[Federal Register Volume 72, Number 223 (Tuesday, November 20, 2007)]
[Notices]
[Pages 65360-65377]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-22331]


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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 October 25, 2007, to November 7, 2007. The 
last biweekly notice was published on November 6, 2007 (72 FR 62685).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.

[[Page 65361]]

    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, 
person(s) 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 
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, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request 
for a hearing or petition for leave to intervene is filed within 60 
days, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order, which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion, which supports 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 
[email protected], 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 Viewer 
TM to access the Electronic Information Exchange (EIE), a 
component of the E-Filing system.
    The Workplace Forms Viewer TM is free and is available 
at http://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 http://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 http://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

[[Page 65362]]

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 http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC 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 
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, the 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, 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].

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

    Date of amendments request: October 17, 2007.
    Description of amendments request: The proposed amendment would 
modify the Technical Specifications (TS) to establish more effective 
and appropriate action, surveillance, and administrative requirements 
related to the inoperability of snubbers in accordance with Nuclear 
Regulatory Commission (NRC)-approved TS Task Force (TSTF) change 
traveler TSTF-372-A, Revision 4. Specifically, the proposed amendment 
would add Limiting Condition for Operation (LCO) 3.0.8. The NRC staff 
issued a ``Notice of Opportunity To Comment on Model Safety Evaluation 
on Technical Specification Improvement To Modify Requirements Regarding 
the Addition of LCO 3.0.8 on the Inoperability of Snubbers Using the 
Consolidated Line Item Improvement Process'' in the Federal Register on 
November 24, 2004 (69 FR 68412). The notice included a model safety 
evaluation (SE) and a model no-significant-hazards-consideration (NSHC) 
determination. The NRC staff issued a ``Notice of Availability of Model 
Application Concerning Technical Specification Improvement To Modify 
Requirements Regarding the Addition of Limiting Condition for Operation 
3.0.8 on the Inoperability of Snubbers Using the Consolidated Line Item 
Improvement Process'' in the Federal Register on May 4, 2005 (70 FR 
23252). The notice included a model application, including a revised 
model SE. In its application dated October 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 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 allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an inoperable snubber if risk is assessed and managed. The 
postulated seismic event requiring snubbers is a low probability 
occurrence and the overall TS system safety function would still be 
available for the vast majority of anticipated challenges. 
Therefore, the probability of an accident previously evaluated is 
not significantly increased, if at all. The consequences of an 
accident while relying on allowance provided by proposed LCO 3.0.8 
are no different than the consequences of an accident while relying 
on the TS required actions in effect without the allowance provided 
by proposed LCO 3.0.8. Therefore, the consequences of an accident 
previously evaluated are not significantly affected by this change. 
The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
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 does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed).

[[Page 65363]]

Allowing delay times for entering supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is 
assessed and managed, 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 addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, 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 allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low probability occurrence and the 
overall TS system safety function would still be available for the 
vast majority of anticipated challenges. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in RG 1.177. A bounding risk assessment was performed to 
justify the proposed TS changes. This application of LCO 3.0.8 is 
predicated upon the licensee's assessment and management of plant 
risk. 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 adopted by the licensee 
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 amendments request involves NSHC.
    Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 
17th floor, Baltimore, MD 21202.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment request: September 14, 2007.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) requirements related to 
control room envelope habitability. The proposed changes include 
revisions to the control room post-accident recirculation system, the 
instrument operating conditions for isolation functions, and a control 
room envelope habitability program. The changes are consistent with TS 
Task Force (TSTF) Change Traveler TSTF-448-A, Revision 3, ``Control 
Room Habitability,'' except for the differential pressure surveillance 
requirements. The availability of this TS improvement was published in 
the Federal Register on January 17, 2007 (72 FR 2022).
    In addition to the changes related to TSTF-448-A, the proposed 
amendment would: (1) Align TS with those delineated in NUREG-1431, 
Revision 3, ``Standard Technical Specifications, Westinghouse Plants,'' 
to the extent necessary to adopt TSTF-448-A, including the adoption of 
the necessary portions of TSTF-51-A, Revision 2, ``Revise Containment 
Requirements During Handling of Irradiated Fuel and Core Alterations,'' 
and TSTF-287-A, Revision 5, ``Ventilation System Envelope Allowed 
Outage Time,'' (2) add TS for control room radiation monitor R-23 
(ventilation system air monitor), and (3) reformat or clarify current 
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?
    No.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility. The proposed changes do not 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. This is a 
revision to the TS for the control room post-accident recirculation 
system and control room isolation function, which are mitigation 
systems designed to minimize unfiltered air in-leakage into the 
control room envelope and to filter the control room envelope 
atmosphere to protect the control room envelope occupants following 
accidents previously analyzed. An important part of the system is 
the control room envelope boundary. The control room envelope post-
accident recirculation system is not an initiator or precursor to 
any accident previously evaluated. Therefore, the probability of any 
accident previously evaluated is not significantly increased.
    Establishing operability requirements for SSCs, performing tests 
and implementing programs that verify the integrity of the control 
room envelope boundary and control room envelope habitability ensure 
that the mitigation features are capable of performing their assumed 
functions. Therefore, the consequences of any accident previously 
evaluated are not significantly increased.
    Therefore, the proposed changes do 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?
    No.
    The proposed changes will not significantly change the 
requirements of the control room envelope ventilation system or its 
function during accident conditions. No new or different accidents 
result from performing the new surveillance or following the new 
program. The changes do 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 changes are consistent with the safety 
analysis assumptions including the revised gas decay tank and volume 
control tank rupture analysis and current plant operating practice.
    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 a margin of safety?
    No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not affected by these changes. The proposed changes will not 
result in plant operation in a configuration outside the design 
basis for an unacceptable period without compensatory measures. The 
proposed changes do not significantly affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition.
    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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, Riverside 2, Richmond, 
VA 23219.
    NRC Acting Branch Chief: Travis L. Tate.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin
    Date of amendment request: October 2, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Sections 3.7, ``Auxiliary 
Electrical Systems'' and 4.6, ``Periodic Testing of Emergency Power 
System,'' to change the testing requirements for ensuring operability 
of the remaining operable emergency diesel generator (EDG) when the 
other EDG is inoperable. In addition, the

[[Page 65364]]

proposed amendment would add a new specification when two EDGs are 
inoperable and revise the surveillance requirements for the EDGs.
    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?
    No.
    The proposed amendment would clarify testing requirements for 
the operable EDG, when one EDG is inoperable, and limit testing to 
only the intended purpose of the requirement. The intended purpose 
of the testing requirement is to provide reasonable assurance that 
when an EDG is inoperable, the opposite EDG is operable. The 
proposed change does not affect the initiators of analyzed events or 
the assumed mitigation of accident or transient events. 
Specifically, testing of the remaining operable diesel will still 
occur unless evaluation of the inoperable EDG confirms that its 
failure is not attributable to a common cause failure mechanism. 
Furthermore, the proposed change clarifies the surveillance testing 
necessary to give reasonable assurance of operability and restricts 
the amount of time to perform the testing (i.e. with two inoperable 
EDGs) to two hours. This ensures no significant increase in the 
probability of a loss-of-power during the period of the confirming 
surveillance concurrent with an opposite train inoperable EDG. 
Elimination of unnecessary testing by acceptable evaluation of the 
operable EDG reduces component wear and promotes overall EDG 
reliability and availability. Clarification of required testing and 
restriction in the amount of time to complete the surveillance to 
confirm operability, reduces the probability and significance of 
common mode failures.
    The proposed amendment would also add a new specification 
allowing two EDGs to be inoperable for up to two hours. This change 
does not significantly increase the initiators of analyzed events or 
the assumed mitigation of any accidents or transients. 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 does not involve a physical alteration of 
the plant or a change in the methods used to respond to any 
evaluated plant accident. No new or different equipment is being 
installed and no installed equipment is being removed or operated in 
a different manner. Only a surveillance test clarification and 
limited two-hour action statement have been added to permit testing 
of the opposite train, operable EDG. Although the diesel generators 
will be tested in a different manner, the proposed changes will 
improve the availability and reliability of the diesel generators 
without creating the possibility of a new or different kind of 
accident from any accident previously evaluated. Furthermore, there 
is no alteration to the parameters within which the plant is 
normally operated or in the setpoints, which initiate protective or 
mitigative actions. Since the diesel generators will continue to be 
operated in the same manner and the proposed test protocol will 
improve diesel generator availability and reliability, no new 
failure modes are introduced by the proposed amendment.
    Therefore, the proposed amendment 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 amendment would add a TS allowing two EDGs to be 
inoperable for up to two hours before the plant must be shut down in 
a controlled manner. Allowing two EDGs to be inoperable for this 
limited period of time, while the normal offsite power source 
remains available, is consistent with Regulatory Guide 1.93 and not 
considered to be a significant reduction in a margin of safety.
    Station operations and EDG surveillance requirements are not 
adversely affected by the proposed change. Furthermore, the proposed 
amendment does not adversely impact the condition or performance of 
structures, systems or components relied upon for accident 
mitigation or any safety analysis assumptions. The proposed 
amendment adds provisions to reduce EDG wear and increase 
availability.
    Therefore, the proposed amendment to the KPS [Kewaunee Power 
Station] TS does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 
120 Tredegar Street, Richmond, VA 23219.
    NRC Acting Branch Chief: Travis L. Tate.

Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: October 16, 2007.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to accommodate plant modifications 
that will address water hammer concerns described in Generic Letter 96-
06, ``Assurance of Equipment Operability and Containment Integrity 
During Design-Basis Conditions,'' dated September 30, 1996.
    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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The requested license amendment seeks approval for the Low 
Pressure Service Water Reactor Building Waterhammer Prevention 
System that is being added to the design of the three Oconee Units 
and the associated revised Technical Specifications. The Low 
Pressure Service Water Reactor Building Waterhammer Prevention 
modification will provide a combination passive and automatic means 
to isolate the Low Pressure Service Water flow stream to the Reactor 
Building Cooling Units, Reactor Building Auxiliary Coolers, and 
Reactor Coolant Pump Motor Coolers on a loss of Low Pressure Service 
Water flow that can lead to a waterhammer should the Low Pressure 
Service Water system become depressurized.
    New check valves and air operated valves are added into an 
Engineered Safeguards flowpath. The existing Low Pressure Service 
Water header that discharges from the Reactor Building Cooling Units 
is to be modified by separating it into two headers and then joining 
back into a common header. Each header will contain two air operated 
valves. The Waterhammer Prevention System maintains the Low Pressure 
Service Water System inside containment water solid during a Loss of 
Offsite Power such that voids, which could later collapse, cannot 
form. The Waterhammer Prevention System will eliminate an Operable 
but degraded/non-conforming condition associated with waterhammers.
    The design of the proposed modification and its associated 
Technical Specifications will provide means to assure that the Low 
Pressure Service Water Reactor Building Waterhammer Prevention 
System operates at a performance level necessary to provide for safe 
operation of the Low Pressure Service Water system following 
installation on each of the three Units. The system is designed such 
that a single active failure will not prevent the system from 
preventing a waterhammer event if power is lost to the Low Pressure 
Service Water pumps (e.g., Loss of Offsite Power), nor will a single 
active failure prevent the Engineered Safeguards flowpath from being 
available if needed during a Loss of Coolant Accident or Main Steam 
Line Break. Evaluations have been performed to assure that the risk 
of adding new hardware is acceptable.
    Therefore, the addition of this modification and associated 
Technical Specifications does not significantly increase the 
probability or consequences of any accident previously evaluated.

[[Page 65365]]

    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed Low Pressure Service Water Reactor Building 
Waterhammer Prevention Modification and its associated Technical 
Specifications will provide a means to assure the mechanical and 
electrical components operate at a performance level necessary to 
provide for safe operation of the modified Low Pressure Service 
Water system flow to the Reactor Building Cooling Units, Reactor 
Building Auxiliary Coolers and Reactor Coolant Pump Motor Coolers.
    The change enhances the plant design by eliminating the 
possibility of significant waterhammers that occur on a loss of Low 
Pressure Service Water flow to the above components.
    The modification does not add any new single active failures 
that would prevent the Low Pressure Service Water System from 
supplying cooling water to the Reactor Building Cooling Units. The 
Reactor Building Cooling Units will be isolated briefly during an 
Engineered Safeguards event; however, the flow path will be restored 
before cooling is required following the event. Since cooling was 
previously not available until after power restoration following a 
Loss of Offsite Power, there is no change in system response 
regarding Low Pressure Service Water flow through the Reactor 
Building Cooling Units when compared to the previous design.
    Therefore, the proposed modification and associated Technical 
Specifications will not create the possibility of a new or different 
kind of accident from any kind of accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change does not adversely affect any plant safety 
limits, setpoints, or design parameters. The change also does not 
adversely affect the fuel, fuel cladding, Reactor Coolant System, or 
Containment Operability. The Reactor Building Cooling Units will be 
isolated briefly during an Engineered Safeguards event; however, the 
flow path will be restored before cooling is required following the 
event.
    Since cooling is currently not available until after power 
restoration following a Loss of Offsite Power, there is no change in 
system response regarding Low Pressure Service Water flow through 
the Reactor Building Cooling Units when compared to the previous 
design.
    The modification mitigates significant waterhammers in the Low 
Pressure Service Water piping to the Reactor Building Cooling Units 
and Reactor Cooling Pump Motor Coolers. The change will maintain the 
ability to provide Low Pressure Service Water flow to safety related 
loads following Loss of Offsite Power events.
    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: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Evangelos C. Marinos.

Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina.

    Date of amendment request: October 22, 2007.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to accommodate the use of AREVA NP 
Mark-B-HTP fuel.
    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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed revisions to the technical specifications and to 
Duke's NRC-approved methodology reports support the use of the AREVA 
NP Mark-B-HTP fuel design. The methodology will be approved by the 
NRC prior to plant operation with the new fuel. The proposed safety 
limit ensures that fuel integrity will be maintained during normal 
operations and anticipated operational transients. The core 
operating limits report will be developed in accordance with the 
approved methodology. The proposed safety limit value does not 
affect the performance of any equipment used to mitigate the 
consequences of an analyzed accident. There is no impact on the 
source term or pathways assumed in accidents previously assumed. No 
analysis assumptions are violated and there are no adverse effects 
on the factors that contribute to offsite or onsite dose as the 
result of an accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed safety limit value does not change the methods 
governing normal plant operation, nor are the methods utilized to 
respond to plant transients altered. The BHTP correlation is not an 
accident/event initiator. No new initiating events or transients 
result from the use of the BHTP correlation or the related safety 
limit change.
    3. Involve a significant reduction in a margin of safety.
    The proposed safety limit value has been established in 
accordance with the methodology for the BHTP correlation to ensure 
that the applicable margin of safety is maintained (i.e. there is at 
least 95% probability at a 95% confidence level that the hot fuel 
rod does not experience DNB). The other reactor core safety limits 
will continue to be met by analyzing the reload using NRC approved 
methods and incorporation of resultant operating limits into the 
Core Operating Limits Report (COLR).

    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: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: August 30, 2007.
    Description of amendment request: The proposed amendment would 
modify Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2) 
Technical Specification (TS) requirements related to control room 
envelope habitability in TS 3.7.10, ``Control Room Emergency 
Ventilation System (CREVS)'' and TS Section 5.5, ``Administrative 
Controls--Programs and Manuals.'' This change is consistent with 
Nuclear Regulatory Commission (NRC)-approved Technical Specification 
Task Force (TSTF) Change Traveler TSTF-448, Revision 3. The 
availability of this TS revision was announced in the Federal Register 
on January 17, 2007 (72 FR 2022) as part of the consolidated line item 
improvement process.
    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 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

[[Page 65366]]

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 analysis adopted by the licensee 
and, based on this review, it appears that the 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: David W. Jenkins, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, 
Akron, OH 44308.
    NRC Branch Chief: Mark G. Kowal.

