[Federal Register Volume 73, Number 214 (Tuesday, November 4, 2008)]
[Notices]
[Pages 65685-65705]
From the Federal Register Online via the Government Publishing 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, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request 
for a hearing or petition for leave to intervene is filed within 60 
days, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which 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 
[email protected], or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate).

[[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 http://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying for a digital ID 
certificate is available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, in accordance 
with 10 CFR 2.302(g), with their initial paper filing requesting 
authorization to continue to submit documents in paper format. Such 
filings must be submitted by: (1) First class mail addressed to the 
Office of the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff; or (2) courier, express mail, or expedited 
delivery service to the Office of the Secretary, Sixteenth Floor, One 
White Flint North, 11555 Rockville, Pike, Rockville, Maryland, 20852, 
Attention: Rulemaking and Adjudications Staff. Participants filing a 
document in this manner are responsible for serving the document on all 
other participants. Filing is considered complete by first-class mail 
as of the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the petition 
and/or request should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later 
than 11:59 p.m. Eastern Time on the due date.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, 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, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

AmerGen Energy Company, LLC, Docket No. 50-461, 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 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.

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

    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low-probability occurrence and the 
overall TS system safety function would still be available for the 
vast majority of anticipated challenges. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in Regulatory Guide 1.177. A bounding risk assessment 
was performed to justify the proposed TS changes. The proposed LCO 
3.0.8 defines limitations on the use of the provision and includes a 
requirement for the licensee to assess and manage the risk 
associated with operation with an inoperable snubber. The net change 
to the margin of safety is insignificant. Therefore, this change 
does not involve a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

[[Page 65694]]

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: July 30, 2008, as supplemented on 
October 2, 2008.
    Description of amendment request: Entergy Operations Inc. (the 
licensee) proposes to modify the technical specifications (TS) 3.6.6, 
``Spray Additive System.'' Specifically, this amendment proposes to 
revise the Sodium Hydroxide (NaOH) tank concentration stated in TS 
3.6.6.3 from between 5.0 percent and 16.5 percent to between 6.0 
percent and 8.5 percent.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    There are no changes to the design or operation of the plant 
that could affect system, component, or accident functions as a 
result of changing the sodium hydroxide (NaOH) tank solution 
concentration limits. In addition, the dose reduction provided by 
maintaining the sump pH above 7.0 is retained, and therefore, dose 
consequences resulting from iodine dissolution remain unchanged. The 
proposed change simply imposes more restrictive operating conditions 
than are within the current TS limits. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of the proposed change. 
Structures, systems, and components previously required for 
mitigation of an accident remain capable of fulfilling their 
intended design function with this change to the TS. The proposed 
change has no new adverse effects on safety-related systems or 
components and does not challenge the performance or integrity of 
safety-related systems. The proposed change simply imposes more 
restrictive operating conditions that are within the current TS 
limits. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change imposes more restrictive operating 
conditions that are within the current TS limits. Revising the NaOH 
tank solution concentration limits reduces the amount of chemical 
precipitates formed under post-loss-of-coolant accident conditions. 
The margin of safety related to ensuring that the sump pH remains 
above 7.0 is not reduced. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: August 21, 2008.
    Description of amendment request: Entergy Operations Inc. (the 
licensee) proposes a one-time amendment for next containment integrated 
leakage rate test (ILRT) or Type A test at the Arkansas Nuclear One, 
Unit No. 2 (ANO-2). The ILRT is required by Technical Specification 
(TS) 6.5.16, ``Containment Leakage Rate Testing Program,'' to be 
performed every ten-years. The amendment would permit the existing ILRT 
frequency to be extended from 120 months (10 years) to approximately 
135 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed exemption involves a one-time extension to the 
current interval for Type A containment testing. The current test 
interval of 120 months (10 years) would be extended on a one-time 
basis to no longer than approximately 135 months from the last Type 
A test. The proposed extension does not involve a physical change to 
the plant or a change in the manner in which the plant is operated 
or controlled. The containment is designed to provide an essentially 
leak tight barrier against the uncontrolled release of radioactivity 
to the environment for postulated accidents. As such, the reactor 
containment itself and the testing requirements invoked to 
periodically demonstrate the integrity of the reactor containment 
exist to ensure the plant's ability to mitigate the consequences of 
an accident, and do not involve the prevention or identification of 
any precursors of an accident. Therefore, this proposed extension 
does not involve a significant increase in the probability of an 
accident previously evaluated nor does it create the possibility of 
a new or different kind of accident.
    This proposed extension is for the Type A containment leak rate 
tests only. The Type B and C containment leak rate tests will 
continue to be performed at the frequency currently required by the 
ANO-2 TS. As documented in NUREG 1493, Type B and C tests have 
identified a very large percentage of containment leakage paths and 
that the percentage of containment leakage paths that are detected 
only by Type A testing is very small. ANO-2's Type A test history 
supports this conclusion.
    The integrity of the reactor containment is subject to two types 
of failure mechanisms which can be categorized as (1) activity based 
and (2) time based. Activity based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment itself combined with the containment inspections 
performed in accordance with ASME, Section XI, the Maintenance Rule, 
and Licensing commitments serve to provide a high degree of 
assurance that the containment will not degrade in a manner that is 
detectable only by a Type A test. Based on the above, the proposed 
extension does not involve a significant increase in the 
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 revision to the TS involves a one-time extension to 
the current interval for Type A containment testing. The reactor 
containment and the testing requirements invoked to periodically 
demonstrate the integrity of the reactor containment exist to ensure 
the plant's ability to mitigate the consequences of an accident and 
do not involve the prevention or identification of any precursors of 
an accident. The proposed TS change does not involve a physical 
change to the plant or the manner in which the plant is operated or 
controlled. 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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to the TS involves a one-time extension to 
the current interval for Type A containment testing. The proposed TS 
change does not involve a physical

