Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 65685-65705 [E8-25882]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 73, Number 214 (Tuesday, November 4, 2008)]
[Notices]
[Pages 65685-65705]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-25882]


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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 9, 2008 to October 22, 2008. The 
last biweekly notice was published on October 21, 2008 (73 FR 370501).

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

[[Page 65686]]

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 6D44, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 
4:15 p.m. Federal workdays. Copies of written comments received may be 
examined at the Commission's Public Document Room (PDR), located at One 
White Flint North, Public File Area O1F21, 11555 Rockville Pike (first 
floor), Rockville, Maryland. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    Within 60 days after the date of publication of this notice, 
person(s) may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
via electronic submission through the NRC E-Filing system for a hearing 
and a petition for leave to intervene. Requests for a hearing and a 
petition for leave to intervene shall be filed in accordance with the 
Commission's ``Rules of Practice for Domestic Licensing Proceedings'' 
in 10 CFR Part 2. Interested person(s) should consult a current copy of 
10 CFR 2.309, which is available at the Commission's PDR, located at 
One White Flint North, Public File Area 01F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland. Publicly available records will be 
accessible from the Agencywide Documents Access and Management System's 
(ADAMS) Public Electronic Reading Room on the Internet at the NRC Web 
site, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request 
for a hearing or petition for leave to intervene is filed within 60 
days, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party. 
Those permitted to intervene become parties to the proceeding, subject 
to any limitations in the order granting leave to intervene, and have 
the opportunity to participate fully in the conduct of the hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for hearing or a petition for leave to intervene must be 
filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve documents over the Internet 
or in some cases to mail copies on electronic storage media. 
Participants may not submit paper copies of their filings unless they 
seek a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
hearingdocket@nrc.gov, or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate).

[[Page 65687]]

    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@nrc.gov.

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

    Date of amendment request: September 2, 2008.
    Description of amendment request: The proposed amendment would 
relocated surveillance requirement (SR) 3.8.3.6 from the technical 
specifications (TSs) to a licensee-controlled document. SR 3.8.3.6 
requires the emergency diesel generator fuel oil storage tanks to be 
drained, sediment removed, and cleaned on a 10-year interval.
    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 FOSTs [fuel oil storage tanks] provide the storage for the 
DG [diesel generator] DG fuel oil, assuring an adequate volume is 
available for each DG to operate for seven days in the event of a 
loss of offsite power concurrent with a loss of coolant accident. 
The relocation of the SR to drain and clean the FOSTs to a licensee-
controlled document will not impact any of the previously analyzed 
accidents. Sediment in the tank, or failure to perform this SR, does 
not necessarily result in an inoperable storage tank. Fuel oil 
quantity and quality are assured by other TS SRs that remain 
unchanged.
    These SRs help ensure tank sediment is minimized and ensure that 
any degradation of the tank wall surface that results in a fuel oil 
volume reduction is detected and corrected in a timely manner. 
Future changes to the licensee-controlled document will be evaluated 
pursuant to the requirements of 10 CFR 50.59, ``Changes, tests, and 
experiments,'' to ensure that such changes do not result in more 
than a minimal increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration or the manner in which the plant is 
operated and maintained. The proposed change does not adversely 
affect the ability of structures, systems or components (SSCs) to 
perform their intended function to mitigate the consequences of an 
initiating event within the assumed acceptance limits.
    The proposed change does not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
Further, the proposed change does not increase the types and amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposure.

[[Page 65688]]

    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 TS change does not involve the addition or 
modification of any plant equipment. Also, the proposed change will 
not alter the design configuration, or method of operation of plant 
equipment beyond its normal functional capabilities. The 
requirements retained in the TS continue to require testing of the 
diesel fuel oil to ensure the proper functioning of the DGs. The 
proposed TS change does not create any new credible failure 
mechanisms, malfunctions or accident initiators.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change does not alter or exceed a design basis or 
safety limit. The requirements retained in the TS continue to 
require testing of the diesel fuel oil to ensure the DGs are able to 
perform their intended function.
    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. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

