpoints, and Expanded Tolerances (ETs) for the instrument. 
These new AVs, trip setpoints,

[[Page 68215]]

and ETs establish increased design margin between the nominal trip 
setpoint and the AV. This increased design margin, combined with 
historical CC data, provides adequate assurance that the component 
will remain operable when necessary for the prevention or mitigation 
of accidents or transients.
    The TS requirements that govern operability or routine testing 
of plant instruments are not assumed to be initiators of any 
analyzed event because these instruments are intended to prevent, 
detect, or mitigate accidents. Therefore, these proposed STI changes 
will not involve an increase in the probability of occurrence of an 
accident previously evaluated. Additionally, these changes will not 
increase the consequences of an accident previously evaluated 
because the proposed changes do not involve any physical changes to 
plant systems, structures or components (SSCs), or the manner in 
which these SSCs are operated. These changes will not alter the 
operation of equipment assumed to be available for the mitigation of 
anticipated operational occurrences (AOOs) by the plant safety 
analysis or licensing basis.
    The proposed deletion of SR 3.3.1.10 is an administrative 
change, since the SR will no longer be applicable to any instrument 
function in DNPS TS Table 3.3.1-1. Therefore, the proposed deletion 
of SR 3.3.1.10 will not impact the testing, calibration, and 
inspection of RPS instrumentation that is necessary to assure that 
the quality of the instrumentation is maintained, that facility 
operation will be within safety limits, and that the limiting 
conditions for operation will be met.
    For these reasons, the proposed DNPS STI changes do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    In summary, the proposed license changes do not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the DNPS and QCNPS Turbine Condenser 
Vacuum--Low scram AV and the DNPS CFT and CC STIs do not affect the 
design, functional performance, or operation of the facility. 
Similarly, the proposed changes do not affect the design or 
operation of any SSCs involved in the mitigation of any accidents, 
nor do they affect the design or operation of any component in the 
facilities such that new equipment failure modes are created.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of these changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed DNPS and QCNPS AV change does not affect the 
acceptance criteria for any analyzed event, nor is there a change to 
any Safety Analysis Limit. There will be no effect on the manner in 
which safety limits, limiting safety system settings, or limiting 
conditions for operation are determined nor will there be any effect 
on those plant systems necessary to assure the accomplishment of 
protection functions. All required design functions are maintained, 
and the AVs, are consistent with NRC-approved methodology and 
guidance for establishment of TS AVs.
    The proposed AV changes do not affect the accident analyses that 
assume operability of the instrument associated with the AV. The 
Turbine Condenser Vacuum--Low scram function is credited in the Loss 
of Main Condenser Vacuum AOO. The loss of main condenser AOO event 
assumes that the main condenser is instantaneously lost while the 
unit is operating at full power. This is classified as a moderate 
frequency event and is described in the UFSAR [updated final safety 
analysis report] as being bounded by the turbine trip with bypass 
failure event.
    The worst case for this AOO would occur if the loss of vacuum 
were instantaneous. In this case, the loss of main condenser event 
would be identical to the turbine trip with bypass failure event. 
During a turbine trip with bypass failure event, the primary system 
relief valves would remove the majority of the stored heat, while 
the IC [isolation condenser] at DNPS and RCIC [reactor core 
isolation cooling] at QCNPS would remove the remaining decay heat. 
Slower losses of condenser vacuum would produce less severe AOOs, 
since the turbine stop valves and bypass valves will still be 
available prior to vacuum levels reaching the nominal trip setpoint 
for the turbine trip and turbine bypass valve closure scram.
    In that the proposed reduction of the Turbine Condenser Vacuum--
Low AV is based upon an AL [analytical limit] that is equal to the 
nominal trip setpoint for the turbine trip, the resulting nominal 
trip setpoint for the Turbine Condenser Vacuum--Low scram will still 
be more conservative than the turbine trip setpoint. Therefore, the 
sequence of events for the loss of main condenser AOO will still 
result in a reactor scram prior to the turbine trip. Since the 
proposed change to the Turbine Condenser Vacuum--Low AV will not 
impact the limiting AOO analysis (i.e., the turbine trip with bypass 
failure event), the proposed change does not reduce any margin of 
safety.
    Therefore, the proposed AV changes do not involve a significant 
reduction in the margin of safety.
    The proposed DNPS CFT STI change is based on an NRC-approved 
generic analysis. This analysis concluded that the proposed CFT STI 
change does not significantly affect the probability of failure or 
availability of the affected instrumentation systems. Therefore, the 
proposed DNPS CFT STI change does not affect the accident analyses 
that assume operability of the instrument associated with the AV.
    The proposed DNPS CC STI change is based on a revised setpoint 
error analysis for the Turbine Condenser Vacuum--Low scram 
instrument function that provides a revised AV, trip setpoint, and 
Expanded Tolerance (ET) for the instrument. The new AV, trip 
setpoint, and ET establish increased design margin between the 
nominal trip setpoint and the AV. This increased design margin, 
combined with historical CC data, provides adequate assurance that 
the component will remain operable when necessary for the prevention 
or mitigation of accidents or transients. Therefore, the proposed 
DNPS CFT STI change does not affect the accident analyses that 
assume operability of the instrument associated with the AV.
    Therefore, the proposed changes to extend the DNPS CFT and CC 
STIs do not involve a significant reduction in the margin of safety.
    In summary, the proposed DNPS and QCNPS AV changes and DNPS STI 
changes do 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. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555. NRC Branch Chief: Russell Gibbs.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
    Date of amendment request: October 9, 2007.
    Description of amendment request: The proposed amendment would 
change the technical specifications (TS) of Dresden Nuclear Power 
Station (DNPS), Units 2 and 3, consistent with TS Task Force (TSTF) 
Change Traveler TSTF-423 to the standard TSs 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 10 CFR 50.65(a)(4). Changes proposed herein will be 
made to the DNPS, Units 2 and 3, TSs for selected Required Action end 
states providing this allowance.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on December 14, 2005 (70 FR 74037), on possible 
license amendments adopting TSTF-423 using the NRC's consolidated line 
item improvement process (CLIIP) for amending licensee's TSs, which 
included a model safety evaluation (SE) and model no significant 
hazards consideration (NSHC) determination. The NRC staff

