[Federal Register Volume 72, Number 242 (Tuesday, December 18, 2007)]
[Notices]
[Pages 71703-71719]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-24284]


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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 November 22, 2007, to December 5, 2007. The 
last biweekly notice was published on December 4, 2007 (72 FR 68206).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area O1F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, 
person(s) may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
via electronic submission through the NRC E-Filing system for a hearing 
and a petition for leave to intervene. Requests for a hearing and a 
petition for leave to intervene shall be filed in accordance with the 
Commission's ``Rules of Practice for Domestic Licensing Proceedings'' 
in 10 CFR Part 2. Interested person(s) should consult a current copy of 
10 CFR 2.309, which is available at the Commission's PDR, located at 
One White Flint North, Public File Area O1F21, 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

[[Page 71704]]

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). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM 
to access the Electronic Information Exchange (EIE), a component of the 
E-Filing system. The Workplace Forms ViewerTM is free and is 
available at 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.

[[Page 71705]]

    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 (CPS), DeWitt County, Illinois

    Date of amendment request: September 27, 2007.
    Description of amendment request: The proposed amendment would 
modify technical specification (TS) by relocating references to 
specific American Society for Testing and Materials (ASTM) standards 
for fuel oil testing to licensee-controlled documents. In the 
referenced letter, AmerGen (the licensee) previously received approval 
for a change to the Unit No. 1, CPS TS that added the water and 
sediment content test as alternative criteria to the ``clear and 
bright'' acceptance test for new fuel oil.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The proposed changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. Requirements to perform testing in 
accordance with applicable ASTM standards are retained in the TS as 
are requirements to perform surveillances of both new and stored 
diesel fuel oil. 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. In addition, the 
``clear and bright'' test used to establish the acceptability of new 
fuel oil for use prior to addition to storage tanks has been 
expanded to recognize more rigorous testing of water and sediment 
content. Relocating the specific ASTM standard references from the 
TS to a licensee-controlled document and allowing a water and 
sediment content test to be performed to establish the acceptability 
of new fuel oil will not affect nor degrade the ability of the 
emergency diesel generators (DGs) to perform their specified safety 
function. Fuel oil quality will continue to meet ASTM requirements.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not adversely 
affect the ability of structures, systems, and components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits.
    The proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated. 
Further, the proposed changes do not increase the types and amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. In addition, the ``clear and bright'' 
test used to establish the acceptability of new fuel oil for use 
prior to addition to storage tanks has been expanded to allow a 
water and sediment content test to be performed to establish the 
acceptability of new fuel oil. The changes do not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. The requirements retained in the TS continue to require 
testing of the diesel fuel oil to ensure the proper functioning of 
the DGs. Therefore, the changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes relocate the specific ASTM standard 
references from the Administrative Controls Section of TS to a 
licensee-controlled document. Instituting the proposed changes will 
continue to ensure the use of applicable ASTM standards to evaluate 
the quality of both new and stored fuel oil designated for use in 
the emergency DGs. Changes to the licensee-controlled document are 
performed in accordance with the provisions of 10 CFR 50.59. This 
approach provides an effective level of regulatory control and 
ensures that diesel fuel oil testing is conducted such that there is 
no significant reduction in a margin of safety.
    The ``clear and bright'' test used to establish the 
acceptability of new fuel oil for use prior to addition to storage 
tanks has been expanded to allow a water and sediment content test 
to be performed to establish the acceptability of new fuel oil. The 
margin of safety provided by the DGs is unaffected by the proposed 
changes since there continue to be TS requirements to ensure fuel 
oil is of the appropriate quality for emergency DG use. The proposed 
changes provide the flexibility needed to improve fuel oil sampling 
and analysis methodologies while maintaining sufficient controls to 
preserve the current margins of safety.

    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 A. Gibbs.

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: November 8, 2007.
    Description of amendments request: The amendment would clarify the 
Technical Specification definitions for Channel Calibration and Channel 
Functional Test. The proposed amendments would incorporate Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler, TSTF-205-A, ``Revision of Channel Calibration, Channel 
Functional Test, and Related Definitions,'' Revision 3, dated July 31, 
2003.
    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. Would not involve a significant increase in the probability 
or consequences of any accident previously evaluated.
    The proposed change clarifies the Technical Specification 
requirements for performance of channel calibrations and channel 
functional tests. Specifically, the proposed change incorporates the 
Nuclear Regulatory Commission-approved Technical Specification Task 
Force Standard Technical Specification Change Traveler, TSTF-205-A, 
``Revision of Channel Calibration, Channel Functional Test, and 
Related Definitions,'' Revision 3, dated July 31, 2003. The change 
does not adversely affect the performance or effectiveness of 
required testing, as testing appropriate to the associated 
Surveillance Requirements will continue to be performed. The 
proposed change does not have a detrimental impact on the condition 
or performance of any plant structure, system, or component that 
could initiate an analyzed

[[Page 71706]]

event. Therefore, the probability of an accident previously 
evaluated is not significantly increased.
    The equipment being calibrated or tested is still required to be 
operable and capable of performing the accident mitigation functions 
assumed in the accident analysis. As a result, the consequences of 
any accident previously evaluated are not significantly affected.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The scope of the proposed change is limited to the clarification 
of existing calibration and test requirements. As such, 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.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. The proposed change will not involve a significant reduction 
in [a] margin of safety.
    The margin of safety in this case is the verification of 
instrument channel operability. The proposed change clarifies 
requirements for the performance of channel calibrations and channel 
functional tests. Specifically, the proposed change incorporates the 
Nuclear Regulatory Commission-approved Technical Specification Task 
Force Standard Technical Specification Change Traveler, TSTF-205-A, 
``Revision of Channel Calibration, Channel Functional Test, and 
Related Definitions,'' Revision 3, dated July 31, 2003. No changes 
of setpoints to plant process limits are involved. The surveillance 
requirements, as revised, will continue to ensure that affected 
equipment is tested in a manner that gives confidence that the 
equipment can perform its appropriate safety function.
    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 
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: November 9, 2007.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 3.8.a.7 related to the movement of 
heavy loads over and in the spent fuel pools and would relocate the 
modified requirements to a licensee-controlled document, the Kewaunee 
Power Station Technical Requirements Manual (TRM). The proposed 
amendment is needed to facilitate future spent fuel cask handling 
activities associated with dry cask spent fuel storage. The proposed 
amendment would incorporate the use of a single-failure-proof lifting 
system for handling of necessary heavy loads over or in the spent fuel 
pool with irradiated fuel in either the fuel storage racks or in the 
just-loaded spent fuel canister in the spent fuel pool. The proposed 
modified TS 3.8.a.7 would then be relocated to the TRM.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment revises Kewaunee Power Station (KPS) 
heavy load handling Technical Specification (TS) 3.8.a.7 
requirements consistent with modifications to the Auxiliary Building 
(AB) crane and the NRC's [Nuclear Regulatory Commission] current 
guidance for single-failure-proof lifting systems. The proposed 
amendment also relocates the affected heavy load handling-related TS 
to a licensee-controlled document, consistent with the NRC's 
regulations.
    The proposed change to TS 3.8.a.7 permits spent fuel cask 
handling in the spent fuel pool, which is required for loading spent 
fuel for dry storage at the on-site Independent Spent Fuel Storage 
Installation (ISFSI). Proposed TS 3.8.a.7 includes a new requirement 
that the AB crane and associated lifting devices meet the applicable 
single-failure-proof criteria.
    Heavy load handling will continue to be conducted in accordance 
with the KPS heavy load handling program, which meets the NRC's 
guidance in NUREG-0612, as described in this LAR, and as augmented 
by Regulatory Information Summary 2005-25. With the upgrade of the 
AB crane load handling system, drops of heavy loads will not be 
considered credible. Notwithstanding the AB crane upgrade, heavy 
loads will still be prohibited from being suspended over irradiated 
fuel in the spent fuel pool storage racks under the revised 
requirements.
    The previously evaluated cask drop accident is not considered 
credible with the upgraded AB crane because the crane trolley is 
being upgraded to a single-failure-proof design, consistent with 
applicable NRC-endorsed guidance. Lifting devices and interfacing 
lifting points associated with spent fuel cask handling will also be 
designed in accordance with applicable NRC guidance pertaining to 
single-failure-proof lifting systems. The result of these design 
upgrades is that the AB crane will retain the lifted load in the 
event of a single failure in the load path, including a failure of a 
wire rope. In addition, the crane will hold the load and the trolley 
and bridge will be designed to stay on their respective rails during 
a design basis seismic event.
    The relocation of TS 3.8.a.7 to the KPS Technical Requirements 
Manual (TRM) is an administrative change that does not affect plant 
operation or heavy load handling.
    Revised TS 3.8.a.7 and its associated Bases will be relocated to 
the TRM after approval of this amendment request. Changes to the KPS 
TRM are controlled by 10 CFR 50.59. Regulation 10 CFR 50.59 requires 
that NRC approval be obtained prior to any change that would result 
in more than a minimal increase in (1) the frequency of occurrence 
of an accident previously evaluated, (2) likelihood of occurrence of 
a malfunction of a SSC important to safety previously evaluated, or 
(3) consequences of a malfunction of a SSC important to safety 
previously evaluated. Accordingly, upon relocation of the 
requirements of TS 3.8.a.7 and associated Bases to the TRM, 
appropriate control of changes will be maintained, based on the 
criteria in 10 CFR 50.59. Administrative relocation of the 
requirements of TS 3.8.a.7 does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, configuration of KPS or the manner in which it is 
operated.
    Therefore, the proposed change does not significantly increase 
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.
    Heavy load handling will continue to be conducted in accordance 
with the KPS heavy load handling program, which meets the NRC's 
guidance in NUREG-0612, as approved for KPS. Drops of heavy loads 
will continue to be very improbable events and the upgrade of the 
KPS AB crane lifting system to a single-failure-proof design 
provides additional defense-in-depth against such events. 
Notwithstanding the AB crane upgrade, heavy loads will still be 
prohibited from being suspended over irradiated fuel in the spent 
fuel pool storage racks under the revised requirements.
    Heavy load handling operations at KPS will continue to be 
conducted as they currently are and no new heavy load handling 
operations are required as a result of this amendment. The 
previously evaluated cask drop accident is not considered credible 
with the upgraded AB crane because the crane trolley is being 
upgraded to a single-failure-proof design, consistent with 
applicable NRC-endorsed guidance. Lifting devices and interfacing 
lifting points associated with spent fuel cask handling will also be 
designed in accordance with applicable NRC guidance pertaining to 
single-failure-proof lifting systems. The result

