[Federal Register Volume 71, Number 195 (Tuesday, October 10, 2006)]
[Notices]
[Pages 59529-59538]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-16560]


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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 September 15, 2006, to September 28, 2006. 
The last biweekly notice was published on September 26, 2006 (71 FR 
56189).

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.

[[Page 59530]]

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 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

[[Page 59531]]

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].

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: May 30, 2006.
    Description of amendment request: The proposed amendment would 
revise technical specification (TS) Definitions 1.14, ``LEAKAGE'', and 
1.16, ``PRESSURE BOUNDARY LEAKAGE''; revise TS 3/4.6.2, ``Reactor 
Coolant System Operational Leakage''; add a new TS 3/4.4.5, ``Steam 
Generator (SG) Tube Integrity;'' add a new TS 6.8.4.g, ``Steam 
Generator (SG) Program;'' and add a new TS 6.9.1.12, ``Steam Generator 
Tube Inspection Report''; as well as administrative and editorial 
changes. These changes are consistent with the NRC-approved Revision 4 
to TS Task Force (TSTF) Standard TS change traveler, TSTF-449, ``Steam 
Generator Tube Integrity.'' The proposed changes are necessary in order 
to implement the guidance for the industry initiative on Nuclear Energy 
Institute (NEI) 97-06, ``Steam Generator Program Guidelines.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments 
concerning TSTF-449, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on May 6, 2005 
(70 FR 24126). The licensee affirmed the applicability of the following 
NSHC determination in its application dated May 30, 2006.
    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:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change requires a SG Program that includes 
performance criteria that will provide reasonable assurance that the 
SG tubing will retain integrity over the full range of operating 
conditions (including startup, operation in the power range, hot 
standby, cooldown and all anticipated transients included in the 
design specification). The SG performance criteria are based on tube 
structural integrity, accident induced leakage, and operational 
LEAKAGE.
    A[n] SGTR [SG tube rupture] event is one of the design basis 
accidents that are analyzed as part of a plant's licensing basis. In 
the analysis of a[n] SGTR event, a bounding primary to secondary 
LEAKAGE rate equal to the operational LEAKAGE rate limits in the 
licensing basis plus the LEAKAGE rate associated with a double-ended 
rupture of a single tube is assumed.
    For other design basis accidents such as [an] MSLB [main steam 
line break], rod ejection, and reactor coolant pump locked rotor[,] 
the tubes are assumed to retain their structural integrity (i.e., 
they are assumed not to rupture). These analyses typically assume 
that primary to secondary LEAKAGE for all SGs is 1 gallon per minute 
or increases to 1 gallon per minute as a result of accident induced 
stresses. The accident induced leakage criterion introduced by the 
proposed changes accounts for tubes that may leak during design 
basis accidents. The accident induced leakage criterion limits this 
leakage to no more than the value assumed in the accident analysis.
    The SG performance criteria proposed change[s] to the TS[s] to 
identify the standards against which tube integrity is to be 
measured. Meeting the performance criteria provides reasonable 
assurance that the SG tubing will remain capable of fulfilling its 
specific safety function of maintaining reactor coolant pressure 
boundary integrity throughout each operating cycle and in the 
unlikely event of a design basis accident. The performance criteria 
are only a part of the SG Program required by the proposed change[s] 
to the TS[s]. The program, defined by NEI 97-06, Steam Generator 
Program Guidelines, includes a framework that incorporates a balance 
of prevention, inspection, evaluation, repair, and leakage 
monitoring. The proposed changes do not, therefore, significantly 
increase the probability of an accident previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT 1[I]-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT 1[I]-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The typical analysis of the limiting design basis 
accident assumes that primary to secondary leak rate after the 
accident is 1 gallon per minute with no more than [500 gallons per 
day or 720 gallons per day] in any one SG, and that the reactor 
coolant activity levels of DOSE EQUIVALENT 1[I]-131 are at the TS 
values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design basis 
accident and is an improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of a[n] SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed performance based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed SG Program will not 
introduce

[[Page 59532]]

any adverse changes to the plant design basis or postulated 
accidents resulting from potential tube degradation. The result of 
the implementation of the SG Program will be an enhancement of SG 
tube performance. Primary to secondary LEAKAGE that may be 
experienced during all plant conditions will be monitored to ensure 
it remains within current accident analysis assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the [a] Margin of Safety

    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the SG Program to manage SG tube 
inspection, assessment, repair, and plugging. The requirements 
established by the SG Program are consistent with those in the 
applicable design codes and standards and are an improvement over 
the requirements in the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-18, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Daniel S. Collins.

