Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 54862-54874 [E8-21925]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 73, Number 185 (Tuesday, September 23, 2008)]
[Notices]
[Pages 54862-54874]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-21925]


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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 August 28, 2008 to September 10, 2008. The 
last biweekly notice was published on September 9, 2008 (73 FR 52412).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, 
person(s) may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
via electronic submission through the NRC E-Filing system for a hearing 
and a petition for leave to intervene. Requests for a hearing and a 
petition for leave to intervene shall be filed in accordance with the 
Commission's ``Rules of Practice for Domestic Licensing Proceedings'' 
in 10 CFR part 2. Interested person(s) should consult a current copy of 
10 CFR 2.309, which is available at the Commission's PDR, located at 
One White Flint North, Public File Area 01F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland. Publicly available records will be 
accessible from the Agencywide Documents Access and Management System's 
(ADAMS) Public Electronic Reading Room on the Internet at the NRC Web 
site, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request 
for a hearing or petition for leave to intervene is filed within 60 
days, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel will rule on the request and/or petition; and 
the Secretary or the Chief Administrative Judge of the Atomic Safety 
and Licensing Board will issue a notice of a hearing or an appropriate 
order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's

[[Page 54863]]

property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for hearing or a petition for leave to intervene must be 
filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve documents over the internet 
or in some cases to mail copies on electronic storage media. 
Participants may not submit paper copies of their filings unless they 
seek a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
hearingdocket@nrc.gov, or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). 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 Viewer\TM\ is free and is 
available at https://www.nrc.gov/site-help/e-submittals/install-
viewer.html. Information about applying for a digital ID certificate is 
available on NRC's public Web site at https://www.nrc.gov/site-help/e-
submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at https://www.nrc.gov/
site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, in accordance 
with 10 CFR 2.302(g), with their initial paper filing requesting 
authorization to continue to submit documents in paper format. Such 
filings must be submitted by: (1) First class mail addressed to the 
Office of the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff; or (2) courier, express mail, or expedited 
delivery service to the Office of the Secretary, Sixteenth Floor, One 
White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, 
Attention: Rulemaking and Adjudications Staff. Participants filing a 
document in this manner are responsible for serving the document on all 
other participants. Filing is considered complete by first-class mail 
as of the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the petition 
and/or request should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later 
than 11:59 p.m. Eastern Time on the due date.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a Presiding Officer. Participants are requested not to include

[[Page 54864]]

