Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 62685-62696 [E7-21435]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 214 (Tuesday, November 6, 2007)]
[Notices]
[Pages 62685-62696]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-21435]


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NUCLEAR REGULATORY COMMISSION


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

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 October 11, 2007, to October 24, 2007. The 
last biweekly notice was published on October 23, 2007 (72 FR 60032).

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 6D44, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area O1F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, 
person(s) may file a request for a hearing with respect

[[Page 62686]]

to issuance of the amendment to the subject facility operating license 
and any person whose interest may be affected by this proceeding and 
who wishes to participate as a party in the proceeding must file a 
written request via electronic submission through the NRC E-Filing 
system for a hearing and a petition for leave to intervene. Requests 
for a hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested person(s) should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and 4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for hearing or a petition for leave to intervene must be 
filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated in August 28, 2007, (72 FR 49139). The E-Filing process 
requires participants to submit and serve documents over the internet 
or in some cases to mail copies on electronic storage media. 
Participants may not submit paper copies of their filings unless they 
seek a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/ requestor 
must contact the Office of the Secretary by e-mail at 
HEARINGDOCKET@NRC.GOV, or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/ 
requestor will need to download the Workplace Forms ViewerTM 
to access the Electronic Information Exchange (EIE), a component of the 
E-Filing system. The Workplace Forms ViewerTM is free and is 
available at https://www.nrc.gov/site-help/e-submittals/install-
viewer.html. Information about applying for a digital ID certificate is 
available on NRC's public Web site at https://www.nrc.gov/site-help/e-
submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at https://www.nrc.gov/
site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC

[[Page 62687]]

technical help line, which is available between 8:30 a.m. and 4:15 
p.m., Eastern Time, Monday through Friday. The help line number is 
(800) 397-4209 or locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, in accordance 
with 10 CFR 2.302(g), with their initial paper filing requesting 
authorization to continue to submit documents in paper format. Such 
filings must be submitted by: (1) First class mail addressed to the 
Office of the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff; or (2) courier, express mail, or expedited 
delivery service to the Office of the Secretary, Sixteenth Floor, One 
White Flint North, 11555 Rockville, Pike, Rockville, Maryland, 20852, 
Attention: Rulemaking and Adjudications Staff. Participants filing a 
document in this manner are responsible for serving the document on all 
other participants. Filing is considered complete by first-class mail 
as of the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the petition 
and/or request should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later 
than 11:59 p.m. Eastern Time on the due date.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a Presiding Officer. Participants are requested not to include personal 
privacy information, such as social security numbers, home addresses, 
or home phone numbers in their filings. With respect to copyrighted 
works, except for limited excerpts that serve the purpose of the 
adjudicatory filings and would constitute a Fair Use application, 
Participants are requested not to include copyrighted materials in 
their submission.
    Non-timely 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 amendment action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan Date of amendment request: July 12, 2007, as supplemented by 
letter dated September 21, 2007.

    Description of amendment request: The proposed amendment would 
revise Surveillance Requirement (SR) 3.3.3.1.2 in Technical 
Specification (TS) 3.3.3.1, ``Post Accident Monitoring 
Instrumentation.'' Specifically, the proposed amendment would delete 
the note which excludes radiation detectors from calibration 
requirements.
    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 change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Primary Containment Radiation Monitors are part of the post 
accident monitoring instrumentation. Deleting the note excluding 
radiation detectors from the channel calibration requirement in TS 
3.3.3.1 surveillance requirement does not adversely affect any of 
the parameters in accident analyses. Revising the detectors 
calibration requirement does not affect the probability or 
consequences of previously evaluated accidents. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Deleting the note excluding the radiation detectors from channel 
calibration requirement in SR 3.3.3.1.2 restores the calibration of 
the primary containment high range radiation monitors to the 
requirements in NUREG-0737 [,''Clarification of TMI Action Plan 
Requirements''], Table II.F.1-3. The revision of the primary 
containment high range radiation monitor calibration provides an 
improved assurance of the accuracy and function of the monitor 
during and following an accident. These monitors provide indication 
of high-range radiation and are primarily used by emergency response 
personnel for evaluating protective action recommendations. These 
monitors are provided for indication only and do not initiate any 
automatic action. Removing the exclusion of radiation detectors from 
the channel calibration requirement in SR 3.3.3.1.2 cannot create a 
new or different kind of accident from previously evaluated 
accidents. 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 change does not involve a significant reduction 
in the margin of safety.
    This proposed license amendment involves a change in the channel 
calibration surveillance of primary containment high range radiation 
monitor in TS 3.3.3.1. The surveillance frequency is unchanged. The 
change in the high range radiation monitor channel calibration only 
removes the exclusion of the detectors from SR 3.3.3.1.2. 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 G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
    NRC Acting Branch Chief: Travis L. Tate.