Tennessee Valley Authority (TVA), Docket No. 50-259 , Browns Ferry 
Nuclear Plant (BFN), Unit 1, Limestone County, Alabama

    Date of amendment request: September 22, 2006 (TS-431).
    Description of amendment request: The proposed amendment 
supplements a June 28, 2004, request to increase the licensed thermal 
power from 3293 megawatt thermal (MWt) to 3952 MWt, an approximate 20 
percent increase in thermal power. This supplement requests interim 
approval of an increase in licensed thermal power from 3293 MWt to 3458 
MWt with an attendant 30-psi increase in reactor pressure. This 
represents an approximate 5 percent increase above the original 
licensed thermal power (OLTP) of 3293 MWt. An interim approval would 
provide for operation at 105 percent power until such time as certain 
steam dryer analyses can be completed. The NRC staff's review of the 
remainder of the June 2004 application would resume upon receipt of the 
satisfactorily completed steam dryer analyses.
    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 probability (frequency of occurrence) of Design Basis 
Accidents occurring is not affected by the increased power level, 
because BFN Unit 1 continues to comply with the regulatory and 
design basis criteria established for plant equipment. An evaluation 
of the Boiling Water Reactor probabilistic risk assessments 
concludes that the calculated core damage frequency does not 
significantly change due to operation at 105% OLTP.
    Scram setpoints (equipment settings that initiate automatic 
plant shutdowns) are established such that there is no significant 
increase in scram frequency due to operation at 105% OLTP. No new 
challenges to safety-related equipment result from operation at 105% 
OLTP.
    The probability of Design Basis Accidents occurring is not 
affected by taking credit for containment overpressure in ensuring 
adequate NPSH [Net Positive Suction Head] for the BFN Unit 1 low 
pressure ECCS pumps. NRC Bulletin 96-03 requested that BWR [Boiling-
Water Reactors] owners implement appropriate measures to minimize 
the potential clogging of the Emergency Core Cooling System (ECCS) 
suppression chamber strainers by potential debris generated by a 
LOCA [loss-of-coolant accident]. TVA installed new, high-capacity 
passive strainers on BFN Unit 1 of the same design as BFN Units 2 
and 3. In addition, TVA's proposed resolution of NRC Bulletin 96-03 
for BFN Unit 1 takes credit for containment overpressure to maintain 
adequate ECCS pump Net Positive Suction Head (NPSH). Containment 
pressure will increase following a pipe break occurring inside 
containment. Crediting containment overpressure in the analysis of 
the consequences of the Loss of Coolant Accident (LOCA) does not 
affect the precursors for the LOCA, nor does it affect the 
precursors for any other accident or transient analyzed in Chapter 
14 of the BFN Updated Final Safety Analysis Report (UFSAR). 
Therefore, there is no increase in the probability of any accident 
previously evaluated.
    The changes in consequences of hypothetical accidents, which 
would occur from 102% of the stretch power uprate reactor thermal 
power compared to those previously evaluated, are in all cases 
insignificant. The stretch power uprate accident evaluations do not 
exceed any of their NRC-approved acceptance limits. The spectrum of 
hypothetical accidents and transients has been investigated, and are 
shown to meet the plant's currently licensed regulatory criteria. In 
the area of core design, for example, the fuel operating limits such 
as Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and 
Safety Limit Minimum Critical Power Ratio (SLMCPR) are still met, 
and fuel reload analyses will show plant transients meet the 
criteria accepted by the NRC. Challenges to fuel (ECCS performance) 
are evaluated, and shown to continue to meet the criteria of 10 CFR 
50.46.
    Challenges to the containment have been evaluated at the 
increased power level, and the containment and its associated 
cooling systems continue to meet the design and licensing criteria. 
Radiological release events (accidents) have been evaluated at the 
increased power level, and shown to be less than the limits of 10 
CFR 50.67.
    The radiological consequences of the design basis accident are 
not increased by taking credit for the post-LOCA suppression chamber 
airspace pressure. The containment will continue to function as 
designed. This proposed change only takes credit for containment 
pressure that would exist following a LOCA. Crediting this pressure 
in ensuring adequate ECCS NPSH will not result in an increase in 
containment leakage assumed in any analysis.
    Therefore, the proposed amendment does not result in a 
significant increase in consequences or a significant increase in 
the probability or consequences of any 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.
    Equipment that could be affected by operation at 105% OLTP has 
been evaluated. No new operating mode, safety-related equipment 
lineup, accident scenario or equipment failure mode was identified. 
The full spectrum of accident considerations has been evaluated and 
no new or different kind of accident has been identified. Operation 
at 105% OLTP uses developed technology, and applies it within the 
capabilities of existing plant safety related equipment in 
accordance with the regulatory criteria, including NRC approved 
codes, standards and methods. No new power dependent accidents have 
been identified.
    The BFN Unit 1 TS [Technical Specifications] require revision to 
implement