[[Page 71707]]

of these design upgrades is that the AB crane will retain the lifted 
load in the event of a single failure in the load path, including a 
failure of a wire rope. In addition, the crane will hold the load 
and the trolley and bridge will be designed to stay on their 
respective rails during a design basis seismic event.
    The relocation of TS 3.8.a.7 to the KPS Technical Requirements 
Manual (TRM) is an administrative change that does not affect plant 
operation or heavy load handling.
    Accordingly, upon relocation of the requirements of TS 3.8.a.7 
and associated Bases to the TRM, appropriate control of changes will 
be maintained, based on the criteria in 10 CFR 50.59. Modification 
of the requirements of TS 3.8.a.7 does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, configuration of KPS or the manner in which it is 
operated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment revises KPS heavy load handling TS 
3.8.a.7 requirements consistent with modifications to the AB crane 
and the NRC's current guidance for single-failure-proof lifting 
systems.
    Heavy load handling will continue to be conducted in accordance 
with the KPS heavy load handling program, which meets the NRC's 
guidance in NUREG-0612, as approved for KPS. Drops of heavy loads 
will continue to be very improbable events and the upgrade of the 
KPS AB crane lifting system to a single-failure-proof design 
provides additional defense-in-depth against such events and an 
increase in overall design margin. Notwithstanding the AB crane 
upgrade, heavy loads will still be prohibited from being suspended 
over irradiated fuel in the spent fuel pool storage racks under the 
revised requirements.
    Further, the relocation of TS 3.8.a.7 to the KPS Technical 
Requirements Manual (TRM) is an administrative change that does not 
affect plant operation or heavy load handling.
    Heavy load handling operations at KPS will continue to be 
conducted as they currently are and no new heavy load handling 
operations are required as a result of this amendment. The 
previously evaluated cask drop accident is less probable with the 
upgraded AB crane because the crane trolley is being upgraded to a 
single-failure-proof design, consistent with applicable NRC-endorsed 
guidance. Lifting devices and interfacing lifting points associated 
with spent fuel cask handling will also be designed in accordance 
with applicable NRC guidance pertaining to single-failure-proof 
lifting systems. The result of these design upgrades is that the AB 
crane will retain the lifted load in the event of a single failure 
in the load path, including a failure of a wire rope. In addition, 
the crane will hold the load and the trolley and bridge will be 
designed to stay on their respective rails during a design basis 
seismic event.
    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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 
120 Tredegar Street, Richmond, VA 23219.
    NRC Acting Branch Chief: Cliff Munson.

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

    Date of amendment request: November 9, 2007.
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Power Station (KPS) Updated Safety Analysis Report 
(USAR) to modify the design and licensing basis for the auxiliary 
building (AB) crane. The proposed amendment would allow the use of a 
methodology for performing the seismic qualification analysis of the 
upgraded crane. The crane is being upgraded to become a single-failure-
proof design. The new methodology includes rolling of the crane bridge 
and trolley wheels during a seismic event.
    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.
    This amendment request pertains solely to an analysis method 
supporting the upgrade of the KPS AB crane from a non-single-
failure-proof design to a single-failure-proof design. The AB crane 
is used to lift and handle loads in the KPS spent fuel pool and 
truck bay areas. The AB crane does not interface with operating 
plant equipment. The design rated load of the AB crane remains the 
same as previously approved. The proposed amendment does not change 
the current heavy load handling practices that are in use at KPS. 
Upgrading the AB crane to a single-failure-proof design will reduce 
the probability of a heavy load drop in the areas where the AB crane 
lifts and handles loads.
    The seismic analysis method proposed for use recognizes the 
inherent propensity for structures not fixed to one another (e.g., 
steel wheels on steel rails) to roll if sufficient lateral force is 
applied to either object. This seismic analysis method is proposed 
for use solely on the AB crane upgrade and not for any other plant 
structures, systems, or components. The recognition of wheel rolling 
between the AB crane trolley and bridge and their respective rails 
reflects the true nature of the installed equipment and its response 
to horizontal forces generated by a seismic event. Consideration of 
rolling reduces the projected analyzed loads on the crane and 
building structures and eliminates the need for unnecessary 
modifications to both.
    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.
    This amendment request pertains to an analysis method supporting 
the upgrade of an existing plant component. Specifically, the 
existing AB crane trolley is being replaced with a state-of-the-art 
design that is single-failure-proof. The AB crane does not interface 
with operating plant equipment. This seismic analysis method is 
proposed for use solely on the AB crane upgrade and not for any 
other plant structures, systems, or components.
    The design rated load of the AB crane remains the same at 125 
tons. This load controls the design and supporting analysis. The 
auxiliary hook design rated load is being increased from 10 tons to 
15 tons. The proposed amendment does not change the currently 
acceptable heavy load handling practices in use at KPS. The number 
and types of lifts made using this crane in support of KPS plant 
operations are not significantly changed from that contemplated 
during original plant licensing. Furthermore, the basic operations 
of the crane (i.e., hoisting and horizontal travel) remain the same, 
although the electronic controls will be upgraded to current 
standards.
    Therefore, the proposed amendment does not create a new or 
different kind of accident from any accident previously evaluated in 
the KPS licensing basis.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Although the proposed change is made specifically to support the 
upgrade of the KPS AB crane from a non-single-failure-proof to a 
single-failure-proof design, the margin of safety under 
consideration in this evaluation is mainly based on that contained 
within the safety analysis (seismic analysis).
    The purpose of this methodology is to determine the stress 
placed on the AB cranes' structural components. The stresses 
determined by this methodology are then compared to the yield 
strength values contained in CMAA-70. If the stresses the structural 
component are analyzed to receive during a postulated seismic event 
are less than the values contained in CMAA-70 the structural 
integrity of the crane is maintained and a suspended load will 
remain suspended during a seismic event. Additional margin has been 
added by reducing the analysis acceptance criteria to 90% of the 
acceptance criteria values contained in CMAA-70, modifying the crane 
support structure

