Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 1712-1723 [E9-35]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 74, Number 8 (Tuesday, January 13, 2009)]
[Notices]
[Pages 1712-1723]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E9-35]


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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 December 18, 2008 to December 30, 2008. The 
last biweekly notice was published on December 30, 2008 (73 FR 79928).

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.
    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

[[Page 1713]]

Administrative Judge of the Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for hearing or a petition for leave to intervene must be 
filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve documents over the internet 
or in some cases to mail copies on electronic storage media. 
Participants may not submit paper copies of their filings unless they 
seek a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
hearingdocket@nrc.gov, or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer\TM\ to 
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ is free and is available 
at https://www.nrc.gov/site-help/e-submittals/install-viewer.html. 
Information about applying for a digital ID certificate is available on 
NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/
apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at https://www.nrc.gov/
site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, in accordance 
with 10 CFR 2.302(g), with their initial paper filing requesting 
authorization to continue to submit documents in paper format. Such 
filings must be submitted by: (1) First class mail addressed to the 
Office of the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff; or (2) courier, express mail, or expedited 
delivery service to the Office of the Secretary, Sixteenth Floor, One 
White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, 
Attention: Rulemaking and Adjudications Staff. Participants filing a 
document in this manner are responsible for serving the document on all 
other participants. Filing is considered complete by first-class mail 
as of the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained

[[Page 1714]]

absent a determination by the Commission, the presiding officer, or the 
Atomic Safety and Licensing Board that the petition and/or request 
should be granted and/or the contentions should be admitted, based on a 
balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii). To 
be timely, filings must be submitted no later than 11:59 p.m. Eastern 
Time on the due date.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a Presiding Officer. Participants are requested not to include personal 
privacy information, such as social security numbers, home addresses, 
or home phone numbers in their filings. With respect to copyrighted 
works, except for limited excerpts that serve the purpose of the 
adjudicatory filings and would constitute a Fair Use application, 
participants are requested not to include copyrighted materials in 
their submission.
    For further details with respect to this amendment action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr.resource@nrc.gov.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: September 30, 2008.
    Description of amendment request: This is an administrative change 
which would reflect the creation of new companies as approved by the 
NRC Order dated July 28, 2008. The amendments would not be implemented 
until the restructuring transactions have been completed. The 
amendments would revise the names on the plant licenses to match the 
names of the new companies. Entergy Nuclear FitzPatrick, LLC would be 
changed to Enexus Nuclear FitzPatrick, LLC. Entergy Nuclear Operations, 
Inc. would be changed to EquaGen Nuclear LLC.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed amendment would only change the names of the 
licensees and reflect the referenced NRC Order requirements; 
principal management and operational staffing for the restructured 
organization remain largely unchanged. The proposed name changes do 
not: (a) Involve a significant increase in the probability or 
consequences of an accident previously evaluated; (b) create the 
possibility of a new or different kind of accident from any accident 
previously evaluated; or (c) involve a significant reduction in a 
margin of safety.

