[Federal Register Volume 74, Number 201 (Tuesday, October 20, 2009)]
[Notices]
[Pages 53774-53784]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-24915]



[[Page 53774]]

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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0456]


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) 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 September 24, 2009, to October 7, 2009. The 
last biweekly notice was published on October 6, 2009 (74 FR 51327).

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 Title 10 of the Code of Federal 
Regulations (10 CFR), Section 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 
and Directives Branch (RDB), TWB-05-B01M, 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 faxed to the RDB at 301-492-3446. Documents may be examined, and/or 
copied for a fee, at the NRC's Public Document Room (PDR), located at 
One White Flint North, Public File Area O1F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's ``Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR part 2. Interested person(s) 
should consult a current copy of 10 CFR 2.309, which is available at 
the Commission's PDR, located at One White Flint North, Public File 
Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases 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

[[Page 53775]]

participate fully in the conduct of the hearing.
    If a hearing is requested, 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.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
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 all adjudicatory 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 an exemption in accordance with the procedures 
described below.
    To comply with the procedural requirements of E-Filing, at least 
ten (10) days prior to the filing deadline, the petitioner/requestor 
should contact the Office of the Secretary by e-mail at 
<a href="/cdn-cgi/l/email-protection#660e0307140f0801480209050d03122608140548010910"><span class="__cf_email__" data-cfemail="c5ada0a4b7acaba2eba1aaa6aea0b185abb7a6eba2aab3">[email protected]</span></a>, 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<SUP>TM</SUP> 
to access the Electronic Information Exchange (EIE), a component of the 
E-Filing system. The Workplace Forms Viewer<SUP>TM</SUP> is free and is 
available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying for a digital ID certificate is 
available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory e-
filing system may seek assistance through the ``Contact Us'' link 
located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC Meta-System Help Desk, which is 
available between 8 a.m. and 8 p.m., Eastern Time, Monday through 
Friday, excluding government holidays. The Meta-System Help Desk can be 
contacted by telephone at 1-866-672-7640 or by e-mail at 
<a href="/cdn-cgi/l/email-protection#c38e908b87ed91a6b0acb6b1a0a683adb1a0eda4acb5"><span class="__cf_email__" data-cfemail="d598869d91fb87b0a6baa0a7b6b095bba7b6fbb2baa3">[email protected]</span></a>.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, 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 request and/
or petition 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).
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a Presiding Officer. Participants are requested not to include personal 
privacy information, such as social security numbers, home addresses, 
or home phone numbers in their filings, unless an NRC regulation or 
other law requires submission of such information. 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 submissions.
    For further details with respect to this license amendment 
application, 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 O1F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the ADAMS Public Electronic Reading Room on the Internet at the NRC Web 
site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have 
access to ADAMS or who encounter problems in accessing the documents 
located in ADAMS, should contact the NRC PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to <a href="/cdn-cgi/l/email-protection#1a6a7e6834687f69756f68797f5a746879347d756c"><span class="__cf_email__" data-cfemail="3a4a5e4814485f49554f48595f7a544859145d554c">[email protected]</span></a>.

[[Page 53776]]

