[Federal Register Volume 72, Number 9 (Tuesday, January 16, 2007)]
[Notices]
[Pages 1779-1783]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-321]


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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 December 22, 2006 to January 4, 2007. The 
last biweekly notice was published on January 3, 2007 (72 FR 147).

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, 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. 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 persons 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 should specifically explain the reasons 
why intervention should be permitted

[[Page 1780]]

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 a hearing or a petition for leave to intervene must 
be filed 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; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this 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].
    Nuclear Management Company, 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: December 14, 2006.
    Description of amendment request: The proposed amendments revise 
the technical specifications to add the FERRET Code as an approved 
methodology for determining reactor coolant system pressure and 
temperature limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The proposed change revises Technical Specification (TS) 5.6.5, 
``Reactor Coolant System (RCS) Pressure and Temperature Limits 
Report (PTLR)'', to add the FERRET Code as an approved methodology 
for determining RCS pressure and temperature limits.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or the manner in which the plant is operated and 
maintained. The proposed change does not alter or prevent the 
ability of structures, systems, and components from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits. Therefore, the proposed 
change does not significantly increase the probability of any 
accident previously evaluated.
    There will be no change to normal plant operating parameters, 
engineered safety feature actuation setpoints, accident mitigation 
capabilities, or accident analysis assumptions or inputs. The 
proposed change does not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
Further, the proposed change does not increase the types or amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.
    Therefore, the probability or consequences of any accident 
previously evaluated will not be significantly increased as a result 
of the proposed change.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a new or different kind 
of accident from any accident previously evaluated.
    The proposed change incorporates the FERRET Code as an approved 
methodology for determining RCS pressure and temperature limits. The 
change does not impose any new or different requirements or

[[Page 1781]]