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 amendment request: October 1, 2007.
    Description of amendment request: The proposed amendments would 
revise the accident source term in the design-basis radiological 
consequences analyses and the associated Technical Specifications 
(TSs), pursuant to Section 50.67 of Part 50 of Title 10 of the Code of 
Federal Regulations (10 CFR 50.67). The proposed amendments would 
revise the licensing basis of Point Beach Nuclear Plant, Units 1 and 2 
(PBNP) to support a full-scope application of an Alternative Source 
Term (AST) methodology. The AST methodology will modify PBNP's 
licensing bases by: (1) Replacing the current accident source term with 
an AST as described in 10 CFR 50.67 for design-basis accidents (DBA) 
radiological consequences, and (2) establishing the 10 CFR 50.67 Total 
Effective Dose Equivalent (TEDE) dose limits as acceptance criteria for 
the radiological consequences of DBAs.
    TS changes associated with the AST methodology change are: TS 1.1, 
a reduction in the definition of the maximum allowable containment leak 
rate. TS 3.4.16, the specific activity of the reactor coolant is 
revised for dose equivalent iodine. TS 3.7.9, a new mode of operation 
for the Control Room Emergency Filtration System (CREFS), which will 
allow operation of the CREFS with filtered outside and filtered 
recirculated air.
    TS 3.7.13, the specific activity of the secondary coolant is 
revised for dose equivalent iodine. In addition, a modification to the 
residual heat removal system, containment spray and their support 
systems, will be made to support operation of the containment spray 
system during containment spray recirculation.
    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 results of the applicable radiological design basis accident 
(DBA) re-evaluation demonstrated that, with the requested changes, 
the dose consequences of these limiting events are within the 
regulatory limits and guidance provided by the NRC in 10 CFR 50.67 
and Regulatory Guide 1.183 for alternative source term (AST) 
methodology. The AST is an input to calculations used to evaluate 
the consequences of an accident and does not by itself affect the 
plant response or the actual pathway of the activity released from 
the fuel. It does, however, better represent the physical 
characteristics of the release such that appropriate mitigation 
techniques may be applied.
    The change from the original source term to the new proposed AST 
is a change in the analysis method and assumptions and has no effect 
on accident initiators or causal factors that contribute to the 
probability of occurrence of previously analyzed accidents. Use of 
an AST to analyze the dose effect of DBAs shows that regulatory 
acceptance criteria for the new methodology continues to be met. 
Changing the analysis methodology does not change the sequence or 
progression of the accident scenario.
    The proposed Technical Specification changes reflect the plant 
configuration that will support implementation of the AST analyses. 
The equipment affected by the proposed changes is mitigative in 
nature and relied upon after an accident has been initiated. The 
operation of various filtration systems, the residual heat removal 
and the containment spray system, including associated support 
systems, has been considered in the evaluations for these proposed 
changes. While the operation of these systems does change with the 
implementation of an AST, the affected systems are not accident 
initiators, and application of the AST methodology itself, is not an 
initiator of a design basis accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 65367]]

    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    As described in Item 1 above, the changes proposed in this 
license amendment request involve the use of a new analysis 
methodology and related regulatory acceptance criteria. The proposed 
Technical Specification changes reflect the plant configuration that 
will support implementation of the new methodology. No new or 
different accidents result from utilizing the proposed changes. 
Although the proposed changes require modifications to the control 
room emergency ventilation system, as well as modifications to the 
residual heat removal system and containment spray system, these 
changes will not initiate a new or different kind of accident since 
they are related to system capabilities that provide protection from 
accidents that have already occurred. As a result, no new failure 
modes are being introduced that could lead to different accidents. 
These changes do not alter the nature of events postulated in the 
Updated Final Safety Analysis Report nor do they introduce any 
unique precursor mechanisms.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of Safety.
    Response: No.
    As described in Item 1, the changes proposed in this license 
amendment involve the use of a new analysis methodology and related 
regulatory acceptance criteria. The proposed Technical Specification 
changes reflect the plant configuration that will support 
implementation of the new methodology. Safety margins and analytical 
conservatisms have been evaluated and have been found to be 
acceptable. The analyzed events have been carefully selected and, 
with plant modifications, margin has been retained to ensure that 
the analyses adequately bound postulated event scenarios. The 
proposed changes continue to ensure that the dose consequences of 
DBAs at the exclusion area and low population zone boundaries and in 
the control room are within the corresponding acceptance criteria 
presented in RG 1.183 and 10 CFR 50.67. The margin of safety for the 
radiological consequences of these accidents is 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.

    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. Antonio Fernandez, Senior Attorney, FPL 
Energy Point Beach, LLC P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Acting Branch Chief: Travis L. Tate.
    Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York
    Date of amendment request: September 27, 2007.
    Description of amendment request: The proposed amendment would 
revise the operability requirements contained in Technical 
Specification (TS) Section 3.2.7, ``Reactor Coolant System Isolation 
Valves,'' and associated requirements contained in TS Section 3.6.2, 
``Protective Instrumentation.'' The proposed changes would modify the 
conditions for which reactor coolant system isolation valves (RCSIVs) 
and associated isolation instrumentation must be operable to include 
the hot shutdown reactor operating condition (i.e., when fuel is in the 
reactor vessel and the reactor coolant temperature is greater than 212 
[deg]F). In addition, new requirements are proposed to require that the 
RCSIVs in the shutdown cooling (SDC) system and associated isolation 
instrumentation be operable during the cold shutdown reactor operating 
condition (fuel is in the reactor vessel and the reactor coolant 
temperature is less than or equal to 212 [deg]F) and the refueling 
reactor operating condition (i.e., when fuel is in the reactor vessel 
and the reactor coolant temperature is less than 212 [deg]F). These 
proposed changes will require operability of RCSIVs during conditions 
other than the power operating condition, and are similar in concept to 
primary containment isolation valve operability requirements contained 
in NUREG-1433, ``Standard Technical Specifications General Electric 
Plants, BWR/4.'' Lastly, TS Section 3.6.2 (Table 3.6.2b) would be 
revised to delete unnecessary operability requirements for the cleanup 
system and SDC system high area temperature isolation instrumentation, 
consistent with the proposed revisions to the RCSIV operability 
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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes provide more stringent requirements for 
operation of NMP1. These include requiring operability of RCSIVs and 
associated isolation instrumentation during the hot shutdown 
condition and requiring RCSIVs in the SDC system and associated 
instrumentation to be operable during the cold shutdown and 
refueling operating conditions. Requiring RCSIV operability during 
the hot shutdown operating condition ensures that reactor coolant 
loss in the event of a rupture of a line connected to the reactor 
coolant system (RCS) is minimized, and the release of radioactive 
material to the environment is consistent with the assumptions used 
in the analyses for design basis accidents. Requiring operability of 
the RCSIVs in the SDC system during the cold shutdown and refueling 
operating conditions provides protection against potential draining 
of the reactor vessel through the SDC system during shutdown 
conditions, which is when the SDC system is normally operated.
    In addition, operability requirements for the cleanup system and 
SDC system high area temperature isolation instrumentation are 
revised to be consistent with the proposed revisions to the RCSIV 
operability requirements and with NUREG-1433. The high area 
temperature isolation instrumentation need not be operable in the 
cold shutdown and refueling conditions, since the probability and 
consequences of design basis accidents are reduced due to the 
pressure and temperature limitations of these operating conditions. 
Also, system isolation on high area temperature would likely not 
occur in the event of system leakage or line break since RCS 
temperature during the cold shutdown and refueling conditions is 
typically maintained below the high area temperature isolation 
setpoints (190[deg]F for the cleanup system area and 170[deg]F for 
the SDC system area).
    The revised operability requirements for the RCSIVs and 
associated isolation instrumentation do not result in operation that 
would make an accident more likely to occur and do not alter 
assumptions relative to mitigation of a previously evaluated 
accident. Therefore, the 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 changes to the TS operability requirements for the 
RCSIVs and associated isolation instrumentation do not alter or 
involve any design basis accident initiators. The proposed changes 
do not involve a physical alteration of the plant (no new or 
different type of equipment will be installed) or changes in the 
methods governing normal plant operation. The proposed changes do 
impose different RCSIV operability requirements that are more 
stringent than existing requirements, and incorporate RCSIV 
isolation instrumentation operability requirements that are 
consistent with the RCSIV requirements and with NUREG-1433. These 
changes continue to be consistent with