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

    Date of amendment request: August 10, 2009.
    Description of amendment request: The proposed amendment would 
revise the RBS Technical Specifications (TSs) to support operation with 
24-month fuel cycles. Specifically, the change addresses certain TS 
Surveillance Requirement (SR) frequencies that are specified as ``18 
months'' by revising them to ``24 months'' in accordance with the 
guidance of U.S. Nuclear Regulatory Commission (NRC) Generic Letter 91-
04, ``Changes in Technical Specification Surveillance Intervals to 
Accommodate a 24-Month Fuel Cycle,'' dated April 2, 1991.
    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 TS changes involve a change in the surveillance 
testing intervals and allowable values to facilitate a change in the 
operating cycle length. The proposed TS changes do not physically 
impact the plant. The proposed TS changes do not degrade the 
performance of, or increase the challenges to, any safety systems 
assumed to function in the accident analysis. The proposed TS 
changes do not impact the usefulness of the SRs in evaluating the 
operability of required systems and components, or the way in which 
the surveillances are performed. In addition, the frequency of 
surveillance testing is not considered an initiator of any analyzed 
accident, nor does a revision to the frequency introduce any 
accident initiators. The specific value of the allowable value is 
not considered an initiator of any analyzed accident. Therefore, the 
proposed change does not involve a significant increase in the 
probability of an accident previously evaluated.
    The consequences of a previously evaluated accident are not 
significantly increased. The proposed change does not affect the 
performance of any equipment credited to mitigate the radiological 
consequences of an accident. Evaluation of the proposed TS changes 
demonstrated that the availability of credited equipment is not 
significantly affected because of other more frequent testing that 
is performed, the availability of redundant systems and equipment, 
and the high reliability of the equipment. Historical review of 
surveillance test results and associated maintenance records did not 
find evidence of failures that would invalidate the above 
conclusions.
    The allowable values have been developed in accordance with [NRC 
Regulatory Guide] 1.105, ``Instrument Setpoints,'' to ensure that 
the design and safety analysis limits are satisfied. The methodology 
used for the development of the allowable values ensures the 
affected instrumentation remains capable of mitigating design basis 
events as described in the safety analyses and that the results and 
radiological consequences described in the safety analyses remain 
bounding. Therefore, the proposed change does not alter the ability 
to detect and mitigate events and, as such, does not involve a 
significant increase in the consequences of an accident previously 
evaluated.

Standby Liquid Control System

    The proposed change in required weight of Boron-10 in [standby 
liquid control (SLC)] does not physically impact the plant, nor does 
it degrade the performance of, or increase the challenges to, any 
safety systems assumed to function in the accident analysis. The 
consequences of a previously evaluated accident are not increased. 
The proposed change does not affect the performance of any equipment 
credited to mitigate the radiological consequences of an accident. 
Evaluation of the proposed TS changes demonstrated that the 
availability of credited equipment is not affected. Therefore, the 
proposed change does not alter the ability to detect and mitigate 
events and, as such, does not involve a significant increase in the 
consequences of an accident previously evaluated.

Loss of Power Instrumentation

    A change to the Allowable Values (AVs) is proposed for Table 
3.3.8.1-1, Item 1.c and Item 2.c. The proposed change is the result 
of application of the RBS Instrument Setpoint Methodology using 
plant-specific drift values and incorporating margins available 
based on a revised off-site reliability study. Application of this 
methodology results in AVs that more accurately reflect total device 
accuracy, as well as that of test equipment and calculated drift 
between surveillances. The proposed change will not result in any 
hardware changes. The instrumentation is not assumed to be an 
initiator of any analyzed event. Existing operating margin between 
plant conditions and actual plant setpoints is not significantly 
reduced due to the proposed changes. The role of the instrumentation 
is in mitigating and thereby, limiting the consequences of 
accidents.
    The AVs were developed to ensure the design and safety analysis 
limits are satisfied. The methodology used for the development of 
the AVs ensures that: (1) The affected instrumentation remains 
capable of mitigating design basis events as described in the safety 
analysis, and (2) the results and radiological consequences 
described in the safety analysis remain bounding. Additionally, the 
proposed change does not alter the plant's ability to detect and 
mitigate events. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The change in the degraded voltage protection voltage AVs allows 
the protection scheme to function as originally designed. The 
proposed allowable values ensure that the Class 1E distribution 
system remains connected to the offsite power system when adequate 
offsite voltage is available and motor starting transients are 
considered.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed TS changes involve a change in the surveillance 
testing intervals and allowable values to facilitate a change in the 
operating cycle length. The proposed TS changes do not introduce any 
failure mechanisms of a different type than those previously 
evaluated, since there are no physical changes being made to the 
facility. No new or different equipment is being installed. No 
installed equipment is being operated in a different manner. As a 
result, no new failure modes are being introduced. The way 
surveillance tests are performed remains unchanged. A historical 
review of surveillance test results and associated maintenance 
records indicated there was no evidence of any failures that would 
invalidate the above conclusions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

Standby Liquid Control System

    The proposed change to the required weight of Boron-10 in SLC 
does not introduce any failure mechanisms of a different type than 
those previously evaluated, since there are no physical changes 
being made to the facility. No new or different equipment is being 
installed. No installed equipment is being operated in a different 
manner. As a result, no new failure modes are being introduced. The 
way surveillance tests are performed remains unchanged. A historical 
review of surveillance test results and associated maintenance 
records indicated there was no evidence of any failures that would 
invalidate the above conclusions.