[[Page 71708]]

through additional welds and material, and confirming the bolts are 
of the proper material.
    DEK [Dominion Energy Kewaunee] is modeling the AB crane to roll 
during a seismic event when the postulated forces exceed the brake 
holding force. This provides a more realistic approach because the 
crane trolley is not fixed to the bridge rails. DEK has provided 
additional conservatisms by doubling the calculated force needed to 
overcome the brake holding force.
    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 Acting Branch Chief: Cliff Munson.

Entergy Operations, Inc., Docket Nos. 50-313, Arkansas Nuclear One, 
Unit 1, Pope County, Arkansas

    Date of amendment request: October 22, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specification (TS) to establish more effective and 
appropriate action, surveillance, and administrative requirements 
related to ensuring the habitability of the control room envelope (CRE) 
in accordance with Nuclear Regulatory Commission (NRC)-approved TS Task 
Force (TSTF) Standard Technical Specification change traveler TSTF-448, 
Revision 3, ``Control Room Habitability.'' Specifically, the proposed 
amendment would modify TS 3.7.9, ``Control Room Emergency Ventilation 
System (CREVS),'' and would establish a CRE habitability (CREH) program 
in TS Section 5.5, ``Administrative Controls--Programs and Manuals.'' 
The NRC staff issued a ``Notice of Availability of Technical 
Specification Improvement to Modify Requirements Regarding Control Room 
Envelope Habitability Using the Consolidated Line Item Improvement 
Process'' associated with TSTF-448, Revision 3, in the Federal Register 
on January 17, 2007 (72 FR 2022). The notice included a model safety 
evaluation, a model no significant hazards consideration (NSHC) 
determination, and a model license amendment request. In its 
application dated October 22, 2007, the licensee affirmed the 
applicability of the model NSHC determination which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee is presented below:

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

    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

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

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the analysis adopted by the licensee 
and, based on this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendments involves NSHC.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

Entergy Operations, Inc., Docket Nos. 50-313, Arkansas Nuclear One, 
Unit 1, Pope County, Arkansas

    Date of amendment request: October 22, 2007.
    Description of amendment request: The proposed amendment would 
modify requirements of Technical Specification (TS) 3.4.12, ``RCS 
Specific Activity,'' and TS 3.7.4, ``Secondary Specific Activity,'' as 
related to the use of an alternate source term (AST) associated with 
accident offsite and control room dose consequences. Implementation of 
AST supports adoption of the control room envelope habitability 
controls in accordance with Nuclear Regulatory Commission (NRC)-
approved TS Task Force (TSTF) Standard Technical Specification change 
traveler TSTF-448, Revision 3, ``Control Room Habitability.''
    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.

[[Page 71709]]

    The use of an AST is recognized in 10 CFR 50.67 and guidance for 
its implementation is provided in RG [Regulatory Guide] 1.183. The 
AST involves quantities, isotopic composition, chemical and physical 
characteristics, and release timing of radioactive material for use 
as inputs to accident dose analyses. As such, the AST cannot affect 
the probability of occurrence of a previously evaluated accident. In 
addition, the reduction is specific activity limits within the TSs 
is unrelated to accident initiators. No facility equipment, 
procedure, or process changes are required in conjunction with 
implementing the AST that could increase the likelihood of a 
previously analyzed accident. The proposed changes in the source 
term and the methodology for the dose consequence analyses follow 
the guidance of RG 1.183. As a result, there is no increase in the 
likelihood of existing event initiators.
    Regarding accident consequences, the reduction in specific 
activity limits within the TSs is more restrictive (more 
conservative) and acts to support the analysis results given the 
application of an AST. The results of accident dose analyses using 
the AST are compared to TEDE [total effective dose equivalent] 
acceptance criteria that account for the sum of deep dose equivalent 
(for external exposure) and committed effective dose equivalent (for 
internal exposure). Dose results were previously compared to 
separate limits on whole body, thyroid, and skin doses as 
appropriate for the particular accident analyzed. The results of the 
revised dose consequences analyses demonstrate that the regulatory 
acceptance criteria are met for each analyzed event. Implementing 
the AST involves no facility equipment, procedure, or process 
changes that could affect the radioactive material actually released 
during an event. Consequently, no conditions have been created that 
could significantly increase the consequences of any of the events 
being evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any of the events 
being 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 AST involves quantities, isotopic composition, chemical and 
physical characteristics, and release timing of radioactive material 
for use as inputs to accident dose analyses. As such, the AST cannot 
create the possibility of a new or different kind of accident. In 
addition, the reduction is specific activity limits within the TSs 
is unrelated to accident initiators. No facility equipment, 
procedure, or process changes have been made in conjunction with 
implementing the AST that could initiate or substantially alter the 
progression of an 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.
    Implementing the AST is relevant only to calculated accident 
dose consequences. The results of the revised dose consequences 
analyses demonstrate that the regulatory acceptance criteria are met 
for each analyzed event. In addition, the reduction is specific 
activity limits within the TSs is unrelated to accident initiators. 
No facility equipment, procedure, or process changes are required in 
conjunction with implementing the AST that could increase the 
exposure of control room or offsite individuals to radioactive 
material. The AST does not affect the transient behavior of non-
radiological parameters (e.g., Reactor Coolant System pressure, 
Containment pressure) that are pertinent to a margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

Entergy Operations, Inc., Docket Nos. 50-313, Arkansas Nuclear One, 
Unit 1, Pope County, Arkansas

    Date of amendment request: October 22, 2007.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications (TS) requirements for mode change 
limitations in Limiting Condition for Operation (LCO) 3.0.4 and 
Surveillance Requirement (SR) 3.0.4. The proposed TS changes are 
consistent with Revision 9 of Nuclear Regulatory Commission (NRC)-
approved Industry TS Task Force (TSTF) Standard TS (STS) change 
traveler, TSTF-359, ``Increase Flexibility in Mode Restraints.'' The 
amendment would also modify other TSs to reflect the revisions to LCO 
3.0.4. The spelling of the word ``not'' is corrected in Section 1.4 of 
the TSs.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 2, 2002 (67 FR 50475), as part of the 
Consolidated Line Item Improvement Process (CLIIP), on possible 
amendments to revise the plant-specific TS to modify requirements for 
model change limitations in LCO 3.0.4 and SR 3.0.4.
    The NRC staff subsequently issued a notice of availability of the 
models for Safety Evaluation and No Significant Hazards Consideration 
Determination for referencing in license amendment applications in the 
Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed 
the applicability of the CLIIP, including the model No Significant 
Hazards Consideration Determination, in its application dated October 
22, 2007.
    The proposed TS changes are consistent with NRC-approved Industry 
TSTF STS change, TSTF-359, Revision 8, as modified by 68 FR 16579. 
TSTF-359, Revision 8, was subsequently revised to incorporate the 
modifications discussed in the April 4, 2003, Federal Register notice 
and other minor changes. TSTF-359, Revision 9, was subsequently 
submitted to the NRC on April 28, 2003, and was approved by the NRC on 
May 9, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the NRC staff's analysis 
of the issue of no significant hazards consideration is presented 
below:

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


    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. 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). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, 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

[[Page 71710]]

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 entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS Limiting 
Conditions for Operation (LCO). The risk associated with this 
allowance is managed by the imposition of required actions that must 
be performed within the prescribed completion times. The net effect 
of being in a TS condition on the margin of safety is not considered 
significant. The proposed change does not alter the required actions 
or completion times of the TS. The proposed change allows TS 
conditions to be entered, and the associated required actions and 
completion times to be used in new circumstances. This use is 
predicated upon the licensee's performance of a risk assessment and 
the management of plant risk. The change also eliminates current 
allowances for utilizing required actions and completion times in 
similar circumstances, without assessing and managing risk. The net 
change to the margin of safety is insignificant. Therefore, this 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff proposes to determine that the request for amendment 
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: Thomas G. Hiltz.