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

FPL Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: November 25, 2008.
    Description of amendment request: Amend Renewed Operating Licenses 
DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP) Units 1 and 2 to 
incorporate new Large-Break LOCA (LBLOCA) analyses using the realistic 
LBLOCA methodology contained in NRC-approved WCAP-16009-P-A, 
``Realistic Large Break LOCA Evaluation Methodology Using Automated 
Statistical Treatment of Uncertainty Method (ASTRUM),'' and to revise 
Technical Specification (TS) 5.6.4.b to include reference to WCAP-
16009-P-A. This request also proposes to implement Technical 
Specification Task Force (TSTF) Traveler-363A. TSTF-363A eliminates the 
revision numbers and dates from the list of topical reports in TS 
5.6.4.b. TS 5.6.4.b provides the analytical methods used to determine 
the core operating limits. Relocation of the complete citations to the 
core operating limits report (COLR) will enable the current revisions 
of these topical reports to be used.
    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.
    This application proposes to incorporate LBLOCA analyses using 
the ASTRUM methodology, documented in WCAP-16009-P-A, ``Realistic 
Large Break LOCA Evaluation Methodology Using the Automated 
Statistical Treatment of Uncertainty Method (ASTRUM)'', in the PBNP 
licensing basis, add reference to WCAP-16009-P-A in the Technical 
Specification 5.6.4.b list of approved methodologies for 
establishing core operating limits, and relocate topical report 
detailed reference citations from TS 5.6.4.b to the COLR.
    Accident analyses are not accident initiators, therefore, this 
proposed licensing basis change does not involve a significant 
increase in the probability of an accident. The analyses using 
ASTRUM demonstrated that the acceptance criteria in 10 CFR 50.46, 
``Acceptance criteria for emergency core cooling systems for light-
water nuclear power reactors,'' were met. The NRC has approved WCAP-
16009-P-A for application to two-loop Westinghouse plants with upper 
plenum injection (UPI). Since the PBNP Units 1 and 2 are two-loop 
Westinghouse plants with UPI and the analysis results meet the 10 
CFR 50.46 acceptance criteria, this change does not involve a 
significant increase in the consequences of an accident.
    Addition of the reference to WCAP-16009-P-A in TS 5.6.4.b and 
relocation of topical report detailed citations to the COLR are 
administrative changes that do not affect the probability or 
consequences of an accident previously evaluated.
    The changes proposed in this license amendment do not involve a 
significant increase in the probability or consequences 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.
    This license amendment request proposes to incorporate LBLOCA 
analyses using the ASTRUM methodology, documented in WCAP-16009-P-A, 
``Realistic Large Break LOCA Evaluation Methodology Using the 
Automated Statistical Treatment of Uncertainty Method (ASTRUM),'' in 
the PBNP licensing basis, add a reference to WCAP-16009-P-A in the 
Technical Specification list of approved methodologies for 
establishing core operating limits, and relocate topical report 
detailed reference citations from TS 5.6.4.b to the COLR in 
accordance with approved TSTF-363A.
    There are no physical changes being made to the plant as a 
result of using the Westinghouse ASTRUM analysis methodology in 
WCAP-16009-P-A for performance of the LBLOCA analyses. No

[[Page 1715]]