personal privacy information, such as social security numbers, home 
addresses, or home phone numbers in their filings. With respect to 
copyrighted works, except for limited excerpts that serve the purpose 
of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submission.
    For further details with respect to this amendment action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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 amendments request: July 7, 2008
    Description of amendments request: The proposed change would revise 
Surveillance Requirement (SR) 3.6.1.6.1 to add a new requirement to 
verify that each vacuum breaker is closed within 6 hours following an 
operation that causes any of the vacuum breakers to open and revises SR 
3.6.1.6.2 by removing the requirement to perform functional testing of 
each vacuum breaker within 12 hours following an operation that causes 
any of the vacuum breakers to open.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR Part 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 involve physical changes to any 
plant structure, system, or component. The suppression chamber-to-
drywell vacuum breakers only provide an accident mitigation 
function. As such, the probability of occurrence for a previously 
analyzed accident is not impacted by the change to the surveillance 
frequency for these components.
    The consequences of a previously analyzed accident are dependent 
on the initial conditions assumed for the analysis, the behavior of 
the fuel during the analyzed accident, the availability and 
successful functioning of the equipment assumed to operate in 
response to the analyzed event, and the setpoints at which these 
actions are initiated. No physical change to suppression chamber-to-
drywell vacuum breakers is being made as a result of the proposed 
change, nor does the change alter the manner in which the vacuum 
breakers operate during an accident. As a result, no new failure 
modes of the suppression chamber-to-drywell vacuum breakers are 
being introduced. The surveillance requirements for the suppression 
chamber-to-drywell vacuum breakers will continue to ensure testing 
of the suppression chamber-to-drywell vacuum breakers following 
plant transients involving the discharge of steam to the suppression 
chamber from the SRVs, and such testing will continue to provide 
assurance that the vacuum breakers are able to perform their design 
function. Based on this evaluation, there is no significant increase 
in the consequences of a previously analyzed event.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to the surveillance requirements for the 
suppression chamber-to-drywell vacuum breakers does not involve any 
physical alteration of plant systems, structures, or components. No 
new or different equipment is being installed. No installed 
equipment is being operated in a different manner. There is no 
alteration to the parameters within which the plant is normally 
operated or in the setpoints that initiate protective or mitigative 
actions. As a result no new failure modes are being introduced. 
Therefore, the proposed change to the surveillance requirements for 
the suppression chamber-to-drywell vacuum breakers does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    Response: No.
    The proposed change revises Surveillance Requirement 3.6.1.6.1 
to add a new requirement to verify each vacuum breaker is closed 
within 6 hours following an operation that causes any of the vacuum 
breakers to open and revises Surveillance Requirement 3.6.1.6.2 by 
removing the requirement to perform functional testing of each 
vacuum breaker within 12 hours following an operation that causes 
any of the vacuum breakers to open. The operability and functional 
characteristics of the suppression chamber-to-drywell vacuum 
breakers remains unchanged. The margin of safety is established 
through the design of the plant structures, systems, and components, 
through the parameters within which the plant is operated, through 
the establishment of the setpoints for the actuation of equipment 
relied upon to respond to an event, and through margins contained 
within the safety analyses. The proposed change to the surveillance 
requirements for the suppression chamber-to-drywell vacuum breakers 
does not impact the condition or performance of structures, systems, 
setpoints, and components relied upon for accident mitigation. The 
proposed change to Surveillance Requirements 3.6.1.6.1 and 3.6.1.6.2 
will avoid unnecessary cycling and wear of the vacuum breaker test 
actuation mechanisms, will improve the reliability of the vacuum 
breakers, and will minimize the potential for a plant shut down due 
to a problem with a vacuum breaker test actuating mechanism from 
excessive wear. The proposed change does not impact any safety 
analysis assumptions or results. Therefore, the proposed change does 
not result in 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 proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.

Dominion Nuclear Connecticut, Inc. Docket Nos. 50-245, 50-336, and 50-
423, Millstone Power Station, Units 1, 2, and 3, New London County, 
Connecticut

    Date of amendment request: August 21, 2008.
    Description of amendment request: The proposed amendment removes 
references to and limits imposed by Nuclear Regulatory Commission 
Generic Letter (GL) 82-12, ``Nuclear Power Plant Staff Working Hours,'' 
from the subject plants'' technical specifications (TS). The guidelines 
have been superseded by the requirements of Title 10 of the Code of 
Federal Regulations, Part 26 (10 CFR 26), Subpart I, ``Managing 
Fatigue.''
    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 changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The removal of references to GL 82-12 will not remove the 
requirement to control work hours and manage fatigue. Removal of TS 
references to GL 82-12 will be performed concurrently with the 
implementation of the more conservative 10 CFR 26, Subpart I, 
requirements.

[[Page 54865]]