Loss of Power Instrumentation

    The proposed change in AVs is the result of application of the 
Instrument Setpoint Methodology using plant-specific drift values 
and does not create the possibility of a new or different kind of 
accident from any accident previously evaluated. This is based upon 
the fact that the method and manner of plant operation are 
unchanged.
    The use of the proposed AVs does not impact safe operation of 
the plant in that the safety analysis limits are maintained. The 
proposed change in AVs involves no system additions. The AVs are 
revised to ensure the affected instrumentation remains capable of 
mitigating accidents and transients. Plant equipment will not be 
operated in a manner different from previous operation, except that 
setpoints may be changed. No additional failure mechanisms are 
introduced as a result of the changes to the allowable values. Since 
operational methods remain unchanged and

[[Page 53777]]

the operating parameters were evaluated to maintain the plant within 
existing design basis criteria, no different type of failure or 
accident is created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed TS changes involve a change in the surveillance 
testing intervals and allowable values to facilitate a change in the 
operating cycle length. The impact of these changes on system 
availability is not significant, based on other more frequent 
testing that is performed, the existence of redundant systems and 
equipment, and overall system reliability. Evaluations have shown 
there is no evidence of time dependent failures that would impact 
the availability of the systems. The proposed changes do not 
significantly impact the condition or performance of structures, 
systems, and components relied upon for accident mitigation. The 
proposed changes in TS instrumentation allowable values are the 
result of application of the RBS setpoint methodology using plant 
specific drift values. The revised allowable values more accurately 
reflect total instrumentation loop accuracy including drift while 
continuing to protect any assumed analytical limit. The proposed 
changes do not result in any hardware changes or in any changes to 
the analytical limits assumed in accident analyses. Existing 
operating margin between plant conditions and actual plant setpoints 
is not significantly reduced due to these changes. The proposed 
changes do not significantly impact any safety analysis assumptions 
or results.

Standby Liquid Control System

    The proposed change in required weight of Boron-10 in SLC is to 
facilitate a change in the operating cycle length. The proposed 
change does not result in any hardware changes or in any changes to 
the analytical limits assumed in accident analyses. Existing 
operating margin between plant conditions and actual plant setpoints 
is not reduced due to this change. The proposed change does not 
impact any safety analysis assumptions or results. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