[[Page 59533]]

operation at 105% OLTP. All revisions have been assessed, and it has 
been determined that the proposed change will not introduce a 
different accident than that previously evaluated.
    The proposed use of the post-LOCA suppression chamber airspace 
pressure in the calculation of NPSH for the ECCS pumps does not 
introduce any new modes of plant operation or make physical changes 
to plant systems. Rather, the post-LOCA suppression chamber airspace 
pressure is a consequence of the conditions that would exist in the 
containment following a large pipe break inside containment. The 
proposed amendment does not introduce new equipment which could 
create a new or different kind of accident. No new external threats, 
release pathways, or equipment failure modes are created.
    Therefore, the change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The calculated loads on all affected structures, systems and 
components will remain within their design allowables for all design 
basis event categories. No NRC acceptance criterion is exceeded. 
Because the BFN Unit 1 configuration and reactions to transients and 
hypothetical accidents does not result in exceeding the presently 
approved NRC acceptance limits, operation at 105% OLTP does not 
involve a significant reduction in a margin of safety.
    The post-LOCA suppression chamber airspace pressure is a 
byproduct of the conditions that will exist in the containment after 
a line break inside containment. Conservative analyses have been 
performed that demonstrate that sufficient post-accident suppression 
chamber airspace pressure will be available to meet the NPSH 
requirements for the low pressure ECCS pumps. By enabling credit of 
these conditions for the low pressure ECCS pumps, adequate NPSH 
margin will be ensured, and accordingly, the ECCS pumps will meet 
their performance requirements. Therefore, the credit for 
containment overpressure 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: L. Raghavan.

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].

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

    Date of application for amendment: December 2, 2005.
    Brief description of amendment: The amendment revised the Oyster 
Creek Nuclear Generating Station Technical Specifications (TSs) to 
increase the allowable as-found main steam safety valve code safety 
function lift setpoint tolerance from 1% to 3%.
    Date of Issuance: September 13, 2006.
    Effective date: As of the date of Issuance to be implemented within 
60 days.
    Amendment No.: 261.
    Facility Operating License No. DPR-16: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: January 17, 2006 (71 FR 
2588).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated September 13, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 7, 2005, as supplemented 
on May 12, 2006.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) to eliminate the use of the defined term 
Core Alterations. The amendments incorporate the changes reflected in 
TS Task Force (TSTF) Travelers 471-T (TSTF-471-T), ``Eliminate use of 
term CORE ALTERATIONS in ACTIONS and Notes,'' and TSTF-51-A, ``Revise 
containment requirements during handling irradiated fuel and core 
alterations.'' In addition, the amendments revise TS 3.9.2, ``Nuclear 
Instrumentation,'' by replacing ``Core Alterations'' with ``positive 
reactivity additions'' in the Required Action for an inoperable source 
range monitor during refueling operations. The limiting conditions for 
operation in TS 3.9.4, ``Shutdown Cooling (SDC) and Coolant 
Recirculation--High Water Level,'' are also revised by replacing ``core 
alterations'' with ``movement of fuel assemblies within containment.''
    Date of issuance: September 21, 2006
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 279 and 256.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Licenses and Technical Specifications.
    Date of initial notice in Federal Register: July 5, 2005 (70 FR 
38716).
    The May 12, 2006, letter 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.