[[Page 65695]]

change to the plant or a change in the manner in which the plant is 
operated or controlled. The specific requirements and conditions of 
the Primary Containment leak Rate Testing Program, as defined in the 
TS, exist to ensure that the degree of reactor containment 
structural integrity and leak-tightness that is considered in the 
plant safety analysis is maintained. The overall containment leak 
rate limit specified by TS is maintained. The proposed change 
involves only the extension of the interval between Type A 
containment leak rate tests. The proposed surveillance interval 
extension is bounded by the 15 month extension currently authorized 
within NEI 94-01, Revision 0. Type B and C containment leak rate 
tests will continue to be performed at the frequency currently 
required by TS. Industry experience supports the conclusion that 
Type B and C testing detects a large percentage of containment 
leakage paths and that the percentage of containment leakage paths 
that are detected only by Type A testing is small. The containment 
inspections performed in accordance with ASME, Section XI and the 
Maintenance Rule serve to provide a high degree of assurance that 
the containment will not degrade in a manner that is detectable only 
by Type A testing. The combination of these factors ensures that the 
margin of safety that is in plant safety analysis is maintained. The 
design, operation, testing methods and acceptance criteria for Type 
A, B, and C containment leakage tests specified in applicable codes 
and standards will continue to be met, with the acceptance of this 
proposed change, since these are not affected by changes to the Type 
A test interval. Therefore, the proposed TS change does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: July 21, 2008.
    Description of amendment requests: The proposed amendments would 
modify the Technical Specification (TS) by adding Limiting Condition 
for Operation (LCO) 3.0.8 on the inoperability of snubbers using the 
Consolidated Line Item Improvement Process (CLIIP). The proposed 
amendments would also make conforming changes to TS LCO 3.0.1. This 
request is consistent with NRC-approved Industry/Technical 
Specification Task Force (TSTF) Traveler No. 372, Revision 4, 
``Addition of LCO 3.0.8, Inoperability of Snubbers.''
    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 affirmed the applicability 
of the following NSHC for Arkansas Nuclear One, Unit 1 in its 
application and as published in the Federal Register.

Criterion 1: The Proposed Changes Do Not Involve a Significant Increase 
in the Probability or Consequences of an Accident Previously Evaluated

    The proposed changes allow 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, these changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

Criterion 2: The Proposed Changes Do Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed). 
Allowing delay times for entering a supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns. Thus, these changes do not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.

Criterion 3: The Proposed Changes Do Not Involve a Significant 
Reduction in the Margin of Safety

    The proposed changes allow a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low-probability occurrence and the 
overall TS system safety function would still be available for the 
vast majority of anticipated challenges. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in NRC Regulatory Guide 1.177. A bounding risk 
assessment was performed to justify the proposed TS changes. The 
application of LCO 3.0.8 is predicated upon the licensee's 
performance of a risk assessment and management of plant risk [which 
is required by the proposed TS 3.0.8]. The net change to the margin 
of safety is insignificant. Therefore, these changes do not involve 
a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael Markley.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: September 17, 2008.
    Description of amendment request: The proposed change will revise 
the Operating License to modify Note 2 of Waterford 3 Technical 
Specification Table 4.3-1. The licensee stated that the proposed change 
will result in the addition of conservatism to Core Protection 
Calculator (CPC) power indications when calibrations are required in 
certain conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?

[[Page 65696]]

