[Federal Register Volume 72, Number 166 (Tuesday, August 28, 2007)]
[Notices]
[Pages 49568-49586]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-16766]



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Part III





Nuclear Regulatory Commission





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Biweekly Notice; Applications and Amendments to Facility Operating 
Licenses Involving No Significant Hazards Considerations; Notice

  Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / 
Notices  

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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 August 2, 2007, to August 15, 2007. The last 
biweekly notice was published on August 14, 2007 (72 FR 45454).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area O1F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted 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 basis 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

[[Page 49569]]

fact. Contentions shall be limited to matters within the scope of the 
amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

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

    Date of amendment request: June 29, 2007.
    Description of amendment request: The proposed license amendment 
would revise the TMI-1 Technical Specifications 3.3.1.3, 3.3.2.1 and 
4.1, to reflect a change to the Reactor Building spray system buffering 
agent from sodium hydroxide to trisodium phosphate dodecahydrate. This 
proposed change is designed to minimize the potential for exacerbating 
sump screen blockage under post loss of coolant event conditions by 
limiting potential adverse chemical interactions between the buffering 
agent and certain insulation materials used in the TMI-1 containment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    For the proposed change, trisodium phosphate dodecahydrate (TSP) 
will be used as a buffer for post-accident pH control and will 
replace the existing buffer. The buffer material and means of 
storage and delivery are not initiators for previously analyzed 
accidents. The accident mitigation function of the replacement 
buffer is the same as the existing buffer. The pH of the water in 
the emergency sump following a loss of coolant accident (LOCA) will 
be adjusted with TSP rather than sodium hydroxide (NaOH) to be 
within a range that will reduce the potential for elemental iodine 
re-evolution and long term stress corrosion during the recirculation 
mode of emergency core cooling system (ECCS) operation. In addition, 
the replacement buffer will reduce the formation of precipitates 
resulting from chemical reactions between the recirculating spray 
solution and insulating materials in the Reactor Building (RB), thus 
reducing the potential for ECCS emergency sump intake screen 
blockage. The proposed sump pH range will not result in an increase 
in post-LOCA hydrogen generation. The proposed isolation of the 
sodium hydroxide tank, and the installation of TSP in baskets has 
been evaluated for impacts on accident effects and the safety 
functions of required systems, structures, and components (SSCs). 
The RB emergency sump solution pH profile resulting from the 
proposed change has been evaluated for impacts on environmental 
qualification of SSCs. The accident mitigation functions of required 
SSCs will not be affected by the proposed change.
    As a part of the proposed change, the radiological consequences 
of a postulated LOCA have been reanalyzed using Standard Review Plan 
(SRP) 6.5.2, ``Containment Spray as a Fission Product Cleanup 
System,'' and the Alternate Source Term (AST) guidance in Regulatory 
Guide 1.183. The analysis considered the use of a plain borated 
water spray during the post-LOCA injection phase and a spray mixture 
with a minimum pH of 7.3 during the recirculation phase. The results 
of the reanalysis show that the consequences of the accident are not 
increased. The calculated doses at the Exclusion Area Boundary, Low 
Population Zone boundary, and in the Control Room remain within 10 
CFR 50.67 AST dose limits.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change will replace the existing spray additive 
design using sodium hydroxide solution stored in a tank with TSP 
contained in baskets located on the floor of the RB. The TSP storage 
and delivery method is passive. The baskets are constructed of 
stainless steel to resist corrosion and are seismically qualified. 
The existing sodium hydroxide tank, associated piping, and valves 
will no longer be used and will be permanently isolated, but their 
structural integrity will be maintained. The RB spray system will 
perform the same function and operate in the same manner for the 
proposed change; however, the sodium hydroxide tank isolation valves 
will no longer be required to open on an engineered safeguards 
actuation

[[Page 49570]]

signal. The accident mitigation function of TSP will be the same as 
the existing buffer, sodium hydroxide. The TSP will act as a 
buffering agent to raise the pH of the water in the containment 
emergency sump to greater than 7.3 for long-term post-LOCA RB spray 
recirculation. The SSCs required for post-LOCA accident mitigation 
have been evaluated for the proposed change including the effects of 
the modified emergency sump solution pH profile. No new accident 
scenarios, failure mechanisms, or single failures are introduced as 
a result of the proposed change.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change from sodium hydroxide to TSP will not reduce 
the effectiveness of the post-LOCA pH control buffer. The TSP will 
buffer the sump water sufficiently to assure that the resulting 
mixture pH is > 7.3 and < 8.0. This pH level will be effective in 
reducing the potential for iodine re-evolution during the 
recirculation phase of a LOCA, preventing long-term stress corrosion 
cracking of austenitic stainless steel, and minimizing post-LOCA 
hydrogen generation. In addition, the use of TSP will reduce the 
formation of precipitates resulting from chemical reactions between 
the recirculating spray solution and insulating materials in the RB, 
thus reducing the potential for ECCS emergency sump intake screen 
blockage. The proposed use of SRP 6.5.2 guidance, which is an NRC-
approved methodology, for post-LOCA dose calculations does not 
result in a reduction in a margin of safety. The proposed change 
does not adversely affect the performance of SSCs required for post-
LOCA mitigation, and does not affect an operating parameter or 
setpoint used in the accident analyses to establish a margin of 
safety. Also, the proposed change does not affect a margin of safety 
associated with containment functional performance.
    Therefore, the proposed change does not involve a significant 
reduction in any margin of safety.

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

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

    Date of amendment request: July 17, 2007.
    Description of amendment request: A change is proposed to the 
standard technical specifications (STS) (NUREGs 1430 through 1434) and 
plant specific technical specifications (TS), to strengthen TS 
requirements regarding control room envelope (CRE) habitability by 
changing the action and surveillance requirements associated with the 
limiting condition for operation operability requirements for the CRE 
emergency ventilation system, and by adding a new TS administrative 
controls program on CRE habitability. Accompanying the proposed TS 
change are appropriate conforming technical changes to the TS Bases. 
The proposed revision to the Bases also includes editorial and 
administrative changes to reflect applicable changes to the 
corresponding STS Bases, which were made to improve clarity, conform 
with the latest information and references, correct factual errors, and 
achieve more consistency among the STS NUREGs. The proposed revision to 
the TS and associated Bases is consistent with STS as revised by TSTF-
448, Revision 3.
    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:

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

    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

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

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Thomas H. Boyce.

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, WI

    Date of amendment request: June 12, 2007.

[[Page 49571]]

    Description of amendment request: The proposed amendment would 
revise the nuclear instrumentation system permissive setpoints in 
Technical Specification (TS) Table 3.5-2, ``Instrument Operation 
Conditions for Reactor Trip,'' revise the Table format, and revise TS 
2.3, ``Instrumentation System,'' to make consistent with other proposed 
changes to the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not change the probability or 
consequences of any previously evaluated accidents in the KPS 
[Kewaunee Power Station] updated safety analysis report (USAR). The 
proposed amendment would modify the TS setpoint values for the P-7 
and P-10 permissives. The actual plant settings will continue to be 
approximately 10% of rated reactor power. The reactor protection 
system (RPS) is designed to monitor various plant parameters and 
initiate a reactor trip in the event these parameters are outside 
predetermined limits. The RPS is not an accident initiator and 
therefore, changing the setpoints for these permissives will not 
increase the probability of an accident previously evaluated.
    The proposed amendment would add a setpoint band to the current 
TS required settings for permissive P-7 and P-10 to accommodate 
proper setting of the permissives. The only previously evaluated 
accident that is potentially affected by the proposed changes is the 
Uncontrolled Rod Cluster Assembly Rod Withdrawal At-Power (RWAP) 
accident analysis. The effects of these setpoint changes have been 
evaluated and determined not to have a significant effect on the 
consequences of the RWAP accident analysis results. The acceptance 
criteria for the RWAP accident analysis continue to be met. 
Therefore the proposed changes would not increase the consequences 
of an accident previously evaluated.
    Therefore the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment modifies the TS setpoint values for 
permissives P-7 and P-10. The actual plant settings will continue to 
be approximately 10% power. The proposed changes affect the power 
level at which RPS trip functions are enabled or blocked to ensure 
proper operation of the RPS. The changes do not add any new systems, 
structures or components (SSCs) or physically modify any existing 
SSCs with the possibility of creating a new accident.
    The proposed amendment does not functionally affect the 
operation of any SSC important to safety or its ability to perform 
its design function. Additionally, the proposed amendment does not 
create the possibility of a new or different kind of accident due to 
credible new failure mechanisms, malfunctions, or accident 
initiators not considered in the design and licensing bases.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment would add a setpoint band to the current 
TS required settings for permissivies P-7 and P-10 to accommodate 
proper setting of the permissives. The safety function of the 
nuclear instrumentation system and the affected permissives are not 
affected by this proposed change.
    The only safety analysis in the KPS USAR potentially affected by 
these proposed changes is the Uncontrolled Rod Cluster Assembly Rod 
Withdrawal At-Power (RWAP) event analysis. Evaluation of the RWAP 
event analysis results demonstrated that the RWAP would not have a 
significant effect on a margin of safety.
    The effects of the proposed change have been evaluated and all 
safety analysis acceptance criteria will continue to be met.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Acting Branch Chief: Travis L. Tate.