[[Page 68216]]

subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on March 26, 2006 (71 FR 14726), which included the resolution of 
public comments on the model SE. The licensee affirmed the 
applicability of the following NSHC determination in its application.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change allows a change to certain required end 
states when the TS Completion Times for remaining in power operation 
will be exceeded. Most of the requested technical specification (TS) 
changes are to permit an end state of hot shutdown (Mode 3) rather 
than an end state of cold shutdown (Mode 4) contained in the current 
TS. The request was limited to: (1) Those end states where entry 
into the shutdown mode is for a short interval, (2) entry is 
initiated by inoperability of a single train of equipment or a 
restriction on a plant operational parameter, unless otherwise 
stated in the applicable technical specification, and (3) the 
primary purpose is to correct the initiating condition and return to 
power operation as soon as is practical. Risk insights from both the 
qualitative and quantitative risk assessments were used in specific 
TS assessments. Such assessments are documented in Section 6 of GE 
NEDC-32988, Revision 2, ``Technical Justification to Support Risk 
Informed Modification to Selected Required Action End States for BWR 
Plants.'' They provide an integrated discussion of deterministic and 
probabilistic issues, focusing on specific technical specifications, 
which are used to support the proposed TS end state and associated 
restrictions. The staff finds that the risk insights support the 
conclusions of the specific TS assessments. Therefore, the 
probability of an accident previously evaluated is not significantly 
increased, if at all. The consequences of an accident after adopting 
proposed TSTF-423, are no different than the consequences of an 
accident prior to adopting TSTF-423. Therefore, the consequences of 
an accident previously evaluated are not significantly affected by 
this change. The addition of a requirement to assess and manage the 
risk introduced by this change will further minimize possible 
concerns. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
If risk is assessed and managed, allowing a change to certain 
required end states when the TS Completion Times for remaining in 
power operation are exceeded, i.e., entry into hot shutdown rather 
than cold shutdown to repair equipment, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change and the commitment by the licensee to adhere to the guidance 
in TSTF-IG-05-02, Implementation Guidance for TSTF-423, Revision 0, 
``Technical Specifications End States, NEDC-32988-A,'' will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change allows, for some systems, entry into hot 
shutdown rather than cold shutdown to repair equipment, if risk is 
assessed and managed. The BWROG's risk assessment approach is 
comprehensive and follows staff guidance as documented in RGs 1.174 
and 1.177. In addition, the analyses show that the criteria of the 
three-tiered approach for allowing TS changes are met. The risk 
impact of the proposed TS changes was assessed following the three-
tiered approach recommended in RG 1.177. A risk assessment was 
performed to justify the proposed TS changes. The net change to the 
margin of safety is insignificant. Therefore, this change does not 
involve a significant reduction in a margin of safety.