Entergy Operations, Inc., Docket Nos. 50-368, Arkansas Nuclear One, 
Unit 2*, Pope County, Arkansas

    Date of amendment request: October 22, 2007.
    Description of amendment request: The proposed amendment would 
modify the Technical Specification (TS) to establish more effective and 
appropriate action, surveillance, and administrative requirements 
related to ensuring the habitability of the control room envelope (CRE) 
in accordance with Nuclear Regulatory Commission (NRC)-approved TS Task 
Force (TSTF) Standard Technical Specification change traveler TSTF-448, 
Revision 3, ``Control Room Habitability.'' Specifically, the proposed 
amendment would modify TS 3.7.6.1, ``Control Room Emergency Ventilation 
and Air Condition System,'' and would establish a CRE habitability 
(CREH) program in TS Section 6.5, ``Administrative Controls--Programs 
and Manuals.'' The NRC staff issued a ``Notice of Availability of 
Technical Specification Improvement to Modify Requirements Regarding 
Control Room Envelope Habitability Using the Consolidated Line Item 
Improvement Process'' associated with TSTF-448, Revision 3, in the 
Federal Register on January 17, 2007 (72 FR 2022). The notice included 
a model safety evaluation, a model no significant hazards consideration 
(NSHC) determination, and a model license amendment request. In its 
application dated October 22, 2007, the licensee affirmed the 
applicability of the model NSHC determination which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee is presented below:

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

    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated. Therefore, the probability of any accident 
previously evaluated is not increased. Performing tests to verify 
the operability of the CRE boundary and implementing a program to 
assess and maintain CRE habitability ensure that the CRE emergency 
ventilation system is capable of adequately mitigating radiological 
consequences to CRE occupants during accident conditions, and that 
the CRE emergency ventilation system will perform as assumed in the 
consequence analyses of design basis accidents. Thus, the 
consequences of any accident previously evaluated are not increased. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

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

    The proposed change does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the analysis adopted by the licensee 
and, based on this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendments involves NSHC.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

Entergy Operations, Inc., Docket Nos. 50-368, Arkansas Nuclear One, 
Unit 2, Pope County, Arkansas

    Date of amendment request: October 22, 2007.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications (TS) requirements for mode change 
limitations in Limiting Condition for Operation (LCO) 3.0.4 and 
Surveillance Requirement (SR) 4.0.4. The proposed TS changes are 
consistent with Revision

[[Page 71711]]

9 of Nuclear Regulatory Commission (NRC)-approved Industry TS Task 
Force (TSTF) Standard TS (STS) change traveler, TSTF-359, ``Increase 
Flexibility in Mode Restraints.'' The amendment would also modify other 
TSs to reflect the revisions to LCO 3.0.4. In addition, a change to TS 
3.4.3 was made which was determined to be equivalent to the TSTF-359 
changes.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 2, 2002 (67 FR 50475), as part of the 
Consolidated Line Item Improvement Process (CLIIP), on possible 
amendments to revise the plant-specific TS to modify requirements for 
model change limitations in LCO 3.0.4 and SR 4.0.4.
    The NRC staff subsequently issued a notice of availability of the 
models for Safety Evaluation and No Significant Hazards Consideration 
Determination for referencing in license amendment applications in the 
Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed 
the applicability of the CLIIP, including the model No Significant 
Hazards Consideration Determination, in its application dated October 
22, 2007.
    The proposed TS changes are consistent with NRC-approved Industry 
TSTF STS change, TSTF-359, Revision 8, as modified by 68 FR 16579. 
TSTF-359, Revision 8, was subsequently revised to incorporate the 
modifications discussed in the April 4, 2003, Federal Register notice 
and other minor changes. TSTF-359, Revision 9, was subsequently 
submitted to the NRC on April 28, 2003, and was approved by the NRC on 
May 9, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the NRC staff's analysis 
of the issue of no significant hazards consideration is presented 
below:

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

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. 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). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, 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 entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS Limiting 
Conditions for Operation (LCO). The risk associated with this 
allowance is managed by the imposition of required actions that must 
be performed within the prescribed completion times. The net effect 
of being in a TS condition on the margin of safety is not considered 
significant. The proposed change does not alter the required actions 
or completion times of the TS. The proposed change allows TS 
conditions to be entered, and the associated required actions and 
completion times to be used in new circumstances. This use is 
predicated upon the licensee's performance of a risk assessment and 
the management of plant risk. The change also eliminates current 
allowances for utilizing required actions and completion times in 
similar circumstances, without assessing and managing risk. The net 
change to the margin of safety is insignificant. Therefore, this 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff proposes to determine that the request for amendment 
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: Thomas G. Hiltz.

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: October 18, 2007.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to change requirements related to 
Emergency Diesel Generator (EDG) fuel oil tank volume, EDG fuel oil 
testing and Reactor Building crane inspections.
    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 changes do not impact the operability of any 
Structure, System or Component that affects the probability of an 
accident or that supports mitigation of an accident previously 
evaluated. The proposed change does not affect reactor operations or 
accident analysis and has no radiological consequences. The 
operability requirements for accident mitigation systems remain 
consistent with the licensing and design basis. Therefore, the 
proposed 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. As such, no 
new or different types of equipment will be installed, and the 
operation of 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. The specified margin for onsite fuel oil storage 
is maintained and the applicable testing standards and methods 
remain unchanged. These changes do not change any existing 
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.


[[Page 71712]]


    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: October 18, 2007.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications applicability requirements related 
to primary containment oxygen concentration and drywell-to-suppression 
chamber differential pressure limits. The associated actions would also 
be revised to be consistent with exiting the applicability for each 
specification.
    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 does not increase the 
probability of an accident since it does not involve the 
modification of any plant equipment or affect how plant systems or 
components are operated, it only changes the requirements for when 
inerting and differential pressure need to be established. Whether 
the containment is inerted or differential pressure is established 
does not impact the likelihood of an accident previously evaluated. 
Therefore, the proposed change does not involve a significant 
increase in the probability of an accident previously evaluated. The 
technical limits (i.e., oxygen concentration and differential 
pressure) imposed by the associated Technical Specifications remain 
unchanged. Brief periods where the requirements for maintaining 
these technical limits are relaxed are currently considered in the 
Technical Specifications and associated licensing basis. The 
proposed change clarifies the definition of these periods however, 
any changes are not considered significant and are supported by 
remaining consistent with the recommended allowances of NUREG 1433, 
Revision 3. The consequences of analyzed events are therefore not 
affected. Therefore, the proposed change 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 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. As such, no 
new or different types of equipment will be installed, and the 
operation of 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. The proposed change does not involve the 
modification of any plant equipment or affect basic plant operation. 
Additionally, the associated limitations remain unchanged. These 
changes do not negate any existing requirement, 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.
    The revised plant conditions reflecting the applicability and 
the duration allowed to restore limits are not credited in any 
design basis event. These changes do not reflect any significant 
adverse impact to the overall risk of operating during brief periods 
without the required primary containment oxygen concentration or 
differential pressure since the total time for any occurrence is 
only marginally extended and reflects times consistent with NUREG-
1433, Revision 3. 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.