    The proposed changes do not impact the physical configuration or 
function of plant structures, systems, or components (SSCs) or the 
manner in which SSCs are operated, maintained, modified, tested, or 
inspected. The proposed changes do not impact the initiators or 
assumptions of analyzed events, nor do they impact the mitigation of 
accidents or transient events.
    Because these new requirements are administrative in nature and 
further, are more conservative with respect to work hour controls 
and fatigue management, the proposed change will not significantly 
increase the probability or consequence of an accident previously 
evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes remove references to GL 82-12 from TS 
consistent with the recently revised Subpart I to 10 CFR 26. These 
regulations are more restrictive than the current guidance and would 
add conservatism to work hour controls and fatigue management. Work 
hours will continue to be controlled in accordance with NRC 
requirements. The new rule continues to allow for deviations from 
controls to mitigate or prevent a condition adverse to safety or 
necessary to maintain the security of the facility. This ensures 
that the new rule will not restrict work hours at the expense of the 
health and safety of the public as well as plant personnel.
    The proposed changes do not alter plant configuration, require 
that new plant equipment be installed, alter assumptions made about 
accidents previously evaluated, add any initiators, or impact the 
function of plant SSCs or the manner in which SSCs are operated, 
maintained, modified, tested, or inspected.
    Because the proposed changes do not remove the station's 
requirement to control work hours and increases the conservatism of 
work hour controls by changing administrative scheduling 
requirements, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    Compliance with the new rule adds conservatism to existing 
fatigue management and contributes to the margin of safety. Deletion 
of references to GL 82-12 in the TS is administrative in nature 
since fatigue management is controlled through the new rule. MPS1, 
MPS2 and MPS3 will continue their fitness-for-duty and behavioral 
observation programs, both of which will be strengthened by 
compliance with the new rule. The proposed changes add conservatism 
to fatigue management and contribute to the margin of safety.
    The proposed changes do not involve any physical changes to 
plant SSCs or the manner in which SSCs are operated, maintained, 
modified, tested, or inspected. The proposed changes do not involve 
a change to any safety limits, limiting safety system settings, 
limiting conditions of operation, or design parameters for any SSC.
    The proposed changes do not impact any safety analysis 
assumptions and do not involve a change in initial conditions, 
system response times, or other parameters affecting an accident 
analysis.
    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 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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
    NRC Branch Chief: Harold K. Chernoff.

Duke Energy Carolinas, LLC, Docket No. 50-269, Oconee Nuclear Station, 
Unit1, Oconee County, South Carolina

    Date of amendment request: June 26, 2008.
    Description of amendment request: The proposed amendment would 
result in a revision of the current licensing basis (LB) in regard to 
high-energy line break (HELB) events occurring outside of containment 
for Oconee Nuclear Station, Unit 1 (ONS-1).
    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.
    Justification: The ONS-1 changes proposed in this LAR [license 
amendment request] include revisions to the current HELB methodology 
and mitigation strategy as documented in a new HELB report. This 
report provides the completed analysis for ONS HELBs including the 
descriptions of the station modifications that have been or will be 
made as a result of this comprehensive HELB reanalysis.
    The modifications associated with the revised HELB LB will be 
designed and installed in accordance with applicable quality 
standards such that the likelihood of failure of new or modified 
SSCs will not initiate failures, malfunctions, or inadvertent 
operations of existing accident mitigating SSCs [structures, 
systems, and components], such as the KHUs [Keowee hydro units], SSF 
[standby shutdown facility], HPI [high-pressure injection], or the 
Central Tie Switchyard 100 kV alternate power systems. For Turbine 
Building HELBs that could adversely affect equipment needed to 
stabilize and cooldown the units, the addition of the PSW [protected 
service water] System provides added assurances that safe shutdown 
can be readily established and maintained beyond the 72-hour SSF 
mission time.
    In conclusion, the changes will collectively enhance the 
station's overall design, safety, and risk margin; therefore, the 
proposed change does not involve a significant increase in the 
probability or consequence 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.
    Justification: The proposed modifications address potential 
adverse consequences from a HELB outside of containment. These 
modifications will be designed and installed in compliance with 
applicable quality standards such that there are reasonable 
assurances that they will neither introduce nor cause new failure 
mechanisms, malfunctions or accident initiators not already 
considered in the current HELB design and licensing basis.
    The overall effect of the changes to the HELB LB is considered 
an enhancement to the station's ability to achieve safe and cold 
shut down following a damaging HELB; therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Justification: The revised HELB LB will collectively enhance the 
station's overall design, safety, risk margin, and the station's 
ability to mitigate a HELB event; therefore, the proposed change 
does not involve a significant reduction in a margin of safety.
    Based on the above, Duke concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significance hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Melanie C. Wong.