[[Page 59534]]

    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated September 21, 2006.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: March 3, 2005, as supplemented 
by letter dated July 6, 2006.
    Brief description of amendment: The amendment revises the 
requirements of Technical Specification (TS) 5.6.5, ``Core Operating 
Limits Report (COLR).''
    Date of issuance: September 20, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No. 209.
    Renewed Facility Operating License No. DPR-23. Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: May 24, 2005 (70 FR 
29787). The supplemental letter provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 20, 2006.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: August 20, 2004, as supplemented 
by letters dated June 22, 2005, June 26, 2006, and September 18, 2006.
    Brief description of amendment: The amendment revises Table 3.3.1-
1, Functions 3, 14, 17.a., 20, and the footnote related to Function 20.
    Date of issuance: September 22, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No. 210.
    Renewed Facility Operating License No. DPR-23. Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: November 23, 2004 (69 
FR 68182). The letters dated June 22, 2005, June 26, 2006, and 
September 18, 2006, provided clarifying information that did not change 
the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 22, 2006.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423, 
Millstone Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: May 27, 2004, as supplemented 
September 27, 2004, October 20, 2004, March 23, 2005, January 30, 2006 
and May 25, 2006.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to incorporate a full-scope application of an 
alternate source term methodology in accordance with 10 CFR 50.67.
    Date of issuance: September 15, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 232.
    Facility Operating License No. NPF-49: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: September 14, 2004 (69 
FR 55468). The supplements contained clarifying information only, and 
did not change the initial no significant hazards consideration 
determination or expand the scope of the initial Federal Register 
notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 15, 2006.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423, 
Millstone Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: September 13, 2005, as 
supplemented by letters dated June 13 and August 14, 2006.
    Brief description of amendment: The amendment revised the Technical 
Specification (TS) surveillance requirements for the recirculation 
spray system.
    Date of issuance: September 20, 2006.
    Effective date: As of the date of issuance and shall be implemented 
prior to entering Mode 1 following refueling outage 3R11.
    Amendment No.: 233.
    Facility Operating License No. NPF-49: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: October 25, 2005 (70 FR 
61657). The supplements dated June 13 and August 14, 2006, 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 amendment is contained 
in a Safety Evaluation dated September 20, 2006.
    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: June 29, 2005, as supplemented 
May 1, 2006, and September 8, 2006.
    Brief description of amendments: The amendments requested 
authorization to revise the Updated Final Safety Analysis Report and 
the emergency operating procedures to allow an additional operator 
action to manually start one containment air return fan in the air 
return system in response to Nuclear Regulatory Commission Bulletin 
2003-01, ``Potential Impact of Debris Blockage on Emergency Sump 
Recirculation at Pressurized-Water Reactors,'' June 6, 2003.
    Date of issuance: September 25, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 231 and 227.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the licenses.
    Date of initial notice in Federal Register: October 25, 2005 (70 FR 
61657).
    The supplement dated May 1, 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 September 25, 2006.
    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: July 25, 2005, as supplemented 
July 28, 2005, and August 1, 2005.
    Brief description of amendments: The amendments revised the 
Technical

[[Page 59535]]

Specifications temperature limit for the standby nuclear service water 
pond from 91.5 [deg]F to 95 [deg]F.
    Date of issuance: September 25, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance September 25, 2006.
    Amendment Nos.: 232 and 228.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: August 4, 2005 (70 FR 
44946).
    The supplements dated July 28, 2005, and August 1, 2005, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 25, 2006.
    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: September 13, 2005, as 
supplemented March 20, 2006.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) to correct a nonconservative TS 
associated with spent fuel storage in the spent fuel pool. The licensee 
identified the nonconservative TS while comparing results from spent 
fuel pool criticality codes.
    Date of issuance: September 27, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance September 27, 2006.
    Amendment Nos.: 233 and 229.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the license and the technical specifications.
    Date of initial notice in Federal Register: November 21, 2005 (70 
FR 70104). The supplement dated March 20, 2006, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the Nuclear 
Regulatory Commission (NRC) staff's original proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 27, 2006.
    No significant hazards consideration comments received: No.