[[Page 65368]]

the assumptions in the safety analyses and licensing basis. 
Therefore, the proposed changes 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 proposed changes to the TS operability requirements for the 
RCSIVs and associated isolation instrumentation ensure that RCSIV 
closure will occur when required to mitigate the consequences of 
design basis accidents. The proposed changes also ensure that SDC 
system isolation can be accomplished to protect against potential 
draining of the reactor vessel through the SDC system during 
shutdown conditions, which is when the SDC system is normally 
operated. The imposition of these revised RCSIV operability 
requirements either has no impact on or increases the margin of 
plant safety. The plant responses to accidents will not be adversely 
affected, and the accident mitigation equipment will continue to 
function as assumed in the accident analyses. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Mark G. Kowal.
    Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit No, 2 (NMP2), Oswego County, New York
    Date of amendment request: September 19, 2007.
    Description of amendment request: The proposed amendment would 
revise NMP2 Limiting Condition for Operation (LCO) 3.10.1 to expand its 
scope to include provisions for temperature excursions greater than 200 
[deg]F as a consequence of inservice leak and hydrostatic testing, and 
as a consequence of scram time testing initiated in conjunction with an 
inservice leak or hydrostatic test, while considering operational 
conditions to be in Mode 4. This change is consistent with Nuclear 
Regulatory Commission (NRC)-approved Revision 0 to Technical 
Specification (TS) Task Force (TSTF) Change Traveler, TSTF-484, ``Use 
of TS 3.10.1 for Scram Time Testing Activities.'' The availability of 
this TS revision was announced in the Federal Register on October 27, 
2006 (71 FR 63050) as part of the consolidated line item improvement 
process. The licensee affirmed the applicability of the model no 
significant hazards consideration determination in its application.
    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 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
    Technical Specifications currently allow for operation at greater 
than [200][deg]F while imposing MODE 4 requirements in addition to the 
secondary containment requirements required to be met. Extending the 
activities that can apply this allowance will not adversely impact the 
probability or consequences of an accident previously evaluated. 
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.
    Technical Specifications currently allow for operation at greater 
than [200][deg]F while imposing MODE 4 requirements in addition to the 
secondary containment requirements required to be met. No new 
operational conditions beyond those currently allowed by LCO 3.10.1 are 
introduced. The changes do 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 changes do not impose any new or different requirements 
or eliminate any existing requirements. The changes do not alter 
assumptions made in the safety analysis. The proposed changes are 
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.
    Criterion 3: The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety
    Technical Specifications currently allow for operation at greater 
than [200][deg]F while imposing MODE 4 requirements in addition to the 
secondary containment requirements required to be met. Extending the 
activities that can apply this allowance will not adversely impact any 
margin of safety. Allowing completion of inspections and testing and 
supporting completion of scram time testing initiated in conjunction 
with an inservice leak or hydrostatic test prior to power operation 
results in enhanced safe operations by eliminating unnecessary 
maneuvers to control reactor temperature and pressure. 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 on 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 amendments involves no significant hazards 
consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment request: September 25, 2007.
    Description of amendment request: The proposed amendment would 
revise the MNGP licensing basis to incorporate the results of a revised 
small-break loss-of-coolant accident (LOCA) analysis to determining the 
Low Pressure Coolant Injection (LPCI) loop select logic minimum 
detectable break area. This analysis showed that a small break, rather 
than the current large recirculation line break LOCA, would become the 
limiting accident with respect to peak cladding temperature (PCT). In 
conjunction with this proposed new licensing basis analysis, the 
licensee proposed to revise the Table 3.3.5.1-1 (regarding emergency 
core cooling system instrumentation) of the Technical Specifications 
(TS) as follows: (1) change the allowable value from the current 24 
inch water column to 100 inch water column for Function 2.j, 
``Recirculation Riser Differential Pressure--High (Break Detection);'' 
and (2) change the associated channel calibration frequency 
Surveillance Requirement (SR) from a nominal 12-month to a 24-month 
interval.
    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). 
The NRC staff reviewed the licensee's analysis, and has performed its 
own as follows:
    1. Do the proposed changes involve a significant increase in the 
probability or

[[Page 65369]]

consequences of an accident previously evaluated?
    No. The proposed changes to the PCT licensing basis and the TS do 
not involve a physical alteration of the plant, i.e., no design change 
to plant system, and no new or different type of equipment will be 
installed. The proposed PCT change is an analysis result which is 
within regulatory acceptance limits, and the proposed TS changes 
reflect the revised analysis. Thus, the proposed changes affect only 
parameters assumed for certain analyses, but do not adversely affect 
accident initiators, precursors, plant design, configuration, or the 
manner in which the plant is operated and maintained. The proposed 
changes do not adversely affect the ability of structures, systems and 
components 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 capability, or radiological consequences of any accident 
previously evaluated. Furthermore, the proposed changes do not increase 
the types and the amounts of radioactive effluent that may be released, 
and do not significantly increase individual or cumulative 
occupational/public radiation exposures. Therefore, the proposed 
changes do not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed 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 plant operation. The 
requirements in the TS will continue to assure operation of the plant 
within its design specifications and safety limits. Therefore, the 
changes do not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in the 
margin of safety?
    No. The proposed amendment would only change the analysis of record 
LOCA PCT, the allowed value of an instrument function, and its 
associated SR frequency. There will be no modification of any TS 
limiting condition for operation, no change to any limit on previously 
analyzed accidents, no change to how previously analyzed accidents or 
transients would be mitigated, no change in any methodology used to 
evaluate consequences of accidents, and no change in any operating 
procedure or process. Therefore, the proposed amendment does not entail 
a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis, and based on 
its own 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: Clifford G. Munson.

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

    Date of amendment request: October 5, 2007.
    Description of amendment request: The proposed amendment requests a 
change to Technical Specification 3.7(1)ci, ``Emergency Power Periodic 
Test,'' related to the surveillance testing of the Fort Calhoun Station 
emergency diesel generators (DGs) to support a modification to the DG 
start circuitry.
    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 removal of the anticipatory (idle speed) diesel generator 
(DG) start signal on a reactor protective system (RPS) reactor trip 
does not adversely affect the design function of the DGs and thus is 
not an initiator of any previously evaluated accidents.
    No Updated Safety Analysis Report (USAR) accident analyses take 
credit for the anticipatory (idle speed) DG start following a design 
basis accident (DBA). The DGs provide emergency power to their 
respective 4.16 KV [Kilovolt] buses and will continue to do so after 
the proposed modification is installed. Upon the occurrence of an 
undervoltage condition on the bus or an engineered safety features 
(ESF) signal, the modification provides a full speed DG start to 
achieve rated voltage and frequency. The safety function of the DGs 
is not altered by the installation of the modification. The 
associated Technical Specification (TS) change allows surveillance 
testing to reflect the way that the DGs start and load onto their 
respective buses following the modification.
    Deletion of a footnote containing historical information 
pertaining to a one-time surveillance interval extension and the 
punctuation correction are administrative changes. These 
administrative changes do not increase 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 removal of the anticipatory (idle speed) diesel generator 
(DG) start signal on an RPS reactor trip does not adversely affect 
the design function of the DGs and thus does not create the 
possibility of a new or different kind of accident. There are no 
USAR accident analyses which take credit for the anticipatory (idle 
speed) DG start following a DBA. The DGs provide emergency power to 
their respective 4.16 KV buses and will continue to do so after the 
proposed modification is installed. Upon the occurrence of an 
undervoltage condition on the bus or an ESF signal, the modification 
provides a full speed DG start to achieve rated voltage and 
frequency. The safety function of the DGs is not altered by the 
installation of this modification. The associated TS change allows 
surveillance testing to reflect the way that the DGs start and load 
onto their respective buses following the modification.
    Deletion of a footnote containing historical information 
pertaining to a one-time surveillance interval extension and the 
punctuation correction are administrative changes that do not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    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 a margin of safety?
    Response: No.
    The removal of the anticipatory (idle speed) diesel generator 
(DG) start signal on an RPS reactor trip does not adversely affect 
the design function of the DGs and thus does not involve a 
significant reduction in a margin of safety. There are no USAR 
accident analyses which take credit for the anticipatory (idle 
speed) DG start following a DBA. The DGs provide emergency power to 
their respective 4.16 KV buses and will continue to do so after 
installation of the proposed modification. Upon the occurrence of an 
undervoltage condition on the bus or an ESF signal, the modification 
provides a full speed DG start to achieve rated voltage and 
frequency. The safety function of the DGs is not altered by the 
installation of this modification. The associated TS change allows 
surveillance testing to reflect the way that the DGs will start and 
load onto their respective buses following the modification.
    Deletion of a footnote containing historical information 
pertaining to a one-time surveillance interval extension and the