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

    Date of amendment request: June 26, 2008.
    Description of amendment request: The proposed amendments would 
result

[[Page 54866]]

in a revision to portions of the Updated Final Safety Analysis Report 
(UFSAR) regarding the tornado licensing basis (LB).
    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:

    (4) Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No.
    Justification: Although a tornado does not constitute a 
previously-evaluated UFSAR Chapter 15 design basis accident or 
transient as described in 10 CFR 50.36(c)(2), it is a design basis 
criterion that is required to be considered in plant equipment 
design. The possibility of a tornado striking the ONS is 
appropriately considered in the UFSAR and Duke has concluded that 
the proposed changes do not increase the possibility that a damaging 
tornado will strike the site or increase the consequences from a 
damaging tornado.
    The modifications associated with the revised tornado LB will be 
designed and installed such that failures in these new or modified 
SSCs [structures, systems, and components will not initiate failures 
or inadvertent operations of existing ONS accident mitigating SSCs, 
such as the KHUs [Keowee hydro units], SSF [standby shutdown 
facility], or HPI [high-pressure injection] systems. The use of the 
NRC-approved TORMIS methodology confirmed that the risk from missile 
damage was acceptably low to vulnerable areas of the SSF structures 
and other SSCs required for SSD [safe shutdown]. As a result, there 
is reasonable assurance that a tornado missile will not prohibit the 
SSF system from fulfilling its tornado LB or other functions.
    Also, there are additional electrical power sources available 
which provide increased assurance that systems used to transition 
the units to SSD can be readily powered following a damaging 
tornado. The PSW [protected service water] System will provide 
additional assurance that SSD can be established and maintained.
    Overall, the changes proposed will increase assurance that 
potential challenges to the integrity of the RCS due to the effects 
of a damaging tornado will not result in a radioactive release to 
the environment. In conclusion, the changes will collectively 
enhance the station's overall design, safety, and risk margin; 
therefore, the probability or consequences of accidents previously 
evaluated are not significantly increased.
    (5) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Justification: Although only the SSF is credited for 
establishing and maintaining SSDHR [secondary side decay heat 
removal] and RCMU [reactor coolant makeup] during the first 72 hours 
following a damaging tornado, there are two relatively independent, 
diverse and redundant systems capable of safely shutting down all 
three units in the revised LB (SSF and PSW). Other modifications 
improve the ability of the SSF and PSW systems to perform their 
functions following a damaging tornado. The modifications will be 
designed and installed such that they will not introduce new failure 
mechanisms, malfunctions or accident initiators not already 
considered in the design and LB.
    In conclusion, the changes to the tornado LB will not degrade 
existing plant systems and will significantly enhance the station's 
ability to achieve SSD following a damaging tornado. The design and 
installation of the PSW system will be such that there is reasonable 
assurance that the system, including new power paths, will not 
contribute to the possibility of new or different kind of accident 
from any accident previously evaluated.
    (6) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Justification: The revised tornado LB will collectively enhance 
the station's overall design, safety, and risk margin; therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Melanie C. Wong.