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station (KPS), Kewaunee County, Wisconsin

    Date of amendment request: July 2, 2007.
    Description of amendment request: The proposed amendment would 
delete operating license (OL) condition 2.C (5), ``Fuel Burnup,'' which 
restricts maximum rod average burnup to 60 giga-watt days per metric 
ton uranium (GWD/MTU). Deletion of the OL condition will provide the 
opportunity to increase maximum rod average burnup to as high as 62 
GWD/MTU and allow fuel management flexibility.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Deletion of KPS OL condition 2.C (5) does not add, delete, or 
modify any KPS systems, structures, or components (SSCs). The 
proposed amendment would effectively allow future increases in the 
KPS maximum rod average burnup limit using currently existing fuel 
management methods and models that have been reviewed and approved 
by the NRC [Nuclear Regulatory Commission].
    Maximum average rod burnup limits will continue to be maintained 
within safe and acceptable limits using these fuel management 
methods and models. Nuclear fuel is the only plant component 
potentially affected by increasing the maximum rod average burnup 
limit. Increasing the KPS maximum rod average burnup limit does not 
affect the thermal hydraulic response or the radiological 
consequences of any previously evaluated accident. The fuel rod 
design criteria will continue to be met at the maximum burnup limits 
allowed under the current fuel management and evaluation processes. 
An increase to the maximum rod average burnup limit will not 
increase the likelihood of a malfunction of nuclear fuel since the 
fuel currently used at KPS has been designed to support a maximum 
rod average burnup well in excess of 62 GWD/MTU.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment would delete a KPS OL condition that 
limits maximum rod average burnup. The proposed amendment would 
effectively allow future increases in the KPS maximum rod average 
burnup limit using currently existing fuel management methods and 
models that have been reviewed and approved by the NRC. Nuclear fuel 
is the only component potentially affected by changes to the maximum 
rod average burnup limit. The proposed amendment does not change the 
design function of the nuclear fuel or create any credible new 
failure mechanisms or malfunctions for nuclear fuel. Fuel rod design 
criteria will continue to be met at the maximum burnup limits 
allowed under the fuel management methods and models that have been 
previously reviewed and approved by the NRC. Therefore, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?

[[Page 49572]]

    Response: No.
    The proposed amendment deletes a KPS OL condition that limits 
maximum rod average burnup. The proposed amendment would effectively 
allow future increases in the KPS maximum rod average burnup limit 
using currently existing methods and models that have been reviewed 
and approved by the NRC. The proposed amendment does not result in 
altering or exceeding a design basis or safety limit for the plant. 
All current fuel design criteria will continue to be satisfied, and 
the safety analysis of record, including evaluations of the 
radiological consequences of design basis accidents, will remain 
applicable.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Acting Branch Chief: Travis L. Tate.

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

    Date of amendment request: July 26, 2007, as superseded by letter 
dated August 8, 2007.
    Description of amendment request: The proposed changes revise the 
requirements of Technical Specification (TS) 3.3.5.2, ``Reactor Core 
Isolation Cooling (RCIC) System Instrumentation,'' and TS 3.5.2, ``ECCS 
[Emergency Core Cooling System]--Shutdown,'' to increase the Condensate 
Storage Tank (CST) level.
    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 operation of Columbia in accordance with the proposed 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated. Neither of 
these changes affects the probability of any accident previously 
evaluated as they do not involve or impact accident initiators.
    The proposed change to TS 3.3.5.2 would ensure that the 
consequences would remain the same as that previously evaluated for 
during any event in which the RCIC pump was utilized. Adequate 
volume would be maintained in the CST whenever the RCIC pump was 
aligned to it to ensure that it did not experience loss of suction 
due to vortexing.
    The proposed changes to TS 3.5.2.2 would ensure that the 
previously assumed volume of water in the CST would still be 
available to inject into the reactor vessel during Modes 4 and 5 
should the suppression pool not meet minimum volume requirements. 
Therefore, operation of Columbia in accordance with the proposed 
amendment will 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 operation of Columbia in accordance with the proposed 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
change will not create a new or different kind of accident since it 
only affects the amount of water held in reserve to support reactor 
vessel inventory loss. The proposed change does not introduce any 
credible mechanisms for unacceptable radiation release nor does it 
require physical modification to the plant. The plant has operated 
well within the existing allowable values. The increased margin 
provided by the increased level will assure no new or different 
kinds of accidents result from the proposed change. Therefore, the 
operation of Columbia in accordance with the proposed amendment 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 operation of Columbia in accordance with the proposed 
amendment will not involve a significant reduction in the margin of 
safety. The proposed amendment provides assurance that the RCIC pump 
suction will be transferred without loss of suction and that 135,000 
gallons of CST inventory will continue to be available for injection 
into the RPV [reactor pressure vessel] under worst case conditions. 
Therefore, operation of Columbia in accordance with the proposed 
amendment will not involve a significant reduction in the margin of 
safety.

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

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

    Date of amendment request: July 30, 2007.
    Description of amendment request: The proposed changes revise 
Technical Specifications (TSs) 1.4, ``Frequency,'' 3.1.5, ``Control Rod 
Scram Accumulators,'' 3.4.1, ``Recirculation Loops Operating,'' 3.5.1, 
``ECCS [Emergency Core Cooling System]--Operating,'' 3.5.2, ``ECCS--
Shutdown,'' 3.7.1, ``Standby Service Water (SW) System and Ultimate 
Heat Sink (UHS),'' 3.8.1, ``AC [Alternating Current] Sources--
Operating,'' 3.8.2, ``AC Sources--Shutdown,'' and 5.5.6, ``Inservice 
Testing Program.'' The proposed changes include updates to adopt 
approved TS Task Force (TSTF) Travelers 284, Revision 3, ``Add `Met' 
vs. `Perform' to Specification 1.4, Frequency,'' TSTF-479, Revision 0, 
``Changes to Reflect Revision of 10 CFR 50.55a,'' and TSTF-485, 
Revision 0, ``Correct Example 1.4-1.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment is administrative in nature and does not 
affect analysis inputs or mitigation of analyzed accidents and 
transients. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment does not 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, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. The 
proposed change does not introduce any new modes of plant operation 
or make any changes to system setpoints. Therefore, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed amendment is administrative in nature and does not 
involve physical changes to plant SSCs [structures, systems, or 
components], or the manner in which these SSCs are operated, 
maintained, modified, tested, or inspected. The proposed amendment 
does not involve a change to any

[[Page 49573]]

safety limit, limiting safety system setting, limiting condition for 
operation, or design parameters for any SSC. The only minor 
alteration to the plant design basis is relative to the application 
of TS 3.4.1. However, as discussed in Section 4 [of the licensee's 
submittal], this alteration biases the operation of the plant in the 
direction of safety. The proposed amendment does not impact any 
safety analysis assumptions and does not involve a change in initial 
conditions, system response times, or other parameters affecting any 
accident analysis. For these reasons, the proposed amendment does 
not involve a significant reduction in the margin of safety.