Loss of Power Instrumentation

    The proposed protection voltage AVs are low enough to prevent 
inadvertent power supply transfer, but high enough to ensure that 
sufficient voltage is available to the required equipment. The 
proposed change does not involve a reduction in a margin of safety. 
The proposed change was developed using a methodology to ensure 
safety analysis limits are not exceeded. As such, this proposed 
change does not involve a significant reduction in a margin of 
safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: August 25, 2009.
    Description of amendment request: The proposed amendment would 
allow for a one-time extension to the ten-year frequency for the next 
Palisades Nuclear Plant (PNP) containment Type A integrated leak rate 
test (ILRT) that is required by Technical Specification (TS) 5.5.14. 
The proposed change would permit the existing ILRT frequency to be 
extended from ten years to approximately 11.25 years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed exemption involves a one-time extension to the 
current interval for Type A containment testing. The current test 
interval of 120 months (10 years) would be extended on a one-time 
basis to no longer than approximately 135 months from the last Type 
A test. The proposed extension does not involve either a physical 
change to the plant or a change in the manner in which the plant is 
operated or controlled. The containment is designed to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. As such, 
the containment and the testing requirements invoked to periodically 
demonstrate the integrity of the containment exist to ensure the 
plant's ability to mitigate the consequences of an accident, and do 
not involve the prevention or identification of any precursors of an 
accident. Therefore, this proposed extension does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    This proposed extension is for the Type A containment leak rate 
tests only. The Type B and C containment leak rate tests would 
continue to be performed at the frequency currently required by the 
PNP TS. As documented in NUREG 1493, Type B and C tests have 
identified a very large percentage of containment leakage paths and 
that the percentage of containment leakage paths that are detected 
only by Type A testing is very small. The PNP Type A test history 
supports this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as (1) activity based and 
(2) time based. Activity based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment combined with the containment inspections performed in 
accordance with ASME Section XI, the Maintenance Rule, and TS 
requirements serve to provide a high degree of assurance that the 
containment would not degrade in a manner that is detectable only by 
a Type A test. Based on the above, the proposed extension does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed revision to the TS involves a one-time extension to 
the current interval for Type A containment testing. The containment 
and the testing requirements invoked to periodically demonstrate the 
integrity of the containment exist to ensure the plant's ability to 
mitigate the consequences of an accident and do not involve the 
prevention or identification of any precursors of an accident. The 
proposed TS change does not involve a physical change to the plant 
or the manner in which the plant is operated or controlled. 
Therefore, the proposed TS change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to the TS involves a one-time extension to 
the current interval for Type A containment testing. The proposed TS 
change does not involve a physical change to the plant or a change 
in the manner in which the plant is operated or controlled. The 
specific requirements and conditions of the TS Containment Leak Rate 
Testing Program exist to ensure that the degree of containment 
structural integrity and leak-tightness that is considered in the 
plant safety analysis is maintained. The overall containment leak 
rate limit specified by TS is maintained. The proposed change 
involves only the extension of the interval between Type A 
containment leak rate tests. The proposed surveillance interval 
extension is bounded by the 15-month extension currently authorized 
within NEI 94-01, Revision 0. Type B and C containment leak

[[Page 53778]]

rate tests would continue to be performed at the frequency currently 
required by TS. Industry experience supports the conclusion that 
Type B and C testing detects a large percentage of containment 
leakage paths and that the percentage of containment leakage paths 
that are detected only by Type A testing is small. The containment 
inspections performed in accordance with ASME Section XI and the 
Maintenance Rule serve to provide a high degree of assurance that 
the containment would not degrade in a manner that is detectable 
only by Type A testing. The combination of these factors ensures 
that the margin of safety in the plant safety analysis is 
maintained. The design, operation, testing methods and acceptance 
criteria for Type A, B, and C containment leakage tests specified in 
applicable codes and standards would continue to be met, with the 
acceptance of this proposed change, since these are not affected by 
changes to the Type A test interval. Therefore, the proposed TS 
change does not involve a significant reduction in a margin of 
safety.

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

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: August 26, 2009.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) Section 6.5 that governs 
administrative controls of High Radiation Areas (HRA) to incorporate 
the HRA administrative controls contained within the Standard Technical 
Specifications, NUREG-1433, Revision 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of Vermont Yankee Nuclear Power Station (VY) in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed amendment does not impact the operability of any 
structure, system or component that affects the probability of an 
accident or that supports mitigation of an accident previously 
evaluated. The proposed amendment does not affect reactor operations 
or accident analysis and has no radiological consequences. The 
operability requirements for accident mitigation systems remain 
consistent with the licensing and design basis. Therefore, the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The operation of VY in accordance with the proposed amendment 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed amendment does not change the design or function of 
any component or system. No new modes of failure or initiating 
events are being introduced. Therefore, operation of VY in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The operation of VY in accordance with the proposed amendment 
will not involve a significant reduction in a margin of safety.
    The proposed amendment does not change the design or function of 
any component or system. The proposed amendment does not involve any 
safety limits, safety settings or safety margins. The TS 
administrative access controls for high radiation areas are being 
replaced with those contained in section 5.7 of NUREG-1433 to 
provide additional requirements and options for the control of these 
areas.
    Therefore, operation of VY in accordance with the proposed 
amendment will not involve a significant reduction in the margin to 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Nancy Salgado.