    Response: No.
    The proposed change will redefine the tolerance band allowed for 
the Reactor Protection System (RPS) linear power, Core Protection 
Calculator (CPC) [Delta]T [Delta Temperature] power, and CPC neutron 
flux power signals, and clarify the intent of the calibration 
requirements for CPC power indications when at less than 15% 
[percent] power, and specify that adjustment limits are percentages 
of RATED THERMAL POWER instead of percentages of current power. 
Redefining the tolerance band is in conformance with the safety 
analysis. The consequences of an accident will be in conformance 
with the safety analysis.
    Clarifying the intent of there being no calibration requirements 
for CPC power indications when at less than 15% power is essentially 
editorial. At this low power level, CPC calculations compensate for 
any potential de-calibration. Specifying that adjustment limits are 
percentages of RATED THERMAL POWER instead of percentages of current 
power is essentially editorial. This change is made to avoid 
confusion in interpreting the requirements. This amendment request 
does not change the design, analysis or operation of any plant 
systems or components.
    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 to Technical Specification power calibration 
tolerance limits is in conformance with the safety analysis. This 
amendment request does not change the design, analysis or operation 
of any plant systems or components. CPC's cannot cause an accident, 
and this change will not create the possibility of a new or 
different type of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to Technical Specification power calibration 
tolerance limits is in conformance with the safety analysis. This 
proposed change maintains the margin of safety for design basis 
events. Therefore, this 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: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: September 17, 2008.
    Description of amendment request: Entergy has proposed to add a 
license condition on the extension of the reactor vessel inservice 
inspection interval. This proposed license condition is the result of a 
condition in the Nuclear Regulatory Commission (NRC) safety evaluation, 
issued by letter dated May 8, 2008, on topical report WCAP-16168-NP-A, 
Revision 2, ``Risk-Informed Extension of the Reactor Vessel In-Service 
Inspection [ISI] Interval,'' dated June 8, 2008. The ISI interval 
extension part of a relief request is being separately evaluated by NRC 
and independent of this amendment request.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will revise the license to require the 
submission of information and analyses to the NRC following 
completion of each ASME Code, Section XI, Category B-A and B-D 
reactor vessel weld inspection. The extension of the ISI interval 
from 10 to 20 years is being evaluated as part of the relief request 
independent from this license change. Submission of the information 
and analyses are administrative in nature and has no impact on any 
plant configuration or system performance relied upon to mitigate 
the consequences of an accident.
    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 a physical alteration of 
any SSC or change the way any SSC is operated. The proposed addition 
of the license condition has no impact on any plant configurations 
or on system performance that is relied upon to mitigate the 
consequences of an accident. The license condition is administrative 
in nature and does not result in a change to the physical plant or 
to the modes of operation defined in the facility license. Entergy 
has demonstrated that the Limitations and Conditions associated with 
the NRC SE will be met.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The addition of the license condition is administrative in 
nature and has no impact on plant operation or equipment or on any 
margin of safety. The license condition to submit information and 
analyses is an administrative tool to assure the NRC has the ability 
to independently review information developed by the Licensee.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment request: June 27, 2007, as supplemented on 
September 4, 2008.
    Description of amendment request: The proposed amendment request 
dated June 27, 2008, would revise Technical Specifications (TS) 
Surveillance Requirements 3.8.1.2, 8, 12, 13, 16, and 19, changing the 
steady state frequency and voltage of all diesel generators (DGs) from 
the currently allowed frequency range of 59.4-61.2 Hz to 59.4-60.5 Hz 
(i.e., a decrease of the upper limit, resulting in narrowing of the 
current range). The licensee stated that the current frequency range is 
nonconservative and could result in undesirable effects such as 
centrifugal charging pump motor brake horsepower exceeding its 
nameplate maximum horsepower, and overloading the DGs. The Commission 
previously noticed this proposed amendment request on August 14, 2007 
(72 FR 45458).
    The scope of the June 27, 2008, proposed amendment request was 
expanded as described in a supplemental letter dated September 4, 2008. 
The expanded scope would revise (1) TS Surveillance Requirements 
3.8.1.8, 13, 16, and 22, changing the minimum voltage and frequency 
that

[[Page 65697]]

the DGs must achieve within 10 seconds after starting from >= 3740 
Volts (V) to >= 3910 V and >= 58.8 Hz to >= 59.4 Hz, respectively, and 
(2) TS Surveillance Requirement 3.8.1.10, changing the maximum DG 
frequency allowed to occur within 2 seconds following a load rejection 
of the single largest post-accident load from <= 61.2 Hz to <= 60.5 Hz. 
The changes proposed by the supplement indirectly affect TS 3.8.2.1 
which requires that TS Surveillance Requirements 3.8.1.8, 10, and 16 be 
met.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration. The 
NRC staff has performed its own analysis, which is presented below:

    (1) Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The more restrictive transient voltage and frequency limits 
ensures that the equipment powered from the DGs will function as 
designed to mitigate an accident as described in the Update Final 
Safety Analysis Report (UFSAR). The DGs and the equipment they power 
are part of the systems required to mitigate accidents; no accident 
analyzed in the UFSAR is initiated by mitigation equipment. 
Therefore, the proposed change to the allowed frequency range of the 
DGs will not have any impact on the probability of an accident 
previously evaluated. Furthermore, other than requiring more 
restrictive transient voltage and frequency limits of DGs, there is 
no other design or operational change. Therefore, the proposed 
change does not increase the probability of malfunction of the DGs 
or the equipment they power.
    The more restrictive DG transient voltage and frequency limits 
will ensure that the equipment powered by the DGs will perform as 
originally designed and analyzed to mitigate the consequences of any 
accident described in the UFSAR. Therefore, the proposed change does 
not involve a significant increase in the consequences of an 
accident previously evaluated in the UFSAR.
    (2) Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    There is no design change associated with the proposed 
amendment. Making an existing DG requirement more restrictive alone 
will not alter plant configuration because no new or different type 
of equipment will be installed, and because no methods governing 
plant operation will be changed. The proposed change to transient 
voltage and frequency limits will not have any effect on the 
assumptions of accident scenarios previously made in the UFSAR. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    Despite the proposed change to the DG transient voltage and 
frequency limits, the DGs and equipment powered by the DGs will 
continue to perform as originally designed, and originally analyzed 
in the UFSAR. There is no associated change to the methods and 
assumptions used to analyze DG performance. The proposed change will 
maintain the required function of the DGs and the equipment powered 
by the DGs to ensure that operation of structures, systems, or 
components is as currently set forth in the UFSAR. Therefore, the 
proposed change does not involve a significant reduction in the 
margin of safety.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on its own analysis, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the proposed amendment involves no 
significant hazards consideration.
    Attorney for licensee: James M. Petro, Jr., One Cook Place, 
Bridgman, MI 49106.
    NRC Branch Chief: Lois M. James.

