Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 68451-68460 [E8-27110]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 73, Number 223 (Tuesday, November 18, 2008)]
[Notices]
[Pages 68451-68460]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-27110]



[[Page 68451]]

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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 23, 2008, to November 5, 2008. The 
last biweekly notice was published on November 4, 2008 (73 FR 65685).

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.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example, in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area O1F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, 
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, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request 
for a hearing or petition for leave to intervene is filed within 60 
days, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one

[[Page 68452]]

contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for hearing or a petition for leave to intervene must be 
filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve documents over the internet 
or in some cases to mail copies on electronic storage media. 
Participants may not submit paper copies of their filings unless they 
seek a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
hearingdocket@nrc.gov, or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms 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 https://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying for a 
digital ID certificate is available on NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/site-help/e-submittals.html or by calling the NRC technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, in accordance 
with 10 CFR 2.302(g), with their initial paper filing requesting 
authorization to continue to submit documents in paper format. Such 
filings must be submitted by: (1) First class mail addressed to the 
Office of the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff; or (2) courier, express mail, or expedited 
delivery service to the Office of the Secretary, Sixteenth Floor, One 
White Flint North, 11555 Rockville, Pike, Rockville, Maryland 20852, 
Attention: Rulemaking and Adjudications Staff. Participants filing a 
document in this manner are responsible for serving the document on all 
other participants. Filing is considered complete by first-class mail 
as of the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the petition 
and/or request should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later 
than 11:59 p.m. Eastern Time on the due date.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a Presiding Officer. Participants are requested not to include personal 
privacy information, such as social security numbers, home addresses, 
or home phone numbers in their filings. With respect to copyrighted 
works, except for limited excerpts that serve the purpose of the 
adjudicatory filings and would constitute a Fair Use application, 
participants are requested not to include copyrighted materials in 
their submission.
    For further details with respect to this amendment action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr.resource@nrc.gov.

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

    Date of amendment request: October 1, 2008.

[[Page 68453]]

    Description of amendment request: The amendment would modify 
Technical Specification (TS) 5.5.16, Containment Leakage Rate Testing 
Program, by adding exceptions to Regulatory Guide (RG) 1.163, 
``Performance-Based Containment Leak-Test Program,'' that would allow 
the next integrated leak rate test (ILRT) (Type A test) to be performed 
at a 15-year interval at Palo Verde Nuclear Generating Station (PVNGS), 
Units 1, 2, and 3. The proposed amendment is risk-informed and follows 
the guidance in RG 1.174, ``An Approach for Using Probabilistic Risk 
Assessment in Risk-Informed Decisions on Plant-Specific Changes to the 
Licensing Basis.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to extend the next ILRT interval from 10 to 
15 years one time does not involve a physical change to PVNGS[,] 
Units 1, 2, and 3, or a change in the manner in which the plant is 
operated or controlled. The containment vessel is designed to 
provide an essentially leak-tight barrier against the uncontrolled 
release of radioactivity to the environment for any postulated 
accidents. As such, the reactor containment itself and the testing 
guidelines invoked to periodically demonstrate the integrity of the 
containment exist to ensure the containment can mitigate the 
consequences of any accident and do not involve the prevention or 
identification of any precursors of any accidents. There is no 
design basis accident that is initiated by a failure of the 
containment leakage mitigation function. The extension of the ILRT 
will not create any adverse interactions with other systems that 
could result in initiation of a design basis accident. Therefore, 
the probability of occurrence of an accident previously evaluated is 
not significantly increased.
    Based on a completed probability risk assessment of the affects 
of this change to the ILRT interval there is a slight increase in 
risk dose. This increase in risk in terms of person-rem year within 
50 miles of the plant resulting from design basis accidents is 
significantly less than one percent and of a magnitude that NUREG-
1493 indicates is imperceptible. The risk assessment also analyzed 
the increase in risk in terms of the frequency of large early 
releases from accidents. The increase in the large early release 
frequency resulting from the proposed extension was determined to be 
within the guidelines published in Regulatory Guide 1.174. 
Additionally, the proposed change maintains defense-in-depth by 
preserving a reasonable balance among prevention of core damage, 
prevention of containment failure, and consequence mitigation. The 
increase in the conditional containment failure probability from 
reducing the ILRT frequency from one test per 10 years to one test 
per 15 years is less than one percent and considered insignificant. 
Continued containment integrity is assured by the history of 
successful ILRTs, and the established programs for local leakage 
rate testing and in-service inspections which are not affected by 
the proposed change. Therefore, the consequences of an accident 
previously analyzed are not significantly increased.
    In summary, the probability of occurrence and the consequences 
of an accident previously evaluated are not significantly increased.
    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 to extend the ILRT interval from 10 to 15 
years does not create any new or different accident initiators or 
precursors. The length of the ILRT interval does not affect the 
manner in which any accident begins. The proposed change does not 
physically change the plant, does not create any new failure modes 
for the containment and does not affect the interaction between the 
containment and any other system. Thus, 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 risk-based margins of safety associated with the containment 
ILRT are those associated with the estimated person-rem per year, 
the large early release frequency, and the conditional containment 
failure probability. The potential effect of the proposed change on 
the parameters have been quantified and it has been determined that 
the effect is considered insignificant. The non-risk-based margins 
of safety associated with the containment ILRT are those involved 
with its structural integrity and leak tightness. The proposed 
change to extend the ILRT interval from 10 to 15 years does not 
adversely affect either of these attributes. The proposed change 
only affects the frequency at which these attributes are verified. 
Therefore, the proposed change does not involve a significant 
reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Michael G. Green, Senior Regulatory Counsel, 
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, 
Phoenix, Arizona 85072-2034.
    NRC Branch Chief: Michael T. Markley.