    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: Russell Gibbs.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa
    Date of amendment request: September 14, 2007.
    Description of amendment request: Duane Arnold Energy Center 
requests a proposed change to plant specific technical specifications 
(TS) 3.3.2.1, ``Control Rod Block Instrumentation,'' to allow the use 
of the improved Banked Position Withdrawal Sequence (BPWS) during 
shutdowns in accordance with NEDO-33091-A, Revision 2, ``Improved BPWS 
Control Rod Insertion Process,'' dated July 2004. The proposed changes 
are consistent with Nuclear Regulatory Commission (NRC)-approved 
Industry Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-476, Revision 1, ``Improved BPWS 
Control Rod Insertion Process (NEDO-33091).''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no-significant-hazards-consideration is presented below:
    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated.
    The proposed changes modify the TS to allow the use of the improved 
banked position withdrawal sequence (BPWS) during shutdowns if the 
conditions of NEDO-33091-A, Revision 2, ``Improved BPWS Control Rod 
Insertion Process,'' July 2004, have been satisfied. The staff finds 
that the licensee's justifications to support the specific TS changes 
are consistent with the approved topical report and TSTF-476, Revision 
1. Since the change only involves changes in control rod sequencing, 
the probability of an accident previously evaluated is not 
significantly increased, if at all. The consequences of an accident 
after adopting TSTF-476 are no different than the consequences of an 
accident prior to adopting TSTF-476. Therefore, the consequences of an 
accident previously evaluated are not significantly affected by this 
change. Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    Criterion 2 --The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident from any Previously Evaluated.
    The proposed change will not introduce new failure modes or effects 
and will not, in the absence of other unrelated failures, lead to an 
accident whose consequences exceed the consequences of accidents 
previously evaluated. The control rod drop accident (CRDA) is the 
design basis accident for the subject TS changes. This change does not 
create the possibility of a new or different kind of accident from an 
accident previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The proposed change, TSTF-476, Revision 1, incorporates the 
improved BPWS, previously approved in NEDO-33091-A, into the improved 
TS. The control rod drop accident (CRDA) is the design basis accident 
for the subject TS changes. In order to minimize the impact of a CRDA, 
the BPWS process was developed to minimize control rod reactivity worth 
for BWR plants. The proposed improved BPWS further simplifies the 
control rod insertion process, and in order to evaluate it, the

[[Page 68217]]