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

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

    Date of amendment request: July 30, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) to establish more effective 
and appropriate action, surveillance, and administrative TS 
requirements related to ensuring the habitability of the control room 
envelope (CRE) in accordance with Nuclear Regulatory Commission (NRC)-
approved TS Task Force (TSTF) Standard Technical Specification change 
traveler TSTF-448, Revision 3, ``Control Room Habitability.'' 
Specifically, the proposed amendment would modify TS 3.7.3, ``Control 
Room Emergency Filtration (CREF) System,'' and add new TS 5.5.14, 
``Control Room Envelope Habitability Program,'' to Section 5.5, 
``Programs and Manuals.''
    The NRC staff issued a ``Notice of Availability of Technical 
Specification Improvement to Modify Requirements Regarding Control Room 
Envelope Habitability Using the Consolidated Line Item Improvement 
Process'' associated with TSTF-448, Revision 3, in the Federal Register 
on January 17, 2007 (72 FR 2022). The notice included a model safety 
evaluation, a model no significant hazards consideration (NSHC) 
determination, and a model license amendment request. In its 
application dated July 30, 2007, the licensee affirmed the 
applicability of the model NSHC determination which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee is presented below:

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

    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE [control room 
envelope] emergency ventilation system, which is a mitigation system 
designed to minimize unfiltered air leakage into the CRE and to 
filter the CRE atmosphere to protect the CRE occupants in the event 
of accidents previously analyzed. An important part of the CRE 
emergency ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

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

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation.
    The proposed change does not alter any safety analysis 
assumptions and is consistent with current plant operating practice. 
Therefore, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the analysis adopted by the licensee 
and, based upon this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendment involves NSHC.
    Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: July 30, 2007.
    Description of amendment request: The proposed changes revise 
Technical Specifications (TSs) 3.3.3.1, ``Post Accident Monitoring 
(PAM) Instrumentation,'' 3.3.6.1, ``Primary Containment Isolation 
Instrumentation,'' 3.6.1.3, ``Primary Containment Isolation Valves 
(PCIVs),'' and 3.6.4.2, ``Secondary Containment Isolation Valves 
(SCIVs).'' The proposed changes adopt the following TS Task Force 
(TSTF) Travelers that have been previously approved by the Nuclear 
Regulatory Commission (NRC): TSTF-45-A, Revision 2, ``Exempt 
Verification of CIVs [containment isolation valves] that are Not 
Locked, Sealed or Otherwise Secured,'' TSTF-46-A, Revision 1, ``Clarify 
the CIV Surveillance to Apply Only to Automatic Isolation Valves,'' 
TSTF-207-A, Revision 5, ``Completion Time for Restoration of Various 
Excessive Leakage Rates,'' TSTF-269-A, Revision 2, ``Allow 
Administrative Means of Position Verification for Locked or Sealed 
Valves,'' TSTF-295-A, Revision 0, ``Modify Note 2 to Actions of PAM 
Table to Allow Separate Condition Entry for Each Penetration,'' TSTF-
306-A, Revision 2, ``Add Action to LCO

[[Page 49574]]

[limiting condition for operation] 3.3.6.1 to Give Option to Isolate 
the Penetration,'' and TSTF-323-A, Revision 0, ``EFCV [excess flow 
check valve] Completion Time to 72 Hours.''
    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. The licensee addressed each 
proposed TSTF separately in its analysis:

TSTF-45-A, Revision 2

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change would exempt manual isolation valves and 
blind flanges located inside and outside the primary containment and 
in the secondary containment that are locked, sealed, or otherwise 
secured in position from the periodic verification of valve position 
required by SRs [surveillance requirements] 3.6.1.3.2 and 3.6.1.3.3, 
and SR 3.6.4.2.1. The exempted valves are verified to be in the 
correct position upon being locked, sealed, or secured. Because the 
valves are in the condition assumed in the accident analysis, the 
proposed change will not affect the initiators or mitigation of any 
accident previously evaluated.
    2. The proposed amendment does not 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 to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change replaces the periodic verification of valve 
position with verification of valve position followed by locking, 
sealing, or otherwise securing the valve in position. Periodic 
verification is also effective in detecting valve mispositioning. 
However, verification followed by securing the valve in position is 
effective in preventing valve mispositioning.

TSTF-46-A, Revision 1

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change would revise the verification of PCIV and 
SCIV closure time to clarify that only power operated, automatic 
valves are required to be tested. PCIVs and SCIVs are not an 
initiator of any accident previously evaluated; rather, they serve 
to mitigate the consequences of evaluated accidents. The proposed 
change does not change the requirement to verify that power 
operated, automatic PCIVs and SCIVs close within the time assumed in 
the accident analysis, but rather, clarifies that non-automatic 
valves, which the accident analysis does not assume close within a 
specified time, are not required to be tested to verify the closure 
time. As a result, the mitigating action of the PCIVs and SCIVs is 
not affected by this change.
    2. The proposed amendment does not 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 to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change would revise the verification of PCIV and 
SCIV closure time to clarify that only power operated, automatic 
valves are required to be tested, and not all power operated valves. 
There is no closure time assumed in the accident analysis for power 
operated PCIVs and SCIVs that are not automatic.

TSTF-207-A, Revision 5

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change revises the Actions of TS 3.6.1.3 to make 
the presentation consistent with similar Conditions in the ISTS 
[Improved Standard TSs]. Part of this change would extend the CT 
[completion time] for hydrostatically tested lines on a closed 
system to 72 hours for
    Condition D. Most of the proposed changes do not affect the 
requirements in the TS and have no effect on the initiation or 
mitigation of any accident previously evaluated. Leakage of 
hydrostatically tested lines on a closed system is not an initiator 
of any accident previously evaluated. The consequences of a 
previously evaluated accident during the extended CT are the same as 
the consequences during the existing CT.
    2. The proposed amendment does not 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 to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed changes are editorial in nature and do not affect 
the requirements of the TS. Extension of the CT for hydrostatically 
tested lines on a closed system to 72 hours does not represent a 
significant reduction in safety given the reliability of closed 
systems. Nonetheless, leakage can be isolated restored by isolating 
the penetration with a valve not exceeding the leakage limits.

TSTF-269-A, Revision 2

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The proposed change modifies TS 3.6.1.3 and TS 
3.6.4.2. Both TS 3.6.1.3 and TS 3.6.4.2 require penetrations with an 
inoperable isolation valve to be isolated and periodically verified 
to be isolated. A Note is added to TS 3.6.1.3, Actions A and C, and 
TS 3.6.4.2, Action A, to allow isolation devices that are locked, 
sealed, or otherwise secured to be verified by use of administrative 
means. The proposed change does not affect any plant equipment, test 
methods, or plant operation, and are not initiators of any analyzed 
accident sequence. The inoperable containment penetrations will 
continue to be isolated, and hence perform their isolation function. 
Operation in accordance with the proposed TS will ensure that all 
analyzed accidents will continue to be mitigated as previously 
analyzed.
    2. The proposed amendment does not 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 to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change will not affect the operation of plant 
equipment or the function of any equipment assumed in the accident 
analysis. The PCIVs and SCIVs will continue to be operable or will 
be isolated as required by the existing specifications.

TSTF-295-A, Revision 0

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change clarifies the separate condition entry Note 
in TS 3.3.3.1 for Function 7, ``PCIV Position.'' The proposed change 
does not affect any plant equipment, test methods, or plant 
operation, and are not initiators of any analyzed accident sequence. 
The actions taken for inoperable PAM channels are not changed. 
Operation in accordance with the proposed TS will ensure that all 
analyzed accidents will continue to be mitigated as previously 
analyzed.

[[Page 49575]]

    2. The proposed amendment does not 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 to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change will not affect the operation of plant 
equipment or the function of any equipment assumed in the accident 
analysis. The PAM channels will continue to be operable or the 
existing, appropriate actions will be followed.

TSTF-306-A, Revision 2

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change revises TS 3.3.6.1 by adding an Actions Note 
that would allow penetration flow paths to be unisolated 
intermittently under administrative controls. Furthermore, the TIP 
[traversing incore probe] isolation system is segregated into a 
separate Function, allowing 24 hours to isolate the penetration. The 
proposed change does not affect any plant equipment, test methods, 
or plant operation, and are not initiators of any analyzed accident 
sequence. The allowance to unisolate a penetration flow path will 
not have a significant effect on the mitigation of any accident 
previously evaluated because the penetration flow path can be 
isolated, if needed, by a dedicated operator. The option to isolate 
a TIP penetration will ensure the penetration will perform as 
assumed in the accident analysis. Operation in accordance with the 
proposed TS will ensure that all analyzed accidents will continue to 
be mitigated as previously analyzed.
    2. The proposed amendment does not 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 to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change will not affect the operation of plant 
equipment or the function of any equipment assumed in the accident 
analysis. The allowance to unisolate a penetration flow path will 
not have a significant effect on a margin of safety because the 
penetration flow path can be isolated manually, if needed. The 
option to isolate a TIP penetration will ensure the penetration will 
perform as assumed in the accident analysis.

TSTF-323-A, Revision 0

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change would revise Action C of TS 3.6.1.3 to 
provide a 72-hour CT instead of a 12 hour CT to isolate an 
inoperable EFCV. PCIVs are not an initiator of any accident 
previously evaluated. The consequences of a previously evaluated 
accident during the extended CT are the same as the consequences 
during the existing CT.
    2. The proposed amendment does not 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 to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The PCIVs serve to mitigate the potential for radioactive 
release from the primary containment following an accident. The 
design and response of the PCIVs to an accident are not affected by 
this change. The revised CT is appropriate given the EFCVs are on 
penetrations that have been found to have acceptable barrier(s) in 
the event that the single isolation valve failed.