Duke Power 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: June 29, 2005, as supplemented 
May 1, 2006.
    Brief description of amendments: The amendments requested 
authorization to revise the Updated Final Safety Analysis Report and 
the emergency operating procedures to allow an additional operator 
action to manually start one containment air return fan in the air 
return system in response to Nuclear Regulatory Commission Bulletin 
2003-01, ``Potential Impact of Debris Blockage on Emergency Sump 
Recirculation at Pressurized-Water Reactors,'' June 6, 2003.
    Date of issuance: September 25, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 234 and 216.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the licenses.
    Date of initial notice in Federal Register: October 25, 2005 (70 FR 
61657).
    The supplement dated May 1, 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 September 25, 2006.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: June 29, 2005, as supplemented 
by letter dated May 18, 2006.
    Brief description of amendment: By letter dated June 29, 2005, 
Entergy Operations, Inc., the licensee for Arkansas Nuclear One, Unit 2 
(ANO-2), requested a license amendment to relocate the shutdown cooling 
(SDC) open permissive interlock (OPI) license condition from the 
operating license to the licensee's technical requirements manual. The 
license condition to maintain OPI operability was previously accepted 
by the NRC staff in a letter to the licensee, dated March 30, 2005, and 
incorporated into ANO-2's operating license. The OPI prevents the two 
SDC suction isolation valves from opening above a selected set point to 
separate the high-pressure reactor coolant system from the low-pressure 
SDC system.
    Date of issuance: September 13, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 267.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: December 6, 2005 (70 FR 
72671). The supplement dated May 18, 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 as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 13, 2006.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of application for amendment: December 14, 2004, revised by 
letter dated August 30, 2006.
    Brief description of amendment: The Technical Specification 
amendment relocates structural integrity requirements to the Final 
Safety Analysis Report.
    Date of issuance: September 14, 2006.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 224.
    Facility Operating License No. DPR-35: The amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: March 1, 2005 (70 FR 
9991).
    The licensee originally requested for additional TS relocations in 
their submittal dated December 14, 2004. The NRC staff found these 
unacceptable. Therefore, the licensee revised the original application 
by letter dated August 30, 2006, reducing the scope of the application 
as originally noticed. Hence, there is no change to the NRC staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a

[[Page 59536]]

Safety Evaluation dated September 14, 2006.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of application for amendment: June 2, 2005, supplemented by 
letter dated June 14, 2006.
    Brief description of amendment: The amendment revised the Technical 
Specification (TS) reactor coolant system leakage detection 
instrumentation requirements and actions.
    Date of issuance: September 20, 2006.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 225.
    Facility Operating License No. DPR-35: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: May 23, 2006 (71 FR 
29676). The supplement dated June 14, 2006, provided additional 
information that 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 amendment is contained 
in a Safety Evaluation dated September 20, 2006.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit No. 2, Beaver County, Pennsylvania

    Date of application for amendment: April 11, 2005, as supplemented 
December 2, 2005, and January 27, April 14, August 16, and September 1, 
2006.
    Brief description of amendment: The amendment revised the scope of 
the steam generator tubesheet inspections and subsequent repair using 
the F* inspection methodology.
    Date of issuance: September 27, 2006.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No: 160.
    Facility Operating License No. NPF-73: The amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: June 7, 2005 (70 FR 
33214). The supplements dated December 2, 2005, and January 27, April 
14, August 16, and September 1, 2006, provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the NRC staff's 
original proposed no significant hazards consideration determination. 
The Commission's issuance of Amendment No. 158 to Facility Operating 
License NPF-73 for BVPS-2, regarding steam generator tube integrity 
(Technical Specification Task Force (TSTF) Item 449) on September 7, 
2006, resulted in renumbering and rewording the requirements as 
originally proposed by the licensee to fit the TSTF-449 format, but did 
not change the scope of the application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 27, 2006.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of application for amendments: November 7, 2005, as 
supplemented April 25, June 1, and August 3, 2006.
    Brief description of amendments: The amendments include changes to 
the definition of leakage, changes to the primary-to-secondary leakage 
requirements, changes to the steam generator (SG) tube surveillance 
program (SG tube integrity), and changes to the SG reporting 
requirements.
    Date of issuance: September 7, 2006.
    Effective date: As of the date of issuance to be implemented within 
90 days for BVPS-1 and prior to entry into Mode 4 following the fall 
2006 refueling outage for BVPS-2.
    Amendment Nos.: 276 and 158.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications and Licenses.
    Date of initial notice in Federal Register: December 20, 2005 (70 
FR 75491). The supplements dated April 25, June 1, and August 3, 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 as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 7, 2006.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of application for amendments: February 17, 2005, as 
supplemented May 12 and August 22, 2006.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) 3.7.7.1 (BVPS-1), ``Control Room 
Emergency Habitability Systems,'' and 3.7.7 (BVPS-2), ``Control Room 
Emergency Air Cleanup and Pressurization System,'' by dividing these 
TSs into two specifications, addressing control room emergency 
ventilation and control room air cooling functions separately. The 
amendments also improved consistency with the Standard TSs and improved 
consistency between the units.
    Date of issuance: September 25, 2006
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos.: 277 and 159
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the License and Technical Specifications.
    Date of initial notice in Federal Register: April 26, 2005 (70 FR 
21458). The supplements dated May 12 and August 22, 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 as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 25, 2006.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, 
Pennsylvania