staff followed the guidelines of Standard Review Plan Section 15.4.9, 
and referred to General Design Criterion 28 of Appendix A to 10 CFR 
Part 50 as its regulatory requirement. The TSTF stated the improved 
BPWS provides the following benefits: (1) Allows the plant to reach the 
all-rods-in condition prior to significant reactor cool down, which 
reduces the potential for re-criticality as the reactor cools down; (2) 
reduces the potential for an operator reactivity control error by 
reducing the total number of control rod manipulations; (3) minimizes 
the need for manual scrams during plant shutdowns, resulting in less 
wear on control rod drive (CRD) system components and CRD mechanisms; 
and, (4) eliminates unnecessary control rod manipulations at low power, 
resulting in less wear on reactor manual control and CRD system 
components. The addition of procedural requirements and verifications 
specified in NEDO-33091-A, along with the proper use of the BPWS will 
prevent a control rod drop accident (CRDA) from occurring while power 
is below the low power setpoint (LPSP). The net change to the margin of 
safety is insignificant. Therefore, this change does not involve a 
significant reduction in a margin of safety.
    Based upon the above discussion of the amendment request, the 
requested change does not involve a significant hazards consideration.
    Attorney for licensee: Marjan Mashhadi, Florida Power & Light 
Company, 801 Pennsylvania Avenue, Suite 220, Washington, DC 20004.
    NRC Acting Branch Chief: Clifford G. Munson.
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: October 12, 2007.
    Description of amendment request: FPL Energy Point Beach, LLC 
(FPLE-PB) proposes to revise Technical Specification (TS) 5.5.1 5 
``Containment Leakage Rate Testing Program,'' for Units 1 and 2. The 
proposed change would allow a one-time interval extension of no more 
than 5 years for the Type A, Integrated Leakage Rate Test (ILRT).
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment proposes to revise the Technical 
Specifications (TS) to allow for the one-time extension of the 
containment integrated leakage rate test interval from 10 to 15 
years. The containment vessel function is to mitigate consequences 
of an accident. There are no design basis accidents initiated by a 
failure of the containment leakage mitigation function. The 
extension of the containment integrated leakage rate test interval 
will not create an adverse interaction with other systems that could 
result in initiation of a design basis accident. Therefore, the 
probability of occurrence of an accident previously evaluated is not 
significantly increased.
    The potential consequences of the proposed change have been 
quantified by analyzing the changes in risk that would result from 
extending the containment integrated leakage rate test interval from 
10 to 15 years. The increase in risk in terms of person-rem per year 
within 50 miles resulting from design basis accidents was estimated 
to be of a magnitude that NUREG-1493 indicates is very small. FPLE-
PB has also analyzed the increase in risk in terms of the frequency 
of large early releases from accidents. The increase in the large 
early release frequency resulting from the proposed extension was 
determined to be within the guidelines published in RG 1.I74. 
Additionally, the proposed change maintains defense-in-depth by 
preserving a reasonable balance among prevention of core damage, 
prevention of containment failure, and consequence mitigation. FPLE-
PB has determined that the increase in conditional containment 
failure probability from reducing the containment integrated leakage 
rate test frequency from one test per 10 years to one test per 15 
years would be small.
    Continued containment integrity is also assured by the history 
of successful containment integrated leakage rate tests, and the 
established programs for local leakage rate testing and IWE 
inservice inspections which are not affected by the proposed change. 
Therefore, the probability of occurrence or the consequences of an 
accident previously analyzed are not significantly increased.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to extend the containment integrated leakage 
rate test interval from 10 to 15 years does not create any new or 
different accident initiators or precursors. The length of the 
containment integrated leakage rate test interval does not affect 
the manner in which any accident begins. The proposed change does 
not create any new failure modes for the containment and does not 
affect the interaction between the containment and any other system. 
Thus, the proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The risk-based margins of safety associated with the containment 
integrated leakage rate test are those associated with the estimated 
person-rem per year, the large early release frequency and the 
conditional containment failure probability. FPLE-PB has quantified 
the potential effect of the proposed change on these parameters and 
determined that the effect is not significant. The non-risk-based 
margins of safety associated with the containment integrated leakage 
rate test are those involved with its structural integrity and leak 
tightness. The proposed change to extend the containment integrated 
leakage rate test interval from 10 to 15 years does not adversely 
affect either of these attributes. The proposed change only affects 
the frequency at which these attributes are verified. Therefore, the 
proposed change does not involve a significant reduction in 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. Antonio Fernandez, Senior Attorney, FPL 
Energy, LLC, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Acting Branch Chief: Cliff Munson.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit No. 2 (NMP2), Oswego County, New York
    Date of amendment request: October 22, 2007.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) 3.6.3.1, Primary Containment 
Hydrogen Recombiners, and references to the hydrogen and oxygen 
monitors in TS 3.3.3.1, Post Accident Monitoring (PAM) Instrumentation. 
The proposed TS changes support implementation of the revisions to 
Title 10 of the Code of Federal Regulations (10 CFR) Section 50.44, 
``Combustible gas control for nuclear power reactors,'' that became 
effective on October 16, 2003. These changes are consistent with 
Nuclear Regulatory Commission (NRC)-approved Revision 1 to TS Task 
Force (TSTF) Change Traveler, TSTF-447, ``Elimination of Hydrogen 
Recombiners and Change to Hydrogen and Oxygen Monitors.'' The 
availability of this TS improvement was announced in the Federal 
Register on September 25, 2003 (68 FR 55416) as part of the 
consolidated line item improvement process. The licensee affirmed the 
applicability of the model no significant hazards consideration 
determination in its application.

[[Page 68218]]