[[Page 65370]]

punctuation correction are administrative changes that do not reduce 
a margin of safety.
    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.
    Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station, Unit No. 1, Washington County, Nebraska
    Date of amendment request: October 12, 2007.
    Description of amendment request: The proposed amendment would 
modify the Fort Calhoun Station, Unit 1 design and licensing basis to 
increase the shutdown cooling (SDC) system entry temperature from 300 
[deg]F to 350 [deg]F (cold leg), and the SDC entry pressure from 250 
psia to 300 psia (indicated at the pressurizer). Additionally, the 
licensee proposes to change to the Updated Safety Analysis Report 
(USAR) described design methodology applied to the SDC heat exchangers.
    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 shutdown cooling (SDC) system provides flow to the reactor 
during long term cooling mode following a large break loss-of-
coolant accident (LOCA). In addition, the SDC system can supply 
cooled sump water to the high pressure safety injection (HPSI) pumps 
for long term core cooling. The SDC system is also designed to 
reduce the temperature of the reactor coolant system (RCS) from 300 
[deg]F to refueling temperature within 24 hours and to maintain the 
proper RCS temperature during refueling. As such, the SDC system is 
not an initiator for any accident previously evaluated.
    The proposed change to increase the SDC entry temperature from 
300 [deg]F to 350 [deg]F affects the inputs to the analysis of the 
Boron Dilution Incident.
    However, re-analysis of this accident with the increased 
temperature does not result in an increase in the probability of the 
accident. The proposed increase in SDC system design and operating 
temperature and pressure has been evaluated for affects on system 
piping and components using appropriate codes and standards. The 
proposed changes do not introduce any failure mechanisms that would 
initiate a previously analyzed accident. Therefore, the proposed 
change to uprate the SDC system entry conditions does not result in 
a significant increase in the probability of a previously evaluated 
accident.
    The potential effect of the proposed change on the consequences 
of a previously evaluated accident has been considered. Re-analysis 
of the Boron Dilution Incident with the proposed increased SDC entry 
temperature does not result in an increase in the consequences of 
the accident.
    In addition, although an increase in the SDC system leakage test 
pressure is proposed, the leakage test acceptance criteria (i.e., 
maximum permitted leakage per hour) will not be affected. Therefore, 
the limit on post-accident leakage to atmosphere from the SDC system 
is unchanged. The proposed increase in SDC system design and 
operating temperature and pressure does not affect the redundancy or 
availability of the SDC system. The design functions of the system 
are not affected by the proposed change. Therefore, the SDC system 
will still be capable of performing the safety functions needed to 
mitigate the consequences of an 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change alters the SDC system entry conditions and 
increases the system leakage test pressure. In the current design, 
the SDC system has been excluded from consideration as a pipe 
rupture initiator since it is not normally in operation. It is used 
for plant shutdown and startup, and for accident mitigation. With 
the proposed change, the operating modes of the system will not be 
affected. The proposed change increases the RCS temperature and 
pressure at which the SDC system can be placed in service during 
shutdown (or removed from service during startup), but the RCS, SDC, 
and other plant systems are not operated in a different manner. The 
increased heat load on the component cooling water (CCW) system 
resulting from normal operation of the SDC at increased SDC 
temperatures has been evaluated. The increased normal operating heat 
load has been determined to be bounded by the post-accident CCW heat 
load. Any adjustments to the cooldown rate needed to accommodate the 
increased SDC entry temperature will be performed using approved 
procedures consistent with current practice and would not require 
operating the plant in a different manner.
    The RCS cooldown rate limitations in the Technical 
Specifications (TS) are not affected by the proposed change. In 
addition, adjustments of CCW heat loads to maintain required CCW 
inlet temperatures for the SDC (Low Pressure Safety Injection 
(LPSI)) pump coolers, when operating at the increased SDC entry 
temperature, will be in accordance with plant procedures and within 
existing system capabilities. The low temperature overpressurization 
(LTOP) analysis has been revised for the proposed change. However, 
there are no effects on existing LTOP setpoints or operating 
limitations, other than the proposed change to TS 2.1.1(11)(b), 
which states that the unit cannot be placed on shutdown cooling 
until the RCS has been cooled to <= 350 [deg]F. The proposed change 
in SDC operating limitations does not introduce the possibility of 
new or different equipment malfunctions or accident precursors.
    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 margins of safety are established through design parameters, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed change increases the SDC system entry 
conditions for plant shutdown, startup and following postulated 
accidents, and the SDC system leakage test pressure. However, the 
accident mitigation function and post-accident operation of the 
system is not affected. The operating limits on temperature and 
pressure will remain below the design temperature and pressure for 
the system. The time interval for operator action after a postulated 
boron dilution event with the SDC system in operation is reduced, 
however, the available time remains greater than the minimum 
required time interval of 15 minutes. The proposed change does not 
affect any design or operating parameter or setpoint used in the 
accident analyses to establish the margin of safety.
    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.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of amendment requests: October 15, 2007.
    Description of amendment requests: The proposed amendments would 
relocate all periodic surveillance frequencies from the technical 
specifications (TS) and place the frequencies under licensee control in

[[Page 65371]]

accordance with a new program, the Surveillance Frequency Control 
Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change involves the relocation of various 
surveillance test intervals from TSs to a licensee-controlled 
program and is administrative in nature. The proposed change does 
not involve the modification of any plant equipment or affect basic 
plant operation. The proposed change will have no impact on any 
safety related structures, systems or components. Surveillance test 
intervals are not assumed to be an initiator of any analyzed event, 
nor are they assumed in the mitigation of consequences of accidents. 
The [Surveillance Requirements] themselves will be maintained in the 
TS along with the applicable Limiting Conditions for Operation 
(LCOs) and Action statements. The surveillances performed at the 
intervals specified in the licensee-controlled program will assure 
that the affected system or component function is maintained, that 
the facility operation is within the Safety Limits, and that the 
LCOs are met.
    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 does not involve any physical alteration of 
plant equipment and does not change the method by which any safety-
related structure, system, or component performs its function or is 
tested. As such, no new or different types of equipment will be 
installed, and the basic operation of installed equipment is 
unchanged. The methods governing plant operation and testing remain 
consistent with current safety analysis assumptions.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change is administrative in nature, does not negate 
any existing requirement, and does not adversely affect existing 
plant safety margins or the reliability of the equipment assumed to 
operate in the safety analysis. As such, there are no changes being 
made to safety analysis assumptions, safety limits or safety system 
settings that would adversely affect plant safety as a result of the 
proposed change. Margins of safety are unaffected by relocation of 
the surveillance test intervals to a licensee-controlled program.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: Thomas G. Hiltz.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendments request: October 17, 2007.
    Description of amendments request: The proposed amendment would 
modify the Technical Specifications (TS) to establish more effective 
and appropriate action, surveillance, and administrative requirements 
related to the inoperability of snubbers in accordance with Nuclear 
Regulatory Commission (NRC)-approved TS Task Force (TSTF) change 
traveler TSTF-372-A, Revision 4. Specifically, the proposed amendment 
would add Limiting Condition for Operation (LCO) 3.0.8. The NRC staff 
issued a ``Notice of Opportunity To Comment on Model Safety Evaluation 
on Technical Specification Improvement To Modify Requirements Regarding 
the Addition of LCO 3.0.8 on the Inoperability of Snubbers Using the 
Consolidated Line Item Improvement Process'' in the Federal Register on 
November 24, 2004 (69 FR 68412). The notice included a model safety 
evaluation (SE) and a model no-significant-hazards-consideration (NSHC) 
determination. The NRC staff issued a ``Notice of Availability of Model 
Application Concerning Technical Specification Improvement To Modify 
Requirements Regarding the Addition of Limiting Condition for Operation 
3.0.8 on the Inoperability of Snubbers Using the Consolidated Line Item 
Improvement Process'' in the Federal Register on May 4, 2005 (70 FR 
23252). The notice included a model application, including a revised 
model SE. In its application dated October 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 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 allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an inoperable snubber if risk is assessed and managed. The 
postulated seismic event requiring snubbers is a low probability 
occurrence and the overall TS system safety function would still be 
available for the vast majority of anticipated challenges. 
Therefore, the probability of an accident previously evaluated is 
not significantly increased, if at all. The consequences of an 
accident while relying on allowance provided by proposed LCO 3.0.8 
are no different than the consequences of an accident while relying 
on the TS required actions in effect without the allowance provided 
by proposed LCO 3.0.8. Therefore, the consequences of an accident 
previously evaluated are not significantly affected by this change. 
The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
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 does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is 
assessed and managed, 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 addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, 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 allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low probability occurrence and the 
overall TS system safety function would still be available for the 
vast majority of anticipated challenges. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in RG 1.177. A bounding risk assessment was performed to 
justify the proposed TS changes. This application of LCO 3.0.8 is 
predicated upon the licensee's assessment

[[Page 65372]]

and management of plant risk. 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 adopted by the licensee 
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 amendments request involves NSHC.
    Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Branch Chief: Mark G. Kowal.