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

    Date of amendment request: August 21, 2007.
    Description of amendment request: The proposed amendment would 
modify technical specification (TS) requirements related to control 
room envelope (CRE) habitability in TS 3.7.10, ``Control Room 
Ventilation Filtration,'' and TS Section 5.5, ``Administrative 
Controls--Programs and Manuals.'' The NRC staff issued a notice of 
opportunity for comment in the Federal Register on October 17, 2006 (71 
FR 61075), on possible amendments in accordance with TSTF-448-A, 
Revision 3, to revise the plant specific TS to strengthen requirements 
regarding CRE habitability by changing

[[Page 62688]]

the action and surveillance requirements for the CRE emergency 
ventilation system, and by adding a new TS administrative controls 
program on CRE habitability.
    The NRC staff subsequently issued a notice of availability of the 
models for referencing in license amendment applications in the Federal 
Register on January 17, 2007 (72 FR 2022). The licensee affirmed the 
applicability of the following no significant hazards determination in 
its application dated August 21, 2007.
    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 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 I 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: 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, Arkansas

    Date of amendment request: October 5, 2007.
    Description of amendment request: The proposed amendment changes 
the containment sump buffering agent in Technical Specification (TS) 
3.6.2.2, ``Trisodium Phosphate (TSP),'' from TSP to sodium tetraborate 
(NaTB).
    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.
    There are no changes to the design or operation of the plant 
that could affect system, component, or accident functions as a 
result of replacing trisodium phosphate (TSP) with sodium 
tetraborate (NaTB). Buffering agents are used to minimize the 
potential consequences of an accident and do not represent an 
accident initiator. Utilizing NaTB as a buffering agent ensures the 
post-loss-of-coolant accident (LOCA) containment sump mixture will 
have an equilibrium pH >= 7.0. Replacing TSP with NaTB, which 
achieves comparable pH buffering results, will maintain the iodine 
retention and corrosion inhibition required by the safety analyses.
    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 accident scenarios, failure mechanisms, or single 
failures are introduced as a result of the proposed change. 
Structures, systems, and components previously required for 
mitigation of an event remain capable of fulfilling their intended 
design function with this change to the TS. The proposed change has 
no new adverse effects on safety-related systems or components and 
does not challenge the performance or integrity of safety-related 
systems. The replacement buffering agent has been evaluated and no 
new accident scenarios or single failures are introduced.
    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 safety analyses assume a post-LOCA equilibrium pH >= 7.0 to 
minimize iodine re-evolution and to minimize corrosion of components 
within containment. Changing the containment sump buffering agent 
requirement from TSP to NaTB and revising the required volume of 
NaTB continues to ensure a containment sump equilibrium pH >= 7.0. 
The margin for pH control is not altered by the proposed changes.
    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.

[[Page 62689]]

    NRC Branch Chief: Thomas G. Hiltz.
    FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
No. 1, Rockingham County, New Hampshire
    Date of amendment request: July 17, 2007.
    Description of amendment request: The proposed amendment would 
revise the Seabrook Station, Unit No. 1 Technical Specifications (TSs) 
related to control room envelope (CRE) habitability consistent with 
Technical Specification Task Force (TSTF) change traveler TSTF-448, 
Revision 3, ``Control Room Habitability.''
    The NRC staff issued a notice of availability of a model safety 
evaluation, a model no significant hazards consideration (NSHC) 
determination, and a model license amendment request in the Federal 
Register on January 17, 2007 (72 FR 2022). In its application dated 
July 17, 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 no significant hazards consideration, is presented below:
    1. The proposed changes do 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.
    2. The proposed changes do 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.
    3. The proposed changes do 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 shutdown 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 proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Harold K. Chernoff.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant (MNGP), Wright County, Minnesota