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

    Date of amendment request: May 27, 2009, as supplemented on August 
28, 2009.
    Description of amendment request: The proposed amendment requests 
an increase in the maximum steady-state power level at NMP2 from 3467 
megawatts thermal (MWt) to 3988 MWt. This represents a 15-percent 
increase over the current licensed thermal power.
    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?
    No, the increase in power level discussed herein will not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    The proposed change will increase NMP2's authorized maximum 
power level from the current licensed thermal power (CLTP) level of 
3467 megawatts thermal (MWt) to 3988 MWt. In support of this 
Constant Pressure Extended Power Uprate (CPPU), a comprehensive 
evaluation was performed for nuclear steam supply system (NSSS) and 
balance of plant (BOP) systems, structures, components, and analyses 
that could be affected by this change. The effect of increasing the 
maximum power level from the CLTP of 3467 MWt to 3988 MWt on the 
NMP2 licensing and design bases was evaluated. The result of this 
evaluation is that all plant components, as modified, will continue 
to be capable of performing their design function at an uprated core 
power of 3988 MWt. In addition, an evaluation of the accident 
analyses concludes that applicable analysis acceptance criteria 
continue to be met. Power level is an input assumption to the 
equipment design and accident analyses, but it is not an initiator 
for any transient or accident. Therefore, no accident initiators are 
affected by this uprate and no challenges to any plant safety 
barriers are created by this change.
    Therefore, operation of the facility in accordance with the 
proposed change does not involve a significant increase in the 
probability of an accident previously evaluated.
    This change does not affect the release paths, the frequency of 
release, or the source term for release for any accidents previously 
evaluated in the Updated Safety Analysis Report (USAR). Structures, 
systems, and components (SSC) required to mitigate transients remain 
capable of performing their design functions, and thus were found 
acceptable. The source terms used to assess radiological 
consequences have been reviewed and determined to bound operation at 
the uprated condition. The results of EPU [extended power uprate] 
accident evaluations do not exceed the U. S. Nuclear Regulatory 
Commission (NRC) approved acceptance limits.
    The spectrum of postulated accidents and transients has been 
investigated and are shown to meet the regulatory criteria to which 
NMP2 is currently licensed. In the area of fuel and core design, the 
Safety Limit Minimum Critical Power ratio (SLMCPR) and other 
applicable Specified Acceptable Fuel Design Limits (SAFDLS) are 
still met. Continued compliance with the SLMCPR and other SAFDLs is 
confirmed on a cycle specific basis consistent with criteria 
accepted by the NRC.
    Challenges to the reactor coolant pressure boundary were 
evaluated at EPU conditions

[[Page 53779]]