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

    Date of amendment request: November 20, 2007.
    Description of amendment request: The proposed amendment would 
revise the values of the safety limit minimum critical power ratio 
(SLMCPR) in Technical Specification (TS) Section 2.1.1, ``Reactor Core 
SLs.'' Specifically, the proposed change would delete the Quad Cities 
Nuclear Power Station (QCNPS) Unit 2 fuel-specific SLMCPR requirements 
for Global Nuclear Fuel (GNF) GE14 fuel and consolidate the Unit 1 and 
Unit 2 SLMCPR requirements into a bounding dual-unit requirement. This 
change is needed to support the next cycle of Unit 2 operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established consistent with NRC-
approved methods to ensure that fuel performance during normal, 
transient, and accident conditions is acceptable. The proposed 
change to delete the QCNPS Unit 2 fuel-specific SLMCPR requirements 
for Global Nuclear Fuel (GNF) GE14 fuel conservatively establishes 
the SLMCPR for QCNPS, Unit 2, Cycle 20 at the SLMCPR value for the 
co-resident Westinghouse SVEA-96 Optima2 fuel, such that the fuel is 
protected during normal operation and during plant transients or 
anticipated operational occurrences (AOOs).
    The proposed change to delete the GE14 SLMCPR and establish the 
requirement at the SLMCPR value for the co-resident Westinghouse 
SVEA-96 Optimal fuel does not increase the probability of an 
evaluated accident. The change does not require any physical plant 
modifications, physically affect any plant components, or entail 
changes in plant operation. Therefore, no individual precursors of 
an accident are affected.
    The proposed change to delete the GE14 SLMCPR and establish the 
requirement at the SLMCPR value for the co-resident Westinghouse 
SVEA-96 Optimal fuel revises the QCNPS Unit 2 SLMCPR requirement to 
protect the fuel during normal operation as well as during plant 
transients or AOOs. Operational limits will be established based on 
the proposed SLMCPR to ensure that the SLMCPR is not violated. This 
will ensure that the fuel design safety criterion (i.e., that at 
least 99.9% of the fuel rods do not experience transition boiling 
during normal operation and AOOs) is met. Since the proposed change 
does not affect operability of plant systems designed to mitigate 
any consequences of accidents, the consequences

[[Page 71713]]

of an accident previously evaluated will not increase.
    The proposed consolidation of the Unit 1 and Unit 2 SLMCPR 
requirements into a bounding dual-unit requirement is 
administrative. As such, the proposed consolidation does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Creation of the possibility of a new or different kind of 
accident requires creating one or more new accident precursors. New 
accident precursors may be created by modifications of plant 
configuration, including changes in allowable modes of operation. 
The proposed changes do not involve any plant configuration 
modifications or changes to allowable modes of operation. The 
proposed change to delete the GE14 SLMCPR and establish the 
requirement at the SLMCPR value for the co-resident Westinghouse 
SVEA-96 Optimal fuel assures that safety criteria are maintained for 
QCNPS, Unit 2, Cycle 20. The proposed consolidation of the Unit 1 
and Unit 2 SLMCPR requirements into a bounding dual-unit requirement 
is administrative.
    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?
    The SLMCPR provides a margin of safety by ensuring that at least 
99.9% of the fuel rods do not experience transition boiling during 
normal operation and AOOs if the SLMCPR limit is not violated. The 
proposed change will ensure the current level of fuel protection is 
maintained by continuing to ensure that at least 99.9% of the fuel 
rods do not experience transition boiling during normal operation 
and AOOs if the SLMCPR limit is not violated. The proposed SLMCPR 
values were developed using NRC-approved methods. Additionally, 
operational limits will be established based on the proposed SLMCPR 
to ensure that the SLMCPR is not violated. This will ensure that the 
fuel design safety criterion (i.e., that no more than 0.1% of the 
rods are expected to be in boiling transition if the MCPR limit is 
not violated) is met.
    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 
requested amendments involve 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.

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

    Date of amendment request: October 29, 2007.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) for Prairie Island Nuclear 
Generating Plant (PINGP) Units 1 and 2 Surveillance Requirement (SR) 
3.8.1.9, to require that the test is performed at or below a power 
factor of 0.85. The proposed amendments fulfill the commitment made in 
Amendments 178 to Unit 1, and 168 to Unit 2, issued on May 30, 2007 
(Agency wide Documents Access and Management System (ADAMS) Accession 
No. ML071310023).
    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.
    This license amendment request proposes more restrictive changes 
to the Technical Specification Surveillance Requirements for the 
emergency diesel generators which will require testing at a 
specified power factor, grid conditions permitting.
    The emergency diesel generators are not accident initiators and 
therefore, these changes do not involve a significant increase in 
the probability of an accident. The proposed changes increase the 
load testing requirements, are consistent with the intent of current 
regulatory guidance for testing emergency diesel generators, and 
will continue to assure that this equipment performs its design 
function. Thus these changes do not involve a significant increase 
in the consequences of an accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This license amendment request proposes more restrictive changes 
to the Technical Specification Surveillance Requirements for the 
emergency diesel generators which will require testing at a 
specified power factor, grid conditions permitting.
    The changes proposed for the emergency diesel generators do not 
change any system operations or maintenance activities. Testing 
requirements will be revised and will continue to demonstrate that 
the Limiting Conditions for Operation are met and the system 
components are functional. These changes do not create new failure 
modes or mechanisms which are not identifiable during testing and no 
new accident precursors are generated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This license amendment request proposes more restrictive changes 
to the Technical Specification Surveillance Requirements for the 
emergency diesel generators which will require testing at a 
specified power factor, grid conditions permitting.
    The current Technical Specification Surveillance Requirements do 
not specify testing at any power factor. The Technical Specification 
Surveillance Requirements proposed in this license amendment request 
are thus more restrictive in that they place additional restraints 
on the test conditions. These changes may make the testing more 
rigorous and thus more difficult for the emergency diesel generators 
to meet the test acceptance criteria. The addition of a power factor 
is consistent with the intent of current regulatory guidance for 
testing emergency diesel generators. Since these changes are an 
increase in the test requirements and are consistent with the intent 
of current regulatory guidance, these changes do not involve a 
significant reduction in a margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Acting Branch Chief: Cliff Munson.

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

    Date of amendment request: November 19, 2007.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications for the PINGP, Units 1 and 2, to 
replace the

[[Page 71714]]

current fixed Frequency for testing the containment spray nozzles in 
Surveillance Requirement 3.6.5.8 with a maintenance or event based 
Frequency.
    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.
    This license amendment request proposes Technical Specification 
Surveillance Requirement changes which will require verification 
that the containment spray system spray nozzles are unobstructed 
following maintenance which could result in nozzle blockage.
    The containment spray system and its spray nozzles are not 
accident initiators and therefore, these changes do not involve a 
significant increase the probability of an accident. The revised 
surveillance requirement will require event based verification in 
lieu of fixed Frequency verification which may require either fewer 
or more verifications of operability. The proposed changes to verify 
system operability following maintenance is considered adequate to 
ensure operability of the containment spray system. Since the system 
continues to be available to perform its accident mitigation 
function, the consequences of accidents previously evaluated are not 
significantly increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This license amendment request proposes Technical Specification 
Surveillance Requirement changes which will require verification 
that the containment spray system spray nozzles are unobstructed 
following maintenance which could result in nozzle blockage.
    The proposed change does not introduce a new mode of plant 
operation and does not involve physical modification to the plant. 
The change does not introduce new accident initiators or impact the 
assumption made in the safety analysis. Testing requirements will be 
revised and will continue to demonstrate that the Limiting 
Conditions for Operation are met and the system components are 
functional.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This license amendment request proposes Technical Specification 
Surveillance Requirement changes which will require verification 
that the containment spray system spray nozzles are unobstructed 
following maintenance which could result in nozzle blockage.
    The containment spray system is not susceptible to corrosion-
induced obstruction or obstruction from sources external to the 
system. Maintenance activities that could introduce foreign material 
into the system would require subsequent verification to ensure 
there is no spray nozzle blockage. The spray header nozzles are 
expected to remain unblocked and available in the event that the 
safety function is required. Therefore, the capacity of the system 
would remain unaffected.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Acting Branch Chief: Clifford G. Munson.

Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power 
Plant (HBPP), Unit 3 Humboldt County, California J00336

    Date of amendment request: November 5, 2007.
    Description of amendment request: The licensee has proposed 
amending the technical specifications (TS) to delete many operational 
and administrative requirements upon transfer of spent nuclear fuel 
assemblies and fuel fragment containers from the Spent Fuel Pool (SFP) 
to the Humboldt Bay Independent Spent Fuel Storage Installation 
(ISFSI). Some TS requirements will be relocated to the HBPP Quality 
Assurance Plan.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes reflect the transfer of spent fuel from the 
Spent Fuel Pool to the Humboldt Bay (HB) Independent Spent Fuel 
Storage Installation. Design basis accidents related to the SFP are 
discussed in the Humboldt Bay Power Plant Unit 3 Defueled Safety 
Analysis Report (DSAR). These postulated accidents are predicated on 
spent fuel being stored in the SFP. With the removal of the spent 
fuel from the SFP, there are no important-to-safety systems, 
structures or components required to function or to be monitored. In 
addition, there are no remaining credible accidents involving spent 
fuel or the SFP that require actions of a Certified Fuel Handler or 
Noncertified Fuel Handler to prevent occurrence or to mitigate 
consequences. The proposed change to the Design Features section of 
the Technical Specifications (TS) clarifies that the spent fuel is 
being stored in dry casks within an ISFSI. The probability or 
consequences of accidents at the ISFSI are evaluated in the HB ISFSI 
Final Safety Analysis Report (FSAR) and are independent of the 
accidents evaluated in the HBPP Unit 3 DSAR. Therefore, the proposed 
changes will 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 evaluated?
    Response: No.
    The proposed changes reflect the reduced operational risks as a 
result of the spent fuel being transferred to dry casks within an 
ISFSI. The proposed changes do not modify any systems, structures or 
components. The plant conditions for which the HBPP Unit 3 DSAR 
design basis accidents relating to spent fuel and the SFP have been 
evaluated are no longer applicable. The aforementioned proposed 
changes do not affect any of the parameters or conditions that could 
contribute to the initiation of an accident. Design basis accidents 
associated with the dry cask storage of spent fuel are already 
considered in the HB ISFSI FSAR. No new accident scenarios are 
created as a result of deleting nonapplicable operational and 
administrative requirements. Therefore, the proposed changes will 
not create the possibility of a new or different kind of accident 
from those previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed changes reflect the reduced operational risks as a 
result of the spent fuel being transferred to dry casks within an 
ISFSI. The design basis and accident assumptions within the HBPP 
Unit 3 DSAR and the TS relating to spent fuel are no longer 
applicable. The proposed changes do not affect remaining plant 
operations, nor structures, systems, or components supporting 
decommissioning activities. In addition, the proposed changes do not 
result in a change in initial conditions, system response time, or 
in any other parameter affecting the course of a decommissioning 
activity accident analysis. Therefore, the proposed changes will 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

[[Page 71715]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Jennifer K. Post, Pacific Gas and 
Electric Company, 77 Beale Street, B30A, San Francisco, CA.
    NRC Branch Chief: Andrew Persinko.

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

    Date of amendment request: November 30, 2007.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Sections TS 5.5.9, ``Steam 
Generator (SG) Program'' and TS 5.6.10, ``Steam Generator Tube 
Inspection Report.'' The proposed changes to TS 5.5.9 modify the 
inspection and plugging requirements for portions of SG tubes within 
the hot leg side of the tubesheet region of the SGs only. The proposed 
changes to TS 5.6.10 will add requirements to report specific data 
related to indications, leakage detected, and calculated accident 
leakage.
    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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The previously analyzed accidents are initiated by the 
failure of plant structures, systems, or components. The proposed 
changes that alter the SG inspection criteria do not have a 
detrimental impact on the integrity of any plant structure, system, 
or component 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. Therefore, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    Of the applicable accidents previously evaluated, the limiting 
transients with consideration to the proposed changes to the SG tube 
inspection criteria, are the SG tube rupture (SGTR) event and the 
steam line break (SLB) accident.
    During the SGTR event, the required structural integrity margins 
of the SG tubes will be maintained by the presence of the SG 
tubesheet. SG tubes are hydraulically expanded in the tubesheet 
area. Tube rupture in tubes with cracks in the tubesheet is 
precluded by the constraint provided by the tubesheet. This 
constraint results from the hydraulic expansion process, thermal 
expansion mismatch between the tube and tubesheet and from the 
differential pressure between the primary and secondary side. Based 
on this design, the structural margins against burst discussed in RG 
1.121 (Reference 4) [Regulatory Guide 1.121, ``Bases for Plugging 
Degraded PWR Steam Generator Tubes,'' dated August 1976], are 
maintained for both normal and postulated accident conditions.
    The proposed changes do not affect other systems, structures, 
components or operational features. Therefore, the proposed changes 
result in no significant increase in the probability of the 
occurrence of a SGTR accident.
    At normal operating pressures, leakage from primary water stress 
corrosion cracking (PWSCC) below the proposed limited inspection 
depth is limited by both the tube-to-tubesheet crevice and the 
limited crack opening permitted by the tubesheet constraint. 
Consequently, negligible normal operating leakage is expected from 
cracks within the tubesheet region. The consequences of a SGTR event 
are affected by the primary-to-secondary leakage flow during the 
event. Primary-to-secondary leakage flow through a postulated broken 
tube is not affected by the proposed change since the tubesheet 
enhances the tube integrity in the region of the hydraulic expansion 
by precluding tube deformation beyond its initial hydraulically 
expanded outside diameter.
    The probability of a SLB is unaffected by the potential failure 
of a SG tube, since this failure is not an initiator for a SLB.
    The consequences of a SLB are also not significantly affected by 
the proposed changes. During a SLB accident, the reduction in 
pressure above the tubesheet on the shell side of the SG creates an 
axially uniformly distributed load on the tubesheet due to the 
reactor coolant system pressure on the underside of the tubesheet. 
The resulting bending action constrains the tubes in the tubesheet, 
thereby restricting primary-to-secondary leakage below the midplane.
    The purpose of the tube-end weld is to ensure the hydraulically 
expanded tube-to-tubesheet joints in Model F SGs are leak-tight. 
Considerations were also made with regard to the potential for 
primary-to-secondary leakage during postulated faulted conditions. 
However, the leak rate during postulated accident conditions would 
be expected to be less than that during normal operation for 
indications near the bottom of the tubesheet based on the evaluation 
(Reference 1) [Westinghouse Electric Company WCAP-16794-P, ``Steam 
Generator Tube Alternate Repair Criteria for the Portion of the Tube 
Within the Tubesheet at the Vogtle 1 & 2 Electric Generating 
Plants,'' dated October 2007] which shows that while the driving 
pressure increases by about a factor of almost two, the flow 
resistance increases, because the tube-to-tubesheet contact pressure 
also increases. Depending on the depth within the tubesheet, the 
relative increase in resistance could easily be larger than that of 
the pressure potential. Therefore, the leak rate under normal 
operating conditions could exceed its allowed value before the 
accident condition leak rate would be expected to exceed its allowed 
value. This approach is termed an application of the ``bellwether 
principle.'' While such a decrease in the leak rate is expected, the 
postulated accident leak rate could conservatively be taken to be 
bounded by twice the normal operating leak rate if the increase in 
contact pressure is ignored.
    Since normal operating leakage is limited by VEGP TS 3.4.13 and 
by NEI 97-06 (Reference 3) [NEI 97-06, ``Steam Generator Program 
Guidelines,'' Revision 2, dated May 2, 2005] to less than 150 gpd 
throughout one SG in the VEGP Units 1 and 2 SGs, the attendant 
accident condition leak rate, assuming all leakage to be from lower 
tubesheet indications, would be bounded by 0.20 gpm in the faulted 
SG which is less than the accident analysis assumption of 0.35 gpm 
to the affected SG included in Section 15.1.5 of the VEGP FSAR. 
Hence, it is reasonable to omit any consideration of inspection of 
the tube, tube end weld, bulges/overexpansions or other anomalies 
below 17 inches from the top of the hot leg tubesheet.
    Based on the above discussion, the proposed changes do not 
involve an increase in the consequences of an accident previously 
evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    No. The proposed changes do not involve the use or installation 
of new equipment and the currently installed equipment will not be 
operated in a new or different manner. No new or different system 
interactions are created and no new processes are introduced. The 
proposed changes will not introduce any new failure mechanisms, 
malfunctions, or accident initiators not already considered in the 
design and licensing bases.
    Based on this evaluation, 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?
    No. The proposed changes maintain the required structural 
margins of the SG tubes for both normal and accident conditions. NEI 
97-06 (Reference 3) and RG 1.121 (Reference 4), are used as the 
bases in the development of the limited tubesheet inspection depth 
methodology for determining that SG tube integrity considerations 
are maintained within acceptable limits. RG 1.121 (Reference 4) 
describes a method acceptable to the NRC for meeting the following 
General Design Criteria (GDC).
     GDC 14, ``Reactor coolant pressure boundary,''
     GDC 15, ``Reactor coolant system design,''
     GDC 31, ``Fracture prevention of reactor coolant 
pressure boundary,'' and,
     GDC 32, ``Inspection of reactor coolant pressure 
boundary.''
    RG 1.121 concludes that by determining the limiting safe 
conditions for tube wall degradation, the probability and 
consequences of a SGTR are reduced. This RG uses safety factors on 
loads for tube burst that are consistent with the requirements of 
Section III of the ASME Code [American Society of Mechanical 
Engineers, Boiler and Pressure Vessel Code].
    Application of the limited tubesheet inspection depth criteria 
will preclude