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: June 26, 2008, as supplemented by 
letters dated August 4 and August 26, 2008.
    Description of amendment request: The proposed amendments would 
make changes to the Technical Specifications that are conforming or 
related to a change in fuel type from Westinghouse 0.400-inch OD 
Vantage+ fuel to Westinghouse 0.422-inch OD Vantage+ fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The requested amendment is related to a change in the reload 
fuel design. The design criteria for the reload fuel are consistent 
with those for the existing fuel and ensure that the reload fuel is 
compatible on the basis of coolant flow and neutronic 
characteristics, as well as DNB and peak cladding temperature 
requirements. The reload fuel design also ensures mechanical 
compatibility with the existing fuel, reactor core, control rods, 
steam supply system, and fuel handling tools and system.
    The reactor fuel and its analysis are not accident initiators. 
Therefore, the change in reload fuel design does not affect accident 
or transient initiation.
    The minimum boron accumulator concentration is also not an 
accident initiator. The proposed change to the minimum accumulator 
boron concentration Technical Specification limit ensures that the 
plant will continue to operate in a manner that provides acceptable 
levels of protection for health and safety of the public. Further, 
all design basis accidents and transients affected by the fuel 
upgrade were re-analyzed or evaluated using representative core 
designs and the results for each fuel type show all acceptance 
criteria will continue to be met.
    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.
    Use of the 422V+ fuel is consistent with current plant design 
bases and does not adversely affect any fission product barrier, nor 
does it alter the safety function of safety significant systems, 
structures and components or their roles in accident prevention or 
mitigation. The operational characteristics of 422V+ fuel are 
bounded by the safety analyses * * *. The 422V+ fuel design performs 
within existing fuel design limits.
    The proposed change to the minimum accumulator boron 
concentration Technical Specification limit ensures that the plant 
will continue to operate in a manner that provides acceptable levels 
of protection for health and safety of the public. Further, all 
design basis accidents and transients affected by the fuel upgrade 
were re-analyzed or evaluated using representative core designs and 
the results for each fuel type show all acceptance criteria will 
continue to be met.
    No equipment additions or modifications are included with the 
proposed change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which applicable 
design basis limits are determined, nor do they result in exceeding 
existing design basis limits. Thus, all licensed safety margins are 
maintained.

[[Page 54867]]

    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 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: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Lois M. James.