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 18, 2007.
    Description of amendment request: The proposed amendments to 
Technical Specification Administrative Controls Section 5.3.1 would 
revise the training and qualifying education and experience eligibility 
requirements for certain unit staff positions to correspond to a 
defined training program. The training program is based on National 
Academy for Nuclear Training guidance documents (ACADs) as described in 
the licensee's October 18, 2007, application. The proposed changes will 
also replace a specific position title with a generic position title 
for the senior individual in charge of Health Physics. An application 
that addressed similar issues was previously submitted on October 30, 
2006, and notice of that application was provided in the Federal 
Register on July 17, 2007 (72 FR 39084). Due to certain changes in the 
specifics of the October 18, 2007, application, from those proposed in 
the October 30, 2006, application, the application is being renoticed 
in its entirety. This notice supersedes the notice published in the 
Federal Register on July 17, 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:

    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 Technical Specifications Administrative 
Controls Section 5.3.1 involves the use of a more generic 
designation for the unit staff position responsible for Health 
Physics without reducing the level of authority required for that 
position. The proposed change also allows the flexibility to use an 
accredited program for qualifying personnel to fill certain unit 
staff positions as stipulated in Enclosure 1 [of October 18, 2007, 
application], which represents an acceptable alternative to the 
qualification requirements for these positions as currently 
specified in the Technical Specifications. Since the proposed 
changes are administrative in nature, they do not involve any 
physical changes to any structures, systems, or components, nor will 
their performance requirements be altered. The proposed changes also 
do not affect the operation, maintenance, or testing of the plant. 
Therefore, the response of the plant to previously analyzed 
accidents will not be affected. Consequently, the proposed changes 
do 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 previously evaluated?
    Response: No.
    The proposed changes to the Technical Specifications will have 
no adverse impact on the overall qualification of the unit staff. 
The use of a more generic designation for the unit staff position 
responsible for Health Physics and the proposed addition [of] a 
statement to Section 5.3.1 that will reference this letter and the 
accreditation information for the positions stipulated in Enclosure 
1 will allow the use of an accredited program that has been endorsed 
by the NRC and will ensure the educational requirements and power 
plant experience for each unit staff position are properly satisfied 
and will continue to fulfill applicable regulatory requirements. 
Also, since no change is being made to the design, operation, 
maintenance, or testing of the plant, no new methods of operation or 
failure modes are introduced by the proposed changes. Therefore, the 
possibility of a new or different kind of accident from any 
previously evaluated is not created.
    3. Does the proposed change involve a significant decrease in 
the margin of safety?
    Response: No
    The proposed changes to the Technical Specifications will have 
no adverse impact on the onsite organizational features necessary to 
assure safe operation of the plant. Lines of authority for plant 
operation are unaffected by the proposed changes. Also, the adoption 
of the more generic designation of the individual responsible for 
Health Physics will reduce the regulatory burden of having to devote 
limited resources to process a license amendment whenever a title 
change for this position is implemented. Accordingly, this reduction 
in regulatory burden and the proposed addition of a statement to 
Section 5.3.1 that will reference this letter and the use of 
accreditation information provided in Enclosure 1, will allow the 
use of an accredited program endorsed by NRC to qualify certain unit 
staff positions and will improve organizational flexibility without 
compromising plant safety. Therefore, the proposed changes do not 
involve a significant decrease 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.

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

    Date of amendment request: August 28, 2007, as supplemented on 
October 9, 2007.
    Description of amendment request: The proposed amendments would 
revise the ``Maximum Power Level'' in paragraph 2.C(1) of the Vogtle 
Electric Generating Plant Facility Operating Licenses NPF-68 and NPF-81 
for Unit 1 and Unit 2, respectively. In addition, the amendments would 
revise the definition of ``Rated Thermal Power (RTP)'' in Technical 
Specification 1.1 for both units to reflect the change to the Maximum 
Power Level. The proposed change increases the RTP from 3565 MWt to 
3625.6 MWt, resulting in an increase of 1.7% from the current reactor 
output. This increase in reactor core power level is referred to as a 
Measurement Uncertainty Recapture (MUR) power uprate.
    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?

Operating License--Maximum Power Level and Technical Specification 
1.1--Definition of Rated Thermal Power

    The increase in Maximum Power Level and Rated Thermal Power 
(RTP) does not involve a significant increase in the probability or 
consequences of an accident previously evaluated, because operation 
at the higher power level will not cause any design or analysis 
acceptance criteria to be exceeded. As a result, structural and 
functional

[[Page 65373]]

integrity of the plant systems is maintained. Power level is an 
input assumption to the equipment design and accident analyses, but 
it is not itself an initiator for any transient. Therefore, the 
probability of occurrence of an accident previously evaluated is not 
affected.
    The radiological consequences of operation at the Measurement 
Uncertainty Recapture (MUR) power uprate conditions have been 
assessed. It was concluded that offsite dose predictions remain 
within the acceptance criteria for each of the accidents affected. 
Therefore, the consequences of an accident previously evaluated are 
not increased.

Technical Specification 1.1--Definition of Dose Equivalent Iodine

    The proposed change to the definition of dose equivalent iodine 
(DEI) impacts the reactor coolant activity surveillance and 
calculations of accident consequences and makes these activities 
consistent with each other. Neither of these functions affects the 
probability of any accident previously evaluated.
    In order to support the MUR power uprate, the accidents 
previously evaluated in the Updated Final Safety Analysis Report 
(UFSAR) were re-analyzed. As part of this reanalysis, the dose 
conversion factors (DCFs) were reviewed, and a consistent set of 
DCFs was used for all re-analyses based on Federal Guidance Report 
No. 11, as suggested by RIS 2001-19. The results of these re-
analyses continue to meet the acceptance limits as currently 
described in the UFSAR.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

Technical Specification 3.3.1, Table 3.3.1-1, Function 16--P-9 Setpoint

    The revised Power Range Neutron Flux P-9 permissive nominal 
setpoint and allowable value do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated, because operation with these revised values will not 
cause any design or analysis acceptance criteria to be exceeded. The 
structural and functional integrity of any plant system is 
unaffected. The P-9 permissive function is part of the transient 
mitigation response and is not itself an initiator for any 
transient. Therefore, the probability of occurrence of an accident 
previously evaluated is not affected.
    The changes to the P-9 nominal setpoint and allowable value do 
not affect the integrity of the fission product barriers utilized 
for the mitigation of radiological dose consequences as a result of 
an accident. The change continues to ensure that the pressurizer 
power operated relief valves (PORVs) are not challenged following a 
turbine trip without a reactor trip which, in turn, minimizes the 
potential for a release. There are no offsite dose predictions for 
this transient. Since it has been determined that the transient 
results are unaffected by the change to the P-9 nominal setpoint and 
allowable value, it is concluded that the consequences of an 
accident previously evaluated are not increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?

Operating License--Maximum Power Level and Technical Specification 
1.1--Definition of Rated Thermal Power

    The increase in Maximum Power Level and RTP does not create the 
possibility of a new or different kind of accident from any 
previously evaluated, because no new operating configuration is 
being imposed that will create a new failure scenario, and no new 
failure modes are being created for any plant equipment. System and 
component design bases have been reviewed. The proposed change does 
not have an adverse effect on safety-related systems or components 
and does not challenge the integrity of any safety-related system. 
Therefore, the types of accidents defined in the UFSAR continue to 
represent the credible spectrum of events to determine safe plant 
operation.