Entergy Nuclear Operations, Inc., Docket No. 50-155, Big Rock Point 
Plant, Charleviox County, Michigan

    Date of amendment request: September 22, 2008.
    Description of amendment request: The proposed amendment would 
amend the facility operating license by changing the names of the 
licensees from Entergy Nuclear Operations, Inc., and Entergy Nuclear 
Palisades, LLC to EquaGen Nuclear LLC and Enexus Nuclear Palisades, 
LLC, respectively.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed amendment would only change the names of the 
licensees and reflect associated order requirements. The proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The proposed 
changes do not create the possibility of a new or different kind of 
accident from an accident previously evaluated. The proposed changes 
do not involve a significant reduction in margin of safety.

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

Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247, and 50-
286, Indian Point Nuclear Generating Unit Nos. 1, 2 and 3, Westchester 
County, New York

    Date of amendment request: September 30, 2008 (2 letters).
    Description of amendment request: This is an administrative change 
which would reflect the creation of new companies as approved by the 
NRC Order dated July 28, 2008. The amendments would not be implemented 
until the restructuring transactions have been completed. The 
amendments would revise the names on the plant licenses to match the 
names of the new companies. Entergy Nuclear Indian Point 2, LLC would 
be changed to Enexus Nuclear Indian Point 2, LLC. Entergy Nuclear 
Indian Point 3, LLC

[[Page 68454]]

would be changed to Enexus Nuclear Indian Point 3, LLC. Entergy Nuclear 
Operations, Inc. would be changed to EquaGen Nuclear LLC.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed amendment would only change the names of the 
licensees and reflect associated order requirements. The proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated. 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: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment request: September 22, 2008.
    Description of amendment request: The proposed amendment would 
amend the renewed facility operating license and Technical 
Specifications Design Features, Section 4, by changing the names of the 
licensees from Entergy Nuclear Operations, Inc. and Entergy Nuclear 
Palisades, LLC to EquaGen Nuclear LLC and Enexus Nuclear Palisades, 
LLC, respectively.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed amendment would only change the names of the 
licensees and reflect associated order requirements. The proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The proposed 
changes do not create the possibility of a new or different kind of 
accident from an accident previously evaluated. The proposed changes 
do not involve a significant reduction in margin of safety.