    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: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Thomas G. Hiltz.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, MI

    Date of amendment request: May 22, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 5.5.7, ``Inservice Testing 
Program'' to: (1) Delete reference to American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code), Section 
XI and incorporate reference to the ASME Code for Operation and 
Maintenance of Nuclear Power Plants (ASME OM Code), and (2) address the 
applicability of Surveillance Requirement (SR) 3.0.2 to other normal 
and accelerated frequencies specified as two years or less in the 
inservice testing (IST) program.
    The proposed amendment incorporates changes based on U.S. Nuclear 
Regulatory Commission (NRC)--approved Technical Specification Task 
Force (TSTF) TSTF-479-A, ``Changes to Reflect Revision of 10 CFR 
50.55a,'' Revision 0, as modified by NRC-approved TSTF-497, ``Limit 
Inservice Testing Program SR 3.0.2 Application to Frequencies of Two 
Years or Less,'' Revision 0. The proposed changes include two 
deviations from the NRC-approved TSTFs that are administrative in 
nature: (1) Addition of ``ASME'' to TS 5.5.7 to make references to 
``ASME OM Code'' and (2) use of the term ``intervals'' instead of 
``frequencies.'' Basis for proposed no significant hazards 
consideration determination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed changes do not have any impact on the 
integrity of any plant system, structure, or component that 
initiates an analyzed event. The proposed changes would not alter 
the operation of, or otherwise increase the failure probability of 
any plant equipment that initiates an analyzed accident. Thus, the 
probability of any accident previously evaluated is not 
significantly increased.
    The proposed changes do not affect the ability to mitigate 
previously evaluated accidents, and do not affect radiological 
assumptions used in the evaluations. The proposed changes do not 
change or alter the design criteria for the systems or components 
used to mitigate the consequences of any design basis accident. The 
proposed amendment does not involve operation of the required 
structures, systems, or components (SSCs) in a manner or 
configuration different from those previously recognized or 
evaluated. Thus, the radiological consequences of any accident 
previously evaluated are not increased.
    Therefore, operation of the facility in accordance with the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

[[Page 49576]]

    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed amendment does not involve a physical 
alteration of any SSC or a change in the way any SSC is operated. 
The proposed amendment does not involve operation of any required 
SSCs in a manner or configuration different from those previously 
recognized or evaluated. No new failure mechanisms would be 
introduced by the changes being requested.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The amendment does not involve a significant reduction in a 
margin of safety. The proposed amendment does not affect the 
acceptance criteria for any safety analysis analyzed accidents or 
anticipated operational occurrences. The proposed amendment does not 
alter the limiting values and acceptance criteria used to judge the 
continued acceptability of components tested by the IST Program. The 
safety function of the affected pumps and valves will be maintained.
    Therefore, the proposed amendment would not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: July 31, 2007.
    Description of amendment request: The proposed amendment will 
revise Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification 
(TS) 6.6.5, Core Operating Limits. The proposed change will add new 
analytical methods to support the implementation of Next Generation 
Fuel (NGF).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the COLR [Core Operating Limits Report] 
TS are administrative in nature and have no impact on any plant 
configuration or system performance relied upon to mitigate the 
consequences of an accident. Changes to the calculated core 
operating limits may only be made using NRC-approved methodologies, 
must be consistent with all applicable safety analysis limits, and 
are controlled by the 10 CFR 50.59 process.
    The proposed change will add the following topical reports to 
the list of referenced core operating analytical methods.

WCAP-16500-P and Final Safety Evaluation (SE)

    Westinghouse topical report WCAP-16500-P describes the methods 
and models that will be used to evaluate the acceptability of CE 
[Combustion Engineering] 16 x 16 NGF at CE plants. Entergy has 
demonstrated that the Limitations and Conditions associated with the 
NRC SE will be met. Prior to implementation of NGF, the new core 
design will be analyzed with applicable NRC staff-approved codes and 
methods.

WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A

    The proposed change allows the use of methods required for the 
implementation of Optimized ZIRLO\TM\ clad fuel rods. Entergy has 
demonstrated that the Limitations and Conditions associated with the 
NRC SE will be met.

WCAP-16523-P and Final Safety Evaluation

    This topical report describes the departure from nucleate 
boiling [DNB] correlations that will be used to account for the 
impact of the CE 16 x 16 NGF fuel assembly design. Entergy has 
demonstrated that the Limitations and Conditions associated with the 
NRC SE will be met. Prior to implementation of NGF, the new core 
design will be analyzed with applicable NRC staff-approved codes and 
methods.

CENPD-387-P-A

    The proposed addition of this topical report provides the [DNB] 
correlation that will be used to evaluate the DNB impact of non-
mixing vane grid spans for CE 16 x 16 standard and NGF assemblies. 
Entergy has demonstrated that the Limitations and Conditions 
associated with the NRC SE will be met.

CENPD-132, Supplement 4-P-A, Addendum 1-P and Final Safety Evaluation

    The addendum provides an optional steam cooling model that can 
be used for Emergency Core Cooling System (ECCS) Performance 
analyses to support the implementation of the CE 16 x 16 NGF fuel 
assembly design. The optional steam cooling model is not being used 
to support implementation of CE 16 x 16 NGF assemblies in ANO-2 at 
this time. However, Entergy has demonstrated that the Limitations 
and Conditions associated with the NRC SE will be met.
    Assumptions used for accident initiators and/or safety analysis 
acceptance criteria are not altered by the addition of these topical 
reports.
    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 identifies changes in the codes used to 
confirm the values of selected cycle-specific reactor physics 
parameter limits. The proposed change allows the use of methods 
required for the implementation of CE 16 x 16 NGF. The proposed 
addition of the referenced topical reports has no impact on any 
plant configurations or on system performance that is relied upon to 
mitigate the consequences of an accident. These changes are 
administrative in nature and do not result in a change to the 
physical plant or to the modes of operation defined in the facility 
license.

WCAP-16500-P and Final Safety Evaluation

    The proposed change adds Westinghouse topical report WCAP-16500-
P, which describes the methods and models that will be used to 
evaluate the acceptability of CE 16 x 16 NGF at CE plants. Entergy 
has demonstrated that the Limitations and Conditions associated with 
the NRC SE will be met. Prior to implementation of NGF, the new core 
design will be analyzed with applicable NRC staff-approved codes and 
methods.

WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A

    The proposed change allows the use of methods required for the 
implementation of Optimized ZIRLO\TM\ clad fuel rods. Entergy has 
demonstrated that the Limitations and Conditions associated with the 
NRC SE will be met.

WCAP-16523-P and Final Safety Evaluation

    This topical report describes the [DNB] correlations that will 
be used to account for the impact of the CE 16 x 16 NGF fuel 
assembly design. Entergy has demonstrated that the Limitations and 
Conditions associated with the SE will be met.

CENPD-387-P-A

    The proposed addition of this topical report provides the [DNB] 
correlation that will be used to evaluate the DNB impact of non-
mixing vane grid spans for CE 16 x 16 standard and NGF assemblies. 
Entergy has demonstrated that the Limitations and Conditions 
associated with the NRC SE will be met.

CENPD-132, Supplement 4-P-A, Addendum 1-P and Final Safety Evaluation

    The addendum provides an optional steam cooling model that can 
be used for ECCS Performance analyses to support the implementation 
of the CE 16 x 16 NGF fuel assembly design. The optional steam 
cooling model is not being used to support implementation of CE 16 x 
16 NGF