    Date of application for amendments: February 1, 2006.
    Brief description of amendments: The amendments revise the 
Technical Specification (TS) testing frequency for the Surveillance 
Requirements in TS 3.1.4, ``Control Rod Scram Times,'' based on the TS 
Task Force (TSTF) change traveler TSTF-222, Revision 1.
    Date of issuance: September 12, 2006.
    Effective date: As of the date of issuance and to be implemented 
within 60 days.
    Amendment Nos.: 237 and 214.

[[Page 59537]]

    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the License and TSs.
    Date of initial notice in Federal Register: May 9, 2006 (71 FR 
27001).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 12, 2006.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, 
Fairfield County, South Carolina

    Date of application for amendment: June 30, 2005, as supplemented 
July 21, 2006.
    Brief description of amendment: The amendment revises the Virgil C. 
Summer Nuclear Station Technical Specifications to permit the use of a 
best estimate methodology in performing loss-of-coolant accident 
analyses.
    Date of issuance: September 7, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No. 176.
    Renewed Facility Operating License No. NPF-12: Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: October 11, 2005 (70 FR 
59087). The supplemental letter provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 7, 2006.
    No significant hazards consideration comments received No.

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

    Date of application for amendments: November 30, 2005, as 
supplemented by letter dated May 30, 2006.
    Brief description of amendments: The proposed amendments revised 
the Technical Specification (TS) requirements related to steam 
generator tube integrity, based on the NRC-approved Revision 4 to TS 
Task Force (TSTF)-449, ``Steam Generator Tube Integrity.'' Date of 
issuance: September 19, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 2--204; Unit 3--196.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: February 14, 2006 (71 
FR 7812). The May 30, 2006, supplemental letter provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 19, 2006.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-259 Browns Ferry Nuclear 
Plant, Unit 1, Limestone County, Alabama

    Date of application for amendment: March 9, 2004 (TS-434) as 
supplemented on November 15, 2004, and March 7, 2006.
    Brief description of amendment: The amendment reduced the Allowable 
Value used for Reactor Vessel Water Level--Low, Level 3, for several 
instrument functions.
    Date of issuance: September 18, 2006.
    Effective date: September 18, 2006.
    Amendment No.: 258.
    Renewed Facility Operating License No. DPR-33: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19575). The supplements dated November 15, 2004, and March 7, 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 as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 18, 2006.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear 
Plant, Unit 1, Limestone County, Alabama

    Date of application for amendments: November 3, 2003, as 
supplemented May 6, 2004, and August 1, 2006.
    Description of amendment request: The amendment revised Technical 
Specification (TS) Table 3.3.1.1 -1, Reactor Protection system 
Instrumentation, Function 7.b.
    Date of issuance: September 19, 2006.
    Effective date: Date of issuance, to be issued within 60 days.
    Amendment No.: 259.
    Renewed Facility Operating License No. DPR-33: Amendment revised 
the TSs.
    Date of initial notice in Federal Register: April 13, 2004 (69 FR 
19575). The supplements dated May 6, 2004, and August 1, 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 as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 19, 2006.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: August 16, 2004, as 
supplemented by letters dated March 11, 2005, November 4, 2005, and 
April 14, 2006.
    Description of amendment request: To extend the channel calibration 
frequency requirements for instrumentation in the high-pressure coolant 
injection, reactor core isolation cooling, and reactor water core 
isolation cooling systems.
    Date of issuance: September 21, 2006.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment Nos.: 260, 297 and 255.
    Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 23, 2006 (71 FR 
29680). The supplemental letters provided clarifying information that 
did not expand the scope of the original application or change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 21, 2006.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear 
Plant, Unit 1, Limestone County, Alabama