[[Page 71716]]

unacceptable primary-to-secondary leakage during all plant 
conditions. The methodology for determining leakage provides for 
large margins between calculated and actual leakage values in the 
proposed limited tubesheet inspection depth criteria.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: October 23, 2007.
    Description of amendment request: The amendments will relocate the 
surveillance test intervals of various Technical Specifications (TSs) 
to a licensee-controlled program (risk-informed Initiative 5(b)) in 
accordance with the Surveillance Frequency Control Program, which is 
being added to the Administrative Controls section of the 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. [Do] the proposed change[s] involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change[s] [involve] the relocation of various 
surveillance test intervals from Technical Specifications (TS) to a 
licensee-controlled program. The proposed change[s] [do] not involve 
the modification of any plant equipment or affect basic plant 
operation. The proposed change[s] will have no impact on the design 
or function of any safety related structures, systems or components. 
Surveillance test intervals are not assumed to be an initiator of 
any analyzed event, nor are they assumed in the mitigation of 
consequences of accidents. The surveillance requirements themselves 
will be maintained in the TS along with the applicable Limiting 
Conditions for Operation (LCOs) and Action statements. The 
surveillances performed at the intervals specified in the licensee-
controlled program will assure that the affected system or component 
function is maintained, that the facility operation is within the 
Safety Limits, and that the LCOs are met.
    Therefore, the proposed change[s] [do] not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. [Do] the proposed change[s] create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change[s] [do] not involve any physical alteration 
of plant equipment and does not change the method by which any 
safety-related structure, system, or component performs its function 
or is tested. As such, no new or different types of equipment will 
be installed, and the basic operation of installed equipment is 
unchanged.
    The methods governing plant operation and testing remain 
consistent with current safety analysis assumptions.
    Therefore, the proposed change[s] will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change[s] [do] not negate any existing requirement, 
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. Margins of safety are unaffected by relocation of the 
surveillance test intervals to a licensee-controlled program.
    Therefore, the proposed change[s] [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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Branch Chief: Thomas G. Hiltz.

U.S. Department of Transportation (USDOT), United States Maritime 
Administration (MARAD), License No. NS-1, Docket No. 50-238, Nuclear 
Ship Savannah (NSS)

    Date of amendment request: October 9, 2007.
    Description of amendment request:
    The proposed license amendment would modify the Technical 
Specification (TS) requirements to clarify the TS and make the 
requirements commensurate with the current ship status and 
decommissioning schedule. Thirty-nine TS changes are proposed. The 
proposed changes modify the TS as follows:
     Delete requirements more appropriate for the Final Safety 
Analysis Report;
     Provide consistent titles and phrases;
     Delete duplicate requirements;
     Organize similar requirements into single locations;
     Remove requirements that can be implemented through 
current regulations;
     Delete archaic requirements;
     Invoke requirements commensurate with current ship status 
and decommissioning schedule;
     Format and renumber, as appropriate;
     Revise requirements to reflect historical practices;
     Revise TS to be consistent with the Decommissioning 
Quality Assurance Plan; and
     Correct errors introduced in License Amendment 13, 
Reference (a).
    The application for license amendment is available electronically 
at the NRC's Electronic Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this site, you can access the NRC's Agencywide 
Document Access and Management System (ADAMS), which provides text and 
image files of NRC's public documents. The ADAMS accession number for 
the October 9, 2007, request is ML072880143.
    If you do not have access to ADAMS, or if there are problems in 
accessing the documents located in ADAMS, contact the NRC Public 
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737, or 
by e-mail to [email protected]. These documents may also be viewed 
electronically on the public computers located at the NRC's PDR, O 1 
F21, One White Flint North, 11555 Rockville Pike, Rockville, MD 20852. 
The PDR reproduction contractor will copy documents for a fee.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes are administrative and do not involve 
modification of any plant equipment or affect basic plant operation. 
The NSS's reactor is not operational and the level of radioactivity 
in the NSS has significantly decreased from the levels that

[[Page 71717]]