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

    Date of amendment request: July 2, 2008.
    Description of amendment request: The proposed amendment would 
correct several typographical errors and make administrative 
clarifications to the Technical Specifications (TS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes correct typographical and administrative 
errors, or make clarifications that more accurately reflect TS 
requirements. Administrative and editorial changes such as these are 
not an initiator of any accident previously evaluated. As a result, 
the probability of an accident previously evaluated is not affected. 
The consequences of an accident with the incorporation of these 
administrative and editorial changes are no different than the 
consequences of the same accident without these changes. As a 
result, the consequences of an accident previously evaluated are not 
affected by these changes.
    The proposed changes do not alter or prevent the ability of 
structures, systems, and components from performing their intended 
function to mitigate the consequences of an initiating event within 
the assumed acceptance limits. The proposed changes do not affect 
the source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of an 
accident previously evaluated.
    Further, the proposed changes do not increase the types or 
amounts of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures. The proposed changes are consistent with the 
safety analysis assumptions and resultant consequences. Therefore, 
the proposed changes do not involve an increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. The proposed changes do not alter any assumptions made in 
the safety analysis. Therefore, the proposed changes do not create 
the possibility of a new or different accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes consist of administrative and editorial 
changes to correct typographical or administrative errors and 
oversights or clarify the meaning of the TS. The changes do not 
alter the manner in which safety limits, limiting safety system 
settings or limiting conditions for operation are determined. The 
safety analysis acceptance criteria are not affected by these 
changes. The proposed changes will not result in plant operation in 
a configuration outside of the design basis. Therefore, the proposed 
changes do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 
K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant 
(WBN), Unit 1, Rhea County, Tennessee

    Date of amendment request: September 18, 2008.
    Description of amendment request: The proposed amendment would 
revise technical specification (TS) 3.8.7,

[[Page 65698]]

``Inverters--Operating.'' The current TS requires one inverter for each 
of the four channels. The proposed amendment would revise TS 3.8.7 to 
require two inverters for each of the four channels.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed revisions to WBN's Vital AC [alternating current] 
Power System do not alter the safety functions of the Vital 
Inverters or the Unit 1 and Unit 2 120V [volt] AC Vital Instrument 
Power Boards. The initial conditions for the DBAs [design-basis 
accidents] defined in the WBN UFSAR [Updated Final Safety Analysis 
Report] assume the ESF [engineered safety feature] systems are 
operable. The vital inverters are designed to provide the required 
capacity, capability, redundancy, and reliability to ensure the 
availability of necessary power to vital instrumentation so that the 
fuel, reactor coolant system, and containment design limits are not 
exceeded. Separating the Unit 2 loads from the Unit 1 inverters does 
not alter the accident analyses. Design calculations document that 
the inverters have adequate capacity to support the loads required 
for Unit 1 operation and no changes are proposed that will impact 
the separation of the Vital AC Power System.
    The inverters and the associated 120V AC Vital Instrument Power 
Boards are utilized to support instrumentation that monitor critical 
plant parameters to aid in the detection of accidents and to support 
the mitigation of accidents, but are not considered to be an 
initiator of design basis accidents. Based on this and the preceding 
information, the proposed amendment 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.
    When implemented, the proposed TS amendment will allow the Unit 
2 Vital Instrument Power Boards to receive their UPS 
[uninterruptible power supply] power from new Unit 2 inverters. 
Calculations have verified that the loads will not affect the 
ability of the inverters to perform their intended safety functions. 
In addition, the inverters and the 120V AC Vital Instrument Power 
Boards are not considered to be an initiator of a DBA. These 
components provide power to instrumentation that supports the 
identification and mitigation of accidents as well as system control 
functions during normal plant operations. The functions of the 
inverters are not altered by the proposed TS change and will not 
create the possibility of a new or different accident. Further, the 
separation of the Unit 2 loads from the Unit 1 inverters is the 
principal change to the inverter system, and this change is bounded 
by previously evaluated accident analyses. 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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The plant setpoints and limits that are utilized to ensure safe 
operation and detect accident conditions are not impacted by the 
proposed TS amendment. The inverters and the 120V Vital Instrument 
Power Boards will continue to provide reliable power to safety-
related instrumentation for the identification and mitigation of 
accidents and to support plant operation. Therefore, the margin of 
safety is not reduced.
    Based on the above, TVA concludes that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: L. Raghavan.

Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of amendment request: September 18, 2008.
    Description of amendment request: The proposed amendment would 
revise technical specification (TS) Table 3.3.2-1, ``Engineered Safety 
Feature Actuation System Instrumentation,'' to modify Mode 1 and 2 
Applicability for Function 6.e, and would revise limiting condition for 
operation (LCO) 3.3.2, Condition J.
    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 design basis events which impose [auxiliary feedwater] AFW 
safety function requirements are loss of normal main feedwater, main 
feed line or main steam line break, loss of offsite power (LOOP), 
and small break loss of coolant accident. These design bases event 
evaluations assume actuation of the AFW due to LOOP signal, low-low 
steam generator level or a safety injection signal. The anticipatory 
AFW auto-start signals from the turbine driven main feedwater 
(TDMFW) pumps are not credited in any design basis accidents and 
are, therefore, not part of the primary success path for postulated 
accident mitigation as defined by 10 CFR 50.36(c)(2)(ii), Criterion 
3. Modifying Mode 1 and 2 Applicability for this function will not 
impact any previously evaluated design basis accidents. 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.
    This TS change allows for an operational allowance during Mode 1 
and 2 for placing TDMFW pumps in service or securing TDMFW pumps. 
This change involves an anticipatory AFW auto-start function that is 
not credited in the accident analysis. Since this change only 
affects the conditions at which this auto-start function needs to be 
operable and does not affect the function that actuates AFW due to 
loss of offsite power, low-low steam generator level or a safety 
injection signal, it will not be an initiator to 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 TS change involves the automatic start of the AFW pumps due 
to trip of both TDMFW pumps, which is not an assumed start signal 
for design basis events. This change does not modify any values or 
limits involved in a safety related function or accident analysis. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, TVA concludes that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: L. Raghavan.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant 
(WBN), Unit 1, Rhea County, Tennessee

    Date of amendment request: September 19, 2008.

[[Page 65699]]

    Description of amendment request: The proposed amendment would 
modify the WBN Final Safety Analysis Report (FSAR) by requiring an 
inspection of the ice condenser within 24 hours of experiencing a 
seismic event greater than or equal to an Operating Basis Earthquake 
(OBE) within the five week period after ice basket replenishment has 
been completed.
    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 analyzed accidents of consideration in regard to changes 
potentially affecting the ice condenser are a loss of coolant 
accident and a steam or feedwater line break inside Containment. The 
ice condenser is an accident mitigator and is not postulated as 
being the initiator of a LOCA [loss-of-coolant accident] or HELB 
[high energy line break]. The ice condenser is structurally designed 
to withstand a Safe Shutdown Earthquake plus a Design Basis Accident 
and does not interconnect or interact with any systems that 
interconnect or interact with the Reactor Coolant, Main Steam, or 
Feedwater systems. Because the proposed changes do not result in, or 
require any physical change to the ice condenser that could 
introduce an interaction with the Reactor Coolant, Main Steam, or 
Feedwater systems, there can be no change in the probability of an 
accident previously evaluated.
    Under the proposed change, there is some finite probability 
that, within 24 hours following a seismic disturbance, a LOCA or 
HELB in Containment could occur within five weeks of the completion 
of ice basket replenishment. However, several factors provide 
defense-in-depth and tend to mitigate the potential consequences of 
the proposed change.
    Design basis accidents are not assumed to occur simultaneously 
with a seismic event. Therefore, the coincident occurrence of a LOCA 
or HELB with a seismic event is strictly a function of the combined 
probability of the occurrence of independent events, which in this 
case is very low. Based on the Probabilistic Risk Assessment model 
and seismic hazard analysis, the combined probability of occurrence 
of a seismic disturbance greater than or equal to an OBE during the 
5 week period following ice replenishment coincident with or 
subsequently followed by a LOCA or HELB during the time required to 
perform the proposed inspection (24 hours) and if required by 
Technical Specifications, complete Unit shutdown (37 hours), is less 
than 3.7E-09 for WBN. This probability is well below the threshold 
that is typically considered credible.
    Even if ice were to fall from ice baskets during a seismic event 
occurring coincident with or subsequently followed by an accident, 
the ice condenser would be expected to perform its intended safety 
function. Due to the ice servicing methodology utilized by WBN, the 
relatively small amount of ice that may potentially fallout from the 
ice baskets to the floor behind the lower inlet doors during the 
seismic event is such that complete blockage of flow into the ice 
condenser is not credible during a LOCA or HELB.
    Based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences. The ice 
condenser is expected to perform its intended safety function under 
all circumstances following a LOCA or HELB in Containment.
    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 provides an alternate methodology to confirm 
the ice condenser lower inlet doors are capable of opening if a 
seismic event occurs within five weeks of ice basket replenishment. 
As previously discussed, the ice condenser is not postulated as an 
initiator of any design basis accident. The proposed change does not 
impact any plant system, structure, or component that is an accident 
initiator. The proposed change does not involve any hardware changes 
to the ice condenser or other changes that could create new accident 
mechanisms. Therefore, there can be no new or different accidents 
created from those previously identified and evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the Reactor Coolant system, and the Containment 
system. The performance of the fuel cladding and the Reactor Coolant 
system will not be impacted by the proposed change.
    The requirement to inspect the ice condensers within 24 hours of 
experiencing seismic activity greater than or equal to an OBE during 
the five (5) week period following the completion of ice basket 
replenishment will confirm whether the ice condenser lower inlet 
doors are capable of opening. This inspection will either confirm 
that the ice condenser doors remain fully capable of performing 
their intended safety function under credible circumstances or that 
a Unit shutdown is required.
    The ice condenser has reasonable assurance of performing its 
intended function during the highly unlikely scenario in which a 
postulated accident (LOCA or HELB) occurs coincident with or 
subsequently following a seismic event.
    The proposed change affects the assumed timing of a postulated 
seismic and design basis accident applied to the ice condenser and 
provides an alternate methodology in confirming the ice condenser 
lower inlet doors are capable of opening. As previously discussed, 
the combined probability of occurrence of a LOCA or HELB and a 
seismic disturbance greater than or equal to an OBE during the 
``period of potential exposure'' is less than 3.7E-09 for WBN. This 
probability is well below the threshold that is considered credible.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety. The WBN ice condenser will 
perform its intended safety function under credible circumstances.
    The changes proposed in this LAR [license amendment request] do 
not make any physical alteration to the ice condensers, nor does it 
affect the required functional capability of the ice condenser in 
any way. The intent of the proposed change to the FSAR is to 
eliminate an overly restrictive waiting period prior to Unit ascent 
to power operations following the completion of ice basket 
replenishment. The required inspection of the ice condenser 
following a seismic event greater than or equal to an OBE will 
confirm whether the ice condenser lower inlet doors will continue to 
fully perform their safety function as assumed in the WBN safety 
analyses.
    Thus, it can be concluded that the proposed change does not 
involve a significant reduction in the margin of safety.
    Based on the above, TVA concludes that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: April 2, 2008.
    Description of amendment request: The proposed change revises 
Technical Specification (TS) Section 5.0, ``Design Features,'' to 
delete certain design details and descriptions included in TS 5.0 that 
are already contained in the Updated Final Safety Analysis Report

