Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 56189-56199 [06-8014]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 186 (Tuesday, September 26, 2006)]
[Notices]
[Pages 56189-56199]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 06-8014]


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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 1, 2006, to September 14, 2006. 
The last biweekly notice was published on September 12, 2006 (71 FR 
53715).

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.

[[Page 56190]]

    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, https://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 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, HearingDocket@nrc.gov; 
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 
OGCMailCenter@nrc.gov. 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, https://
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 pdr@nrc.gov.

[[Page 56191]]

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

    Date of amendments request: July 20, 2006.
    Description of amendments request: The proposed amendments would 
revise Technical Specification 3.1.6, ``Shutdown Control Element 
Assembly (CEA) Insertion 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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Safety analyses require that the shutdown CEAs insert into the 
core at least 90% within 4 seconds of the safety signal initiating 
the shutdown sequence with the assumption that the shutdown CEAs' 
starting positions are at 150 inches withdrawn. This assumption will 
not be altered with the new proposed withdrawal limit.
    The positioning of control rods (shutdown CEAs) to a new limit 
of >=147.75 inches withdrawn is not a precursor to any accident 
analyzed at Palo Verde nor do these conditions affect any accident 
precursor; thus, initial control rod position does not change the 
probability of an accident previously evaluated.
    To assess the effect control rod position would have on the 
safety analyses with the rods positioned at the new limit, several 
events and specific parameters were analyzed. The events were chosen 
because of their sensitivity to rod position. The specific 
parameters were analyzed to determine if, with the rods positioned 
at the new limit, the power distribution in the core was still 
within the assumptions made in the safety analyses.
    Since none of the related safety analyses resulted in a 
significant change in the previously calculated values and the 
limiting parameters associated for those analyses were not exceeded, 
the consequences of these accidents remain unchanged. Therefore, the 
new insertion limit for the shutdown CEAs will not increase the 
consequences of any accident analyzed in our licensing bases 
documents.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different accident from any accident previously evaluated?
    Response: No.
    PVNGS [Palo Verde Nuclear Generating Station] licensing bases 
documents describe the design function of the control rods as 
components that include a positive means (gravity) for inserting the 
control rods and are capable of reliably controlling the nuclear 
reactor to assure that under conditions of normal operation, 
including anticipated accidents, fuel design limits are not 
exceeded. The proposed amendment, new control rod (shutdown CEA) 
insertion limit, does not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
nor does it affect the control rods ability to perform its design 
function.
    Control rods placed at the new insertion limit will not cause 
fuel design limits to be exceeded during normal operations or 
accidents. Placing the control rods at the new insertion limit in no 
way impedes their insertion due to gravity. These CEAs are tested to 
ensure that they will insert greater than 90% into the core in less 
than 4 seconds from a completely withdrawn position (150 inches) and 
this requirement will continue to be met.
    Establishing a new insertion limit for the control rods does not 
modify any of the existing components or systems used to position 
the control rods. The new insertion limit will also satisfy the 
assumptions made in the safety analyses.
    In conclusion, the new insertion limit stills [sic] allows the 
control rods to fulfill their design function and does not create a 
new or different accident than is already described in the licensing 
bases documents. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment, new shutdown CEA insertion limit, does 
not involve a reduction in the margin of safety. The new shutdown 
CEA insertion limit does not affect any of the limits used to 
determine the acceptability of newly designed cores. The safety 
analyses in the licensing bases documents remain acceptable when 
this new (more restrictive) shutdown CEA insertion limit is applied. 
Additionally, the design basis of the control rods is unaffected by 
the new insertion limit. The design function of the control rods is 
to provide a positive means (gravity) for inserting the control rods 
and is capable of reliably controlling the nuclear reactor to assure 
that under conditions of normal operation, including anticipated 
accidents, fuel design limits are not exceeded. Since the bounding 
safety analyses limits used remain the same and the control rod 
design basis is unaffected, the fuel design limits associated with 
the clad material; which houses the fuel; and the design limits of 
the coolant system; which houses the fuel assemblies; remain 
unchanged. Therefore, the margin of safety is not reduced.
    In conclusion, since the bounding limits used for safety 
analyses are unaffected by the new shutdown CEA insertion limit, the 
safety limits associated with the fuel and the coolant system remain 
unchanged. The design basis on the control rods is to ensure the 
fuel safety limits are not exceeded and since they remain unchanged, 
the design basis is still achieved. Therefore, there is no reduction 
in the margin of safety.
    Therefore, APS [Arizona Public Service] has concluded that the 
proposed license amendment request does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Janet S. Mueller, Director, Law Department, 
Arizona Public Service Company, P.O. Box 52034, Mail Station 8695, 
Phoenix, Arizona 85072-2034.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: August 2, 2006.
    Description of amendment request: The proposed change would delete 
Waterford 3 Technical Specification Surveillance Requirement (SR) 
4.6.1.7.2. This SR is the augmented testing requirement for containment 
purge supply and exhaust isolation valves with resilient seal materials 
and allows the surveillance intervals to be set in accordance with the 
Containment Leakage Rate Testing Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This change deletes the augmented testing requirement for these 
containment isolation valves and allows the surveillance intervals 
to be set in accordance with the Containment Leakage Rate Testing 
Program. This change does not affect the system function or design. 
The purge valves are not an initiator of any previously analyzed 
accident. Leakage rates do not affect the probability of the 
occurrence of any accident. Operating history has demonstrated that 
the valves do not degrade and cause leakage as previously 
anticipated. Because these valves have been demonstrated to be 
reliable, these valves can be expected to perform the containment 
isolation function as assumed in the accident analyses. Therefore, 
there is no significant increase in the consequences of any 
previously evaluated accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 56192]]