existed when the 1976 Possession-only License was issued. No aspect 
of any of proposed changes is and initiator of any accident 
previously evaluated. Consequently, the probability of an accident 
previously evaluated is not significantly increased.
    Therefore, the proposed changes no 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 evaluated?
    Response: No.
    All of the proposed changes are administrative and do not 
involve physical alteration of plant equipment that was not 
previously allowed by Technical Specifications. These proposed 
changes do not change the method by which any safety-related system 
performs its function. As such, no new or different types of 
equipment will be installed, and the basic operation of installed 
equipment is unchanged. The methods governing plant operation and 
testing remain consistent with current safety analysis assumptions.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    All of the proposed changes are administrative in nature. No 
margins of safety exist that are relevant to the ship's defueled and 
partially dismantled reactor. 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 changes. The proposed changes involve movement of the ship, 
changes in the performance of responsibilities and reflect 
significantly improved radiological conditions since 1976.
    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 upon 
the staff's review of the licensee's analysis, as well as the staff's 
own evaluation, the staff concludes 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.
    Senior Technical Advisor, N.S. Savannah: Erhard W. Koehler, MARAD, 
Office of Ship Disposal Programs.
    NRC Branch Chief: Andrew Persinko.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: October 24, 2007.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) 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.
    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?
    The proposed amendment does not involve a significant increase 
in the probability or consequence of an accident previously 
analyzed. There is no change to how or under what conditions the 
inverters or 120 VAC vital buses are operated, nor are there any 
changes to acceptable operating parameters. Operability 
requirements, which are consistent with current operation of the 
inverters and vital buses, are being established for the inverters 
and vital buses associated with shared systems. The proposed change 
will ensure that there is an operable electrical control circuit for 
the Auxiliary Building Central Exhaust subsystem filter and bypass 
dampers for each train of the [Emergency Core Cooling System Pump 
Room Exhaust Air Cleanup System] ECCS PREACS which will ensure that 
the evaluated dose consequences for [design basis accidents] DBAs 
will not be exceeded.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. [Does the proposed amendment] create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The implementation of the proposed changes does not create the 
possibility of an accident of a different type than was previously 
evaluated in the [Updated Final Safety Analysis Report] UFSAR. There 
is no change to how or under what conditions the inverters or 120 
VAC vital buses are operated nor are there any changes to acceptable 
operating parameters. The proposed operability requirements, which 
are consistent with current operation of the inverters and vital 
buses, are being established for the inverters and vital buses 
associated with shared systems. The proposed changes ensure vital 
120 VAC power is available to support operation of the Auxiliary 
Building Central Exhaust subsystems. These changes do not alter the 
nature of events postulated in the UFSAR nor do they introduce any 
unique precursor mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. [Does the proposed amendment] involve a significant reduction 
in the margin of safety?
    The implementation of the proposed changes does not reduce the 
margin of safety. The proposed changes for the 120 VAC Vital Bus 
System and Inverters do not affect the ability of these systems or 
components to perform their intended safety functions to provide 
power to required safety and monitoring systems or components. 
Operability requirements, which are consistent with current 
operation of the inverters and vital buses, are being established 
for the inverters and vital buses associated with shared systems. 
These changes provide additional assurance that the Auxiliary 
Building Central Exhaust subsystems will operate to maintain the 
margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Evangelos C. Marinos.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance

[[Page 71718]]

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].

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina.

    Date of application for amendments: January 22, 2007, as 
supplemented by letter dated September 28, 2007.
    Brief Description of amendments: The amendments change the 
Technical Specifications (TSs) related to the fuel design description 
and the fuel criticality methods to accommodate the transition to AREVA 
NP fuel.
    Date of issuance: November 27, 2007.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment Nos.: 243 and 271.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
changed the TSs.
    Date of initial notice in Federal Register: August 29, 2007 (72 FR 
49742). The supplement dated September 28, 2007, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 27, 2007.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 17, 2007.
    Brief description of amendment: The amendment revised the date for 
performing the ``Type A test'' in the River Bend Station, Unit 1, 
Technical Specification 5.5.13, ``Primary Containment Leak Rate Testing 
Program,'' from ``prior to December 14, 2007,'' to ``prior to April 14, 
2008.''
    Date of issuance: December 3, 2007.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 155.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: September 11, 2007 (72 
FR 51857). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 3, 2007.
    No significant hazards consideration comments received: No.

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

    Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1 
and 2, Will County, Illinois.
    Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 
2 and 3, Grundy County, Illinois.
    Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 
2, LaSalle County, Illinois.
    Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
Units 1 and 2, Rock Island County, Illinois.
    EGC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach 
Bottom Atomic Power Station, Units 2 and 3 (PBAPS), York and Lancaster 
Counties, Pennsylvania.
    Date of application for amendments: December 15, 2006.
    Brief description of amendments: The amendments modify the 
technical specifications (TSs) by replacing the term ``plant-specific'' 
with ``generic'' when discussing job titles in TS Section 5.2.1.a. This 
revision will ensure the TS description is consistent with the licensee 
Quality Assurance Topical Report (QATR). The proposed amendment will 
also revise the PBAPS TS Section 5.2.1.a to replace the reference to 
the Updated Final Safety Analysis Report with reference to the EGC 
QATR. This change aligns the PBAPS TS wording with the rest of the 
licensee fleet.
    Date of issuance: November 19, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of the date of issuance.
    Amendment Nos.: 152, 152, 147, 147, 225, 217, 187, 174, 265, 269, 
236, and 231.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72, NPF-77, 
DPR-19, DPR-25, NPF-11, NPF-18, DPR-29, DPR-30, DRP-44, and DPR-56: The 
amendments revised the Technical Specifications and Operating Licenses.
    Date of initial notice in Federal Register: March 13, 2007 (72 FR 
11387).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 19, 2007.
    No significant hazards consideration comments received: No.

Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell 
County, Texas

    Date of amendment request: December 19, 2006.
    Brief description of amendments: Amendments revise the requirements 
in Technical Specification (TS) 5.5.8, ``Inservice Testing Program,'' 
to update references to the American Society of Mechanical Engineers 
(ASME) Boiler and Pressure Vessel Code, Section XI, as the source of 
requirements for the inservice testing of ASME Code Class 1, 2, and 3 
pumps and valves, and address the applicability of Surveillance 
Requirement 3.0.2 to other normal and accelerated frequencies specified 
as 2 years or less in the Inservice Testing Program.
    Date of issuance: December 4, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of the date of issuance.
    Amendment Nos.: Unit 1-140; Unit 2-140.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: May 22, 2007 (72 FR 
28724). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 4, 2007.
    No significant hazards consideration comments received: No.

[[Page 71719]]

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: August 16, 2007, as supplemented by 
letter dated November 5, 2007.
    Brief description of amendment: The amendment revised Technical 
Specification 5.5.6, ``Inservice Testing Program,'' to allow a one-time 
extension of the 5-year frequency requirement for setpoint testing of 
safety valve MS-RV-70ARV.
    Date of issuance: December 4, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 228.
    Facility Operating License No. DPR-46: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: September 25, 2007 (72 
FR 54476). The supplement dated November 5, 2007, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
initially published in the Federal Register. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
December 4, 2007.
    No significant hazards consideration comments received: No.

Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco 
Nuclear Generating Station, Sacramento County, California

    Date of application for amendment: April 12, 2006, and supplemented 
November 21, 2006.
    Brief description of amendment: The amendment incorporates the 
Nuclear Regulatory Commission (NRC) approved, License Termination Plan 
(LTP), and associated addendum, into the Rancho Seco license and 
specifies limits on the changes the licensee is allowed to make to the 
approved LTP without prior NRC review and approval.
    Date of issuance: November 26, 2007.
    Effective date: November 26, 2007.
    Amendment No: 133.
    Facility Operating License No. DPR-54: The amendment revised the 
License.
    Date of initial notice in Federal Register: February 13, 2007 (72 
FR 6789).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 26, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 14, 2006, as supplemented 
by letters dated June 28, September 26, and November 2, 2007.
    Brief description of amendments: The amendments incorporate a 
description of the parent tube inspection limitation adjacent to the 
nickel band portion of the lower sleeve joint and provide the basis for 
the structural and leakage integrity of the joint being ensured with 
the existing inspection of the parent tube adjacent to the nickel band 
region.
    Date of issuance: November 29, 2007.
    Effective date: As of its date of issuance, to be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 2-215; Unit 3-207.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: September 12, 2006 (71 
FR 53720). The supplements dated June 28, September 26, and November 2, 
2007, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated November 29, 2007.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 5, 2007, as supplemented 
June 11, 2007.
    Brief description of amendments: The amendments revised the 
Technical Specifications testing frequency for surveillance requirement 
3.1.4, ``Control Rod Scram Times,'' from ``120 days cumulative 
operation in MODE 1'' to ``200 days cumulative operation in MODE 1.''
    Date of issuance: November 26, 2007.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days from the date of issuance.
    Amendment Nos.: 254, 198.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: July 17, 2007, (72 FR 
39084).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 26, 2007.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 10th day of December 2007.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
 [FR Doc. E7-24284 Filed 12-17-07; 8:45 am]
BILLING CODE 7590-01-P