    Date of amendment request: September 17, 2007.
    Description of amendment request: The proposed amendment would 
revise the MNGP Technical Specifications (TS) by adding an Action 
Statement to the Limiting Condition for Operation (LCO) for 
Specification 3.7.5, ``Control Room Ventilation System.'' The new 
Action Statement will allow a finite time (72 hours) to restore one 
control room ventilation subsystem to operable status when both 
ventilation subsystems are inoperable, provided the control room 
temperature is verified to be less than 90 degrees once every 4 hours.
    The proposed amendment is consistent with the NRC's approved 
Technical Specification Task Force (TSTF) Improved Standard Technical 
Specifications Change Traveler, TSTF-477, Revision 3 (March 26, 2007; 
72 FR 14143).
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Part 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration (NSHC) by 
referencing the NRC staff's model NSHC analysis published on December 
18, 2006 (71 FR 75774). The NRC staff's model NSHC analysis is 
reproduced 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 is described in Technical Specification Task 
Force (TSTF) Standard TS Change Traveler TSTF-477[; it] adds an 
action statement for two inoperable control room subsystems.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
The proposed changes add an action statement for two inoperable 
control room subsystems. The equipment qualification temperature of 
the control room equipment is not affected. Future changes to the 
Bases or licensee-controlled document will be evaluated pursuant to 
the requirements of 10 CFR 50.59, ``Changes, test and experiments,'' 
to ensure that such changes do not result in more than a minimal 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not adversely 
affect the ability of structures, systems and

[[Page 62690]]

components (SSCs) to perform their intended safety function to 
mitigate the consequences of an initiating event within the assumed 
acceptance limits. The proposed changes do not affect the source 
term, containment isolation, or radiological consequences of any 
accident previously evaluated. Further, the proposed changes do not 
increase the types and the amounts of radioactive effluent that may 
be released, nor significantly increase individual or cumulative 
occupation/public radiation exposures.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated.

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

    The proposed changes add an action statement for two inoperable 
control room subsystems. The changes do not involve a physical 
altering of the plant (i.e., no new or different type of equipment 
will be installed) or a change in methods governing normal pant 
operation. The requirements in the TS continue to require 
maintaining the control room temperature within the design limits.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.

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

    The proposed changes add an action statement for two inoperable 
control room subsystems. Instituting the proposed changes will 
continue to maintain the control room temperature within design 
limits. Changes to the Bases or licensee controlled document are 
performed in accordance with 10 CFR 50.59. This approach provides an 
effective level of regulatory control and ensures that the control 
room temperature will be maintained within design limits.
    The proposed changes maintain sufficient controls to preserve 
the current margins of safety. Based upon the reasoning above, the 
NRC staff concludes that the amendment request involves no 
significant hazards consideration.

    The NRC staff has reviewed the licensee's referenced analysis, and 
has found that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the proposed 
amendment 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, Nebraska