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

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: September 22, 2008.
    Description of amendment request: The proposed amendment would 
relocate the contents of the Vermont Yankee (VY) Technical 
Specification (TS) relating to the Reactor Building crane to the VY 
Technical Requirements Manual (TRM).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of Vermont Yankee Nuclear Power Station (VY) in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    This proposed change relocates the VY TS and associated Bases 
related to the Reactor Building crane to the VY TRM. The proposed 
amendment does not impact the operability of any structure, system 
or component that affects the probability of an accident or that 
supports mitigation of an accident previously evaluated. The 
proposed amendment does not affect reactor operations or accident 
analysis and has no radiological consequences. The operability 
requirements for accident mitigation systems remain consistent with 
the licensing and design basis. Therefore, the proposed amendment 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station (VY) in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    This proposed change relocates the VY TS and associated Bases 
related to the Reactor Building crane to the VY TRM. The proposed 
amendment does not change the design or function of any component or 
system. No new modes of failure or initiating events are being 
introduced. Therefore, operation of VY in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station (VY) in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    This proposed change relocates the VY TS and associated Bases 
related to the Reactor Building crane to the VY TRM. The proposed 
amendment does not change the design or function of any component or 
system. The proposed amendment does not involve any safety limits, 
safety settings or safety margins. The ability of the Reactor 
Building crane to perform its intended functions will continue to be 
required in accordance with the VY TRM.
    Since the proposed controls are adequate to ensure the 
operability of the Reactor Building crane, there will still be high 
assurance that the components are operable and capable of performing 
their respective functions. Therefore, operation of VY in accordance 
with the proposed amendment will not involve a significant reduction 
in [a] margin to safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: September 22, 2008.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) to change requirements related 
to Battery Systems specified in TS Section 3.10 resulting in removing 
the Limiting Condition for Operation pertaining to 345 kV switchyard 
batteries, chargers and associated direct current distribution panel.
    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 change does not impact the function of any Structure, System 
or Component (SSC)

[[Page 68455]]

that affects the probability of an accident or that supports 
mitigation or consequences of an accident previously evaluated. The 
proposed change removes unnecessary information from the Technical 
Specifications that is not required in accordance with 10 CFR 50.36. 
The proposed change does not affect any plant equipment operation or 
accident analysis and has no radiological consequences. 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 system performs their function. The proposed change removes 
unnecessary information from the Technical Specifications that is 
not required in accordance with 10 CFR 50.36. As such, no new or 
different types of equipment will be installed or removed from the 
facility. Operation of existing installed equipment is unchanged. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not change any existing design or operational 
requirements 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 operation as a result of the 
proposed change. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment request: June 3, 2008.
    Description of amendment request: The proposed amendment would 
revise the analysis methodology in the Final Safety Analysis Report, 
Section 5.4.3, ``Structural Design Criteria,'' and Section 5.4.5.3, 
``Missile Analysis.'' The amendment would allow the licensee to use the 
yield line theory methodology to qualify the east wall of the Auxiliary 
Building for tornado wind and missile loading.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed LAR [license amendment request] will revise the 
methodology used to qualify the east wall of the CR-3 [Crystal River 
Unit 3 Nuclear Generating Plant] Auxiliary Building for all expected 
and postulated loads including tornado wind and missile loading. The 
Yield Line Theory methodology is an industry standard that is used 
for the design and analysis of concrete slabs. The Yield Line Theory 
methodology is used for investigating the failure mechanisms of flat 
reinforced concrete slabs at the ultimate limit (failure point). In 
other words, this methodology determines either the moments in a 
slab at the point of failure or the load at which the slab will 
fail. A change in the methodology of an analysis used to verify 
qualification of an existing structure will not have any impact on 
the probability of accidents previously evaluated.
    The analysis performed demonstrates that the CR-3 Auxiliary 
Building east wall will remain structurally intact following the 
worst case loadings assumed in the calculation. Therefore, this 
proposed change does not involve a significant increase in the 
probability or consequences previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The function of the CR-3 Auxiliary Building wall is to house and 
protect the equipment that is important to safety from damage during 
normal operation, transients and design basis accidents. The use of 
the Yield Line Theory methodology for qualifying the east wall of 
the CR-3 Auxiliary Building has no impact on the capability of the 
structure. A calculation that uses the Yield Line Theory methodology 
demonstrated that the structure meets required design criteria. This 
ensures that the wall is capable of performing its design function 
without alteration or compensatory actions of any kind. No changes 
to any plant system, structure, or component (SSC) are proposed. No 
changes to any plant operating practices, procedures, computer 
firmware/ software will occur.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does not involve a significant reduction in a margin on 
safety.
    The design basis of the plant requires structures to be capable 
of withstanding normal and accident loads including those from a 
design basis tornado. The Yield Line Theory methodology, as applied 
in an approved plant calculation, has demonstrated that the east 
wall of the CR-3 Auxiliary Building is capable of performing its 
design function. There is a slight reduction in conservatism between 
the method used for the remaining Class 1 structures, American 
Concrete Institute (ACI) standard 318-63 and the Yield Line Theory 
methodology, but the calculation performed with the Yield Line 
Theory methodology validates the requirement that the east wall of 
the CR-3 Auxiliary Building will protect the important to safety 
SSCs located in proximity to the wall from damage.
    ACI 318-63 utilizes conservative methods, and due to the 
assumptions and technique, results in a Code defined value for 
strength that is not the maximum limit. The Yield Line Theory 
methodology uses assumptions and techniques that will define the 
failure point. However, the calculation performed for the east wall 
of the CR-3 Auxiliary Building demonstrates that there is margin to 
this ``failure point,'' and the strength of the wall exceeds the 
applied loads, including the tornado wind and pressure drop loads, 
and will not fail due to tornado missile impact.
    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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Thomas H. Boyce.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: July 31, 2008.
    Description of amendment request: The proposed amendments would 
change the PPL Susquehanna, LLC (PPL) Units 1 and 2 Technical 
Specification (TSs) 3.6.1.3 ``Primary Containment Isolation Valves 
(PCIVs).'' It proposes to revise the Secondary Containment Bypass 
Leakage (SCBL) limit in Surveillance Requirement 3.6.1.3.11 from ``less 
than or equal to 9 standard cubic foot/feet per hour (scfh)'' to ``less 
than or equal to 15 scfh when pressurized to greater than or equal to 
Pa.''
    Basis for proposed no significant hazards consideration 
determination:

[[Page 68456]]

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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The structures, systems and components affected by the proposed 
change act as mitigators to the consequences of accidents. These 
components are not initiators of any accident analyzed in the Final 
Safety Analysis Report (FSAR). As such, the proposed change does not 
increase the probability of an accident previously evaluated. Based 
on the revised analysis, the proposed change does revise the 
performance requirement; however, the proposed change does not 
involve a revision to the parameters or conditions that could 
contribute to the initiation of a DBA [design-basis accident] 
discussed in Chapter 15 of the FSAR.
    Plant-specific radiological analysis has been performed using 
the increased Secondary Containment Bypass Leakage (SCBL) limit. 
This analysis demonstrates that the CRHE [control room habitability 
envelope] dose consequences meet the regulatory guidance provided 
for use with the Alternative Source Term (AST), and the offsite 
doses are well within acceptable limits (10 CFR 50.67, Regulatory 
Guide (RG) 1.183, and Standard Review Plan Section (SRP) 15.0.1).
    Therefore, the proposed amendment does not result in a 
significant increase in the consequences of any previously evaluated 
accident.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
any plant equipment. No new equipment is being introduced, and 
installed equipment is not being operated in a new or different 
manner. There are no setpoints, at which protective or mitigative 
actions are initiated, affected by this change. This change does not 
alter the manner in which equipment operation is initiated, nor will 
the function demands on credited equipment be changed. No 
alterations in the procedures that ensure the plant remains within 
analyzed limits are being proposed, and no changes are being made to 
the procedures relied upon to respond to an off-normal event as 
described in the FSAR. As such, no new failure modes are being 
introduced. The change does not alter assumptions made in the safety 
analysis and licensing basis.
    Therefore, the proposed 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 a 
margin of safety?
    Response: No.
    The results of the revised accident analysis are subject to the 
acceptance criteria in 10 CFR 50.67. The revised Secondary 
Containment Bypass Leakage rate limit is used in the LOCA [loss-of-
coolant accident] radiological analysis. The analysis has been 
performed using conservative methodologies. Safety margins and 
analytical conservatisms have been evaluated and have been found 
acceptable. The analyzed LOCA event has been carefully selected and 
margin has been retained to ensure that the analysis adequately 
bounds postulated event scenarios. The dose consequences of the 
limiting event is within the acceptance criteria presented in 10 CFR 
50.67, RG 1.183, and SRP 15.0.1. The effect of the revision to the 
Technical Specification requirements has been analyzed and doses 
resulting from the pertinent design basis accident have been found 
to remain within regulatory limits. The change continues to ensure 
that the doses at the exclusion area and low population zone 
boundaries, as well as the control room, are within the 
corresponding regulatory limits. 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief : Mark Kowal.