    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Extending the test intervals has no influence on, nor does it 
contribute in any way to, the possibility of a new or different kind 
of accident or malfunction from those previously analyzed. No change 
has been made to the design, function or method of performing 
leakage testing. Leakage acceptance criteria have not changed. No 
new accident modes are created by extending the testing intervals. 
No safety-related equipment or safety functions are altered as a 
result of this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The only margin of safety that has the potential of being 
impacted by the proposed change involves the offsite dose 
consequences of postulated accidents which are directly related to 
the containment leakage rate. The proposed change does not alter the 
method of performing the tests nor does it change the leakage 
acceptance criteria. Sufficient data has been collected to 
demonstrate these resilient seals do not degrade at an accelerated 
rate.
    Because of this demonstrated reliability, this change will 
provide sufficient surveillance to determine an increase in the 
unfiltered leakage prior to the leakage exceeding that assumed in 
the accident analysis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1700 
K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York

    Date of amendment request: May 11, 2006.
    Description of amendment request: The proposed change would revise 
Technical Specification (TS) 3.1.7, ``Standby Liquid Control (SLC) 
System,'' to change the minimum required SLC pump discharge pressure 
specified in surveillance requirement (SR) SR 3.1.7.7 from 1235 psig to 
1320 psig. This change is in response to Nuclear Regulatory Commission 
Information Notice 2001-13, ``Inadequate Standby Liquid Control System 
Relief Valve Margin.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the surveillance requirements for 
the SLC system to correspond to the maximum expected pressure in the 
reactor pressure vessel for an ATWS [anticipated transient without 
scram] event. This proposed increase in the specified SLC pump 
discharge pressure involves only the SLC system. No other NMP2 
structures, systems, or components are affected. The SLC system is 
provided to mitigate ATWS events and, as such, is not considered to 
be an initiator of an ATWS event or any other analyzed accident. The 
revised TS surveillance requirement, and the associated change to 
the SLC pump discharge relief valve set pressure (not described in 
the TS), neither reduce the ability of the SLC system to respond to 
and mitigate an ATWS event nor increase the likelihood of a system 
malfunction that could increase the consequences of an accident. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to the SLC pump TS surveillance requirement, 
and the associated change to the SLC pump discharge relief valve set 
pressure (not described in the TS), are consistent with the 
functional requirements of the ATWS rule (10 CFR 50.62). The 
proposed change does not involve the installation of any new or 
different type of equipment, does not introduce any new modes of 
plant operation, and does not change any methods governing normal 
plant operation. The proposed change does not introduce any new 
accident initiators, and therefore does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not alter any assumptions, initial 
conditions or results from any accident analyses. The proposed 
change to the SLC pump TS surveillance requirement, and the 
associated changes to the SLC pump discharge relief valve set 
pressure (not described in the TS), are consistent with the 
functional requirements of the ATWS rule (10 CFR 50.62). The ability 
of the SLC system to respond to and mitigate an ATWS event is not 
affected. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Richard J. Laufer.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York