Technical Specification 1.1--Definition of Dose Equivalent Iodine

    The proposed change to the definition of Dose Equivalent Iodine 
(DEI) ensures the reactor coolant activity surveillances are 
consistent with the assumptions for initial conditions used in the 
accident analyses. The proposed change does not involve the addition 
or modification of any plant equipment. Neither does it alter the 
design, configuration or method of operation of the plant.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

Technical Specification 3.3.1, Table 3.3.1-1, Function 16--P-9 Setpoint

    The revised Power Range Neutron Flux P-9 permissive nominal 
setpoint and allowable value do not create the possibility of a new 
or different kind of accident from any previously evaluated, because 
these changes do not affect accident initiation sequences. No new 
operating configuration is being imposed by the P-9 nominal setpoint 
and allowable value changes that will create a new failure scenario. 
In addition, no new failure modes are being created for any plant 
equipment. Therefore, the types of accidents defined in the UFSAR 
continue to represent the credible spectrum of events to determine 
safe plant operation.
    3. Does the proposed change involve a significant decrease in a 
margin of safety?

Operating License--Maximum Power Level and Technical Specification 
1.1--Definition of Rated Thermal Power

    The increase in Maximum Power Level and RTP does not involve a 
significant reduction in a margin of safety, because power level is 
one of the inherent assumptions that determine the safe operating 
range defined by the accident analyses, which are in turn protected 
by the Technical Specifications. The acceptance criteria for the 
accident analyses are conservative with respect to the operating 
conditions defined by the Technical Specifications. The engineering 
reviews performed for the MUR power uprate confirmed that the 
accident analyses criteria are met at the revised value of MPL and 
RTP. Therefore, the adequacy of the revised Facility Operating 
Licenses and Technical Specifications to maintain the plant in a 
safe operating range is also confirmed, and the increase in MPL and 
RTP do not involve a significant decrease in a margin of safety.

Technical Specification 1.1--Definition of Dose Equivalent Iodine

    The proposed change to the definition of dose equivalent iodine 
(DEI) has the potential to affect the dose consequences offsite and 
in the control room. However, the results of the re-analyses of the 
accidents previously evaluated demonstrate the dose consequences at 
all locations remain within the regulatory acceptance limits, and 
the margin of safety as defined by 10 CFR 100 and GDC 19 has not 
been significantly reduced.

Technical Specification 3.3.1, Table 3.3.1-1, Function 16--P-9 Setpoint

    The change to the P-9 nominal setpoint and allowable value does 
not involve a significant reduction in a margin of safety because 
the margin of safety associated with the P-9 setpoint, as verified 
by the results of the applicable transient analyses, is within 
acceptable limits. The adequacy of the revised Technical 
Specification values to maintain the plant in a safe operating range 
has been confirmed. Therefore, the change to the P-9 nominal 
setpoint and allowable value does not involve a significant decrease 
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. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Branch Chief: Evangelos C. Marinos.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: August 27, 2007.
    Description of amendment request: The amendments would revise the 
licensee's fire protection program requirements as documented in the 
licensee's Fire Hazard's Analysis Report. Specifically, the licensee 
requests the use of reactor operator manual actions in lieu of meeting 
protection requirements of circuit separation.
    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:


[[Page 65374]]


    1. [Do] the proposed amendment[s] involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No.
    The design function of structures, systems and component[s] are 
not Impacted by the proposed change. The proposed change involves 
operator manual actions in response to a fire and will not initiate 
an event. The proposed actions do not increase the probability of 
occurrence of a fire or any other accident previously evaluated.
    The proposed actions are feasible and reliable and demonstrate 
that the unit can be safely shutdown in the event of a fire. No 
significant consequences result from the performance of the proposed 
actions.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. [Do] the proposed amendment[s] create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The design function of structures, systems and component[s] are 
not impacted by the proposed amendment[s]. The proposed change 
involves operator manual actions in response to a fire. [It does 
not] involve new failure mechanisms or malfunctions that can 
initiate 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. [Do] the proposed amendment[s] involve a significant 
reduction in a margin of safety?
    Response: No.
    Adequate time is available to perform the proposed operator 
manual actions to account for uncertainties in estimates of the time 
available and in estimates of how long it takes to diagnose and 
execute the actions. The actions are straightforward and do not 
create any significant concerns. The actions have been verified that 
they can be performed through demonstration and they are 
proceduralized. The proposed actions are feasible and reliable and 
demonstrate that the unit can be safely shutdown in the event of a 
fire.
    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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Branch Chief: Thomas G. Hiltz.

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

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

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

    Date of amendment request: October 22, 2007.
    Brief description of amendment request: The proposed amendment 
would allow an alternate methodology from that previously approved in 
Topical Report DOM-NAF-3-0.0-P-A, GOTHIC Methodology for Analyzing the 
Response to Postulated Pipe Ruptures Inside Containment, as discussed 
in the Surry Power Station, Unit Nos. 1 and 2, Updated Final Safety 
Analysis Report.
    Date of publication of individual notice in Federal Register: 
October 30, 2007 (72 FR 61406).
    Expiration date of individual notice: Public comment period 
expiration date, November 13, 2007; Hearing period expiration date, 
January 31, 2008.

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, 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].

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

    Date of application for amendment: December 15, 2006.
    Brief description of amendment: The amendment incorporates changes 
to the technical specifications (TSs) associated with previously-
approved industry initiatives. The first change relocates the actions 
for a safety limit violation from the administrative controls TS 
section to the safety limit TS section and deletes notification 
requirements, as approved by TS Task Force (TSTF) Change Traveler TSTF-
05-A, ``Deletion of Safety Limit Violation Notification Requirements.'' 
The second change incorporates generic position titles, as approved by 
TSTF-65-A, ``Use of Generic Titles for Utility Positions,'' and 
incorporates items approved by Nuclear Regulatory Commission 
Administrative Letter 95-06, ``Relocation of Technical

[[Page 65375]]

Specification Administrative Controls Related to Quality Assurance.''
    Date of issuance: October 31, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 193.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: March 13, 2007 (72 FR 
11386) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 31, 2007.
    No significant hazards consideration comments received: No.

Duke Power Company LLC, et al., Docket No. 50-413, Catawba Nuclear 
Station, Unit 1, York County, South Carolina

    Date of application for amendments: November 22, 2006.
    Brief description of amendments: The amendment revises the Catawba 
Unit 1 Facility Operating License (FOL) to add a license condition 
requiring a specific date by which the modifications to the Emergency 
Core Cooling Systems (ECCS) sump in response to 2004 Generic Letter 
(GL) 2004-02, ``Potential Impact of Debris Blockage on Emergency 
Recirculation During Design Basis Accidents at Pressurized Water 
Reactors.'' The changes add a license condition which requires that (1) 
Catawba Nuclear Station, Unit 1 will enter Mode 5 for the outage to 
install the sump strainer modification no later than May 19, 2008, and 
that (2) the Unit 1 sump strainer modification will be completed prior 
to entry into Mode 4 after May 19, 2008.
    Date of issuance: October 31, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 237.
    Facility Operating License Nos. NPF-35: Amendment revises the 
license.
    Date of initial notice in Federal Register: March 13, 2007 (72 FR 
11386)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 31, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 4, 2007.
    Brief description of amendment: The amendment revised Technical 
Specifications (TSs) for the Limiting Conditions for Operation and 
Surveillance Requirements for Control Rod Operability, Scram Insertion 
Times, and Control Rod Accumulators.
    Date of issuance: November 5, 2007.
    Effective date: As of the date of issuance, and shall be 
implemented within 120 days.
    Amendment No.: 230.
    Facility Operating License No. DPR-35: The amendment revised the 
License and TSs.
    Date of initial notice in Federal Register: April 24, 2007 (72 FR 
20381).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 5, 2007.
    No significant hazards consideration comments received: No.