    The proposed amendment would also relocate, from the Renewed 
Facility Operating License to the NMP2 Updated Safety Analysis Report, 
License paragraph 2.C.(11a), Additional Condition 3, which requires 
establishing containment hydrogen monitoring within 90 minutes of 
initiating emergency core cooling following a loss-of-coolant accident.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration adopted by the licensee is 
presented below:
    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The revised 10 CFR 50.44 no longer defines a design-basis loss-of-
coolant accident (LOCA) hydrogen release, and eliminates requirements 
for hydrogen control systems to mitigate such a release. The 
installation of hydrogen recombiners and/or vent and purge systems 
required by 10 CFR 50.44(b)(3) was intended to address the limited 
quantity and rate of hydrogen generation that was postulated from a 
design-basis LOCA. The Commission has found that this hydrogen release 
is not risk-significant because the design-basis LOCA hydrogen release 
does not contribute to the conditional probability of a large release 
up to approximately 24 hours after the onset of core damage. In 
addition, these systems were ineffective at mitigating hydrogen 
releases from risk-significant accident sequences that could threaten 
containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen [and oxygen] monitors are no longer required to mitigate 
design-basis accidents and, therefore, the hydrogen monitors do not 
meet the definition of a safety-related component as defined in 10 CFR 
50.2. RG 1.97 Category 1 is intended for key variables that most 
directly indicate the accomplishment of a safety function for design-
basis accident events. The hydrogen [and oxygen] monitors no longer 
meet the definition of Category 1 in RG 1.97. As part of the rulemaking 
to revise 10 CFR 50.44 the Commission found that Category 3, as defined 
in RG 1.97, is an appropriate categorization for the hydrogen monitors 
because the monitors are required to diagnose the course of beyond 
design-basis accidents. [Also, as part of the rulemaking to revise 10 
CFR 50.44, the Commission found that Category 2, as defined in RG 1.97, 
is an appropriate categorization for the oxygen monitors, because the 
monitors are required to verify the status of the inert containment.]
    The regulatory requirements for the hydrogen [and oxygen] monitors 
can be relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite releases 
of radioactivity, and establishing protective action recommendations to 
be communicated to offsite authorities. Classification of the hydrogen 
monitors as Category 3, [classification of the oxygen monitors as 
Category 2] and removal of the hydrogen [and oxygen] monitors from TS 
will not prevent an accident management strategy through the use of the 
SAMGs [severe accident management guidelines], the emergency plan (EP), 
the emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner requirements 
and relaxation of the hydrogen [and oxygen] monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident from Any [Accident] Previously 
Evaluated
    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen [and oxygen] monitor requirements, including 
removal of these requirements from TS, will not result in any failure 
mode not previously analyzed. The hydrogen recombiner and hydrogen [and 
oxygen] monitor equipment was intended to mitigate a design-basis 
hydrogen release. The hydrogen recombiner and hydrogen [and oxygen] 
monitor equipment are not considered accident precursors, nor does 
their existence or elimination have any adverse impact on the pre-
accident state of the reactor core or post accident confinement of 
radionuclides within the containment building.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any [accident] previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in [a] Margin of Safety
    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen [and oxygen] monitor requirements, including 
removal of these requirements from TS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results in 
a neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this hydrogen 
release is not risk-significant because the design-basis LOCA hydrogen 
release does not contribute to the conditional probability of a large 
release up to approximately 24 hours after the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can be 
adequately met without reliance on safety-related hydrogen monitors.
    [Category 2 oxygen monitors are adequate to verify the status of an 
inerted containment.]
    Therefore, this change does not involve a significant reduction in 
[a] margin of safety. [The intent of the requirements established as a 
result of the TMI, Unit 2 accident can be adequately met without 
reliance on safety-related oxygen monitors.] Removal of hydrogen [and 
oxygen] monitoring from TS will not result in a significant reduction 
in their functionality, reliability, and availability.
    The NRC staff has reviewed the analysis adopted by the licensee 
and, based on this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendments involves no significant hazards 
consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station, 
Unit No. 2, Salem County, New Jersey
    Date of amendment request: October 17, 2007.
    Description of amendment request: The proposed amendment would 
allow a one-time revision to the requirements for fuel decay time prior 
to commencing movement of irradiated fuel in the

[[Page 68219]]