    Date of amendment request: September 21, 2007.
    Description of amendment request: The proposed amendment will 
revise the Technical Specifications (TS) safety limit (SL) requirements 
related to the use of a non-cycle specific peak linear heat rate (PLHR) 
SL of 22 kilowatts per foot (kW/ft) to fuel centerline melt. The 
proposed change is consistent with the Technical Specification Task 
Force (TSTF) 445-A, Revision 1. Because these Limiting Safety System 
Setting (LSSS) values appear in the Fort Calhoun Station, Unit 1 (FCS), 
TS Bases section of TS 1.3, ``Limiting Safety System Settings, Reactor 
Protective System,'' TS 1.0, ``Safety Limits and Limiting Safety System 
Settings,'' will be revised to more clearly align with the Combustion 
Engineering (CE) Standard Technical Specifications (STS) 2.0 in 
content. Therefore, TS Section 1.1, ``Safety Limits--Reactor Core,'' 
will be revised to incorporate the TSTF-445-A, Revision 1, peak fuel 
centerline temperature (PFCT) criteria, and TS 1.2, ``Safety Limits--
Reactor Coolant System Pressure,'' will be revised to incorporate the 
SL violation action which is currently delineated in administrative 
control TS 5.7.1. TS Section 1.3 will be relocated to the currently 
unused TS Section 2.13 to be more consistent with the content of the CE 
STS (i.e., the LSSS will be located in the Limiting Conditions for 
Operation (LCO) section of the FCS TS which is similar to the LCO/
Surveillance Requirements Section 3.0 of the STS). As noted above, the 
administrative control in TS 5.7.1, ``Safety Limit Violation,'' will be 
relocated to TS Section 2.13. Also, administrative control TS 5.9.5, 
``Core Operating Limits Report (COLR),'' item a., will be revised to 
add TS 2.13, ``RPS Limiting Safety System Settings,'' Table 2-11, Items 
6, 8, and 9, to the list of items that shall be documented in the COLR. 
The TS table of contents (TOC) will be updated to reflect the deletion 
and subsequent renumbering of TS 1.3 and Table 1-1 to TS 2.13 and Table 
2-11, respectively. In addition, the TOC will be updated to delineate 
the new TS subsections 1.1.1 and 1.1.2, provide the revised titles for 
TS 1.0, 1.1, 1.2, and 2.13, and to reflect TS 5.7.1 as ``Not used.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change does not require any physical change to any 
plant systems, structures, or components nor does it require any 
change in systems or plant operations. The proposed change does not 
require any change in safety analysis methods or results. The change 
to establish the PFCT as the safety limit is consistent with the FCS 
licensing basis for ensuring that the fuel design limits are met. 
Operations and analysis will continue to be in compliance with NRC 
regulations.
    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 accident analyses indicate that the peak linear heat rate 
may exceed the present Limiting Safety System Setpoint of 22 kW/ft 
during the Control Element Assembly (CEA) Drop, Excess Load, and 
Loss of Feedwater Heating events. The analyses for these 
[anticipated operational occurrences] indicate that the PFCT is not 
significantly challenged or exceeded. The existing analyses remain 
unchanged and do not affect any accident initiators that would 
create a new accident.
    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 proposed change does not require any change in accident 
analysis methods or results. Therefore, by changing the SL from PLHR 
to peak fuel centerline melt temperature, the margin, as established 
in the current licensing basis, remains unchanged. The proposed 
administrative change relocates descriptive information from one 
section of the TS to another TS section, thereby maintaining the 
information in the FCS TS, which has no effect on safety margins.
    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.

[[Page 62691]]

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of amendment request: October 3, 2007.
    Description of amendment request: The Edwin I. Hatch Nuclear Plant, 
Units 1 and 2, requests adoption of an approved change to the Standard 
Technical Specifications (STSs) for Boiling Water Reactor (BWR) plants 
(NUREG-1433 and NUREG-1434) and plant-specific technical specifications 
(TSs), to allow the use of the improved banked position withdrawal 
sequence (BPWS) during shutdowns in accordance with NEDO-33091-A, 
Revision 2, ''Improved BPWS Control Rod Insertion Process,'' dated July 
2004. The changes are consistent with NRC-approved Industry/Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler, TSTF-476, Revision 1.
    The Nuclear Regulatory Commission (NRC) staff published a notice of 
availability of a model safety evaluation and model no significant 
hazards consideration (NSHC) determination for referencing in license 
amendment applications in the Federal Register on May 23, 2007, (72 FR 
29004) as apart of the consolidated line-term improvement process 
(CLIIP), and a correction on May 30, 2007, (72 FR 30043). The licensee 
affirmed the applicability of the model NSHC determination in its 
application dated October 3, 2007.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed changes modify the TS to allow the use of the 
improved banked position withdrawal sequence (BPWS) during shutdowns 
if the conditions of NEDO-33091-A, Revision 2, ''Improved BPWS 
Control Rod Insertion Process,'' July 2004, have been satisfied. The 
staff finds that the licensee's justifications to support the 
specific TS changes are consistent with the approved topical report 
and TSTF-476, Revision 1. Since the change only involves changes in 
control rod sequencing, the probability of an accident previously 
evaluated is not significantly increased, if at all. The 
consequences of an accident after adopting TSTF-476 are no different 
than the consequences of an accident prior to adopting TSTF-476. 
Therefore, the consequences of an accident previously evaluated are 
not significantly affected by this change.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