    Date of amendment request: August 11, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3.2.1, ``Control Rod Block 
Instrumentation,'' to revise the number of startups allowed with the 
rod worth minimizer (RWM) inoperable from one per calendar year to two 
per operating cycle (approximately 2 years).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change redefines the frequency at which plant 
startup is permitted without using the RWM. The relevant design 
basis accident is the control rod drop accident (CRDA), which 
involves multiple failures to initiate the event. This 
administrative change does not increase the probability of 
occurrence of any of the failures that are necessary for a CRDA to 
occur. Use of the RWM or the alternate use of a qualified human 
checker to ensure the correct control rod withdrawal sequence is not 
in itself an accident initiator, and redefining the startup 
allowance frequency does not involve any plant hardware changes or 
new operator actions that could serve to initiate a CRDA. The 
proposed change will have no adverse effect on plant operation, or 
the availability or operation of any accident mitigation equipment. 
Also, since the banked position withdrawal sequence (BPWS) will 
continue to be enforced by either the RWM or verification by a 
second qualified individual, the initial conditions of the CRDA 
radiological consequence analysis presented in the U[F]SAR [Updated 
Final Safety Analysis Report] are not affected.

[[Page 56193]]

Therefore, there will be no increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not introduce any new modes of plant 
operation and will not result in a change to the design function or 
operation of any structure, system, or component that is used for 
accident mitigation. The proposed redefinition of the frequency at 
which plant startup is permitted without using the RWM does not 
result in any credible new failure mechanisms, malfunctions, or 
accident initiators not considered in the design and licensing 
basis. This administrative change does not affect the ability of 
safety-related systems and components to perform their intended 
safety functions. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change redefines the frequency at which plant 
startup is permitted without using the RWM. This administrative 
change does not affect the overall frequency of use of the 
allowance. The proposed change will have no adverse affect on plant 
operation or equipment important to safety. The relevant design 
basis accident is the control rod drop accident (CRDA), which 
involves multiple failures to initiate the event. The CRDA analysis 
consequences and related initial conditions remain unchanged when 
invoking the proposed change. The plant response to the CRDA will 
not be affected and the accident mitigation equipment will continue 
to function as assumed in the accident analysis. Therefore, there 
will be no 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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Richard J. Laufer.

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

    Date of amendment request: July 5, 2006.
    Description of amendment request: The proposed change would revise 
the main control room (MCR) and emergency switchgear room (ESGR) air 
conditioning system (ACS) Technical Specifications (TSs) to reflect the 
completion of permanent modifications to the equipment and associated 
power supply configuration. The revisions include the addition of 
requirements and/or action statements addressing the inoperability of 
two or more air handling units (AHUs) on a unit, as well as AHUs 
powered from an H emergency bus. The proposed change, paralleling 
requirements in the Improved Technical Specifications (ITS), also adds 
MCR and ESGR ACS requirements during refueling operations and 
irradiated fuel movement in the fuel building. In addition, the 
proposed change clarifies the service water (SW) requirements for the 
ACS chillers that serve the MCR and ESGRs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed change does not impact the condition or performance 
of any plant structure, system, or component. The proposed change 
does not affect the initiators of analyzed events or the assumed 
mitigation of accident or transient events. No physical changes to 
the ACS or SW System are involved, and accident operation of the ACS 
will not change. As a result, the proposed change to the Surry 
Technical Specifications does not involve any significant increase 
in the probability or the consequences of any accident or 
malfunction of equipment important to safety previously evaluated 
since neither accident probabilities nor consequences are being 
affected by this proposed change.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The proposed change does not involve a physical alteration of 
the plant or a change in the methods used to respond to plant 
transients. No new or different equipment is being installed, and no 
installed equipment is being removed. There is no alteration to the 
parameters with which the plant is normally operated or in the 
setpoints, which initiate protective or mitigative actions. The ACS 
will continue to perform its required function. Consequently, no new 
failure modes are introduced by the proposed change. Therefore, the 
proposed change to the Surry Technical Specifications does not 
create the possibility of a new or different kind of accident or 
malfunction of equipment important to safety from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed TS change does not impact any plant structure, 
system, or component that is relied upon for accident mitigation. 
Margin of safety is established through the design of the plant 
structures, systems, and components, the parameters within which the 
plant is operated, and the establishment of the setpoints for the 
actuation of equipment relied upon to respond to an event. Since ACS 
performance is not affected by the proposed change, the ACS will 
continue to be available to perform its required function. 
Furthermore, the change does not affect the condition or performance 
of structures, systems, or components relied upon the accident 
mitigation or any safety analysis assumptions. Therefore, the 
proposed change to the Surry Technical Specifications does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Branch Chief: Evangelos C. Marinos.