reactor pressure vessel (RPV). Currently, Technical Specification (TS) 
3/4.9.3, ``Decay Time'' requires that: (a) The reactor has been 
subcritical for at least 100 hours prior to movement of irradiated fuel 
in the RPV between October 15th through May 15th; and (b) the reactor 
has been subcritical for at least 168 hours prior to movement of 
irradiated fuel in the RPV between May 16th and October 14th. The 
calendar approach is based on average river water temperature which is 
cooler in the fall through spring months. The proposed amendment would 
revise TS 3/4.9.3 to allow fuel movement to commence at 86 hours after 
the reactor is subcritical. The proposed change would only be 
applicable to Salem Nuclear Generating Station, Unit No. 2 refueling 
outage 2R16, which is scheduled to commence on March 4, 2008.
    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 change involve a significant increase in the 
probability [ ] or consequences of an accident previously evaluated?
    Response: No.
    The proposed license amendment would allow fuel assemblies to be 
removed from the reactor core and be stored in the Spent Fuel Pool 
[SFP] in less time after subcriticality than currently allowed by 
the TSs. Decreasing the decay time of the fuel affects the 
radionuclide make-up of the fuel to be offloaded as well as the 
amount of decay heat that is present from the fuel at the time of 
offload. The accident previously evaluated that is associated with 
the proposed license amendment is the fuel handling accident [FHA]. 
Allowing the fuel to be offloaded in less time after subcriticality 
using actual heat loads does not impact the manner in which the fuel 
is offloaded. The accident initiator is the dropping of the fuel 
assembly. Since earlier offload does not affect fuel handling, there 
is no increase in the probability of occurrence of a [FHA]. The time 
frame in which the fuel assemblies are moved has been evaluated 
against the [Title 10 of the Code of Federal Regulations (10 CFR) 
Section 50.67] dose limits for members of the public, licensee 
personnel and control room. Additionally, the guidance provided in 
[Regulatory Guide (RG)] 1.183 was used for the selective application 
of Alternative Source Term. All dose limits are met with the reduced 
core offload times; and significant margin is maintained, as the 
minimum decay time prior to movement of fuel for the FHA analysis is 
24 hours.
    Therefore, the proposed license amendment does not significantly 
increase the probability [ ] or the consequences of accidents 
previously evaluated.
    2. [Does the change] [c]reate the possibility of a new or 
different kind of accident from any accident previously evaluated[?]
    Response: No.
    The proposed license amendment would allow core offload to occur 
in less time after subcriticality which affects the radionuclide 
makeup of the fuel to be offloaded as well as the amount of decay 
heat that is present from the fuel at the time of offload. The 
radionuclide makeup of the fuel assemblies and the amount of decay 
heat produced by the fuel assemblies do not currently initiate any 
accident. A change in the radionuclide makeup of the fuel at the 
time of core offload or an increase in the decay heat produced by 
the fuel being offloaded will not cause the initiation of any 
accident. The accident previously evaluated that is associated with 
fuel movement is the [FHA]; no new accidents are introduced. There 
is no change to the manner in which fuel is being handled or in the 
equipment used to offload or store the fuel. The effects of the 
additional decay heat load have been analyzed. The analysis 
demonstrates that the existing [SFP] cooling system and associated 
systems under worst-case circumstances would maintain licensing 
limits and the integrity of the [SFP].
    Therefore, the proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The margin of safety pertinent to the proposed changes is the 
dose consequences resulting from a [FHA]. The shorter decay time 
prior to fuel movement has been evaluated against 10 CFR 50.67 and 
all limits continue to be met. All dose limits are met with the 
reduced core offload times; and significant margin is maintained, as 
the minimum decay time prior to movement of fuel for the FHA 
analysis is 24 hours. Decay heat-up calculations performed prior to 
the refueling outage as part of the IDHM [Integrated Decay Heat 
Management] program ensure that planned spent fuel transfer to the 
SFP will not result in maximum SFP temperature exceeding the design 
basis limit of 149[deg]F (with both heat exchangers available) or 
180[deg]F (with one heat exchanger alternating between the two 
pools). As stated above, the changes in radionuclide makeup and 
additional heat load do not impact any safety settings and do not 
cause any safety limit to not be met. In addition, the integrity of 
the [SFP] is maintained.
    The time frame in which the fuel assemblies are moved has been 
evaluated against the 10 CFR 50.67 dose limits for members of the 
public, licensee personnel and control room. Additionally, the 
guidance provided in [RG] 1.183 was used. Calculations performed 
conclude that expected dose limits following a [FHA] are met with 
the proposed decay time prior to commencing fuel movement.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Unit Nos. 1 and 2, Hamilton County, Tennessee
    Date of amendment request: October 27, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) to establish more effective 
and appropriate action, surveillance, and administrative requirements 
related to ensuring the habitability of the control room envelope (CRE) 
in accordance with Nuclear Regulatory Commission (NRC)-approved 
Technical Specification Task Force (TSTF) Standard Technical 
Specification change traveler TSTF-448, Revision 3, ``Control Room 
Habitability.'' Specifically, the proposed amendment would modify TS 
3.7.7, ``Control Room Emergency Ventilation System,'' and TS Section 6, 
``Administrative Controls.'' The NRC staff issued a ``Notice of 
Availability of Technical Specification Improvement to Modify 
Requirements Regarding Control Room Envelope Habitability Using the 
Consolidated Line Item Improvement Process associated with TSTF-448, 
Revision 3, in the Federal Register on January 17, 2007 (72 FR 2022). 
The notice included a model safety evaluation, a model no significant 
hazards consideration (NSHC) determination, and a model license 
amendment request. In its application dated October 27, 2007, Tennessee 
Valley Authority (the licensee) affirmed the applicability of the model 
NSHC determination which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration adopted by the licensee is 
presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability

[[Page 68220]]

of structures, systems, and components to perform their intended 
function to mitigate the consequences of an initiating event within 
the assumed acceptance limits. The proposed change revises the TS 
for the CRE emergency ventilation system, which is a mitigation 
system designed to minimize unfiltered air leakage into the CRE and 
to filter the CRE atmosphere to protect the CRE occupants in the 
event of accidents previously analyzed. An important part of the CRE 
emergency ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident From Any Accident Previously 
Evaluated.
    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the analysis adopted by the licensee 
and, based on this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendments involves no significant hazards 
consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Thomas H. Boyce.

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.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2, New London County, Connecticut
    Date of amendment request: February 16, 2007.
    Brief description of amendment request: The proposed amendment 
would revise Technical Specification 3/4.4.3, ``Reactor Coolant System, 
Relief Valves'' to modify the method of testing the pressurizer Power 
Operated Relief Valves (PORVs). Specifically the requirement for bench 
testing the valves is changed to accommodate testing of the PORVs while 
installed in the plant. The change is requested due to the installation 
of new PORVs that are welded to the piping rather than bolted into the 
system.
    Date of publication of individual notice in Federal Register: 
November 19, 2007.
    Expiration date of individual notice: December 19, 2007 (public 
comment), January 18, 2008 (hearing requests).

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 at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

[[Page 68221]]

Duke Power Company LLC, et. al., Docket No. 50-414, Catawba Nuclear 
Station, Unit 2, York County, South Carolina
    Date of application for amendments: April 30, 2007.
    Brief description of amendments: The amendment revised Technical 
Specification (TS) 5.5.9, ``team Generator (SG) Tube Surveillance 
Program,'' regarding the required SG inspection scope for Catawba Unit 
2 during the End of Cycle 15 Refueling Outage and Operating Cycle 16. 
The changes modified the tube repair criteria for portions of the SG 
tubes within the hot leg tubesheet region of the SGs.
    Date of issuance: October 31, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 233.
    Renewed Facility Operating License No. NPF-52: Amendments revised 
the licenses and the technical specifications.
    Date of initial notice in Federal Register: August 13, 2007 (72 FR 
45272).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 31, 2007.
    No significant hazards consideration comments received: No.
Duke Power Company LLC, et. al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina
    Date of application for amendments: March 29, 2007, as supplemented 
September 7, 2007, October 9 and October 12, 2007.
    Brief description of amendments: The amendments revised the Catawba 
1 and 2, Technical Specifications 3.5.2.8, and authorized changes to 
the updated final safety analysis report concerning modifications to 
the emergency core cooling system sump.
    Date of issuance: November 8, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 238, 234.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: August 13, 2007 (72 FR 
45274). The supplements dated September 7, 2007, October 9, and October 
12, 2007, 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 November 8, 2007.
    No significant hazards consideration comments received: No.
Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
    Date of application of amendments: November 16, 2006, supplemented 
May 9 and August 28, 2007.
    Brief description of amendments: The amendments authorized revision 
of the Updated Final Safety Analysis Report to describe the flood 
protection measures for the auxiliary building.
    Date of Issuance: November 14, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days after completion of the flood protection measures for 
the auxiliary building.
    Amendment Nos.: 357, 359, and 358.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the licenses.
    Date of initial notice in Federal Register: January 3, 2007 (72 FR 
151). The supplements dated May 9 and August 28, 2007, 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 November 14, 2007.
    No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
    Date of amendment request: July 16, 2007, as supplemented by letter 
dated August 7, 2007.
    Brief description of amendment: The proposed amendment revised the 
facility operating license (FOL), Paragraph 2.C, and technical 
specifications (TS) 3.7.2 and TS 5.5 for River Bend Station, Unit 1.
    Date of issuance: November 16, 2007.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 154.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: September 11, 2007 (72 
FR 51857). The supplement dated August 7, 2007, 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 on September 11, 2007 (72 FR 51857). 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated November 16, 2007.
    No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
    Date of application for amendment: November 27, 2006, as 
supplemented by letter dated August 24, 2007.
    Brief description of amendment: This amendment revises multiple TSs 
relating to testing of the Emergency Diesel Generators (EDGs). 
Specifically, the changes eliminate various accelerated testing 
requirements, eliminate the EDG test schedule table based on failure 
rates, relax acceptance criteria associated with the ``fast start'' and 
load rejection tests and eliminate the EDG failure report.
    Date of issuance: November 6, 2007.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 189 and 150.
    Facility Operating License Nos. NPF-39 and NPF-85: This amendment 
revised the license and Technical Specifications.
    Date of initial notice in Federal Register: July 31, 2007 (72 FR 
41784).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 6, 2007.
    No significant hazards consideration comments received: No.
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 application for amendments: June 29, 2007.
    Brief description of amendments: The amendments would modify the 
Technical Specifications (TSs) 3.7.2, by removing the specific 
isolation time for the main steam isolation valves from the