[[Page 65700]]

(UFSAR), or are redundant to existing TS requirements, and are not 
required to be included in the TSs pursuant to Title 10 of the Code of 
Federal Regulations (10 CFR), Part 50, Section 50.36(d)(4). The 
proposed change also revises the format of, and incorporates design 
descriptions into, TS 5.0 consistent with Nuclear Regulatory Commission 
(NRC) policy and NUREG-1431, ``Standard Technical Specifications 
Westinghouse Plants, Revision 3.0,'' to the extent practical. An 
editorial change is also proposed to address a minor TS discrepancy 
introduced by a previous license amendment. More specifically, the 
proposed change includes removing Section 5.2, ``Containment,'' from 
the TSs in its entirety. This section contains the minimum spray flows 
for the Containment Spray (CS) and Recirculation Spray (RS) Subsystems. 
The proposed change also removes the statement describing how draining 
of the spent fuel pool is prevented, and includes a statement in the TS 
that would limit draining the spent fuel pool below the elevation of 41 
feet, 2 inches mean sea level. Additionally, the licensee proposes to 
incorporate the spent fuel pool storage capacity of 1044 assemblies 
into the TSs. This limit was previously established by Amendment Nos. 
37 and 36 to Surry Power Station, Unit Nos. 1 and 2, respectively.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration. The 
NRC staff has performed its own analysis, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to Section 5.0, ``Design Features,'' 
removes certain details from the TSs that are not required to be 
maintained in the TSs by 10 CFR 50.36(d)(4), or are adequately 
controlled by other existing TSs, incorporates previously approved 
TS limits that meet the 10 CFR 50.36(d)(4) inclusion criteria, and 
revises the TSs for consistency with NUREG-1431. An additional 
change addresses a minor editorial discrepancy introduced by a 
previous amendment. The minimum spray flow values for the CS and RS 
Subsystems are removed, but operability and performance of both 
subsystems are adequately controlled by existing TSs ensuring they 
will continue to perform their design functions. The proposed 
changes remove the statement describing how draining of the spent 
fuel pool is prevented (does not meet the criteria of 10 CFR 
50.36(d)(4)for inclusion in the TSs) and includes a statement in the 
TS that would limit draining the spent fuel pool below the elevation 
of 41 feet, 2 inches mean sea level (as analyzed in the UFSAR and 
consistent with the content and format of NUREG-1431). The proposed 
change incorporates the spent fuel pool storage capacity of 1044 
assemblies into the TSs. This limit was evaluated in previously 
approved Amendment Nos. 37 and 36 to Surry Power Station, Unit Nos. 
1 and 2, respectively. The proposed changes are considered 
administrative in nature and do not affect initiators of previously 
analyzed events or assumed mitigation of accident or transient 
events. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    There is no physical alteration of the plant (no new or 
different type of equipment will be installed) associated with the 
proposed amendment. The proposed changes will not have any effect on 
the assumptions of accident scenarios previously made in the UFSAR. 
The proposed changes do not alter or prevent the ability of 
structures, systems, and components to perform their intended 
function to mitigate the consequences of an initiating event. The 
proposed changes are considered administrative in nature. Therefore, 
the proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    Response: No.
    The spent fuel pool and the CS and RS Subsystems will continue 
to perform as designed and analyzed in the UFSAR. There is no 
associated change to the methods and assumptions used to analyze 
their performance. Their required function will be maintained as 
currently set forth in the UFSAR and existing TSs. The proposed 
changes do not result in plant operation in a configuration outside 
the design basis. The proposed changes do not adversely affect 
systems that respond to safely shutdown the plant and to maintain 
the plant in a safe shutdown condition. The dose analysis is also 
not affected. The proposed changes are considered administrative in 
nature and do not alter the manner in which safety limits, limiting 
safety system settings or limiting conditions for operation are 
determined. 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 
its own analysis, 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, Esq., Senior Counsel, 
Dominion Resources Services, Inc., 120 Tredegar Street, RS-2 Richmond, 
VA 23219.
    NRC Branch Chief: Melanie C. Wong.