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

    Date of amendment request: August 29, 2008.
    Description of amendment request: The amendments would modify 
Technical Specification (TS) 5.6.5, Core Operating Limits Report 
(COLR), by updating TS 5.6.5b to reflect the current analytical methods 
used to determine the core operating limits in Palo Verde Nuclear 
Generating Station (PVNGS), Units 1, 2, and 3. The proposed amendment 
is an administrative change and all of the analytical methods have been 
previously reviewed and approved by the Nuclear Regulatory Commission 
(NRC).
    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 the list of methodologies used at PVNGS 
[PVNGS, Units 1, 2, and 3] to determine the various COLR limits is 
an administrative change which updates the list in the TS to include 
NRC reviewed and approved COLR methodologies for PVNGS. It does not 
add or modify any previously used methodologies; it updates the list 
to include those already approved for use. This change does not make 
any physical changes to any structure, system or component, and it 
does not affect any design basis accident evaluation.
    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 to the list of methodologies used at PVNGS 
to determine the various COLR limits is an administrative change 
which updates the list in the TS to include all of the NRC reviewed 
and approved COLR methodologies for PVNGS. This change does not 
create any new failure modes or affect the interaction between any 
structure, system or component.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to the list of methodologies used at PVNGS 
to determine the various COLR limits is an administrative change 
which updates the list in the TS to include all of the NRC reviewed 
and approved COLR methodologies for PVNGS. This change does not 
modify any 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 
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.

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: August 29, 2008.
    Description of amendments request: The amendment would revise 
Calvert Cliffs Nuclear Power Plant (CCNPP) Operating License Nos. DPR-
53 and DPR-69 and Technical Specifications (TSs) by increasing the 
licensed core power of CCNPP, Unit Nos. 1 and 2 by 1.38 percent to 2737 
MWt. The power uprate amendment request is based on the use of the 
Caldon Leading Edge Flow Measurement (LEFM) CheckPlus system for more 
accurate determination of main feedwater flow and the associated 
determination of reactor power through the performance of the power 
calorimetric calculation currently required by CCNPP TSs.
    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. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    In support of this measurement uncertainty recapture (MUR) power 
uprate, a comprehensive evaluation was performed for Nuclear Steam 
Supply System (NSSS), balance of plant systems and components, and 
analyses that could be affected by this change. A power calorimetric 
uncertainty calculation was performed, and the impact of increasing 
plant power by 1.38 percent on the plant's design and licensing 
basis was evaluated. The result of these evaluations is that 
structures, systems, and components required to mitigate transients 
will continue to be capable of performing their design function at 
an uprated core power of 2737 MWt. In addition, an evaluation of the 
accident analyses demonstrates that applicable analysis acceptance 
criteria continue to be met. No accident initiators are affected by 
this uprate and no challenges to any plant safety barriers are 
created by this change. Therefore, operation of the facility in 
accordance with the proposed change will not involve a significant 
increase in the probability of an accident previously evaluated.
    The proposed change does not affect the radiological release 
paths, the frequency of release, or the source-term for release for 
any accidents previously evaluated in the Updated Final Safety 
Analysis Report. Structures, systems, and components required to 
mitigate transients remain capable of performing their design 
functions, and thus were found acceptable. The reduced uncertainty 
in the feedwater flow input to the power calorimetric measurement 
ensures that

[[Page 65689]]