[[Page 68222]]

associated TS surveillance requirements and by replacing it with the 
requirement to verify the valve isolation time is within limits.
    Date of issuance: November 16, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 230, 235.
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revised the Technical Specifications/License.
    Date of initial notice in Federal Register: September 11, 2007 (72 
FR 51865).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 16, 2007.
    No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota
    Date of application for amendment: July 9, 2007.
    Brief description of amendment: The amendment revised the Technical 
Specifications by removing the Table of Contents.
    Date of issuance: November 8, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 152.
    Facility Operating License No. DPR-22.
    Amendment revised the Technical Specifications. Date of initial 
notice in Federal Register: August 14, 2007 (72 FR 45459).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 8, 2007.
    No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, SalemNuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
    Date of application for amendments: August 15, 2007, as 
supplemented on September 6, 2007.
    Brief description of amendments: The amendments revise the 
licensing basis, as described in Appendix 3A of the Salem Updated Final 
Safety Analysis Report (UFSAR), regarding the method of calculating the 
net positive suction head available for the emergency core cooling 
system and containment heat removal system pumps. These changes to the 
Salem licensing basis relate to issues associated with Generic Letter 
2004-02, ``Potential Impact of Debris Blockage on Emergency 
Recirculation During Design Basis Accidents at Pressurized-Water 
Reactors.''
    Date of issuance: November 15, 2007.
    Effective date: As of the date of issuance, to be implemented by 
December 31, 2007.
    Amendment Nos.: 285 and 268.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revise the UFSAR.
    Date of initial notice in Federal Register: September 11, 2007 (72 
FR 51866). The letter dated September 6, 2007, provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the application beyond 
the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 15, 2007.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 51.12(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 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the

[[Page 68223]]

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's 
(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 [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. Within 60 days after the date 
of publication of this notice, person(s) may file a request for a 
hearing with respect to issuance of the amendment to the subject 
facility operating license and any person whose interest may be 
affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request via electronic submission 
through the NRC E-Filing system for a hearing and a petition for leave 
to intervene. 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 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland, and electronically on the Internet at the NRC Web 
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. 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 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the 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.\1\ 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 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.
---------------------------------------------------------------------------

    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
---------------------------------------------------------------------------

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    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. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    A request for hearing or a petition for leave to intervene 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 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 a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
[email protected], 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 petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to

[[Page 68224]]

download the Workplace Forms Viewer\TM\ to access the Electronic 
Information Exchange (EIE), a component of the E-Filing system. The 
Workplace Forms Viewer\TM\ 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 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 technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, 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 petition 
and/or request 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). To be timely, filings must be submitted no later 
than 11:59 p.m. Eastern Time on the due date.
    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, 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. 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.
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: October 22, 2007, as 
supplemented November 2 and November 9, 2007.
    Brief Description of amendments: This amendment adds a new license 
condition, P.(3), to license Nos. DPR-32 and DPR-37, which authorize 
the licensee to modify the GOTHIC code as described in the Updated 
Final Safety Analysis Report (UFSAR) and update the UFSAR as required 
by 10 CFR 50.71(e).
    Date of issuance: November 15, 2007.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment Nos.: 256, 255.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments revise the licenses.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. The notice provided an opportunity to submit 
comments (by November 13, 2007) on the Commission(s proposed NSHC 
determination. No comments have been received. The notice also provided 
an opportunity to request a hearing (by December 31, 2007), but 
indicated that if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment. The 
Commission's related evaluation of the amendment, finding of exigent 
circumstances, state consultation, and final NSHC determination are 
contained in a safety evaluation dated November 15, 2007.
    Attorney for licensee: Ms. Lillian M. Cuoco, Esq.
    NRC Branch Chief: Evangelos C. Marinos.

    Dated at Rockville, Maryland, this 23rd day of November, 2007.

    For the Nuclear Regulatory Commission.

Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E7-23225 Filed 12-3-07; 8:45 am]
BILLING CODE 7590-01-P