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

    Date of amendment request: June 27, 2008.
    Description of amendment request: These proposed changes consist of 
Proposed Change Number 583 (PCN-583) and are in support of the 
replacement of the steam generators (SGs) at SONGS Units 2 and 3. The 
proposed changes reflect revised SG inspection and repair requirements, 
and revised peak containment post-accident pressure resulting from 
installation of the replacement SGs.
    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 changes will reflect installation of Replacement 
Steam Generators (RSGs) at San Onofre Nuclear Generating Station 
(SONGS) Units 2 and 3. The proposed changes involve revising the 
Steam Generator (SG) tube inspection and repair [requirements] and 
revising the peak containment post-accident pressure.
    The proposed change to revise the SG tube inspection and repair 
[requirements] affect Technical Specifications (TSs) 3.4.17, ``Steam 
Generator (SG) Tube Integrity,'' 5.5.2.11, ``Steam Generator (SG) 
Program,'' and 5.7.2.c, ``Special Reports.'' The proposed TS 3.4.17, 
5.5.2.11, and 5.7.2.c revisions remove the repair method (sleeving), 
and Alternate Repair Criteria (ARC). The revisions replace the 44% 
tube repair criterion applicable to the original SGs, with a 35% 
(preliminary) tube repair criterion applicable to the RSGs. The 
revisions replace inspection requirements applicable to the tubing 
material of the original SGs with inspection requirements applicable 
to the tubing material of the RSGs, thus maintaining consistency 
with applicable material-specific regulatory guidance (TSTF-449, 
Revision 4). Overall, these revisions will ensure that all RSG tubes 
found by inservice inspection to contain flaws with a depth equal to 
or exceeding 35% (preliminary) of the nominal tube wall thickness 
will be plugged as required by revised TS 5.5.2.11.c.1.
    The TS 5.5.2.11.b SG structural integrity, accident induced 
leakage, and operational leakage performance criteria are unchanged 
and will continue to be met for the RSGs. Meeting the SG performance 
criteria provides reasonable assurance that the SG tubing will 
remain capable of maintaining reactor coolant pressure boundary 
integrity throughout each operating cycle and in the unlikely event 
of a design basis accident.
    The proposed change to the SG tube inspection and repair 
[requirements] will not affect the probability of any accident 
initiators. There will be no degradation in the performance of, or 
an increase in the number of challenges imposed on, safety-related 
equipment assumed to function during an accident. There will be no 
change to accident mitigation performance. The proposed change will 
not alter any assumptions or change any mitigation actions in the 
radiological consequence evaluations in the Updated Final Safety 
Analysis Report (UFSAR).
    The proposed change to the peak containment post-accident 
pressure will revise TS 5.5.2.15, ``Containment Leakage Rate Testing 
Program,'' by changing the stated values for peak containment 
internal pressure for the design-basis Loss-of-Coolant Accident 
(LOCA) and Main Steam Line Break (MSLB) accidents. The current LOCA 
value of 45.9 psig would be changed to 48.0 psig and the current 
MSLB value of 56.5 psig would be changed to 51.5 psig.
    The proposed change does not affect the probability of 
occurrence of an accident previously evaluated because it relates 
solely to the consequences of hypothesized accidents given that the 
accident has already occurred.
    The proposed change increases the calculated peak containment 
internal pressure for the LOCA events from 45.9 psig to 48.0 psig. 
The revised post-LOCA peak containment pressure is bounded by the 
existing and revised post-MSLB peak containment pressure and the 
containment design pressure of 60 psig. Despite the increase in the 
post-LOCA peak containment pressure, any post-accident containment 
leakage will still be limited to less than 0.1% containment air 
volume per day, consistent with current TS 5.5.2.15. Therefore, 
there is no increase in the radiological consequences of a LOCA as a 
result of the change to the post-LOCA peak containment pressure.
    The post-MSLB peak containment pressure decreases from 56.5 psig 
to 51.5 psig. Thus, the peak containment post-accident pressure is 
decreased as a result of this change, and there is no resulting 
increase in the consequences of a previously evaluated 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 change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    [Response: No.]
    The proposed change to the SG tube inspection and repair 
[requirements] deletes the repair method (sleeving) and the ARC 
applicable to the original SGs, and provides repair criteria and 
inspection requirements applicable to the RSGs. This will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
primary-to-secondary leakage that may be experienced during all 
plant conditions will be monitored to ensure it remains within 
current accident analysis assumptions. The proposed change does not 
adversely affect the method of operation of the SGs or the primary 
or secondary coolant chemistry controls and does not impact other 
plant systems or components.
    The proposed change to the peak containment post-accident 
pressure relates to two accidents, LOCA and MSLB, which are already 
evaluated in the Updated Final Safety Analysis Report (UFSAR).
    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.
    For the proposed change to the SG inspection and repair 
[requirements], the safety function of the SGs is maintained by 
ensuring the integrity of the tubes. SG tube integrity is a function 
of the design, environment, and the physical condition of the SG 
tubes. The proposed change, which deletes the repair method 
(sleeving) and the ARC applicable to the original SGs, and provides 
repair criteria and inspection requirements applicable to the RSGs, 
does not adversely affect the SG tube design or operating 
environment. SG tube integrity will continue to be maintained by 
implementing the TS 5.5.2.11 SG Program to manage SG tube 
inspection, assessment, and plugging. The requirements established 
by the TS 5.5.2.11 SG Program are consistent with those in the 
applicable design codes and standards.
    For the change to the peak containment post-accident pressure, 
the proposed change increases the calculated peak containment 
internal pressure for the LOCA events from 45.9 psig to 48.0 psig. 
The revised post-LOCA peak containment pressure is bounded by the 
existing and revised post-MSLB peak containment pressure. The post-
MSLB peak containment pressure decreases from 56.5 psig to 51.5 
psig. The proposed peak containment internal pressure for the MSLB 
accident is less than the containment design pressure of 60 psig and 
less than the previously calculated pressure.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based on the above, SCE concludes that the proposed amendments 
present no significant hazards consideration under the standards set 
forth in 10 CFR 50.92(c), and accordingly,