new modes of plant operation are being introduced. The 
configuration, operation and accident response of the structures or 
components are unchanged by utilization of the new analysis 
methodology. Analyses of transient events have confirmed that no 
transient event results in a new sequence of events that could lead 
to a new accident scenario. The parameters assumed in the analysis 
are within the design limits of existing plant equipment.
    In addition, employing the Westinghouse ASTRUM LBLOCA analysis 
methodology does not create any new failure modes that could lead to 
a different kind of accident. The design of all systems remains 
unchanged and no new equipment or systems have been installed which 
could potentially introduce new failure modes or accident sequences. 
No changes have been made to any reactor protection system or 
emergency safeguards features instrumentation actuation setpoints.
    Based on this review, it is concluded that no new accident 
scenarios, failure mechanisms or limiting single failures are 
introduced as a result of the proposed methodology changes.
    Addition of the reference to WCAP-16009-P-A in the Technical 
Specifications is an administrative change that does not create the 
possibility of a new or different kind of accident. Relocation of 
topical report detailed citations from the Technical Specifications 
to the core operating limits report in accordance with approved 
TSTF-363A is an administrative change that does not create the 
possibility of a new or different kind of accident.
    The licensing basis and Technical Specification changes proposed 
in this license amendment do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    This application proposes to incorporate LBLOCA analyses using 
the ASTRUM methodology, documented in WCAP-16009-P-A, ``Realistic 
Large Break LOCA Evaluation Methodology Using the Automated 
Statistical Treatment of Uncertainty Method (ASTRUM)'', in the PBNP 
licensing basis, add a reference to WCAP-16009-P-A in the Technical 
Specifications list of approved methodologies for establishing core 
operating limits, and relocate topical report detailed reference 
citations from Technical Specification 5.6.4.b to the COLR.
    The analyses using ASTRUM demonstrated that the applicable 
acceptance criteria in 10 CFR 50.46, ``Acceptance criteria for 
emergency core cooling systems for light-water nuclear power 
reactors'' are met. Margins of safety for LBLOCAs include 
quantitative limits for fuel performance established in 10 CFR 
50.46. These acceptance criteria and the associated margins of 
safety are not being changed by this proposed new methodology. The 
NRC has approved WCAP-16009-P-A for application to two-loop 
Westinghouse plants with UPI. Since the PBNP is a two-loop 
Westinghouse plant with UPI and the analysis results meet the 10 CFR 
50.46 acceptance criteria, this change does not involve a 
significant reduction in a margin of safety. The analysis results 
using this methodology improve the margin of safety of PBNP.
    Addition of the reference to WCAP-16009-P-A in the Technical 
Specifications and implementation of TSTF-363A are administrative 
changes that do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Antonio Fernandez, Esquire, Senior Attorney, 
FPL Energy Point Beach, LLC, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Lois M. James.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: October 21, 2008.
    Description of amendment request: The proposed change would revise 
Sequoyah Nuclear Plant's Updated Final Safety Analysis Report (UFSAR) 
to require an inspection of each ice condenser within 24 hours of 
experiencing a seismic event greater than or equal to an operating 
basis earthquake (i.e., \1/2\ of a safe shutdown earthquake) within the 
5-week period after ice basket replenishment is completed. This will 
confirm that ice condenser lower inlet doors have not been blocked by 
ice fallout.
    The proposed amendment provides a procedural requirement to confirm 
the ice condenser maintains the ice condenser generic qualification as 
set forth in the UFSAR. Justification for the use of the proposed 
procedural requirement is based on reasonable assurance that the ice 
condenser lower inlet doors will open following a seismic event during 
the 5-week period and the low probability of a seismic event occurring 
coincident with or subsequently followed by a design basis accident.
    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 analyzed accidents of consideration in regard to changes 
potentially affecting the ice condenser are a loss-of-coolant 
accident (LOCA) and a steam or feedwater line break inside 
containment. The ice condenser is an accident mitigator and is not 
postulated as being the initiator of a LOCA or high energy line 
break (HELB). The ice condenser is structurally designed to 
withstand a Safe Shutdown Earthquake (SSE) plus a Design Basis 
Accident (DBA) and does not interconnect or interact with any 
systems that interconnect or interact with the reactor coolant, main 
steam or feedwater systems. Because the proposed changes do not 
result in, or require any physical change to the ice condenser that 
could introduce an interaction with the reactor coolant, main steam 
or feedwater systems, there can be no change in the probability of 
an accident previously evaluated.
    Under the current licensing basis, the ice condenser ice baskets 
would be considered fully fused prior to power ascension and the ice 
condenser would perform its accident mitigation function even if a 
safe shutdown seismic event occurred coincident with or just 
preceding the accident. Under the proposed change, there is some 
finite probability that, within 24 hours following a seismic 
disturbance, a LOCA or HELB in containment could occur within 5 
weeks of the completion of ice basket replenishment. However, 
several factors provide defense-in-depth and tend to mitigate the 
potential consequences of the proposed change.
    DBAs are not assumed to occur simultaneously with a seismic 
event. Therefore, the coincident occurrence of a LOCA or HELB with a 
seismic event is strictly a function of the combined probability of 
the occurrence of independent events, which in this case is very 
low. Based on the Probabilistic Risk Assessment model and seismic 
hazard analysis, the combined probability of occurrence of a seismic 
disturbance greater than or equal to an OBE [operating basis 
earthquake] during the 5-week period following ice replenishment 
coincident with or subsequently followed by a LOCA or HELB during 
the time required to perform the proposed inspection (24 hours) and 
if required by technical specifications, complete unit shutdown (37 
hours), is less than 3.89E-09 for Sequoyah [Nuclear Plant]. This 
probability is well below the threshold that is typically considered 
credible.
    Even if ice were to fall from ice baskets during a seismic event 
occurring coincident with or subsequently followed by an accident, 
the ice condenser would be expected to perform its intended safety 
function. There is reasonable assurance that the ice condenser would 
function properly following a seismic event within the 5-week period 
due to inherent conservatisms in the 1974 test data, the low 
likelihood of flow channel and floor drain blockage, and improbable 
blocking of the lower inlet doors by any potential fallout.
    Based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The ice condenser is expected to 
perform its intended safety function under all circumstances 
following a LOCA or HELB in containment.