[[Page 49577]]

assemblies in ANO-2 at this time. However, Entergy has demonstrated 
that the Limitations and Conditions associated with the NRC SE will 
be met.
    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 proposed changes do not amend the cycle-specific parameter 
limits located in the COLR from the values presently required by the 
TS. The individual specifications continue to require operation of 
the plant within the bounds of the limits specified in COLR.
    The addition of the following topical reports to the list of 
analytical methods referenced in the COLR is administrative in 
nature:
    [squ] WCAP-16500-P and Final Safety Evaluation for Westinghouse 
Electric Company (Westinghouse) Topical Report (TR) WCAP-16500-P, 
Revision 0, ``CE [Combustion Engineering] 16 x 16 Next Generation 
Fuel [(NGF)] Core Reference Report''
    [squ] WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A
    [squ] WCAP-16523-P and Final Safety Evaluation for Westinghouse 
Electric Company (Westinghouse) Topical Report (TR), WCAP-16523-P, 
``Westinghouse Correlations WSSV and WSSV-T for Predicting Critical 
Heat Flux in Rod Bundles with Side-Supported Mixing Vanes''
    [squ] CENPD-387-P-A
    [squ] CENPD-132, Supplement 4-P-A, Addendum 1-P and Final Safety 
Evaluation for Westinghouse Electric Company (Westinghouse) Topical 
Report (TR) CENPD-132 Supplement 4-P-A, Addendum 1-P, ``Calculative 
Methods for the CE [Combustion Engineering] Nuclear Power Large 
Break LOCA Evaluation Model--Improvement to 1999 Large Break LOCA EM 
Steam Cooling Model for Less Than 1 in/sec Core Reflood''
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of application for amendments: November 17, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Surveillance Requirement 3.3.1.1.8 to 
increase the frequency interval between Local Power Range Monitor 
(LPRM) calibrations from 1000 megawatt days per ton (MWD/T) average 
core exposure to 2000 MWD/T average core exposure. The LPRM system 
provides signals to associated nuclear instrumentation systems that 
serve to detect conditions in the core that have the potential to 
threaten the overall integrity of the fuel barrier. The LPRM system 
also incorporates features designed to diagnose and display various 
system trip and inoperative conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No
    The proposed amendment revises the surveillance interval for the 
LPRM calibration from 1000 MWD/T average core exposure to 2000 MWD/T 
average core exposure. Increasing the frequency interval between 
required LPRM calibrations is acceptable due to improvements in core 
monitoring processes and nuclear instrumentation and therefore, the 
revised surveillance interval continues to ensure that the LPRM 
detector signal is adequately calibrated.
    This change will not alter the operation of process variables, 
structures, systems, or components as described in the PBAPS Updated 
Final Safety Analysis Report (UFSAR). The proposed change does not 
alter the initiation conditions or operational parameters for the LPRM 
system and there is no new equipment introduced by the extension of the 
LPRM calibration interval. The performance of the APRM, OPRM and RBM 
systems is not significantly affected by the proposed surveillance 
interval increase. As such, the probability of occurrence of a 
previously evaluated accident is not increased.
    The radiological consequences of an accident can be affected by the 
thermal limits existing at the time of the postulated accident; 
however, LPRM chamber exposure has no significant effect on the 
calculated thermal limits since LPRM accuracy does not significantly 
deviate with exposure. For the LPRM extended calibration interval, the 
total nodal power uncertainty remains less than the uncertainty assumed 
in the thermal analysis basis safety limit, maintaining the accuracy of 
the thermal limit calculation. Therefore, the thermal limit calculation 
is not significantly affected by LPRM calibration frequency, and thus 
the radiological consequences of any accident previously evaluated are 
not increased.
    Therefore, based on the above information, the proposed change does 
not involve a significant increase in the probability or consequences 
of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No
    The performance of the APRM, OPRM and RBM systems is not 
significantly affected by the proposed LPRM surveillance interval 
increase. The proposed change does not affect the control parameters 
governing unit operation or the response of plant equipment to 
transient conditions. The proposed change does not change or introduce 
any new equipment, modes of system operation or failure mechanisms.
    Therefore, based on the above information, the proposed change does 
not create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No
    The proposed change has no impact on equipment design or 
fundamental operation, and there are no changes being made to safety 
limits or safety system allowable values that would adversely affect 
plant safety as a result of the proposed LPRM surveillance interval 
increase. The performance of the APRM, OPRM and RBM systems is not 
significantly affected by the proposed change. The margin of safety can 
be affected by the thermal limits existing at the time of the 
postulated accident; however, uncertainties associated with LPRM 
chamber exposure have no significant effect on the calculated thermal 
limits. The thermal limit calculation is not significantly affected 
since LPRM sensitivity with exposure is well defined. LPRM accuracy 
remains within the total nodal power uncertainty assumed in the thermal 
analysis basis; thereby maintaining thermal limits and the safety 
margin. The proposed change does not affect safety analysis assumptions 
or initial conditions and

[[Page 49578]]

therefore, the margin of safety in the original safety analyses are 
maintained.
    Therefore, based on the above information, 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: Mr. J. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

Florida Power and Light Company (FPL), Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Units 1 and 2, St. Lucie County, FL

    Date of amendment request: July 16, 2007.
    Description of amendment request: The proposed amendment would 
modify the technical specification (TS) requirements related to control 
room envelope (CRE) habitability in accordance with Technical 
Specification Task Force (TSTF) Traveler TSTF-448, Revision 3, 
``Control Room Habitability,'' published in the Federal Register on 
January 17, 2007 (Volume 72, Number 10), as part of the consolidated 
line item improvement process. Specifically by modifying Unit 1 TS 
3.7.7.1, ``Control Room Emergency Ventilation System (CREVS),'' and 
Unit 2 TS 3.7.7,'' Control Room Emergency Air Cleanup System 
(CREACS),'' and adding a new Unit 1 and Unit 2 TS Section 6.8.4.m.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

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

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Thomas H. Boyce.

Florida Power and Light Company (FPL), Docket Nos. 50-335, St. Lucie 
Plant, Unit 1, St. Lucie County, FL

    Date of amendment request: July 16, 2007.
    Description of amendment request: The proposed amendment would 
modify the facilities operating licensing bases to adopt the 
alternative source term (AST) as allowed in 10 CFR 50.67 and described 
in Regulatory Guide (RG) 1.183. The licensee proposes to revise the 
plant licensing basis through reanalysis of the following radiological 
consequences of the Updated Final Safety Analysis Report (UFSAR) 
Chapter 15 accidents: Loss-of-Coolant Accident, Fuel Handling Accident, 
Main Steam Line Break, Steam Generator Tube Rupture, Reactor Coolant 
Pump Shaft Seizure, Control Element Assembly Ejection, and Inadvertent 
Opening of a Main Steam Safety Valve.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Alternative source term calculations have been 
performed for St. Lucie Unit No. 1 which demonstrate that the dose 
consequences remain below limits specified in NRC Regulatory Guide 
1.183 and 10 CFR 50.67. The proposed changes do not modify the 
design or operation of the plant. The use of the AST only changes 
the regulatory assumptions regarding the analytical treatment of the 
design basis accidents and has no direct effect on the probability 
of any accident. The AST has been utilized in the analysis of the 
limiting design basis accidents listed above. The results of the 
analyses, which include the proposed changes to the Technical 
Specifications [TSs], demonstrate that the dose consequences of 
these limiting events are all within the regulatory limits.
    With the exception of the deletion of SRs 4.6.6.1.c.[3].b and 
4.7.8.1.c.[3].b, the proposed Technical Specification changes are 
consistent with, or more restrictive than, the current TS 
requirements. The proposed filter testing requirements continue to 
ensure

[[Page 49579]]

that the associated filtration systems function as described in the 
UFSAR and as assumed in the accident analyses. None of the affected 
systems, components or programs are related to accident initiators.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not affect any plant structures, 
systems, or components. The operation of plant systems and equipment 
will not be affected by this proposed change. Neither implementation 
of the alternative source term methodology, establishing more 
restrictive TS requirements, nor deleting SRs 4.6.6.1.c.[3].b and 
4.7.8.1.c.[3].b have the capability to introduce any new failure 
mechanisms or cause any analyzed accident to progress in a different 
manner.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed implementation of the alternative source term 
methodology is consistent with NRC Regulatory Guide 1.183. With the 
exception of the deletion of SRs 4.6.6.1.c.[3].b and 
4.7.8.1.c.[3].b, the proposed Technical Specification changes are 
consistent with, or more restrictive than, the current TS 
requirements. The proposed TS requirements support the AST revisions 
to the limiting design basis accidents. The proposed filter testing 
requirements continue to ensure that the associated filtration 
systems function as described in the UFSAR and as assumed in the 
accident analyses. As such, the current plant margin of safety is 
preserved. Conservative methodologies, per the guidance of RG 1.183, 
have been used in performing the accident analyses. The radiological 
consequences of these accidents are all within the regulatory 
acceptance criteria associated with use of the alternative source 
term methodology.
    The proposed changes continue to ensure that the doses at the 
exclusion area and low population zone boundaries and in the Control 
Room are within the corresponding regulatory limits of RG 1.183 and 
10 CFR 50.67. The margin of safety for the radiological consequences 
of these accidents is considered to be that provided by meeting the 
applicable regulatory limits, which are set at or below the 10 CFR 
50.67 limits. An acceptable margin of safety is inherent in these 
limits.

    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: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Thomas H. Boyce.