applicable accident analyses acceptance criteria continue to be met 
in support of operation at a core power of 2737 MWt. Analyses 
performed to assess the effects of mass and energy remain valid. The 
source-terms used to assess radiological consequences have been 
reviewed and determined to bound operation at the uprated condition. 
Therefore, operation of the facility in accordance with the proposed 
change will not involve a significant increase in 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. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    No new accident scenarios, failure mechanisms, or single-
failures are introduced as a result of the proposed changes. The 
installation of the Caldon LEFM CheckPlus feedwater flow 
instrumentation system has been analyzed, and failures of this 
system will have no adverse effect on any safety-related system or 
any structures, systems, and components required for transient 
mitigation. All structures, systems and components previously 
required for the mitigation of a transient remain capable of 
fulfilling their intended design functions. The proposed changes 
have no adverse effects on any safety-related system or component 
and do not challenge the performance or integrity of any safety-
related system.
    This change does not adversely affect any current system 
interfaces or create any new interfaces that could result in an 
accident or malfunction of a different kind than was previously 
evaluated. Operating at a core power level of 2737 MWt does not 
create any new accident initiators or precursors. The reduced 
uncertainty in the feedwater flow input to the power calorimetric 
measurement ensures that applicable accident analyses acceptance 
criteria continue to be met to support operation at a core power of 
2737 MWt. Credible malfunctions continue to be bounded by the 
current accident analysis of record or evaluations that demonstrate 
that applicable acceptance criteria continue to be met.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The margins of safety associated with the MUR power uprate are 
those pertaining to core power. This includes those associated with 
the fuel cladding, Reactor Coolant System pressure boundary, and 
containment barriers. A comprehensive engineering review was 
performed to evaluate the 1.38 percent increase in the licensed core 
power from 2700 MWt to 2737 MWt. The 1.38 percent increase required 
that revised NSSS design thermal and hydraulic parameters be 
established, which then served as the basis for all of the NSSS 
analyses and evaluations. This engineering review concluded that no 
design modifications are required to accommodate the revised NSSS 
design conditions. The NSSS components were evaluated and it was 
concluded that the NSSS components have sufficient margin to 
accommodate the 1.38 percent power uprate. The NSSS accident 
analyses were evaluated for the 1.38 percent power uprate. In all 
cases, the evaluations demonstrate that the applicable analyses 
acceptance criteria continue to be met. As a result, the margins of 
safety continue to be bounded by the current analyses of record for 
this 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 
amendments request involves no significant hazards consideration.
    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 11, 2008.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications, extending the 15-year interval 
between containment Type A tests specified by Specification 4.4.a, 
``Integrated Leak Rate Test,'' by 6 months. The current Type A test 
interval expires at the end of April 2009. The proposed amendment would 
extend this interval, on a one-time basis, to October 2009 to coincide 
with completion of the next scheduled refueling outage.
    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 probability or consequences of accidents previously 
evaluated in the Updated Safety Analysis Report are unaffected by 
this proposed change. There is no change to any equipment response 
or accident mitigation scenario, and this change results in no 
additional challenges to fission product barrier integrity. The 
proposed change does not alter the design, configuration, operation, 
or function of any plant system, structure, or component. As a 
result, the probabilities of previously evaluated accidents are 
unaffected. The proposed extension to the Type A test interval does 
not involve a significant increase in consequences because, as 
discussed in NUREG-1493, Performance Based Containment Leak Rate 
Test Program, Type B and C tests identify the vast majority 
(approximately 97 percent) of all potential leakage paths. Further, 
Type A tests identify only a few potential leakage paths that cannot 
be identified through Type B and C testing, and leaks found by Type 
A testing have been only marginally greater than existing 
requirements. The frequency and methods of performance of Type B and 
Type C testing are unaffected by this proposed change. In addition, 
periodic inspections of containment required by the ASME [American 
Society of Mechanical Engineers] code and the maintenance rule, 
which are capable of detecting any significant degradation, are 
unaffected by the proposed change.
    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.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The proposed change does not challenge the performance or integrity 
of any safety-related system. The proposed change does not install 
or remove any plant equipment. The proposed change does not alter 
the design, physical configuration, or mode of operation of any 
plant structure, system, or component. No physical changes are being 
made to the plant, so no new accident causal mechanisms are being 
introduced. The proposed change only changes the frequency of 
performing the next Type A test; the Type A test implementation and 
acceptance criteria are unchanged. Type B and Type C testing 
frequency and method of performance are not affected by this 
proposed change.
    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 margin of safety associated with the acceptance criteria of 
any accident is unchanged. The proposed change will have no affect 
on the availability, operability, or performance of the safety-
related systems and components. The proposed change does not alter 
the design, configuration, operation, or function of any plant 
system, structure, or component. The ability of operable structures, 
systems, and components to perform their designated safety function 
is unaffected by this proposed change. NUREG-

[[Page 65690]]

1493 concluded that reducing the frequency of Type A tests to one-
in-20 years resulted in an imperceptible increase in risk. Type B 
and Type C testing frequency and method of performance are 
unaffected by this proposed change. Also, [other] inspections of 
containment required by the ASME code and the maintenance rule 
[will] provide reasonable assurance that containment will not 
degrade in a manner that is only detectable by Type A testing. In 
addition, the inherent risk of an additional plant shutdown would be 
eliminated by the proposed amendment, further ensuring no 
significant reduction in safety margin.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana

    Date of amendment request: July 28, 2008.
    Description of amendment request: The proposed amendment would: (1) 
Delete Technical Specification (TS) surveillance requirement (SR) 
3.1.3.2 and revise SR 3.1.3.3, (2) remove reference to SR 3.1.3.2 from 
Required Action A.2 of TS 3.1.3, ``Control Rod OPERABILITY,'' (3) 
clarify the requirement to fully insert all insertable rods for the 
limiting condition for operation (LCO) in TS 3.3.1.2, required Action 
E.2, ``Source Range Monitoring Instrumentation,'' and (4) revise 
Example 1.4-3 in Section 1.4, ``Frequency,'' to clarify the 
applicability of the 1.25 surveillance test interval extension.
    The NRC staff issued a notice of opportunity to comment in the 
Federal Register on August 16, 2007 (72 FR 46103), on possible 
amendments to revise the plant-specific TSs, modify TS control rod SR 
testing frequency, clarify TS control insertion requirements, and 
clarify SR frequency discussions, including a model safety evaluation 
and model no significant hazards consideration (NSHC) determination, 
using the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on November 13, 2007 (72 FR 63935). The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
July 28, 2008.
    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 generically implements TSTF-475, Revision 1, 
``Control Rod Notch Testing Frequency and SRM [Source Range Monitor] 
Insert Control Rod Action.'' TSTF-475, Revision 1 modifies NUREG-
1433 (BWR/4) and NUREG-1434 (BWR/6) STS. The changes: (1) revise TS 
testing frequency for surveillance requirement (SR) 3.1.3.2 in TS 
3.1.3, ``Control Rod OPERABILITY,'' (2) clarify the requirement to 
fully insert all insertable control rods for the limiting condition 
for operation (LCO) in TS 3.3.1.2, Required Action E.2, ``Source 
Range Monitoring Instrumentation'' (NUREG-1434 only), and (3) revise 
Example 1.4-3 in Section 1.4 ``Frequency'' to clarify the 
applicability of the 1.25 surveillance test interval extension. The 
consequences of an accident after adopting TSTF-475, Revision 1 are 
no different than the consequences of an accident prior to adoption. 
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 Accident Previously 
Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
proposed change will not introduce new failure modes or effects and 
will not, in the absence of other unrelated failures, lead to an 
accident whose consequences exceed the consequences of accidents 
previously analyzed. Thus, 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

    TSTF-475, Revision 1 will: (1) Revise the TS SR 3.1.3.2 
frequency in TS 3.1.3, ``Control Rod OPERABILITY,'' (2) clarify the 
requirement to fully insert all insertable control rods for the 
limiting condition for operation (LCO) in TS 3.3.1.2, ``Source Range 
Monitoring Instrumentation,'' and (3) revise Example 1.4-3 in 
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25 
surveillance test interval extension. The GE [General Electric] 
Nuclear Energy Report, ``CRD [Control Rod Drive] Notching 
Surveillance Testing for Limerick Generating Station,'' dated 
November 2006, concludes that extending the control rod notch test 
interval from weekly to monthly is not expected to impact the 
reliability of the scram system and that the analysis supports the 
decision to change the surveillance frequency. Therefore, the 
proposed changes in TSTF-475, Revision 1 are acceptable and do not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the analysis adopted by the licensee 
and, based upon this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: July 21, 2008.
    Description of amendment request: The proposed amendment would 
support a proposed change to the in-service inspection program that is 
based on topical report WCAP-16168-NP-A, Revision 2, ``Risk-Informed 
Extension of the Reactor Vessel In-Service Inspection Interval.'' In 
the referenced safety evaluation of the topical report, the NRC 
required licensees to amend their licenses to require that the 
information and analyses requested in Section (e) of the final 10 CFR 
50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275 prior to 
issuance of the final 10 CFR 50.61a) be submitted for NRC staff review 
and approval within one year of completing the required reactor vessel 
weld inspection. Entergy Nuclear Operations, Inc., proposes to add a 
new license condition to provide this information.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment changes the renewed facility operating 
license by adding a license condition to require that the