[[Page 1716]]

    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change affects the assumed timing of a postulated 
seismic and DBA applied to the ice condenser and provides an 
alternate methodology to confirm the ice condenser lower inlet doors 
are capable of opening. As previously discussed, the ice condenser 
is not postulated as an initiator of any DBA. The proposed change 
does not impact any plant system, structure or component that is an 
accident initiator. The proposed change does not involve any 
hardware changes to the ice condenser or other changes that could 
create new accident mechanisms. Therefore, there can be no new or 
different accidents created from those previously identified and 
evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The performance of the fuel cladding and the reactor coolant 
system will not be impacted by the proposed change.
    The requirement to inspect the ice condensers within 24 hours of 
experiencing seismic activity greater than or equal to an OBE during 
the 5-week period following the completion of ice basket 
replenishment will confirm that the ice condenser lower inlet doors 
are capable of opening. This inspection will confirm that the ice 
condenser doors remain fully capable of performing their intended 
safety function under credible circumstances.
    The proposed change affects the assumed timing of a postulated 
seismic and DBA applied to the ice condenser and provides an 
alternate methodology in confirming the ice condenser lower inlet 
doors are capable of opening. As previously discussed, the combined 
probability of occurrence of a LOCA or HELB and a seismic 
disturbance greater than or equal to an OBE during the ``period of 
potential exposure'' is less than 3.89E-09 for Sequoyah. This 
probability is well below the threshold that is typically considered 
credible.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety. The SQN [Sequoyah Nuclear Plant] 
ice condenser will perform its intended safety function under 
credible circumstances.
    The changes proposed in this license amendment request (LAR) do 
not make any physical alteration to the ice condensers, nor does it 
affect the required functional capability of the ice condenser in 
any way. The intent of the proposed change to the UFSAR is to 
eliminate an overly restrictive waiting period prior to unit ascent 
to power operations following the completion of ice basket 
replenishment. The required inspection of the ice condenser 
following a seismic event greater than or equal to an OBE will 
confirm that the ice condenser lower inlet doors will continue to 
fully perform their safety function as assumed in the SQN safety 
analyses.
    Thus, it can be concluded that the proposed change does not 
involve a significant reduction in the margin of safety.

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

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: October 23, 2008.
    Description of amendment request: The requested change is a partial 
adoption of Technical Specification Task Force Change Traveler No. 491 
(TSTF-491), Revision 2, ``Removal of Main Steam and Feedwater Valve 
Isolation Times.'' The proposed change only revises TS 3.7.1.5, ``Main 
Steam Line Isolation Valves,'' by relocating the main steam isolation 
valve closure time from Surveillance Requirement (SR) 4.7.1.5.1 to the 
Bases. The proposed amendment deviates from TSTF-491 in that the 
current Sequoyah Nuclear Plant (SQN) TS 3.7.1.6, ``Main Feedwater 
Isolation, Regulating, and Bypass Valves,'' and associated surveillance 
requirements do not include the main feedwater valve closure times, and 
thus, TSTF-491 changes to TS 3.7.1.6 would not apply to the SQN TSs 
without modification. Because of this deviation from TSTF-491, the 
proposed amendment will be processed as a typical amendment.
    Basis for proposed no significant hazards consideration 
determination: To satisfy the requirements of 10 CFR 50.91(a), the 
licensee's amendment request incorporates by reference the proposed no 
significant hazards consideration (NSHC) published in the Federal 
Register on October 5, 2006 (71 FR 58884), as part of the Consolidated 
Line Item Improvement Process associated with TSTF-491. That NSHC is 
reproduced below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change allows relocating main steam and main 
feedwater valve isolation times to the Licensee Controlled Document 
that is referenced in the Bases. The proposed change is described in 
Technical Specification Task Force (TSTF) Standard TS Change 
Traveler TSTF-491 related to relocating the main steam and main 
feedwater valves isolation times to the Licensee Controlled Document 
that is referenced in the Bases and replacing the isolation time 
with the phrase, ``within limits.''
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
The proposed changes relocate the main steam and main feedwater 
isolation valve times to the Licensee Controlled Document that is 
referenced in the Bases. The requirements to perform the testing of 
these isolation valves are retained in the TS. Future changes to the 
Bases or licensee-controlled document will be evaluated pursuant to 
the requirements of 10 CFR 50.59, `` Changes, test and 
experiments,'' to ensure that such changes do not result in more 
than minimal increase in the probability or consequences of an 
accident previously evaluated.
    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 consequences of any accident previously 
evaluated. Further, the proposed changes do not increase the types 
and the amounts of radioactive effluent that may be released, nor 
significantly increase individual or cumulative occupation/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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes relocate the main steam and main feedwater 
valve isolation times to the Licensee Controlled Document that is 
referenced in the Bases. In addition, the valve isolation times are 
replaced in the TS with the phrase, ``within limits.'' The changes 
do not involve a physical altering of the plant (i.e., no new or 
different type of equipment will be installed) or a change in 
methods governing normal plant operation. The requirements in the TS 
continue to require testing of the main steam and main feedwater 
isolation valves to ensure the proper functioning of these isolation 
valves.