Florida Power and Light Company (FPL), Docket No. 50-389, St. Lucie 
Plant, Unit 2, St. Lucie County, FL

    Date of amendment request: July 16, 2007.
    Description of amendment request: The proposed amendment would 
modify the facilities operating licensing bases to adopt the 
alternative source term (AST) as allowed in 10 CFR 50.67 and described 
in Regulatory Guide (RG) 1.183. The licensee proposes to revise the 
plant licensing basis through reanalysis of the following radiological 
consequences of the Updated Final Safety Analysis Report Chapter 15 
accidents: Loss-of-Coolant Accident, Fuel Handling Accident, Main Steam 
Line Break, Steam Generator Tube Rupture, Reactor Coolant Pump Shaft 
Seizure, Control Element Assembly Ejection, Letdown Line Break, and 
Feedwater Line Break.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Alternative source term calculations have been performed for St. 
Lucie Unit No. 2 which demonstrate that the dose consequences remain 
below limits specified in NRC Regulatory Guide 1.183 and 10 CFR 
50.67. The proposed changes do not modify the design or operation of 
the plant. The use of the AST only changes the regulatory 
assumptions regarding the analytical treatment of the design basis 
accidents and has no direct effect on the probability of any 
accident. The AST has been utilized in the analysis of the limiting 
design basis accidents listed above. The results of the analyses, 
which include the proposed changes to the Technical Specifications 
[TSs], demonstrate that the dose consequences of these limiting 
events are all within the regulatory limits.
    The proposed Technical Specification Changes are consistent 
with, or more restrictive than, the current TS requirements. None of 
the affected systems, components or programs are related to accident 
initiators.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not affect any plant structures, 
systems, or components. The operation of plant systems and equipment 
will not be affected by this proposed change. Neither implementation 
of the alternative source term methodology nor establishing more 
restrictive TS requirements have the capability to introduce any new 
failure mechanisms or cause any analyzed accident to progress in a 
different manner.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed implementation of the alternative source term 
methodology is consistent with NRC Regulatory Guide 1.183. The 
proposed Technical Specification changes are consistent with, or 
more restrictive than, the current TS requirements. These TS 
requirements support the AST revisions to the limiting design basis 
accidents. As such, the current plant margin of safety is preserved. 
Conservative methodologies, per the guidance of RG 1.183, have been 
used in performing the accident analyses. The radiological 
consequences of these accidents are all within the regulatory 
acceptance criteria associated with use of the alternative source 
term methodology.
    The proposed changes continue to ensure that the doses at the 
exclusion area and low population zone boundaries and in the Control 
Room are within the corresponding regulatory limits of RG 1.183 and 
10 CFR 50.67. The margin of safety for the radiological consequences 
of these accidents is considered to be that provided by meeting the 
applicable regulatory limits, which are set at or below the 10 CFR 
50.67 limits. An acceptable margin of safety is inherent in these 
limits.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    Based on the above discussion, FP&L has determined that the 
proposed change does not involve a significant hazards 
consideration.

    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: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Thomas H. Boyce.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, MN

    Date of amendment request: July 3, 2007.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications

[[Page 49580]]

(TSs) for the Prairie Island Nuclear Generating Plant (PINGP), Units 1 
and 2 to:
    1. Revise TS 1.4, ``Frequency'' to modify the second paragraph of 
Example 1.4-1 to be consistent with the requirements of Surveillance 
Requirement (SR) 3.0.4 and incorporate the changes in Technical 
Specification Task Force (TSTF) industry traveler TSTF-485, ``Correct 
Example 1.4-1.''
    2. Revise TS 5.5.7.a, to modify references to Section XI of the 
American Society of Mechanical Engineers (ASME) Code with references to 
the ASME Code for Operation and Maintenance of Nuclear Power Plants 
(ASME OM Code), to be consistent with TSTF-479, ``Changes to Reflect 
Revision of 10 CFR [Code of Federal Regulations] 50.55a.
    3. Revise TS 5.5.7.b, to restrict extension of Frequencies to those 
Frequencies specified as 2 years or less, and take exception to the 
limitation in SR 3.0.2 which does not apply the 1.25 times extension to 
Frequencies of 24 months, to be consistent with TSTF-479 and TSTF-497, 
``Limit Inservice Testing Program SR 3.0.2 Application to Frequencies 
of 2 Years or Less.''
    4. Revise TS 5.5.7.d, to modify the referenced ASME Code to be 
consistent with TSTF-479.
    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:

TSTF-479

    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 Improved Standard Technical 
Specification (ISTS) Inservice Testing Program for consistency with 
the requirements of 10 CFR 50.55a(f)(4) for pumps and valves which 
are classified as American Society of Mechanical Engineers (ASME) 
Code Class 1, Class 2 and Class 3. The proposed change incorporates 
revisions to the ASME Code that result in a net improvement in the 
measures for testing pumps and valves.
    The proposed change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. They do not involve the addition or removal of any 
equipment, or any design changes to the facility. Therefore, this 
proposed change does not represent a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the improved Standard Technical 
Specification (ISTS) Inservice Testing Program for consistency with 
the requirements of 10 CFR 50.55a(f)(4) for pumps and valves which 
are classified as American Society of Mechanical Engineers (ASME) 
Code Class 1, Class 2 and Class 3. The proposed change incorporates 
revisions to the ASME Code that result in a net improvement in the 
measures for testing pumps and valves.
    The proposed change does not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. Additionally, there is no 
change in the types or increases in the amounts of any effluent that 
may be released off-site and there is no increase in individual or 
cumulative occupational exposure. Therefore, this proposed change 
does not create the possibility of an accident of a different kind 
than previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the improved Standard Technical 
Specification (ISTS) Inservice Testing Program for consistency with 
the requirements of 10 CFR 50.55a(f)(4) for pumps and valves which 
are classified as American Society of Mechanical Engineers (ASME) 
Code Class 1, Class 2 and Class 3. The proposed change incorporates 
revisions to the ASME Code that result in a net improvement in the 
measures for testing pumps and valves. The safety function of the 
affected pumps and valves will be maintained. Therefore, this 
proposed change does not involve a significant reduction in a margin 
of safety.

TSTF-485

    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 Section 1.4, Frequency, Example 1.4-
1, to be consistent with Surveillance Requirement (SR) 3.0.4 and 
Limiting Condition for Operation (LCO) 3.0.4. This change is 
considered administrative in that it modifies the example to 
demonstrate the proper application of SR 3.0.4 and LCO 3.0.4. The 
requirements of SR 3.0.4 and LCO 3.0.4 are clear and are clearly 
explained in the associated Bases. As a result, modifying the 
example will not result in a change in usage of the Technical 
Specifications (TS). The proposed change does not adversely affect 
accident initiators or precursors, the ability of structures, 
systems, and components (SSCs) to perform their intended function to 
mitigate the consequences of an initiating event within the assumed 
acceptance limits, or radiological release assumptions used in 
evaluating the radiological consequences of an accident previously 
evaluated. Therefore, this change is considered administrative and 
will have no effect on the probability or consequences of any 
accident previously evaluated. 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.
    No new or different accidents result from utilizing the proposed 
change. The change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the change does not impose any new or 
different requirements or eliminate any existing requirements. The 
change does not alter assumptions made in the safety analysis. The 
proposed change is consistent with the safety analysis assumptions 
and current plant operating practice.
    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 change is administrative and will have no effect on 
the application of the Technical Specification requirements. 
Therefore, the margin of safety provided by the Technical 
Specification requirements is unchanged. Therefore, the proposed 
change does not involve a significant reduction in a margin of 
safety.

TSTF-497

    This Traveler is considered an administrative change to the ISTS 
NUREGs. Therefore, a regulatory analysis is not provided.

    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 requests involve no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Acting Branch Chief: Travis L. Tate.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Rivers, Manitowoc 
County, WI

    Date of amendment request: July 12, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.6.3, ``Containment Isolation 
Valves.'' The revision would delete Surveillance

[[Page 49581]]

Requirement (SR) 3.6.3.1, which is no longer required due to the 
containment purge supply and exhaust valve isolation function being 
replaced with blind flanges. The proposed amendment would also support 
a change to the Final Safety Analysis Report (FSAR) to revise the 
requirement to leak check the purge supply and exhaust valves.
    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 to the containment purge supply penetration 
and the containment exhaust penetration presents no change in the 
probability or the consequence of an accident. The penetrations 
continue to conform to the TS requirements for containment and will 
be appropriately tested as required by 10 CFR 50 Appendix J. The 
blind flanges are passive devices not susceptible to an active 
failure or malfunction that could result in a loss of isolation or 
leakage that exceeds the limits assumed in the safety analyses. The 
blind flanges are leak rate tested in accordance with the 
containment leakage rate testing program. Containment isolation is 
not lessened by this change.
    The change to the containment purge system does not affect the 
design basis limit for any fission product barrier.
    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 containment purge supply penetration 
and the containment exhaust penetration does not change the function 
of the system and does not alter containment isolation. The 
penetrations continue to conform to the TS requirements for 
containment isolation and will be appropriately tested as required 
by 10 CFR 50 Appendix J. No new accident scenarios, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed changes.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed change will not alter any assumptions, initial 
conditions or results specified in any accident analysis. The 
containment purge supply and exhaust penetrations will continue to 
conform to the TS requirements for containment and will be 
appropriately tested as required by 10 CFR 50 Appendix J. The blind 
flanges are passive devices not susceptible to an active failure or 
malfunction that could result in a loss of isolation or leakage that 
exceeds limits assumed in the safety analysis. The blind flanges are 
leak rate tested in accordance with the containment leakage rate 
testing program. Containment isolation is not lessened by this 
change. Therefore, there is no reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Acting Branch Chief: Travis L. Tate.