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

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

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-410, Nine 
Mile Point Nuclear Station Unit No. 2, Oswego County, New York

    Date of amendment request: July 30, 2007, as supplemented on April 
7 and September 8, 2008.
    Description of amendment request: This amendment would modify 
Technical Specification 3.7.3, ``Control Room Envelope Air Conditioning 
(AC) System,'' by adding an Action Statement to the Limiting Conditions 
for Operation. The new Action Statement allows a finite time to restore 
one control room envelope AC subsystem to operable status and requires 
verification that the control room temperature remains <90 [deg]F every 
4 hours.
    Date of publication of individual notice in Federal Register: (73 
FR 55166) September 24, 2008.
    Expiration date of individual notice: November 23, 2008.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating

[[Page 65701]]

License, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing in connection with these actions was 
published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (First Floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: October 18, 2007, as supplemented by 
letter dated July 3, 2008.
    Description of amendment request: The amendment changed the Oyster 
Creek Technical Specifications Section 4.5.M.1.e.1 regarding the 
mechanical snubber functional test acceptance test acceptance criteria. 
Specifically, the change replaced the snubber breakaway test with the 
drag force test.
    Date of issuance: October 10, 2008.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 270.
    Facility Operating License No. DPR-16: The amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: June 17, 2008 (73 FR 
34339). The supplement dated July 3, 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 amendment is contained 
in a Safety Evaluation dated October 10, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 29, 2007.
    Brief description of amendment: The amendment consists of changes 
to Technical Specification Section 3.6.8, ``Isolation Valve Seal Water 
(IVSW) System.'' The amendment revises Surveillance Requirements (SR) 
3.6.8.2 and 3.6.8.6 related to IVSW tank volume and header flow rates. 
Specifically, the change clarifies the wording of SR 3.6.8.2, and 
revises SR 3.6.8.6 to provide a total flow rate limit from all four 
headers in place of the individual header limits.
    Date of issuance: October 3, 2008.
    Effective date: Effective as of the date of issuance and shall be 
implemented within 60 days.
    Amendment No. 220.
    Renewed Facility Operating License No. DPR-23: The amendment 
revises the technical specifications and facility operating license.
    Date of initial notice in Federal Register: January 15, 2008 (73 FR 
2548). The Commission's related evaluation of the amendment is 
contained in a safety evaluation dated October 3, 2008.
    No significant hazards consideration comments received: No.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602-1551.
    NRC Branch Chief: Thomas H. Boyce.

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

    Date of application for amendment: July 17, 2007, as supplemented 
by letters dated August 7, 2007, and September 2, 2008.
    Brief description of amendment: The amendment added a new license 
condition (43) on the control room envelope habitability program, 
revised Technical Specification (TS) requirements related to the 
control room envelope habitability in TS 3.7.3, ``Control Room Fresh 
Air (CRFA) System,'' and added the new TS 5.5.13, ``Control Room 
Envelope Habitability Program.''
    Date of issuance: October 14, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No: 178.
    Facility Operating License No. NPF-29: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: September 25, 2007 (72 
FR 54473). The supplemental letters dated August 7, 2007, and September 
2, 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 amendment is contained 
in a Safety Evaluation dated October 14, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 5, 2007, as 
supplemented by letters dated July 21 and August 28, 2008.
    Brief description of amendment: The amendment changed Technical 
Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR),'' to 
add a reference to an analytical method that will be used to determine 
core operating limits. The new reference, NEDC-33383P, ``GEXL97 
Correlation Applicable to ATRIUM-10 Fuel,'' will allow the licensee to 
use a Global Nuclear Fuel method to determine fuel assembly critical 
power of AREVA ATRIUM-10 fuel. Additionally, the amendment made an 
administrative change to an existing reference in TS 5.6.5.
    Date of issuance: October 16, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 179.