    Date of application for amendment: July 9, 2004 (TS 436).
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Surveillance Requirement 3.6.1.3.10 to increase the

[[Page 59538]]

allowed main steam isolation valve (MSIV) leak rate from 11.5 standard 
cubic feet per hour (scfh) per valve to 100 scfh for individual MSIVs 
with a 150 scfh combined leakage for all four main steam lines.
    Date of issuance: September 27, 2006.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment No.: 261.
    Facility Operating License No. DPR-33: Amendment revised the TSs.
    Date of initial notice in Federal Register: May 23, 2006 (71 FR 
29680).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 27, 2006.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: August 14, 2004, as 
supplemented on April 11, 2005, and July 11, 2006 (TS-02-01).
    Brief description of amendments: The amendments revise Technical 
Specifications (TSs) relating to the reactor protection system and 
engineered safety features instrumentation. The Trip Setpoint column of 
TS Tables 2.2-1 and 3.3-4 will be renamed Nominal Trip Setpoint; 
inequality signs in TS Tables 2.2-1 and 3.3-4 will be removed; the trip 
setpoint and allowable value for the Intermediate Range Neutron Flux P-
6 permissive will be revised; Minimum Channels Operable in TS Table 
3.3-3 will be revised; editorial changes will be made to TS Table 3.3-4 
to replace  signs with inequalities; and a correction will 
be made to an alarm/trip setpoint in TS Table 3.3-6.
    Date of issuance: September 13, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos. 310 and 299.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the technical specifications.
    Date of initial notice in Federal Register: October 12, 2004 (69 FR 
60688). The supplemental letters provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 13, 2006.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: September 30, 2004, as 
supplemented on May 25, 2006.
    Brief description of amendments: The amendments revise the 
technical specifications to relocate the requirements for the emergency 
diesel generator start loss of power instrumentation and associated 
actions in the engineering safety features tables to a new limiting 
condition for operation (LCO). In addition, an upper allowable value 
limit has been added to the voltage sensors for loss of voltage and 
degraded voltage consistent with Technical Specification Task Force 
(TSTF) Item, TSTF-365, along with a lower allowable value limit for the 
degraded voltage diesel generator start and load shed timer. The 
auxiliary feedwater loss of power start setpoints and allowable values 
have been relocated to this new LCO.
    Date of issuance: September 14, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos. 311 and 300.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the technical specifications.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2900). The supplemental letter provided clarifying information that was 
within the scope of the initial notice and did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 14, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 29, 2006.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.4.15, ``RCS [Reactor Coolant System] Leakage 
Detection Instrumentation.'' The TS changes delete the containment 
atmosphere gaseous radioactivity monitor from TS 3.4.15 and revise the 
existing conditions, required actions, completion times, and 
surveillance requirements in TS 3.4.15 to account for the monitor being 
deleted. The June 29, 2006, letter superceded the license amendment 
request in the August 26, 2005, letter to authorize changes to the 
Final Safety Analysis Report.
    Date of issuance: September 26, 2006.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days from the date of issuance.
    Amendment No.: 175.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 24, 2006 (71 FR 
41843). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 26, 2006.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: June 26, 2006.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.4.15, ``RCS [Reactor Coolant System] Leakage 
Detection Instrumentation.'' The TS changes delete the monitor from TS 
3.4.15 and revise the existing conditions, required actions, completion 
times, and surveillance requirements in TS 3.4.15 to account for the 
monitor being deleted. The June 26, 2006, letter superceded the license 
amendment request in the August 26, 2005, letter to authorize changes 
to the Updated Safety Analysis Report.
    Date of issuance: September 26, 2006.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 166.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 24, 2006 (71 FR 
41848).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 26, 2006.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 2nd day of October 2006.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
 [FR Doc. E6-16560 Filed 10-6-06; 8:45 am]
BILLING CODE 7590-01-P