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

    Date of amendment request: July 30, 2007.
    Description of amendment request: The proposed amendment by Omaha 
Public Power District requests changes to the Fort Calhoun Station Unit 
No.1 Operating License No. DPR-40 to modify the containment spray 
system actuation logic to preclude automatic start of the containment 
spray pumps for a loss-of-coolant accident.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The containment spray (CS) system and the containment air 
cooling and filtering system (CACFS) are not initiators of any 
accident previously evaluated at the Fort Calhoun Station (FCS). 
Both systems are accident mitigation systems. Their licensing basis 
functions are to limit the containment pressure rise and reduce the 
leakage of airborne radioactivity from the containment by providing 
a means for cooling the containment following a loss-of-coolant 
accident (LOCA) or main steam line break (MSLB) inside containment. 
The proposed modification to the CS system logic shifts the function 
of containment pressure and temperature control during a LOCA from 
the [CS] system to the equally capable and reliable containment air 
coolers. The change in the CS actuation logic does not impact the 
containment response to the MSLB analysis of record (AOR). The CACFS 
provides the design heat removal capabilities for the containment 
during the postulated LOCA. The system is operated to remove 
atmospheric heat loads from the containment during normal plant 
operation. Since system components are only lightly loaded during 
normal operation, system availability and reliability are enhanced. 
In the unlikely event that normal power sources are lost and one 
emergency diesel generator fails to operate, one containment air 
cooling and filtering unit and one containment air cooling unit will 
operate.
    The component cooling water (CCW) system, on which the CACFS is 
dependent, has sufficient capacity for all normal and shutdown 
operating modes. In addition, the system is capable of satisfying 
the design criteria under post design-basis accident (DBA) 
conditions with the single failure of an active component and a loss 
of instrument air. Analyses demonstrate that CCW flowrates to 
essential equipment would be adequate for removing post accident 
design-basis heat loads.
    Following implementation of the proposed change, at least one 
train of containment air coolers will be available to mitigate a 
LOCA. Analyses show that one train of coolers can maintain the 
containment pressure and temperature below the design values; 
therefore, the proposed change will have no adverse effect on the 
containment pressure analysis following a LOCA.
    Analyses have also shown that one train of containment high-
efficiency particulate air (HEPA) filters maintains the radiological 
consequences doses within regulatory limits; therefore, the proposed 
change will have no adverse effect on the radiological consequences 
following a LOCA.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The CACFS was designed to remove heat released to containment 
atmosphere during the [DBA] to the extent necessary to maintain the 
structure below the design pressure. The proposed modification to 
the CS system logic shifts the function of containment pressure and 
temperature control from the [CS] system to the equally capable and 
reliable containment air coolers. The use of CACFS, as a means of 
containment pressure control, has been evaluated for the LOCA event 
and found to result in an acceptable peak containment pressure (peak 
pressure less than 60 psig [pounds per square inch gauge]). 
Radiological consequences were evaluated for the use of CACFS in 
this application using the guidance provided in Regulatory Guide 
(RG) 1.183. This radiological analysis demonstrates that the dose 
consequences are in compliance with applicable regulatory 
requirements. The estimated dose consequences at the exclusion area 
boundary (EAB), low population zone (LPZ), and control room (CR) 
remain within the acceptance criteria of 10 CFR 50.67 as 
supplemented by RG 1.183 and the standard review plan (SRP) 15.0.1. 
The assessment also demonstrates that the dose consequences in the 
technical support center (TSC) remain compliant with regulatory 
guidance provided in Supplement 1 of NUREG-0737.

[[Page 49582]]

    No credible new failure mechanisms, malfunctions, or accident 
initiators not considered in the design and licensing basis have 
been created and none of the initial condition assumptions of any 
accident evaluated in the safety analysis are impacted.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The containment building and associated penetrations are 
designed to withstand an internal pressure of 60 psig at 305 [deg]F 
[degrees Fahrenheit], including all thermal loads resulting from the 
temperature associated with this pressure, with a leakage rate of 
0.1 percent by weight or less of the contained volume per 24 hours. 
The containment air coolers are credited for maintaining containment 
pressure and temperatures within design limitations, and assure that 
the release of fission products to the environment following a [DBA] 
will not exceed regulatory guidelines for a large break (LB) LOCA.
    The [CS] system and containment air coolers continue to be 
credited for limiting peak containment pressure for an MSLB.
    Adequate NPSH [net positive suction head] margin is maintained 
for the HPSI [high-pressure safety injection] pumps during the 
recirculation phase of a[n] LBLOCA due to the reduction in ECCS 
[emergency core cooling system] sump strainer pressure drop.
    The CACFS operates independently of the CS system to remove heat 
from the containment atmosphere. The CACFS consists of two redundant 
trains, each train with one air cooling and filtering unit and one 
air cooling unit, for a total of four cooling units. Operation of 
the CACFS, in accordance with analyses completed for the 2006 steam 
generator replacement, is and will continue to be credited in the 
MSLB containment pressure analysis. The operation and maintenance of 
the CACFS are not impacted by this proposed change. Therefore, the 
containment heat removal licensing basis is not adversely affected 
by the proposed change. The ability to maintain containment peak 
pressure and temperature, as well as long-term containment pressure 
and temperature, is maintained.
    The LBLOCA 10 CFR 50.46 analysis assumes that there will be 
three CS pumps operating when evaluating the effects of containment 
pressure on ECCS performance. This assumption minimizes containment 
pressure, to conservatively evaluate ECCS performance in response to 
a LOCA. Eliminating operation of the CS pumps improves ECCS 
performance and thus increases margin to 10 CFR 50.46 limits on peak 
clad temperature, therefore, the existing analysis remains bounding 
as is.
    In summary, following implementation of the proposed change:
    [squ] Peak containment pressure for analyzed DBAs remains within 
design limits;
    [squ] Radiological releases remain within the limits of 10 CFR 
50.67; and
    [squ] The currently calculated peak clad temperature following a 
LOCA remains bounding.
    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: James R. Curtiss, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Thomas G. Hiltz.

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

    Date of amendment request: July 31, 2007.
    Description of amendment request: The proposed amendment will 
modify Technical Specification (TS) requirements to support a planned 
inverter modification to be installed during the 2008 refueling outage. 
The inverter modification will require revisions to TS 2.7(1), 2.7(2), 
and 3.7(5), and the associated Bases sections to allow for the addition 
of two safety-related swing inverters.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The addition of two safety-related swing inverters to the 120 V a-c 
[Volts alternating current] vital instrument buses is not an initiator 
of any previously evaluated accidents. The swing inverters will not 
prevent safety systems from performing any of the accident mitigation 
functions assumed in the safety analysis. The revisions proposed for 
the Technical Specifications (TS) take advantage of the operational 
flexibility provided by the swing inverters yet maintain current TS 
requirements that four inverters be operable.
    Similarly, the change maintains the current TS allowance for one of 
the required inverters to be inoperable for up to twenty-four hours 
provided all current TS requirements for operability are met.
    Although continued operation for up to twenty-four hours with one 
of the required inverters inoperable is allowed, the addition of the 
two safety-related swing inverters is expected to decrease the amount 
of time that the station must operate with less than four inverters. 
This is because the design allows the inoperable inverter to be 
replaced by its associated swing (or non-swing) inverter. Reducing the 
need to shut the station down due to an inoperable inverter also 
reduces the risk associated with mode transition to shutdown.
    The correction of two typographical errors and correcting spacing 
inconsistencies in the text are administrative changes that do not 
involve a significant increase in the probability or consequences of 
any accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The design function of the safety-related inverters is unchanged. 
The addition of the safety-related swing inverters and their bypass 
sources to the 120 volt a-c vital instrument distribution system allows 
preventative maintenance, repair and for testing to be performed 
online. If a safety-related inverter becomes inoperable or is otherwise 
out-of-service, its instrument bus is manually transferred to the 
associated swing inverter. If a required inverter should fail, the time 
that the station will operate with less than the four inverters 
required by TS 2.7(1)j should, in most cases, be less due to the 
ability to place an associated inverter online. Reducing the need to 
shut the station down due to an inoperable inverter also reduces the 
risk associated with mode transition to shutdown.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction in 
the margin of safety?
    Response: No
    The design function of the safety-related inverters is unchanged. 
The addition of the safety-related swing inverters to the 120 volt a-c 
vital instrument distribution system allows preventative maintenance or 
repair of a safety-related inverter to be performed online since its 
instrument bus can be manually transferred to the associated swing 
inverter. Installation of the safety-related swing inverters does not 
require changes to accident analyses or results. The revisions proposed 
for the TS