eliminate any existing requirements. The proposed change is 
consistent with the safety analysis assumptions and current plant 
operating practice.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. Equipment important to safety will continue 
to operate as designed. The change does not result in any event 
previously deemed incredible being made credible. The change does 
not result in adverse conditions or result in any increase in the 
challenges to safety systems. Therefore, the proposed change does 
not create the possibility of a new or different type of accident 
from any accident previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not result in a significant reduction 
in a margin of safety.
    The proposed change incorporates the FERRET Code as an approved 
methodology for determining RCS pressure and temperature limits. The 
proposed change does not alter safety limits, limiting safety system 
settings, or limiting conditions for operation. The setpoints at 
which protective actions are initiated are not altered by the 
proposed change.
    There are no new or significant changes to the initial 
conditions contributing to accident severity or consequences. The 
proposed amendment will not otherwise affect the plant protective 
boundaries, will not cause a release of fission products to the 
public, nor will it degrade the performance of any other structures, 
systems or components (SSCs) important to safety. Therefore, the 
requested change will not result in 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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Acting Branch Chief: L. Raghavan.
    Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri.
    Date of amendment request: August 17, 2006.
    Description of amendment request: The amendment would revise 
Technical Specifications (TSs) 2.1.1, ``Reactor Core SLs [Safety 
Limits],'' 3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' 3.4.1, 
RCS [reactor coolant system] Pressure, Temperature, and Flow Departure 
from Nucleate Boiling (DNB) Limits,'' and 5.6.5, ``Core Operating 
Limits Report (COLR).'' The changes would (1) relocate certain 
operating cycle-specific parameters limits, including TS Figure 2.1.1-
1, ``Reactor Core Safety Limits,'' from the above TSs to the plant 
COLR, (2) add two new safety limits for departure from nucleate boiling 
ratio (DNBR) and peak fuel centerline temperature, and (3) add several 
topical reports to TS 5.6.5 and have the reports in TS 5.6.5 cited by 
only the report title and number. The TSs would state that the limits 
to be met or the values of denoted parameters are specified in the 
COLR. The existing TS 5.6.5 has seven core operating limits that are 
listed in the specification, and this would be expanded to include the 
three additional limits from TSs 2.1.1, 3.3.1, and 3.4.1. The changes 
are consistent with NRC-approved Standard Technical Task Force (TSTF) 
Traveler TSTF-339, Revision 2, ``Relocate TS Parameters to COLR,'' and 
TSTF-363, Revision 0, ``Relocate Topical Report References in ITS 
[Improved Technical Specification] 5.6.5, COLR.''
    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 change[s] involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no design [or equipment] changes. The design of the reactor trip 
system (RTS) instrumentation and engineered safety feature actuation 
system (ESFAS) instrumentation will be unaffected and these 
protection systems will continue to function in a manner consistent 
with the plant design basis. All design, material, and construction 
standards that were applicable prior to this amendment request will 
be maintained.
    The proposed changes will not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed changes will not 
alter or prevent the ability of structures, systems, and components 
(SSCs) from performing their intended [safety] functions to mitigate 
the consequences of an initiating event within the assumed 
acceptance limits.
    The proposed changes are programmatic and administrative in 
nature. These changes do not physically alter safety-related systems 
nor affect the way in which safety-related systems perform their 
functions. Additional Safety Limits on the DNB [departure from 
nucleate boiling] design basis and peak fuel centerline temperature 
are being imposed in TS 2.1.1, ``Reactor Core Safety Limits,'' and 
the Reactor Core Safety Limits figure is being relocated to the 
COLR. The additional Safety Limits are consistent with the values 
stated in the FSAR [Final Safety Analysis Report for the Callaway 
Plant]. The proposed changes do not, by themselves, alter any of the 
relocated limits. The removal of the cycle-specific parameter limits 
from the TS[s] does not eliminate existing requirements to comply 
with the parameter limits. [The value of the limits is relocated to 
the COLR, but the requirement to follow that limit remains in the 
TSs by the reference to the limits or values in the COLR, and the 
values of the limits are not being changed by this amendment.] TS 
5.6.5.b continues to ensure that the analytical methods used to 
determine the core operating limits meet NRC reviewed and approved 
methodologies [by the requirement stated in TS 5.6.5.b that ``the 
analytical methods used to determine the core operating limits shall 
be those previously reviewed and approved by the NRC'']. TS 5.6.5.c, 
[which is] unchanged by this application, will continue to ensure 
that applicable limits of the safety analyses are met [by continuing 
to state this as a requirement in the TSs].
    The proposed changes to reference only the Topical Report number 
and title do not alter the use of the analytical methods used to 
determine core operating limits that have been reviewed and approved 
by the NRC. [This remains a requirement stated in TS 5.6.5.b.] This 
[proposed] method of referencing Topical Reports would allow the use 
of current [NRC-approved] Topical Reports to support [the] limits in 
the COLR without [the licensee] having to submit an amendment to the 
operating license. Implementation of revisions to Topical Reports 
for Callaway Plant applications would still be reviewed in 
accordance with 10 CFR 50.59(c)(2)(viii) and, where required, 
receive prior NRC review and approval. [The criteria in the 
regulation governing changes to the plant without NRC approval, 10 
CFR 50.59, would have to be met before the licensee could use a 
later version of an NRC-approved Topical Report that is listed in TS 
5.6.5.b.]
    The cycle-specific parameter limits being transferred from the 
TS[s] to the COLR will continue to be controlled under existing 
programs and procedures. The FSAR accident analyses will continue to 
be examined with respect to future changes in the cycle-specific 
parameters using NRC reviewed and approved reload design 
methodologies [(i.e., NRC reviewed and approved Topical Reports)], 
ensuring that the evaluation of new reload designs under 10 CFR 
50.59 is bounded by previously accepted analyses.
    All accident analysis acceptance criteria will continue to be 
met with the proposed changes. The proposed changes will not affect 
the source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of an 
accident previously evaluated. The proposed changes will not alter 
any assumptions or change any mitigation actions in the radiological 
consequence evaluations in the FSAR. The applicable radiological 
dose acceptance criteria will continue to be met.
    [The proposed changes do not alter any requirements in the TSs, 
but they do add two

[[Page 1782]]