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

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

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of amendment request: July 20, 2006.
    Brief description of amendment request: The proposed amendment 
would revise the Vogtle Electric Generating Plant (VEGP), Units 1 and 
2, Technical Specifications (TS) 5.5.9,

[[Page 56194]]

``Steam Generator (SG) Tube Surveillance Program,'' to incorporate 
changes in the SG inspection scope for VEGP, Unit 1 during Refueling 
Outage 13 and the subsequent operating cycle, and VEGP Unit 2 during 
Refueling Outage 12 and the subsequent operating cycle. The proposed 
changes modify the inspection requirements for portions of SG tubes 
within the tubesheet region of the SGs.
    Date of publication of individual notice in Federal Register: July 
31, 2006 (71 FR 43225).
    Expiration date of individual notice: 30-day August 30, 2006; 60-
day, September 29, 2006.

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, https://
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 pdr@nrc.gov.

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

    Date of application for amendment: October 18, 2005, as 
supplemented by letter dated May 26, 2006.
    Brief description of amendment: The amendment revised the Oyster 
Creek Nuclear Generating Station Technical Specifications (TSs) 
Surveillance Requirement (SR) 4.4.B.1 to provide an alternative means 
for testing the electromatic relief valves located on the main steam 
system. The revised SR allows demonstration of the capability of the 
valves to perform their function without requiring that the valves be 
cycled with steam pressure while installed.
    Date of Issuance: September 1, 2006.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 260.
    Facility Operating License No. DPR-16: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: December 20, 2005 (70 
FR 75490). The May 26, 2006, letter provided clarifying information 
within the scope of the original application and did not change the 
staff's initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated September 1, 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: October 27, 2004.
    Brief description of amendments: The amendments revised the 
facility operating licenses by removal of Section 2.E, that lists 
reporting requirements with regard to Maximum Power Level, Updated, 
Fire Protection, Protection of the Environment (Unit 2 only) and 
Physical Protection.
    Date of issuance: September 7, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 233 and 215.
    Renewed Facility Operating License Nos. NPF 9 and NPF-17: 
Amendments revised the licenses.
    Date of initial notice in Federal Register: July 5, 2005 (70 FR 
38717).
    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.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: April 27, 2005, as supplemented 
by letters dated November 22, 2005, and August 1, 2006. The August 1, 
2006, submittal reduced the scope of the changes to only revise 
Technical Specification Limiting Condition for Operation 3.8.4, ``DC 
Sources-Operating.''
    Brief description of amendment: The amendment revises the Technical 
Specifications to allow a battery charger to be out of service for up 
to 7 days.
    Date of issuance: September 14, 2006.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 286.
    Facility Operating License No. DPR-59: The amendment revised the 
License and the Technical Specifications.
    Date of initial notice in Federal Register: July 19, 2005 (70 FR 
41444). The November 22, 2005, and August 1, 2006, supplements 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 as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 14, 2006.
    No significant hazards consideration comments received: No.

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

    Will County, Illinois
    Date of application for amendment: February 15, 2005, as 
supplemented by

[[Page 56195]]

letters dated November 28 and December 9, 2005 (two letters), and 
January 27, February 13, March 17 and July 14, 2006.
    Brief description of amendment: The amendments fully implement an 
alternative source term.
    Date of issuance: September 8, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment Nos.: 147, 147, 140 and 140.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications and License.
    Date of initial notice in Federal Register: May 10, 2005 (70 FR 
24650). The November 28 and December 9, 2005 (two letters), and January 
27, February 13, March 17 and July 14, 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 September 8, 2006.
    No significant hazards consideration comments received: No.