[[Page 49583]]

maintain current TS requirements that four inverters be operable. 
Should a required inverter fail, the time that the station will operate 
with less than the four inverters required by TS 2.7(1)j should, in 
most cases, be less due to the ability to place an associated inverter 
online. Reducing the need to shut the station down due to an inoperable 
inverter also reduces the risk associated with mode transition to 
shutdown. In addition, administrative controls are in place to ensure 
the current station battery capacity is not degraded and to ensure 
battery margin is adequately maintained as a result of the inverter 
modification.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: July 26, 2007.
    Description of amendment request: The proposed amendment would add 
a new reference to Technical Specification 6.9.1.14.a, which lists 
documents that have been approved by the U.S. Nuclear Regulatory 
Commission for use in determining the core operating limits. The new 
reference is the Areva NP, Inc. topical report EMF-2103P-A, ``Realistic 
Large Break LOCA Methodology for Pressurized Water Reactors.''
    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 adds an approved analytical method for 
evaluating large break loss of coolant accidents (LOCAs). The 
proposed change will not affect previously evaluated accidents 
because they continue to be analyzed by NRC approved methodologies 
to ensure required safety limits are maintained. The acceptance 
criteria of the SQN Final Safety Analysis Report analyzed accidents 
and anticipated operational occurrences are not affected by the 
proposed addition of the realistic large break LOCA methodology. As 
the evaluations for accidents and operation occurrences are not 
adversely affected, the proposed change will not increase the 
consequences of a postulated event. The proposed change does not 
result in any modification of the plant equipment or operating 
practices and therefore, does not alter plant conditions or plant 
response prior to or after postulated events. 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.
    As previously noted, the proposed change does not result in any 
modification of the plant equipment or operating practices and 
therefore, does not alter plant conditions or plant response prior 
to or after postulated events. 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 proposed change does not alter plant equipment including the 
automatic accident mitigation setpoints designed to mitigate the 
affects of a postulated accident. The accident analyses and plant 
safety limits continue to be acceptable as evaluated by NRC approved 
methodologies. The proposed application of the realistic large break 
LOCA methodology ensures acceptable margins and limits for fuel core 
designs. 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Thomas H. Boyce.

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, NJ

    Date of application for amendment: November 27, 2006.
    Brief description of amendment: The amendment revised the required 
submittal date for the Annual Radioactive Effluent Release Report. 
Specifically, the required submittal date is revised from ``within 60 
days after January 1, each year,'' to ``prior to May 1 of each year.''
    Date of Issuance: August 8, 2007.

[[Page 49584]]

    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No: 264.
    Facility Operating License No. DPR-16: The amendment revised the 
license and the Technical Specifications.
    Date of initial notice in Federal Register: May 8, 2007 (72 FR 
26174).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated August 8, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 12, 2006, as supplemented 
by letters dated April 25, May 23, June 15, June 20, and June 29, 2007.
    Brief description of amendment: The amendment modifies Conditions, 
Required Actions and Completion Times in Technical Specification (Ts) 
3.8.1, ``AC Sources-Operating,'' associated with the Required Actions 
when emergency diesel generators are declared inoperable.
    Date of issuance: August 1, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No: 175.
    Facility Operating License No. NPF-43: Amendment revised the TSs 
and License.
    Date of initial notice in Federal Register: August 29, 2006 (71 FR 
51225). The April 25, May 23, June 15, June 20, and June 29, 2007, 
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 amendment is contained 
in a Safety Evaluation dated August 1, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 26, 2007.
    Brief description of amendment: The amendment adds a Limiting 
Condition for Operation (LCO) 3.0.9 to the Technical Specifications 
(TS), allowing a delay time for entering a supported system TS, when 
the inoperability is due solely to an unavailable barrier, if risk is 
assessed and managed. Additionally, the amendment makes editorial 
changes to LCO 3.0.8 to be consistent with terminology of LCO 3.0.9.
    Date of issuance: August 1, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 176.
    Facility Operating License No. NPF-43: Amendment revised the TS and 
License.
    Date of initial notice in Federal Register: April 10, 2007 (72 FR 
17945).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 1, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 30, 2006, as supplemented by 
letters dated April 24, 2007, and June 27, 2007.
    Brief description of amendment: This amendment revises the existing 
SG tube surveillance program to be consistent with the Nuclear 
Regulatory Commission's approved TS Task Force (TSTF) Standard TS 
Change Traveler, TSTF-449, ``Steam Generator Tube Integrity.'' A notice 
of availability for this TS improvement using the consolidated line 
item improvement process was published in the Federal Register on May 
6, 2005 (70 FR 24126). The amendment is also the modification of the SG 
portion of the TSs requested in NRC Generic Letter (GL) 2006-01, 
``Steam Generator Tube Integrity and Associated Technical 
Specification.''
    Date of issuance: July 31, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 276.
    Facility Operating License No. NPF-3: The amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: October 10, 2006 (71 FR 
59531). The April 24, 2007, and June 27, 2007 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 amendment is contained 
in a Safety Evaluation dated July 31, 2007.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit No. 2, Oswego County, NY

    Date of application for amendment: March 8, 2007.
    Brief description of amendment: The amendment revises the Technical 
Specification requirements for inoperable snubbers by adding Limiting 
Condition for Operation 3.0.8 using the Consolidated Line Item 
Improvement Process.
    Date of issuance: July 30, 2007.
    Effective date: As of the date of issuance to be implemented within 
180 days.
    Amendment No.: 118.
    Renewed Facility Operating License No. NPF-69: Amendment revises 
the License and Technical Specifications.
    Date of initial notice in Federal Register: April 24, 2007 (72 FR 
20384).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 30, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 25, 2007, as supplemented 
by letters dated July 3 and 26, 2007 (TS-461).
    Brief description of amendment: The amendment deletes License 
Condition 2.G.(2) as the result of completion of power uprate large 
transient testing.
    Date of issuance: August 14, 2007.
    Effective date: The date of issuance, to be implemented within 30 
days.
    Amendment No.: 272.
    Renewed Facility Operating License No. DPR-33: Amendment revised 
the renewed operating license.
    Date of initial notice in Federal Register: July 13, 2007 (72 FR 
38627). The July 3 and 26, 2007, supplemental letters provided 
clarifying information that did not expand the scope of the 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 August 14, 2007.
    No significant hazards consideration comments received: No.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the

[[Page 49585]]

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, 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 email to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. Within 60 days after the date 
of publication of this notice, 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, and electronically on the Internet at the NRC Web site, 
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a 
request for a hearing or petition for leave to intervene is filed by 
the above date, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion. The 
petition must include sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or 
fact.\1\

[[Page 49586]]

Contentions shall be limited to matters within the scope of the 
amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner/requestor 
who fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
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    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
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    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by email 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 or 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).

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

    Date of amendment request: June 29, 2007, as supplemented by 
letters dated August 1, 2007 and August 2, 2007.
    Description of amendment request: The amendments revised the 
maximum allowed Technical Specification (TS) temperature limit, 
contained in TS Surveillance Requirement 3.7.3.1, of the cooling water 
supplied to the plant from the Core Standby Cooling System (CSCS) pond 
(i.e., the Ultimate Heat Sink) from 100 [deg]F to 101.25 [deg]F.
    Date of issuance: August 2, 2007.
    Effective date: August 2, 2007.
    Amendment Nos.: 183 and 170.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications and License.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, state consultation, and final NSHC 
determination are contained in a safety evaluation dated August 2, 
2007.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

    Dated at Rockville, Maryland, this 20th day of August 2007.

    For The Nuclear Regulatory Commission.
John W. Lubinski,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. E7-16766 Filed 8-27-07; 8:45 am]
BILLING CODE 7590-01-P