new safety limits to TS 2.2.1. The changes also relocate certain 
limits or parameter values from the TSs to the COLR; however, these 
limits and values are still required to be met and be determined 
from NRC-approved methodologies that apply to the Callaway Plant. 
Therefore, there are no changes to accident analyses previously 
evaluated and described in the FSAR.]
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. [Do] the proposed change[s] create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    There are no proposed design changes nor are there any changes 
in the method by which any safety-related plant SSC performs its 
safety function. Th[ese] change[s] will not affect the normal method 
of plant operation or change any operating parameters. No equipment 
performance requirements will be affected. The proposed changes will 
not alter any assumptions made in the safety analyses.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of this amendment. There will be no adverse effect or 
challenges imposed on any safety-related system as a result of this 
amendment. [No equipment is being added to the plant by the 
amendment.]
    The proposed amendment will not alter the design or performance 
of the 7300 Process Protection System, Nuclear Instrumentation 
System, or Solid State Protection System used in the plant 
protection systems.
    Relocation of cycle-specific parameter limits has no influence 
on, nor does it contribute in any way to, the possibility of a new 
or different kind of accident. The relocated cycle-specific 
parameter limits will continue to be calculated using the NRC 
reviewed and approved methodologies. The proposed changes do not 
alter assumptions made in the safety analyses. Operation within the 
core operating limits will continue to be observed.
    Therefore, the proposed changes do not create the possibility of 
a new or different accident from any accident previously evaluated.
    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety?
    Response: No.
    There will be no effect on those plant systems necessary to 
assure the accomplishment of protection functions. There will be no 
impacts on the overpower limit, departure from nucleate boiling 
ratio (DNBR) limits, heat flux hot channel factor (FQ), 
nuclear enthalpy rise hot channel factor (F[Delta]H), loss of 
coolant accident peak cladding temperature (LOCA PCT), peak local 
power density, or any other margin of safety. The applicable 
radiological dose consequence acceptance criteria will continue to 
be met.
    The proposed changes do not eliminate any surveillances or alter 
the frequency of surveillances [(i.e., the surveillance test 
intervals)] required by the Technical Specifications. The nominal 
RTS and ESFAS trip setpoints will remain unchanged. None of the 
acceptance criteria for any accident analysis will be changed.
    The development of cycle-specific parameter limits for future 
reload designs will continue to conform to NRC reviewed and approved 
methodologies, and will be performed pursuant to 10 CFR 50.59 to 
assure that plant operation [is] within [these] cycle-specific 
parameter limits.
    The proposed changes will have no impact on the radiological 
consequences of a design basis accident.
    [The proposed changes do not alter any requirements in the TSs. 
They relocate certain limits or parameter values from the TSs to the 
COLR; however, these limits and values are still required to be met 
and be determined from NRC-approved methodologies that apply to the 
Callaway Plant.]
    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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: David Terao.

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 51.12(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 
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].
    Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin.
    Date of application for amendment: September 25, 2006.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 4.2.a, ``ASME [American Society of Mechanical 
Engineers] Code Class 1, 2, 3, and MC Components and Supports.'' The 
revised TS 4.2.a.2, references the ASME Code for Operation and 
Maintenance of Nuclear Power Plants.
    Date of issuance: December 14, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 189
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 24, 2006 (71 FR 
62308).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 14, 2006.
    No significant hazards consideration comments received: No.

    DukePower Company LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina.
    Date of application for amendments: December 20, 2005, as 
supplemented May 4 and August 31, 2006.
    Brief description of amendments: The amendments revised the McGuire 
1 and

[[Page 1783]]

2 licensing basis to adopt a selective implementation of the 
alternative source term radiological analysis methodology. The 
amendments also revised Technical Specification 3.9.4, ``Containment 
Penetrations.''
    Date of issuance: December 22, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 236, 218
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: August 24, 2006 (71 FR 
50105)
    The supplements dated May 4 and August 31, 2006, 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.
    No significant hazards consideration comments received: No.

    Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, 
LaSalle County Station, Units 1 and 2, LaSalle County, Illinois.
    Date of application for amendments: December 9, 2004, as 
supplemented by letters dated August 16, August 24, September 13, and 
October 12, 2006.
    Brief description of amendments: The amendments adopt Technical 
Specification Task Force (TSTF) Standard Technical Specification (STS) 
Change Traveler 360 (TSTF-360), Revision 1, ``DC Electric Rewrite.'' 
The amendment revised Technical Specification (TS) Section 3.8.4, ``DC 
Sources-Operating,'' TS 3.8.5, ``DC Sources-Shutdown,'' TS 3.8.6, 
``Battery Cell Parameters,'' and adds a new TS Section 5.5.14, 
``Battery Monitoring and Maintenance Program.''
    Date of issuance: December 19, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 179/165.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: April 12, 2005 (70 FR 
19115)
    The August 16, August 24, September 13, and October 12, 2006 
supplements contained clarifying information and did not change the NRC 
staff's initial proposed finding of no significant hazards 
consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 19, 2006.
    No significant hazards consideration comments received: No.
    Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia.
    Date of application for amendment: May 30, 2006, as supplemented by 
letter dated June 30, 2006.
    Brief description of amendment: The proposed amendments would 
relocate the American Society for Testing and Materials (ASTM) standard 
being used to test the total particulate concentration of the stored 
fuel oil to the Technical Specification (TS) Bases. This proposed 
change is described in TS Task Force (TSTF) Standard TS Change Traveler 
TSTF-374, Rev. 0, ``Revision to TS 5.5.13 and Associated TS Bases for 
Diesel Fuel Oil.'' In addition, the licensee has proposed to use a 
``water and sediment test'' instead of the ``clear and bright'' test 
provided in TSTF-374.
    Date of issuance: December 11, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 249, 229.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
change the licenses and the technical specifications.
    Date of initial notice in Federal Register: August 15, 2006 (71 FR 
46941)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 11, 2006.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 5th day of January 2007.

    For The Nuclear Regulatory Commission.
John W. Lubinski,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
 [FR Doc. E7-321 Filed 1-12-07; 8:45 am]
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