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

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit No. 1, DeWitt County, Illinois

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

Exelon Generation Company, LLC, Docket No. 50-352 and No. 50-353, 
Limerick Generating Station, Unit 1 and 2, Montgomery County, 
Pennsylvania

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, Pennsylvania

    Date of application for amendments: April 12, 2007.
    Brief description of amendments: The amendments modify technical 
specification (TS) requirements related to control room envelope 
habitability in accordance with TS Task Force (TSTF) Traveler TSTF-448, 
Revision 2, ``Control Room Habitability.''
    Date of issuance: October 31, 2007.
    Effective date: As of the date of issuance, to be implemented 
within 180 days.
    Amendment Nos.: 150, 150, 145, 145, 178, 186, 173, 188, 149, 264, 
and 268.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72, NPF-77, 
NPF-62, NPF-11, NPF-18, NPF-39, NPF-85, DPR-44, and DPR-56: The 
amendments revised the Technical Specifications and the Operating 
Licenses.
    Date of initial notice in Federal Register: June 5, 2007 (72 FR 
31100).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 31, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 10, 2007.
    Brief description of amendment: The amendments revise the value of 
the safety limit minimum critical power ratio for the Dresden Nuclear 
Power Station (DNPS), Unit 2 technical specifications (TSs). The 
amendment also made conforming changes that clarify the wording of the 
DNPS, Unit 3 TSs.
    Date of issuance: November 6, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 224/216.
    Renewed Facility Operating License Nos. DPR-19 and DPR-25: The 
amendments revised the Technical Specifications and License.
    Date of initial notice in Federal Register: July 31, 2007 (72 FR 
41783), and September 5, 2007 (72 FR 50986).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 6, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 7, 2007, as 
supplemented by letter dated January 24, 2007.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) Surveillance Requirement (SR) 3.4.3.1 to increase 
the allowable as-found main steam safety valve lift setpoint tolerance 
from 1 percent to 3 percent. In addition, the 
amendments revise TS SR 3.1.7.10 to increase the enrichment of sodium 
pentaborate used in the standby liquid control system from >=30.0 atom 
percent boron-10 to >=45.0 atom percent boron-10.
    Date of issuance: November 1, 2007.
    Effective date: As of the date of issuance and shall be implemented 
prior to main steam safety valve testing during the next refueling 
outage currently scheduled for May 2009 for Unit 1 and May 2008 for 
Unit 2.

[[Page 65376]]

    Amendment Nos.: 235/230.
    Renewed Facility Operating License Nos. DPR-29 and DPR-30: The 
amendments revised the Technical Specifications and License.
    Date of initial notice in Federal Register: January 30, 2007 (72 FR 
4307) The January 24, 2007, supplement contained clarifying information 
and did not change the NRC staff(s initial proposed finding of no 
significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 1, 2007.
    No significant hazards consideration comments received: No.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida.

    Date of application for amendment: October 11, 2006.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.7.7, ``Nuclear Services Closed Cycle Cooling Water 
(SW) System,'' to reduce the allowed outage time when one of the 
required SW heat exchangers is out of service.
    Date of issuance: October 23, 2007.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment No.: 225.
    Facility Operating License No. DPR-72: Amendment revised the TSs.
    Date of initial notice in Federal Register: February 13, 2007 (72 
FR 6783).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 23, 2007.
    No significant hazards consideration comments received: No.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: February 8, 2007, as 
supplemented by letter dated August 23, 2007.
    Brief description of amendment: The amendment changes the basis for 
protection of the spent fuel stored in the spent fuel pool (SFP) in 
order to eliminate the Final Safety Analysis Report commitment for 
maintaining the SFP missile shields.
    Date of issuance: October 24, 2007.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment No.: 226.
    Facility Operating License No. DPR-72: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 13, 2007 (72 FR 
11381). The supplement dated August 23, 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.
    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.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: October 5, 2006, as supplemented 
by letters dated April 4 and July 19, 2007.
    Brief description of amendment: The amendment changes the 
restrictions on fuel storage in the spent fuel pool.
    Date of issuance: October 25, 2007.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment No.: 227.
    Facility Operating License No. DPR-72: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 21, 2006 (71 
FR 67394). The supplements dated April 4 and July 19, 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.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 25, 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: April 22, 2007.
    Brief description of amendments: Amendments delete Section 3.H of 
Facility Operating License Nos. DPR-67 and NPF-16, which require 
reporting of violations of the requirements of Sections 3.A, 3.D, 3.F 
and 3.G of the operating license.
    Date of Issuance: October 31, 2007.
    Effective Date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 203 and 150.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the operating license conditions and Technical 
Specifications.
    Date of initial notice in Federal Register: June 19, 2007 (72 FR 
33783).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 31, 2007.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of application for amendments: May 4, 2007.
    Brief description of amendments: The proposed amendment would 
incorporate the administrative changes to Technical Specification (TS) 
6.2.1.a, ``On and Offsite Organization'' and 6.8.1.a, ``Procedures and 
Programs.''
    Date of issuance: November 2, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos: 236 and 231.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the TSs.
    Date of initial notice in Federal Register: July 3, 2007 (72 FR 
36522).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 2, 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, New York

    Date of application for amendment: July 23, 2007.
    Brief description of amendment: The amendment modifies Technical 
Specification 3.3.2.1, ``Control Rod Block Instrumentation,'' to allow 
a new banked position withdrawal sequence for shutdown, using the 
Consolidated Line Item Improvement Process.
    Date of issuance: October 26, 2007.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 120.
    Renewed Facility Operating License No. NPF-69: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: September 25, 2007 (72 
FR 54477).
    The Commission's related evaluation of the amendment is contained 
in a

[[Page 65377]]

Safety Evaluation dated October 26, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 10, 2007.
    Brief description of amendments: The requested changes are a 
partial adoption of Technical Specification Task Force (TSTF)-491, 
Revision 2, ``Removal of Main Steam and Feedwater Valve Isolation 
Times'' which was proposed by the TSTF by letter on May 18, 2006. The 
proposed changes revise Technical Specification (TS) 3.7.2 ``Main Steam 
Valves Closure Times'' by relocating the isolation valve closure times 
to a licensee-controlled document identified as a Bases reference. The 
proposed amendments deviate from TSTF-491 in that the current PINGP TS 
(3.7.3) and associated surveillance requirements for the main feedwater 
isolation valves do not include valve closure times, and thus, the 
changes to TS 3.7.3 provided for in TSTF-491 are not applicable to the 
PINGP TSs and are not adopted. TSTF change traveler TSTF-491, Revision 
2, was announced for availability in the Federal Register on December 
29, 2006, as part of the consolidated line item improvement process.
    Date of issuance: October 31, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 181 and 171.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 17, 2007 (72 FR 
39083).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 31, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 17, 2007.
    Brief description of amendments: The amendment modified Technical 
Specifications requirements related to control room envelope 
habitability in accordance with Technical Specifications Task Force 
448, Revision 3, using the Consolidated Line Item Improvement Process.
    Date of issuance: October 31, 2007.
    Effective date: as of its date of issuance, to be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 2-214; Unit 3-206.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: May 22, 2007 (72 FR 
28722). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 31, 2007.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

    Date of application for amendment: May 21, 2007, as supplemented by 
letter dated June 11, 2007.
    Brief description of amendment: The amendment modified the 
technical specification (TS) requirements for inoperable snubbers by 
adding Limited Condition for Operation 3.0.8, using the Consolidated 
Line Item Improvement Process. The change is based on TS Task Force 
(TSTF) TSTF-372, Revision 4.
    Date of issuance: October 17, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 251, 231.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
change the licenses and the technical specifications.
    Date of initial notice in Federal Register: June 19, 2007 (72 FR 
33785) The supplement dated July 11, 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. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated October 17, 2007.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

    Date of application for amendment: May 29, 2007.
    Brief description of amendment: The amendments modify the Technical 
Specification requirements related to control room habitability, using 
the Technical Specification Task Force traveler, TSTF-448, revision 3.
    Date of issuance: October 31, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days from the date of issuance.
    Amendment Nos.: 252, 232.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
change the licenses and the technical specifications.
    Date of initial notice in Federal Register: July 3, 2007 (72 FR 
36523).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 31, 2007.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 8th day of November 2007.

    For The Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E7-22331 Filed 11-19-07; 8:45 am]
BILLING CODE 7590-01-P