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

    The proposed change will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The control rod drop accident (CRDA) 
is the design basis accident for the subject TS changes. This change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The proposed change, TSTF-476, Revision 1, incorporates the 
improved BPWS, previously approved in NEDO-33091-A, into the 
improved TS. The control rod drop accident (CRDA) is the design 
basis accident for the subject TS changes. In order to minimize the 
impact of a CRDA, the BPWS process was developed to minimize control 
rod reactivity worth for BWR plants. The proposed improved BPWS 
further simplifies the control rod insertion process, and in order 
to evaluate it, the staff followed the guidelines of Standard Review 
Plan Section 15.4.9, and referred to General Design Criterion 28 of 
Appendix A to 10 CFR part 50 as its regulatory requirement. The TSTF 
stated the improved BPWS provides the following benefits: (1) Allows 
the plant to reach the all-rods-in condition prior to significant 
reactor cool down, which reduces the potential for re-criticality as 
the reactor cools down; (2) reduces the potential for an operator 
reactivity control error by reducing the total number of control rod 
manipulations; (3) minimizes the need for manual scrams during plant 
shutdowns, resulting in less wear on control rod drive (CRD) system 
components and CRD mechanisms; and (4) eliminates unnecessary 
control rod manipulations at low power, resulting in less wear on 
reactor manual control and CRD system components. The addition of 
procedural requirements and verifications specified in NEDO-33091-A, 
along with the proper use of the BPWS will prevent a control rod 
drop accident (CRDA) from occurring while power is below the low 
power setpoint (LPSP). The net change to the margin of safety is 
insignificant. Therefore, this change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the 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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Branch Chief: Evangelos C. Marinos.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of amendment request: October 5, 2007.
    Description of amendment request: The proposed amendments would add 
a 72-hour Completion Time (CT) for an inoperable swing diesel 
generator. Currently, the Technical Specifications (TSs) provide a 14-
day CT which may be used provided that planned maintenance on certain 
plant components is restricted prior to entering, and for the duration 
of, the 14-day CT. The 72-hour CT and the 14-day CT are explicitly 
addressed in the TS Limiting Condition for Operation (LCO) for four of 
the five emergency diesel generators. In addition, the 14-day CT is 
explicitly addressed in the TS LCO for the fifth diesel generator, the 
swing diesel generator. Further, the existing 14-day CT and the 
proposed 72-hour CT are currently described in the TS Bases for LCO 
3.8.1. This proposed change will provide an explicit reference to the 
72-hour CT in the actual TS for the swing diesel.
    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?
    The proposed change will explicitly add, to the Technical 
Specifications (TS), a 72 hour CT for the swing diesel generator in 
addition to the 14 day CT already listed. The 72 hour CT is 
currently imposed on the swing DG until it can be verified that 
planned maintenance restrictions are in place. Mention of the 
planned restrictions is also being added to the specifications for 
the dedicated DGs for consistency. This TS change does not propose 
any physical changes to systems or components that are important to 
safety, including those systems that are designed to prevent 
previously evaluated accidents, or to mitigate the consequences of 
those accidents.
    Additionally, this proposed TS change does not change any safety 
analyses for LOCA/LOSP [loss-of-coolant accident/loss-

[[Page 62692]]

of-offsite power] with respect to diesel generator availability or 
capabilities. This change does not request an increase to the diesel 
generator out of service CT, in fact, it acts to enforce a 72 hour 
CT for the swing diesel.
    Consequently, this TS change does not significantly increase the 
probability or consequences of a previously evaluated accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    This proposed TS change explicitly adds a 72 hour CT to the 
swing diesel generator ``Completion Time'' column of LCO 3.8.1, and 
adds mention of the planned maintenance restrictions to the same 
column for the dedicated DGs as well. These TS changes will reflect 
the current practices of the operating staff with respect to the 
handling of inoperable diesel generators. No requests are being made 
to increase the CTs for the diesel generators; instead, the more 
restrictive 72 hour CT for the swing diesel is being explicitly 
added to the TS, which currently only includes the 14 day CT.
    No changes are being made to the operations, maintenance, or 
testing of plant equipment. No new modes of operation are proposed 
and therefore, no new failure modes are introduced.
    Consequently, 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 
the margin of safety?
    This TS change will include a more restrictive 72 hour CT for 
the swing diesel generator in addition to the 14 day CT currently 
listed in the TS. The 72 hour CT will reflect that planned 
maintenance restrictions must be in place before using the 14 day CT 
on the swing DG. For consistency, mention of the maintenance 
restrictions is being added to the CT for the dedicated DGs as well. 
These changes are more restrictive than what is currently included 
in the TS. Accordingly, the changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Branch Chief: Evangelos C. Marinos.

 Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of application for amendment: March 22, 2007, as supplemented 
by letter dated July 25, 2007.
    Brief description of amendment: The amendment consists of changes 
to various technical specifications (TSs) related to the variable low 
reactor coolant system pressure-temperature core protection safety 
limit, which is being changed to accommodate the introduction of AREVA 
NP's Mark-B-HTP fuel design in the TMI-1 cycle 17 reload (fall 2007).
    Date of issuance: October 15, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No. 262.
    Facility Operating License No. DPR-50. Amendment revised the 
license and the TSs.
    Date of initial notice in Federal Register: April 24, 2007 (72 FR 
20377). The supplement dated July 25, 2007, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed and did not change the NRC 
staff's original proposed no significant hazards determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated October 15, 2007.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of application for amendment: November 1, 2006, as 
supplemented by letters dated November 22, 2006, April 4, May 7, August 
16, and September 21, 2007.
    Brief description of amendment: The change revised the Grand Gulf 
Nuclear Station, Unit 1, Technical Specification Surveillance 
Requirement 3.3.1.1.7 for the surveillance interval of the local power 
range monitor calibrations from 1000 megawatt-days per ton (MWD/T) 
(approximately every 36 days) to 2000 MWD/T (approximately every 72 
days).
    Date of issuance: October 24, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 177.
    Facility Operating License No. NPF-29: The amendment revises the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 5, 2006 (71 FR 
70559). The supplements dated November 22, 2006, April 4, May 7, August 
16, and September 21, 2007, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.

[[Page 62693]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 24, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 17, 2006.
    Brief description of amendments: The proposed amendment would 
modify the technical specifications (TS) to replace references to 
Section XI of the American Society of Mechanical Engineers Boiler and 
Pressure Vessel Code (ASME Code) with a reference to the ASME Code of 
Operation and Maintenance of Nuclear Power Plants in TS 5.5.7, 
``Inservice Testing Program.'' These changes are consistent with the 
implementation of the LSCS, Units 1 and 2, for the third 10-year IST 
program.
    Date of issuance: October 12, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 185/172.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: April 10, 2007 (72 FR 
17948). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 12, 2007.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et. al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: October 19, 2006.
    Brief description of amendments: The amendments revise various 
technical specifications (TSs) to address requirements that should have 
been changed as part of amendments previously approved by the NRC, as 
well as to correct some typographical errors.
    Date of Issuance: October 22, 2007.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 202 and 149.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised TSs.
    Date of initial notice in Federal Register: January 3, 2007 (72 FR 
00153).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 22, 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, Ohio

    Date of application for amendment: February 12, 2007.
    Brief description of amendment: This amendment revised Technical 
Specification 3/4.9.4, ``Containment Penetrations,'' to allow 
containment penetrations that provide direct access from the 
containment atmosphere to the outside to be open during refueling 
activities if appropriate administrative controls are established.
    Date of issuance: October 17, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 277.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: April 10, 2007 (72 FR 
17949).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 17, 2007.
    No significant hazards consideration comments received: No.

FPL Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point Beach 
Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc County, 
Wisconsin

    Date of application for amendments: December 14, 2006, as 
supplemented by letter dated June 13, 2007.
    Brief description of amendments: The amendments revised Technical 
Specification 5.6.5 (Reactor Coolant System (RCS) Pressure and 
Temperature Limits Report( to add the FERRET Code as an approved 
methodology for determining RCS pressure and temperature limits.
    Date of issuance: October 18, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 229 and 234.
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: 
Amendments revised the Technical Specifications/License.
    Date of initial notice in Federal Register: January 16, 2007 (72 FR 
1780).
    The June 13, 2007, supplement, contained clarifying information and 
did not change the staff(s initial proposed finding of no significant 
hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 18, 2007.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315, Donald C. Cook 
Nuclear Plant, Units 1 and 2 (DCCNP-1 and DCCNP-2), Berrien County, 
Michigan

    Date of application for amendments: June 27, 2007, as supplemented 
on September 21, 2007.
    Brief description of amendments: The amendments made changes to 
Sections 3.3.3, ``Post Accident Monitoring Instrumentation,'' 3.5, 
``Emergency Core Cooling Systems,'' and 3.6.14, sbull I11``Containment 
Recirculation Drains,'' of the DCCNP-1 and DCCNP-2 Technical 
Specifications to reflect resolution of issues raised by Generic Letter 
(GL) 2004-02, ``Potentia