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

    Date of amendment request: July 18, 2008.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) 2.1.1.2 to decrease the safety 
limit minimum critical power ratio (SLMCPR) from 1.11 to 1.09 for 
single recirculation loop operation and from 1.09 to 1.07 for two 
recirculation loop operation.
    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 Technical Specification change involve a 
significant increase in the probability or consequences of an 
accident previously evaluated?
    No. The proposed amendment establishes a revised SLMCPR value 
for single and two recirculation loop operation. The probability of 
an evaluated accident is derived from the probabilities of the 
individual precursors to that accident. The proposed SLMCPR values 
preserve the existing margin to transition boiling and the 
probability of fuel damage is not increased. Since the change does 
not require any physical plant modifications or physically affect 
any plant components, no individual precursors of an accident are 
affected and the probability of an evaluated accident is not 
increased by revising the SLMCPR values.
    The consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. The revised SLMCPR values have been determined using 
NRC-approved methods and procedures. The basis of the MCPR Safety 
Limit is to ensure no mechanistic fuel damage is calculated to occur 
if the limit is not violated. These calculations do not change the 
method of operating the plant and have no effect on the consequences 
of an evaluated accident. Therefore, the proposed TS change does not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed Technical Specification change create the 
possibility of a new or different kind of accident from any accident 
previously evaluated?
    No. The proposed license amendment involves a revision of the 
SLMCPR value for single and two recirculation loop operation based 
on the results of an analysis of the Unit 1 Cycle 8 core. Creation 
of the possibility of a new or different kind of accident would 
require the creation of one or more new precursors of that accident. 
New accident precursors may be created by modifications of the plant 
configuration, including changes in the allowable methods of 
operating the facility. This proposed license amendment does not 
involve any modifications of the plant configuration or changes in 
the allowable methods of operation. Therefore, the proposed TS 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed Technical Specification change involve a 
significant reduction in a margin of safety?
    No. The margin of safety as defined in the TS bases will remain 
the same. The new SLMCPR values were calculated using referenced 
fuel vendor methods and procedures, which are in accordance with the 
fuel design and licensing criteria. The SLMCPR remains high enough 
to ensure that greater than 99.9 percent of all fuel rods in the 
core are expected to avoid transition boiling if the limit is not 
violated, thereby preserving the fuel cladding integrity. Therefore, 
the proposed TS change does not involve a 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.

[[Page 68457]]

    NRC Branch Chief: Thomas H. Boyce.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendment: September 27, 2007, as 
supplemented by letter dated September 5, 2008.
    Brief description of amendment: The amendment modified the 
technical specifications (TS) by relocating references to specific 
American Society for Testing and Materials standards for fuel oil 
testing to licensee-controlled documents as part of the implementation 
of Technical Specification Task Force (TSTF) Traveler No. 374. This 
proposed change to the standard technical specifications was submitted 
by the TSTF in TSTF-374, ``Revision to TS 5.5.13 and Associated TS 
Bases for Diesel Fuel Oil,'' and is applicable to all nuclear power 
reactors.
    Date of issuance: October 30, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 182.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: December 18, 2007 (72 
FR 71705). The September 5, 2008 supplement, contained clarifying 
information, did not expand the scope of the application as originally 
noticed, and did not change the NRC staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 30, 2008.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut

    Date of application for amendment: March 25, 2008, as supplemented 
by letter dated September 30, 2008.
    Brief description of amendment: The amendment revises the reactor 
coolant system (RCS) specific activity to utilize a new indicator, Dose 
Equivalent Xenon-133 and only take into account the noble gas activity 
in the primary coolant, instead of using the average disintegration 
energy (E Bar). Specifically, the Technical Specification 3.4.8, 
``Specific Activity,'' limit on RCS gross specific activity has a new 
limit on RCS noble gas specific activity. The changes are based on 
Technical Specification Task Force (TSTF) change traveler TSTF-490, 
``Deletion of E Bar Definition and Revision to RCS Specific Activity 
Tech. Spec. [Technical Specification].''
    Date of issuance: October 27, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 307 and 246.
    Renewed Facility Operating License Nos. DPR-65 and NPF-49: 
Amendment revised the License and Technical Specifications.
    Date of initial notice in Federal Register: July 29, 2008 (73 FR 
43955-43956). The supplement dated September 30, 2008, 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 amendment is contained 
in a Safety Evaluation dated October 27, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application of amendments: October 16, 2007, as 
supplemented May 7, September 2 and October 23, 2008.
    Brief description of amendments: The amendments revised the 
Technical Specifications to accommodate plant modifications that 
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.
    Date of Issuance: October 29, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 363, 365, 364.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: November 20, 2007 (72 
FR 65364). The supplements dated May 7, September 2 and October 23, 
2008, 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 29, 2008.

[[Page 68458]]

    No significant hazards consideration comments received: No.