[[Page 54868]]

a finding of no significant hazards consideration is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: June 27, 2008.
    Description of amendment request: SONGS Units 2 and 3 requests 
adoption of an approved change to the standard technical specifications 
(STS) for Combustion Engineering Pressurized Water Reactor (PWR) Plants 
(NUREG-1432) and plant-specific technical specifications (TS), to allow 
replacing the departure from nucleate boiling (DNB) parameter limits 
with references to the core operating limits report (COLR) in 
accordance with Generic Letter 88-16, ``Removal of Cycle Specific 
Parameter Limits from Technical Specifications,'' dated October 4, 
1988. The changes are consistent with NRC approved Industry/Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler, TSTF-487, Revision 1, using the consolidated line-item 
improvement process (CLIIP).
    The NRC staff issued a notice of availability in the Federal 
Register on June 5, 2007 (72 FR 31108), including a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination, using the CLIIP process. The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
June 27, 2008.
    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:

    Criterion 1: Does the Proposed Change Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated?
    Response: No.
    The proposed amendment replaces the limit values of the reactor 
coolant system (RCS) DNB parameters (i.e., pressurizer pressure, RCS 
cold leg temperature, and RCS flow rate) in TS with references to 
the COLR, in accordance with the guidance of Generic Letter 88-16, 
to allow these parameter limit values to be recalculated without a 
license amendment. The proposed amendment does not involve operation 
of any required structures, systems, or components (SSCs) in a 
manner or configuration different from those previously recognized 
or evaluated. The cycle-specific values in the COLR must be 
calculated using the NRC-approved methodologies listed in TS 5.6.3, 
``Core Operating Limits Report (COLR).'' Replacing the RCS DNB 
parameter limits in TS with references to the COLR will maintain 
existing operating fuel cycle analysis requirements. Because these 
parameter limits are determined using the NRC approved 
methodologies, the acceptance criteria established for the safety 
analyses of various transients and accidents will continue to be 
met. Therefore, neither the probability nor consequences of any 
accident previously evaluated will be increased by the proposed 
change.
    Therefore, operation of the facility in accordance with the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident preciously evaluated.
    Criterion 2: Does the Proposed Change Create the Possibility of 
a New or Different Kind of Accident from any Previously Evaluated?
    Response: No.
    The proposed amendment to replace the RCS DNB parameter limits 
in TS with references to the COLR does not involve a physical 
alteration of the plant, nor a change or addition of a system 
function. The proposed amendment does not involve operation of any 
required SSCs in a manner or configuration different from those 
previously recognized or evaluated. No new failure mechanisms will 
be introduced by the proposed change. Therefore, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    Criterion 3: Does the Proposed Change Involve a Significant 
Reduction in the Margin of Safety?
    Response: No.
    The proposed amendment to replace the RCS DNB parameter limits 
in TS with references to the COLR will continue to maintain the 
margin of safety. The DNB parameter limits specified in the COLR 
will be determined based on the safety analyses of transients and 
accidents, performed using the NRC-approved methodologies that show 
that, with appropriate measurement uncertainties of these parameters 
accounted for, the acceptance criteria for each of the analyzed 
transients are met. This provides the same margin of safety as the 
limit values currently specified in the TS. Any future revisions to 
the safety analyses that require prior NRC approval are identified 
per the 10 CFR 50.59 review process.
    Therefore, the proposed amendment would not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: Michael T. Markley.

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

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

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

    Date of application for amendments: October 26, 2007.
    Brief description of amendments: The proposed amendment would 
revise the Technical Specification requirements related to control room 
envelope habitability in accordance with the NRC-approved Revision 3 of 
Technical Specification Task Force (TSTF) Standard Technical 
Specifications Change Traveler TSTF-448, ``Control Room Habitability.''
    Date of publication of individual notice in the Federal Register: 
August 29, 2008 (73 FR 51014).
    Expiration date of individual notice: September 29, 2008 (Public 
Comments) and October 28, 2008 (Requests for Hearing).