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

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

    Date of application for amendments: October 10, 2002, as 
supplemented by letters dated March 21, March 28, August 4, September 
15 and October 31, 2003, and June 30, August 6, September 3, September 
10, September 22, November 2 and November 5, 2004, and March 3, August 
22, September 3 and September 27, 2005, and February 17 and May 25, 
2006.
    Brief description of amendments: The amendments adopt the 
alternative source term methodology as prescribed in Title 10 to the 
Code of Federal Regulations Section 50.67.
    Date of issuance: September 11, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment Nos.: 221/212, 233/229.
    Renewed Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and 
DPR-30. The amendments revised the Technical Specifications, 
Surveillance Requirements and Licenses.
    Date of initial notice in Federal Register: August 19, 2003 (68 FR 
49816). The supplements dated March 21, March 28, August 4, September 
15 and October 31, 2003, and June 30, August 6, September 3, September 
10, September 22, November 2, and November 5, 2004, and March 3, August 
22, September 3 and September 27, 2005, and February 17 and May 25, 
2006, contained clarifying information and did not change the NRC 
staff's initial proposed finding of no significant hazards 
consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 11, 2006.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of application for amendment: April 10, 2006, as supplemented 
by letters dated April 12, 13 (2 letters), and June 27, 2006.
    Brief description of amendment: The amendment revised Surveillance 
Requirement 3.8.1.11 of the DCCNP-1 Technical Specifications, raising 
the diesel generator load rejection voltage test limit from 5000 volts 
to 5350 volts.
    Date of issuance: September 1, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 295.
    Facility Operating License No. DPR-58: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 1, 2006 (71 FR 
43534). The supplemental letters contained clarifying information and 
did not change the initial no significant hazards consideration 
determination, and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 1, 2006.
    No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: September 29, 2005, as supplemented by 
letters dated January 16, and April 7, 2006.
    Brief description of amendment: The amendment eliminated 
operability requirements for secondary containment, secondary 
containment isolation valves, the standby gas treatment system, and 
secondary containment isolation instrumentation when handling 
irradiated fuel that has decayed for 24 hours since critical reactor 
operations, and when performing core alterations. Similar technical 
specification relaxations are granted for the Control Room Emergency 
Filter System and its initiation instrumentation after a decay period 
of 7 days.
    Date of issuance: September 5, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 222.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 3, 2006 (71 FR 
149). The supplements dated January 16 and April 17, 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 5, 2006.
    No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: March 7, 2006, as supplemented by letter 
dated May 10, 2006.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Section 5.5.6, ``Inservice Testing Program,'' by 
replacing references to Section Xl of the American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code with ASME 
Code for Operation and Maintenance of Nuclear Power Plants (OM Code). 
Section 50.55a of Title 10 of the Code of Federal Regulations (CFR) 
requires that the Inservice Testing (IST) Program be updated to the 
latest Edition and Addenda of the Code incorporated by reference in 10 
CFR 50.55a(b) 12 months before the start of the applicable 10-year 
interval. Section Xl of the ASME Boiler and Pressure Vessel Code has 
been replaced with the ASME OM Code as the code of reference for IST 
programs. Thus, the ASME OM Code is the code of reference for the IST 
Program for the 10-year interval that began March 1, 2006. In addition, 
the amendment expanded the scope of frequencies specified to be within 
the applicability

[[Page 56196]]