[[Page 65702]]

    Facility Operating License No. NPF-29: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 31, 2007 (72 
FR 74358). The supplements dated July 21 and August 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 amendment is contained 
in a Safety Evaluation dated October 16, 2008.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of application for amendment: December 27, 2007, as 
supplemented by letter dated July 14, 2008.
    Brief description of amendment: The amendment revised Technical 
Specifications (TS) Section 3.4.1, ``RCS [Reactor Coolant System] 
Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) 
Limits,'' to increase the minimum RCS flow rate from 341,100 to 354,000 
gallons per minute. The increased flow rate supports a new analysis of 
a large break loss-of-coolant accident (LOCA). The new analysis is 
performed using an NRC-approved methodology set forth in Westinghouse 
Topical Report WCAP-16009-P-A, ``Realistic Large-Break LOCA Evaluation 
Methodology Using the Automated Statistical Treatment of Uncertainty 
Method (ASTRUM).'' This methodology will be endorsed and reflected by a 
revision to TS Section 5.6.5, ``Core Operating Limits Report (COLR).''
    Date of issuance: October 17, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 306.
    Facility Operating License No. DPR-58: Amendment revised the 
Renewed Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 29, 2008 (73 FR 
5223). The supplement dated July 14, 2008, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staffs 
original proposed no significant hazards consideration determination 
published in the Federal Register on January 29, 2008.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 17, 2008.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 28, 2008, as supplemented by 
letters dated July 28 and September 25, 2008.
    Brief description of amendments: The amendments revised (1) Action 
5 in Technical Specification (TS) 3.3.1, ``Reactor Trip 
Instrumentation,'' for one inoperable channel of extended range neutron 
flux instrumentation and (2) Action c in TS 3.4.1.4.2, ``Reactor 
Coolant System, Cold Shutdown--Loops Not Filled.'' The amendments do 
not complete the Nuclear Regulatory Commission staff's review of the 
licensee's proposed TS changes in the application. The remaining 
proposed TS changes to Action 5 will be addressed in a future letter to 
the licensee.
    Date of issuance: October 16, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: Unit 1-187; Unit 2-174.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: March 25, 2008 (73 FR 
15788). The supplemental letters dated July 28 and September 25, 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 16, 2008.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendment: October 26, 2007.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to adopt TS Task Force (TSTF) Change Traveler TSTF-
448, Revision 3, ``Control Room Envelope Habitability.''
    Date of issuance: October 8, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 70.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications and License.
    Date of initial notice in Federal Register: August 29, 2008 (73 FR 
51014). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 8, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 24, 2007, as 
supplemented by letter dated August 7, 2008.
    Brief description of amendment: The amendments change Technical 
Specifications (TSs) Limiting Condition for Operations (LCO) 3.8.7 and 
3.8.9, pertaining to electrical power systems and distribution 
associated with the 120 Volt AC vital bus inverters. The TS changes are 
intended to support operability of components shared between Unit 1 and 
Unit 2. The proposed changes will add new Conditions, Required Action 
statements and Completion Times for LCO 3.8.7 and LCO 3.8.9 to address 
shared components.
    Date of issuance: October 9, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 253, 234.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
change the licenses and the technical specifications.
    Date of initial notice in Federal Register: December 18, 2007 (72 
FR 71717). The supplement dated August 7, 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 9, 2008.
    No significant hazards consideration comments received: No.

[[Page 65703]]

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room (PDR), located at One White Flint North, Public File Area 01F21, 
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly 
available records will be accessible from the Agencywide Documents 
Access and Management System's (ADAMS) Public Electronic Reading Room 
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems 
in accessing the documents located in ADAMS, contact the PDR Reference 
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. Within 60 days after the date 
of publication of this notice, 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, and electronically on the Internet at the NRC Web 
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are 
problems in accessing the document, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a 
request for a hearing or petition for leave to intervene is filed by 
the above date, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: ( 1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases

[[Page 65704]]

for the contention and a concise statement of the alleged facts or 
expert opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion. The 
petition must include sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or 
fact.\1\ Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner/requestor 
who fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
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    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
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    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    A request for hearing or a petition for leave to intervene must be 
filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve documents over the internet 
or in some cases to mail copies on electronic storage media. 
Participants may not submit paper copies of their filings unless they 
seek a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
[email protected], or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer\TM\ to 
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ is free and is available 
at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. 
Information about applying for a digital ID certificate is available on 
NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, in accordance 
with 10 CFR 2.302(g), with their initial paper filing requesting 
authorization to continue to submit documents in paper format. Such 
filings must be submitted by: (1) First class mail addressed to the 
Office of the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff; or (2) courier, express mail, or expedited 
delivery service to the Office of the Secretary, Sixteenth Floor, One 
White Flint North, 11555 Rockville, Pike, Rockville, Maryland 20852, 
Attention: Rulemaking and Adjudications Staff. Participants filing a 
document in this manner are responsible for serving the document on all 
other participants. Filing is considered complete by first-class mail 
as of the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the petition 
and/or request should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later 
than

[[Page 65705]]

11:59 p.m. Eastern Time on the due date.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, 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.

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

    Date of amendment request: October 13, 2008.
    Description of amendment request: The amendment revised the 
surveillance frequency for Technical Specification Surveillance 
Requirement 3.8.1.10 for the endurance test conducted every 2 years on 
the diesel generators.
    Date of issuance: October 20, 2008.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 255.
    Facility Operating License No. DPR-26: Amendment revises the 
Technical Specifications and License.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. Public notice of the proposed amendment was 
published in The Journal News newspaper, located in Westchester County, 
New York on October 17 and October 18, 2008. The notice provided an 
opportunity to submit comments on the Commission's proposed NSHC 
determination. No comments have been received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a safety evaluation dated October 20, 2008.
    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.

    Dated at Rockville, Maryland, this 24th day October 2008.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E8-25882 Filed 11-3-08; 8:45 am]
BILLING CODE 7590-01-P