[[Page 1717]]

    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes relocate the main steam and main feedwater 
valve isolation times to the Licensee Controlled Document that is 
referenced in the Bases. In addition, the valve isolation times are 
replaced in the TS with the phrase, ``within limits.'' Instituting 
the proposed changes will continue to ensure the testing of main 
steam and main feedwater isolation valves. Changes to the Bases or 
License Controlled Document are performed in accordance with 10 CFR 
50.59. This approach provides an effective level of regulatory 
control and ensures that main steam and feedwater isolation valve 
testing is conducted such that there is no significant reduction in 
the margin of safety.
    The margin of safety provided by the isolation valves is 
unaffected by the proposed changes since there continue to be TS 
requirements to ensure the testing of main steam and main feedwater 
isolation valves. The proposed changes maintain sufficient controls 
to preserve the current margins of safety.

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

Notice of Issuance of Amendments to Facility Operating Licenses

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

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: June 19, 2008, as supplemented 
by letter dated October 1, 2008.
    Brief description of amendments: The amendments (1) revise the 
technical specifications (TS) control rod notch surveillance 
requirement (SR) frequency in TS 3.1.3, ``Control Rod Operability,'' 
and (2) revise Example 1.4-3 in Section 1.4, ``Frequency,'' to clarify 
the applicability of the 1.25 surveillance test extension. The licensee 
is proposing to adopt the approved Technical Specification Task Force 
(TSTF) change traveler TSTF-475, Revision 1, ``Control Rod Notch 
Testing Frequency.'' A notice of availability of TSTF-475, Revision 1, 
was published in the Federal Register on November 13, 2007 (72 FR 
63935).
    In addition, the proposed amendment would remove Note 2 associated 
with SR 3.1.3.3 for Unit 1, which is a cycle-specific note and has 
expired. This change is administrative in nature and does not affect 
the no significant hazards consideration determination.
    Date of issuance: December 15, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 250 and 278.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the technical specifications.
    Date of initial notice in Federal Register: October 7, 2008 (73 FR 
58671). The supplemental letter dated October 1, 2008, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments are contained 
in a safety evaluation dated December 15, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 21, 2008.
    Brief description of amendment: The amendments remove 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. 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.''
    Date of issuance: December 17, 2008.
    Effective date: As of the date of issuance and shall be implemented 
no later thanOctober 1, 2009.
    Amendment Nos.: 116; 308; and 247.
    Facility Operating License No. DPR-21, Renewed Facility Operating 
License No. DPR-65, and Renewed Facility Operating License No. NPF-49: 
Amendments revised the License and Technical Specifications.
    Date of initial notice in Federal Register: September 23, 2008 (73 
FR 54864).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 17, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 18, 2007, as 
supplemented by

[[Page 1718]]

letters dated September 18 and October 28, 2008.
    Brief description of amendment: The amendments revise the Technical 
Specifications (TSs) by adding a Control Room Habitability Program and 
revising the TS on the Control Room Ventilation System in accordance 
with Technical Specifications Task Force (TSTF) change traveler TSTF-
448, ``Control Room Habitability.'' License conditions are added 
regarding the initial performance of the new surveillance.
    Date of issuance: December 22, 2008.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 258 and 239.
    Facility Operating License Nos. DPR-26 and DPR-64: The amendment 
revised the License and the Technical Specifications.
    Date of initial notice in Federal Register: March 25, 2008 (73 FR 
15785). The September 18 and October 28, 2008, supplements provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 22, 2008.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3), Westchester 
County, New York