[[Page 65691]]

information and analyses requested in Section (e) of the final 10 
CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275 
prior to issuance of the final 10 CFR 50.61a) will be submitted for 
NRC staff review and approval within one year of completing the 
required reactor vessel weld inspection. The proposed amendment does 
not involve operation of the required structures, systems or 
components (SSCs) in a manner or configuration different from those 
previously recognized or evaluated.
    The proposed changes are administrative and have no impact on 
plant operation or equipment.
    Therefore, operation of the facility in accordance with the 
proposed amendment would 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 license amendment does not involve a physical 
alteration of any SSC or change the way any SSC is operated. The 
proposed license amendment does not involve operation of any 
required SSCs in a manner or configuration different from those 
previously recognized or evaluated.
    The proposed changes are administrative and have no impact on 
plant operation or equipment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes are administrative and have no impact on 
plant operation or equipment or on any margin of safety.
    Therefore, the proposed amendment would not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: August 28, 2008.
    Description of amendment request: The proposed amendment would 
change Technical Specifications (TS) Administrative Controls section 5 
to incorporate NRC-approved Technical Specification Task Force (TSTF) 
Improved Technical Specification (ITS) TSTF-363, ``Revise Topical 
Report references in ITS 5.6.5, [Core Operating Limits Report] COLR,'' 
revision 0. ENO also proposes to make an administrative change to the 
plant staff qualifications section.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed changes are administrative or provide 
clarification only.
    The proposed changes do not have any impact on the integrity of 
any plant system, structure, or component (SSC) that initiates an 
analyzed event. The proposed changes will not alter the operation 
of, or otherwise increase the failure probability of any plant 
equipment that initiates an analyzed accident. Thus, the probability 
of any accident previously evaluated is not significantly increased.
    The proposed changes do not affect the ability to mitigate 
previously evaluated accidents, and do not affect radiological 
assumptions used in the evaluations. The proposed changes do not 
change or alter the design criteria for the systems or components 
used to mitigate the consequences of any design-basis accident. The 
proposed amendment does not involve operation of the required SSCs 
in a manner or configuration different from those previously 
recognized or evaluated. Thus, the radiological consequences of any 
accident previously evaluated are not increased.
    Therefore, operation of the facility in accordance with the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed amendment does not involve a physical 
alteration of any SSC or a change in the way any SSC is operated. 
The proposed amendment does not involve operation of any required 
SSCs in a manner or configuration different from those previously 
recognized or evaluated. No new failure mechanisms will be 
introduced by the changes being requested.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The amendment does not involve a significant reduction in a 
margin of safety. The proposed amendment does not affect any margin 
of safety. The proposed amendment does not involve any physical 
changes to the plant or manner in which the plant is operated.
    Therefore, the proposed amendment would not involve a 
significant reduction in a margin of safety.

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

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

    Date of amendment request: September 30, 2008.
    Description of amendment request: The proposed amendment would 
revise the Facility Operating License and Technical Specification 
Section 4.0 by changing the names of the licensees to Enexus Nuclear 
Pilgrim LLC and 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 the referenced NRC Order requirements. 
Principal management and operational staffing for the restructured 
organization remain largely unchanged. The proposed changes do not: 
(a) Involve a significant increase in the probability or 
consequences of an accident previously evaluated; (b) create the 
possibility of a new or different kind of accident from any accident 
previously evaluated; or (c) 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

[[Page 65692]]

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 4, 2008.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) Section 5.1, ``Site,'' to 
remove the restriction on the sale and lease of site property and 
replace the restriction with a requirement to retain complete authority 
to determine and maintain sufficient control of all activities, 
including the authority to exclude or remove personnel and property, 
within the minimum exclusion area as described in 10 CFR 100.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    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 that affects the probability of an 
accident or that supports mitigation or consequences of an accident 
previously evaluated. The proposed change establishes requirements 
for sale or lease of property within the exclusion area. 
Additionally, ENO [Entergy Nuclear Operations, Inc.] will retain 
authority to determine all activities within the exclusion area and 
to remove personnel and property from the area as necessary to 
ensure the regulatory exposure limits are met.
    The proposed change does not affect reactor operations or 
accident analysis and there is no change to the radiological 
consequences of a previously analyzed accident. The operability 
requirements for accident mitigation systems remain consistent with 
the licensing and design basis. 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 its function. The proposed 
change establishes requirements for sale or lease of property within 
the exclusion area. Any additional activities performed within the 
exclusion area will be reviewed by ENO and verified to not represent 
a new hazard or that they have been accommodated in the plant 
licensing and design basis. As such, no new or different types of 
equipment will be installed or operated without additional review 
and approval by ENO. Operation of existing installed equipment is 
unchanged. The methods governing plant operation and testing remain 
consistent with current safety analysis assumptions. 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. These changes do not change any existing design or 
operational requirements, and do 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. 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.