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

[[Page 54869]]

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

Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: September 27, 2007.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) TS 3.7.2, ``Main Steam Isolation 
Valves,'' and TS 3.7.3, ``Main Feedwater Isolation Valves, Main 
Feedwater Control Valves, Associated Bypass Valves and Tempering 
Valves,'' by removing the specific isolation time for the isolation 
valves from the associated surveillance requirements.
    Date of issuance: September 8, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 244 and 238.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: February 26, 2008 (73 
FR 10 10297).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 8, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 13, 2008.
    Brief description of amendment: The amendment replaces the current 
Arkansas Nuclear One, Unit No. 2 (ANO-2) TS 3.4.8, ``RCS [reactor 
coolant system] Specific Activity,'' limit on RCS gross specific 
activity with a new limit on RCS noble gas specific activity. The noble 
gas specific activity limit would be based on a new dose equivalent Xe-
133 (DEX) definition that would replace the current E Bar average 
disintegration energy definition. In addition, the current dose 
equivalent I-131 (DEI) definition would be revised to allow the use of 
additional thyroid dose conversion factors (DCFs). This request adopted 
Technical Specification Task Force (TSTF) change traveler TSTF-490, 
Revision 0, ``Deletion of E Bar Definition and Revision to RCS [reactor 
coolant system] Specific Activity Technical Specification'' (Agencywide 
Documents Access and Management System Accession No. ML052630462), for 
pressurized water reactor Standard Technical Specifications (STS) for 
ANO-2.
    Date of issuance: September 8, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: Unit 2-282.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications and license.
    Date of initial notice in Federal Register: May 6, 2008 (73 FR 
25039).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 8, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 22, 2007, as 
supplemented by letters dated April 22, and July 8, 2008.
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) Limiting Condition for Operation (LCO) 3.0.4 and 
Surveillance Requirement (SR) 4.0.4 to adopt the provisions of 
Industry/TS Task Force (TSTF) change TSTF-359, ``Increased Flexibility 
in Mode Restraints.'' This operating license improvement was made 
available by the U.S. Nuclear Regulatory Commission (NRC) on April 4, 
2003, as part of the consolidated line item improvement process. The 
proposed TS changes also include an additional application of LCO 
3.0.4.c for TS 3.4.3, ``Pressurizer Spray Valves.''
    Date of issuance: August 28, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: Unit 2-281.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: December 18, 2007 (72 
FR 71710). The supplements dated April 22, and July 8, 2008, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated August 28, 2008.
    No significant hazards consideration comments received: No.

[[Page 54870]]

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

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

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

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

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

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania

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

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of application for amendments: August 8, 2007.
    Brief description of amendments: The amendment replaces references 
to Section XI of the American Society of Mechanical Engineers (ASME) 
Boiler and Pressure Vessel Code with references to the ASME Code for 
Operation and Maintenance of Nuclear Power Plants (OM Code) in the 
applicable technical specification (TS) section for the Inservice 
Testing Program (IST) for the Exelon Generation Company, LLC, and 
AmerGen Energy Company, LLC, plants that have implemented industry 
Improved Technical Specifications. The changes are based on Technical 
Specification Task Force (TSTF) 479, Revision 0, ``Changes to Reflect 
Revision of 10 CFR 50.55a.'' For all units except Oyster Creek and TMI-
1, the amendments also incorporate TSTF-497, Revision 0, ``Limit 
Inservice Testing Program SR [Surveillance Requirement] 3.0.2 
Application to Frequencies of 2 Years or Less,'' which adds a provision 
in the applicable TS section to only apply the extension allowance of 
SR 3.0.2 to the frequency table listed in the TS as part of the IST 
program and to normal and accelerated inservice testing frequencies of 
two years or less, as applicable.
    Date of issuance: August 28, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 153, 153, 157, 157, 229, 222, 194, 1