    Date of application for amendment: March 13, 2008.
    Brief description of amendment: The amendment revises the licensing 
basis for passive failures in fluid systems for IP2 and IP3 such that 
the loss-of-coolant accident (LOCA) recirculation phase single passive 
failure is assumed to occur 24 hours or greater following initiation of 
a LOCA. Also, the IP2 single passive failure licensing basis for the 
component cooling water system is revised such that a passive failure 
is assumed to occur 24 hours or greater following initiation of a LOCA.
    Date of issuance: December 4, 2008.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 257 and 238.
    Facility Operating License Nos. DPR-26 and DPR-64: The amendment 
revised the License and the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: July 1, 2008 (73 FR 
37503).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 4, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 31, 2008, as 
supplemented by letter dated July 30, 2008.
    Brief description of amendment: The amendment revises the 
description of fuel assemblies specified in Technical Specification 
(TS) 4.2.1 and adds the approved AREVA licensed topical report BAW-
10240(P)-A, ``Incorporation of M5 Properties in Framatome ANP Approved 
Methods,'' to the analytical methods referenced in TS 5.6.5.b to permit 
the use of M5 alloy and supporting analytical methods in future reload 
designs.
    Date of issuance: December 12, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 234.
    Facility Operating License No. DPR-20: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 25, 2008 (73 FR 
15786). The supplement dated July 30, 2008, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the initial Federal Register notice. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated December 12, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 5, 2008.
    Brief description of amendment: The amendment would revise renewed 
facility operating license DPR-20 to correct an error, generated during 
Palisades license transfer approval on April 11, 2007, and also remove 
several outdated license conditions pertaining to surveillance 
requirements.
    Date of issuance: December 15, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 235.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 9, 2008 (73 
FR 52416).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 15, 2008.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of application for amendment: December 21, 2007, as 
supplemented on August 1, 2008.
    Brief description of amendment: The amendment revises several 
Technical Specification (TS) sections to allow relaxations of various 
Reactor Trip System/Engineered Safety Feature (RTS/ESF) logic 
completion times, bypass test times, allowable outage times, and 
surveillance testing intervals that were previously reviewed and 
approved by NRC under Westinghouse Reports WCAP-14333-P-A, 
``Probabilistic Risk Analysis of RPS [reactor protection system] and 
ESFAS [ESF Actuation System] Test Times and Completion Times,'' and 
WCAP-15376-P-A, ``Risk-Informed Assessment of the RTS and ESFAS 
Surveillance Test Intervals and Reactor Trip Breaker Test and 
Completion Times.''
    Date of issuance: December 29, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 282 and 166.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revises the License and Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2008 (73 FR 
32745). The August 1, 2008, supplemental letter provided clarifying 
information that was within the scope of the initial notice and did not 
change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 29, 2008.
    No significant hazards consideration comments received: No.

[[Page 1719]]

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

    Date of application for amendment: December 27, 2007, as 
supplemented on July 28, 2008.
    Brief description of amendment: The amendment establishes more 
effective and appropriate action, surveillance, and administrative 
requirements related to ensuring habitability within the control room 
envelope in accordance with the NRC-approved Technical Specification 
Task Force Traveler (TSTF)-448, Revision 3, and changes the technical 
specifications (TS) related to the control room emergency ventilation 
system in TS Section 3.7.10, ``Control Room Emergency Ventilation 
(CREV) System,'' and TS Section 5.5.16, ``Control Room Envelope 
Habitability Program.'' The amendment also adds a license condition to 
support implementation of the TS change.
    Date of issuance: December 30, 2008.
    Effective date: As of the date of issuance to be implemented within 
180 days.
    Amendment Nos.: 307 (CNP-1), 289 (CNP-2).
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Renewed Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: January 29, 2008 (73 FR 
5224). The supplemental letter dated July 28, 2008, provided 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the Nuclear 
Regulatory Commission staff's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 30, 2008.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220 and 50-410, 
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2 (NMP 1 and 2), 
Oswego County, New York