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 remove the requirement to 
perform quarterly closure time testing of the Main Steam Isolation 
Valves (MSIVs) by deleting TS Surveillance Requirement 4.7.D.1.c. 
Operability testing of the MSIVs will continue to be required by the 
Vermont Yankee Inservice Test 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. 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 deletes the specific surveillance 
requirement to exercise the MSIVs once per quarter from the TS. 
Following implementation of the proposed change, the VY TS still 
will require operability testing of the MSIVs by reference to the VY 
IST program. The quarterly exercise involves a timed full stroke 
closure of each individual MSIV and subsequent reopening to the full 
open position. Details of MSIV testing requirements will continue to 
be contained in the VY IST program. The MSIV closure time setpoint 
values related to the safety functions of the MSIVs will continue to 
be contained in the VY UFSAR [Updated Final Safety Analysis Report] 
and the VY TRM [Technical Requirements Manual]. Changes to the VY 
UFSAR and TRM are evaluated per the requirements of 10 CFR 50.59. 
These controls are adequate to ensure the required inservice testing 
is performed to verify the MSIVs are operable and capable of 
performing their safety functions. The proposed amendment introduces 
no new equipment or changes to how equipment is operated. Therefore, 
the proposed amendment will 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.
    The proposed amendment deletes the specific surveillance 
requirement to exercise the MSIVs once per quarter from the TS. 
Following implementation of the proposed change, the VY TS still 
will require operability testing of the MSIVs by reference to the VY 
IST program. The quarterly exercise involves a timed full stroke 
closure of each individual MSIV and subsequent reopening to the full 
open position. 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.
    The proposed amendment deletes the specific surveillance 
requirement to exercise the MSIVs once per quarter from the TS. 
Following implementation of the proposed change, the VY TS still 
will require operability testing of the MSIVs by reference to the VY 
IST program. The quarterly exercise involves a timed full stroke 
closure of each individual MSIV and subsequent reopening to the full 
open position. The proposed amendment does not change the design or 
function of any component or system. The proposed amendment does not 
involve any safety limits or safety settings. The ability of the 
MSIVs to perform their safety function will continue to be tested in

[[Page 65693]]

accordance with the IST Program, through TS SR 4.7.D.1.b.
    Since the proposed controls are adequate to ensure the required 
inservice testing is performed, there will still be high assurance 
that the components are operable and capable of performing their 
respective safety functions, and that the systems will respond as 
designed to mitigate the subject events. 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 30, 2008.
    Description of amendment request: The proposed amendment would 
revise the Facility Operating License and Technical Specification 
Section 5.0 by changing the names of the licensees to EquaGen Nuclear 
LLC and Enexus Nuclear Vermont Yankee 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 the referenced NRC Order requirements; 
principal management and operational staffing for the restructured 
organization remain largely unchanged. The proposed changes do not: 
(a) Involve a significant increase in the probability or 
consequences of an accident previously evaluated; (b) create the 
possibility of a new or different kind of accident from any accident 
previously evaluated; or (c) 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.

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

    Date of amendment requests: July 21, 2008.
    Description of amendment request: 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 consistent with the program in place for 
complying with the requirements of 10 CFR 50.65(a)(4). Limiting 
Condition for Operation (LCO) 3.0.8 is added to the TS to provide this 
allowance and define the requirements and limitations for its use.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff 
issued a notice of opportunity for comment in the Federal Register on 
November 24, 2004 (69 FR 68412), on possible amendments concerning 
TSTF-372, including a model safety evaluation and model no significant 
hazards consideration (NSHC) determination, using the consolidated line 
item improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on May 4, 2005 (70 FR 23252).
    Basis for proposed no significant hazards consideration 
determination: Entergy Operations, Inc. (Entergy) has reviewed the 
proposed NSHC determination published in the Federal Register as part 
of the CLIIP. Entergy has concluded that the proposed NSHC 
determination presented in the Federal Register notice is applicable to 
Arkansas Nuclear One, Unit 2 and 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 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 
asse