of Surveillance Requirement (SR) 3.0.2 by adding mention of other 
normal and accelerated frequencies specified in the IST Program. This 
will eliminate any confusion regarding the applicability of SR 3.0.2 to 
IST Program Frequencies.
    Date of issuance: September 6, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 223.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 5, 2006 (71 FR 
38184).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 6, 2006.
    No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: July 1, 2005, as supplemented on 
September 16, 2005, November 15, 2005, December 14, 2005, February 16, 
2006, and July 6, 2006.
    Brief description of amendment: The amendment revises the Updated 
Safety Analysis Report, Section 14.10, ``Malfunctions of the Feedwater 
System,'' to describe an existing Emergency Operating Procedure 
operator action to isolate the steam generator blowdown within 15 
minutes of a reactor trip during a loss-of-main feedwater event.
    Date of issuance: September 11, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 242.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Updated Safety Analysis Report.
    Date of initial notice in Federal Register: August 2, 2005 (70 FR 
44403). The September 16, 2005, November 15, 2005, December 14, 2005, 
February 16, 2006, and July 6, 2006, supplemental letters provided 
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 amendment is contained 
in a safety evaluation dated September 11, 2006.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of application for amendments: September 26, 2005, as 
supplemented by letter dated June 28, 2006.
    Brief description of amendments: The proposed amendments revised 
the Salem Technical Specifications (TSs) to eliminate certain 
Surveillance Requirements (SRs) for containment isolation valves. The 
changes deleted SR 4.6.3.1.1 and SR 4.6.3.1 for Salem Unit Nos. 1 and 
2, respectively. These SRs require a complete valve stroke and stroke 
time measurement when a valve is returned to service after maintenance, 
repair, or replacement work. The changes are intended to minimize 
unnecessary testing and plant transients. Other Salem TS containment 
isolation valve SRs ensure that the valves remain operable.
    Date of issuance: August 31, 2006.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos.: 274 and 255.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the License and Technical Specifications.
    Date of initial notice in Federal Register: July 18, 2006 (71 FR 
40739).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 31, 2006.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of application for amendments: January 27, 2005, as 
supplemented by letters dated September 30, 2005, and January 25 and 
May 5, 2006.
    Brief description of amendments: The amendments revised the 
Technical Specifications by extending the surveillance test interval 
for components of the reactor protection system.
    Date of issuance: September 1, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 145 and 125
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: November 8, 2005 (70 FR 
67751). The supplements dated September 30, 2005, and January 25 and 
May 5, 2006, provided clarifying information that did not change the 
scope of the January 27, 2005, application nor the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 1, 2006.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of application for amendments: July 20, 2006, as supplemented 
by letter dated August 4, 2006.
    Brief description of amendments: The amendments revised Technical 
Specification 5.5.9, ``Steam Generator (SG) Tube Surveillance 
Program,'' regarding the required SG inspection scope for Vogtle, Unit 
1, during Refueling Outage 13 and the subsequent operating cycle and 
Vogtle, Unit 2, during Refueling Outage 12, and the subsequent 
operating cycle. The proposed changes modify the inspection 
requirements for portions of the SG tubes within the hot leg tubesheet 
region of the SGs.
    Date of issuance: September 12, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 146 and 126.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: July 31, 2006 (71 FR 
43225). The supplement dated August 4, 2006, provided clarifying 
information that did not expand the scope of the July 20, 2006, 
application nor the initial proposed no significant hazards 
consideration determination.
    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.

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: January 10, 2006 as 
supplemented by letters April 14, August 1, September 5 and 14, 2006.

[[Page 56197]]

    Description of amendment request: The amendments revised Technical 
Specifications 3.3.1.1 and 3.3.5.1 to specify the methodology used for 
determining, setting, and evaluating as-found setpoints for drift-
susceptible instruments that are necessary to ensure compliance with a 
Safety Limit or are critical in ensuring the fuel peak cladding 
temperature acceptance criterion are met.
    Date of issuance: September 14, 2006.
    Effective date: Date of issuance, to be implemented within 90 days.
    Amendment Nos.: 257, 296 and 254.
    Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 28, 2006 (71 FR 
15487). The supplements dated April 14, August 1, September 5 and 14, 
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 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 14, 2006.
    No significant hazards consideration comments received: No.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendments request: December 16, 2005, as supplemented by 
letter dated June 7, 2006.
    Brief description of amendments: The amendments revised the steam 
generator tube surveillance program technical specifications (TSs) to 
be consistent with TS Task Force (TSTF) traveler TSTF-449, Revision 4, 
``Steam Generator Tube Integrity.''
    Date of issuance: September 12, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 128/128.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 14, 2006 (71 FR 
13181). The supplement dated June 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 amendments is contained 
in a Safety Evaluation dated September 12, 2006.
    No significant hazards consideration comments received: No.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room (PDR), located at One White Flint North, Public File Area 01F21, 
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly 
available records will be accessible from the Agencywide Documents 
Access and Management System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC Web site, https://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,

[[Page 56198]]

(301) 415-4737 or by e-mail to pdr@nrc.gov.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. Within 60 days after the date 
of publication of this notice, 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 Com