    Date of application for amendment: June 24, 2008.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) by (1) replacing the references to 
Section XI of the American Society of Mechanical Engineers (ASME) 
Boiler and Pressure Vessel Code (Code) with references to the ASME Code 
for Operation and Maintenance of Nuclear Power Plants (OM Code) in the 
applicable TS section for the Inservice Testing (IST) Program for NMP 1 
TS 6.5.4 and NMP 2 TS 5.5.6; and (2) revising the allowance to extend 
IST frequencies by 25 percent to clearly state that the allowance is 
applicable to IST frequencies of 2 years or less. The proposed changes 
are based on TS Task Force (TSTF) Standard Technical Specification 
Change Traveler 479-A, Revision 0, ``Changes to Reflect Revision of 10 
CFR 50.55a,'' as modified by TSTF-497-A, Revision 0, ``Limit Inservice 
Testing Program SR 3.0.2 Application to Frequencies of 2 Years or 
Less.''
    Date of issuance: December 22, 2008.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 199 and 129.
    Renewed Facility Operating License Nos. DPR-063 and NPF-069: The 
amendments revise the License and TSs.
    Date of initial notice in Federal Register: October 7, 2008 (73 FR 
58674).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 22, 2008.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: December 26, 2007, as 
supplemented onNovember 25, 2008.
    Brief description of amendments: The amendments revised the action 
and surveillance requirements in Technical Specification (TS) 3.7.10, 
``Control Room Ventilation System (CRVS),'' and add a new 
administrative controls program, TS 5.5.18, ``Control Room Envelope 
Habitability Program.'' The amendments are consistent with the TS 
traveler TSTF-448, ``Control Room Habitability,'' Revision 3.
    Date of issuance: December 23, 2008.
    Effective date: As of its date of issuance and shall be implemented 
within 180 days from the date of issuance.
    Amendment Nos.: Unit 1-201; Unit 2-202.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: January 29, 2008 (73 FR 
5227). The supplement dated November 25, 2008, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 23, 2008.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of application for amendment: July 30, 2008, as supplemented 
by letters dated September 29, and October 20, 2008.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.8.3, ``Onsite Power Distribution Systems,'' to 
establish a separate TS Action statement for inoperable inverters 
associated with the 120 volt alternating current distribution panels. 
The amendment extends the allowed outage time for inoperable inverters 
from 8 hours to 24 hours.
    Date of issuance: December 18, 2008.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 175.
    Facility Operating License No. NPF-57: The amendment revised the 
TSs and the License.
    Date of initial notice in Federal Register: September 9, 2008 (73 
FR 52421). The letters dated September 29, and October 20, 2008, 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination or expand 
the application beyond the scope of the original Federal Register 
notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 18, 2008.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: September 4, 2008.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to adopt TS Task Force (TSTF) Change Traveler TSTF-
447, Revision 1, ``Elimination of Hydrogen Recombiners and Change to 
Hydrogen and Oxygen Monitors.''
    Date of issuance: December 23, 2008.

[[Page 1720]]

    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 72.
    Facility Operating License No. NPF-90: Amendment revises the TSs 
and Facility Operating License.
    Date of initial notice in Federal Register: October 21, 2008 (73 FR 
62569).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 23, 2008.
    No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: December 28, 2007.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.7.2, ``Main Steam Isolation Valves (MSIVs),'' and 
TS Table 3.3.2-1, ``Engineered Safety Feature Actuation System 
Instrumentation.''
    Date of issuance: December 18, 2008.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 189.
    Facility Operating License No. NPF-30: The amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register : March 25, 2008 (73 FR 
15790).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 18, 2008.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket No. 50-281, Surry 
Power Station, Unit 2, Surry County, Virginia

    Date of application for amendment: December 17, 2007, as 
supplemented on April 24, 2008, and June 27, 2008.
    Brief Description of amendment: The amendment revised Technical 
Specification (TS) 4.4, pertaining to the containment leak rate testing 
program. The TS change permitted a one-time 5-year extension to the 
once per 10-year frequency of the performance-based leakage rate 
testing program for Type A tests, which are done in accordance with 
Regulatory Guide (RG) 1.163, ``Performance-Based Containment Leak-Test 
Program.'' This one time exception to the RG 1.163 requirement allows 
the next Type A test to be performed no later than October 26, 2015.
    Date of issuance: December 18, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 263.
    Renewed Facility Operating License No. DPR-37: Amendment changed 
the license and the technical specifications.
    Date of initial notice in Federal Register: January 15, 2008 (73 FR 
2551).
    The supplemental letters dated April 24, 2008, and June 27, 2008, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated December 18, 2008.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: January 15, 2008, as supplemented by 
letter dated October 27, 2008.
    Brief description of amendment: The amendment modified 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 U.S. Nuclear Regulatory Commission (NRC)-approved TS 
Task Force (TSTF) Standard Technical Specification change traveler 
TSTF-448, Revision 3, ``Control Room Habitability.'' Specifically, the 
amendment modified TS 3.7.10, ``Control Room Emergency Ventilation 
System (CREVS),'' and established a CRE habitability program in TS 
Section 5.5, ``Administrative Controls--Programs and Manuals.''
    Date of issuance: December 24, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 179.
    Facility Operating License No. NPF-42. The amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: February 12, 2008 (73 
FR 8072). The supplemental letter dated October 27, 2008, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 24, 2008.
    No significant hazards consideration comments received: No.

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

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