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

    Date of application of amendments: October 22, 2007, supplemented 
July 14, September 17, and October 27, 2008.
    Brief description of amendments: The amendments revised the 
Technical Specifications related to accommodate the use of AREVA NP 
Mark-B-HTP fuel.
    Date of Issuance: October 29, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 362, 364, 363.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: November 20, 2007 (72 
FR 65365). The supplements dated July 14, September 17, and October 27, 
2008, 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 29, 2008.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: December 13, 2007, as 
supplemented by letter dated July 10, 2008.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) by adding three Emergency Core Cooling System 
(ECCS) valves and removing four ECCS valves from a TS surveillance 
requirement for checking valve position every 7 days.
    Date of issuance: October 29, 2008.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 256.
    Facility Operating License No. DPR-26: The amendment revised the 
License and the TSs.
    Date of initial notice in Federal Register: March 25, 2008 (73 FR 
15784). The July 10, 2008, supplement provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the NRC staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 29, 2008.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendment: October 19, 2007, supplemented 
by letters dated March 14, 2008, March 26, 2008, and July 18, 2008.
    Brief description of amendment: The amendments consist of changes 
to the technical specifications of each unit, increasing the allowed 
surveillance interval for local power range monitor calibrations from 
1000 effective full power hours (EFPH) to 2000 EFPH.
    Date of issuance: October 28, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 195 and 156.
    Facility Operating License Nos. NPF-39 and NPF-85. These amendments 
revised the license and the technical specifications.
    Date of initial notice in Federal Register: July 8, 2008 (73 FR 
39055). The supplements dated March 14, 2008, March 26, 2008 and July 
18, 2008, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed and did not change the NRC staff's original proposed no 
significant hazards determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 28, 2008.
    No significant hazards consideration comments received: No.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of application for amendment: April 4, 2008, as supplemented 
by letter dated August 6, 2008.
    Brief description of amendment: The amendment revised the Technical 
Specifications by adding a new Limiting Condition for Operation (LCO), 
LCO 3.0.9. This LCO establishes conditions under which systems would 
remain operable when required physical barriers are not capable of 
providing their related support function. This amendment is consistent 
with approved Technical Specification Task Force (TSTF) Improved 
Standard Technical Specifications Change Traveler, TSTF-427, Revision 
2.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days following startup from the 2009 Refueling Outage.
    Amendment No.: 157.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 9, 2008 (73 
FR 52418).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 22, 2008.
    No significant hazards consideration comments received: None.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: October 29, 2007, as 
supplemented by letters dated April 24 and June 13, 2008.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) for Prairie Island Nuclear Generating 
Plants, Units 1 and 2. The amendments revise TS 3.8.1 ``AC Sources--
Operating'' by revising Surveillance Requirement 3.8.1.9 to require the 
emergency diesel generator 24-hour load test be performed at or below a 
power factor of 0.85.
    Date of issuance: October 21, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 189, 178.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 18, 2007 (72 
FR 71713). The supplemental letters contained clarifying information 
and did not change the initial no significant hazards consideration 
determination, and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in Safety Evaluation dated October 21, 2008.
    No significant hazards consideration comments received: No.

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 application for amendments: October 15, 2007, as 
supplemented by letter dated July 8, 2008.

[[Page 68459]]

    Brief description of amendments: The amendments relocate 
surveillance frequencies of most surveillance tests from the Technical 
Specifications (TS) to a licensee-controlled document, the Surveillance 
Frequency Control Program (SFCP). Once relocated, changes to the 
surveillance frequencies may be made using a risk-informed methodology, 
Nuclear Energy Institute (NEI) document NEI 04-10 Rev. 1, as specified 
in the Administrative Controls of the TS. The NRC staff has previously 
approved NEI 04-10 Rev. 1, as acceptable for referencing in licensing 
applications.
    Date of issuance: October 30, 2008.
    Effective date: As of its date of issuance and shall be implemented 
within 360 days from the date of issuance.
    Amendment Nos.: Unit 1-200; Unit 2-201.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: November 20, 2007 (72 
FR 65370). The supplement dated July 8, 2008, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 30, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 17, 2008, as 
supplemented August 15, 2008.
    Brief description of amendment: The amendment will strengthen the 
control room envelope habitability requirements, adds a new 
administrative controls program, and adds an additional condition as 
described in Technical Specification Task Force traveler 448, Revision 
3, ``Control Room Habitability.''
    Date of issuance: October 27, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 180.
    Renewed Facility Operating License No. NPF-12: Amendment revises 
the Appendix A Technical Specifications and the Appendix C Additional 
Conditions.
    Date of initial notice in Federal Register: February 12, 2008 (73 
FR 8071). The supplement dated August 18, 2008, prov