(pressure, temperature, flow, and radiation) and found to meet the 
acceptance criteria for allowable stresses. Adequate overpressure 
margin is maintained.
    Challenges to the containment have been evaluated and the 
containment and its associated cooling system continue to meet 
applicable regulatory requirements. The increase in the calculated 
post Loss of Coolant Accident (LOCA) suppression pool temperature 
above the current peak temperature was evaluated and determined to 
be acceptable.
    Radiological release events (accidents) have been evaluated and 
shown to meet the requirements of 10 CFR 50.67.
    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?
    No, the increase in power level discussed herein will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    The proposed change will increase NMP2's authorized maximum 
power level from the CLTP level of 3467 MWt to 3988 MWt. Equipment 
that could be affected by EPU has been evaluated. No new operating 
mode, safety-related equipment lineup, accident scenario, or 
equipment failure mode was identified. The full spectrum of accident 
considerations has been evaluated and no new or different kind of 
accident has been identified. This Constant Pressure Extended Power 
Uprate utilizes a standard evaluation methodology applied to known 
technology employed within the range of current or modified plant 
capabilities. As such, the plant safety-related equipment continues 
to operate in accordance with regulatory criteria. Evaluations were 
performed using NRC approved codes, standards and methods. No new 
accidents or event precursors have been identified.
    All structures, systems and components previously required for 
the mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes do not adversely 
affect safety-related systems or components and do not challenge the 
performance or integrity of any safety-related system. This change 
does not adversely affect any current system interfaces or create 
any new interfaces that could result in an accident or malfunction 
of a different kind than was previously evaluated. Operating at a 
core power level of 3988 MWt does not create any new accident 
initiators or precursors.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. [Does the proposed amendment] involve a significant reduction 
in a margin of safety?
    No, the increase in power level discussed herein will not 
involve a significant reduction in a margin of safety.
    Comprehensive analyses of the proposed changes have concluded 
that relevant design and safety acceptance criteria will be met 
without a significant reduction in margins of safety. The analyses 
supporting EPU have demonstrated that the NMP2 SSCs are capable of 
safely performing at EPU conditions. The analyses identified and 
defined the major input parameters to the NSSS, analyzed NSSS design 
transients, and evaluated the capabilities of the NSSS fluid 
systems, NSSS/BOP interfaces, NSSS control systems, and NSSS and BOP 
components, as appropriate. Radiological consequences of design 
basis events remain within regulatory limits and are not increased 
significantly. The analyses confirmed that NSSS and BOP SSCs are 
capable, some with modifications, of achieving EPU conditions 
without significant reduction in margins of safety.
    Analyses have shown that the integrity of primary fission 
product barriers will not be significantly affected as a result of 
the power increase. Calculated loads on SSCs important to safety 
have been shown to remain within design allowables under EPU 
conditions for all design basis event categories. Plant response to 
transients and accidents do not result in exceeding acceptance 
criteria. As appropriate, the evaluations that demonstrate 
acceptability of EPU have been performed using methods that have 
either been reviewed and approved by the NRC staff, or that are in 
compliance with regulatory review guidance and standards established 
for maintaining adequate margins of safety. These evaluations 
demonstrate that there are no significant reductions in the margins 
of safety.
    Maximum power level is one of the inherent inputs that determine 
the safe operating range defined by the accident analyses. The 
Technical Specifications ensure that NMP2 is operated within the 
bounds of the inputs and assumptions used in the accident analyses. 
The acceptance criteria for the accident analyses are conservative 
with respect to the operating conditions defined by the Technical 
Specifications. The engineering reviews performed for the constant 
pressure extended power uprate confirm that the accident analyses 
criteria are met at the revised maximum allowable thermal power 
level of 3988 MWt, as well as at the rated thermal power (RTP) 
levels specified in the Facility Operating License and Technical 
Specifications. Therefore, the adequacy of the revised Facility 
Operating Licenses and Technical Specifications to maintain the 
plant in a safe operating range is also confirmed, and the increase 
in maximum allowable power level does not involve a significant 
decrease in a margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Nancy L. Salgado.

Nine Mile Point Nuclear Station, LLC (NMPNS) Docket No. 50-410, Nine 
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York

    Date of amendment request: June 29, 2009, as supplemented on August 
13, 2009.
    Description of amendment request: The proposed amendment would 
revise the NMP2 Technical Specification (TS) 5.5.12 by replacing the 
reference to Regulatory Guide (RG) 1.163 with a reference to Nuclear 
Energy Institute (NEI) Topical Report NEI 94-01, Revision 2-A, as the 
implementation document used by NMPNS to develop the NMP2 performance-
based leakage testing program in accordance with Option B of Title 10 
of the Code of Federal Regulations (10 CFR) Part 50. The proposed 
amendment would allow the next primary containment integrated leak rate 
test (ILRT) to be performed within 15 years from the last ILRT as 
opposed to the current 10-year interval, and would allow successive 
ILRTs to be performed at 15-year intervals.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment involves changes to the NMP2 10 CFR 50 
Appendix J Testing Program Plan. The proposed amendment does not 
involve a physical change to the plant or a change in the manner in 
which the plant is operated or controlled. The primary containment 
function is to provide an essentially leak tight barrier against the 
uncontrolled release of radioactivity to the environment for 
postulated accidents. As such, the containment itself and the 
testing requirements to periodically demonstrate the integrity of 
the containment exist to ensure the plant's ability to mitigate the 
consequences of an accident, and do not involve any accident 
precursors or initiators. Therefore, the probability of occurrence 
of an accident previously evaluated is not significantly increased 
by the proposed amendment.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 2, for development of the NMP2 performance-based 
leakage testing program. Implementation of these guidelines 
continues to provide adequate assurance that during design basis 
accidents, the primary containment and its components will limit

[[Page 53780]]

leakage rates to less [then] the values assumed in the plant safety 
analyses. The potential consequences of extending the ILRT interval 
from 10 years to 15 years have been evaluated by analyzing the 
resulting changes in risk. The increase in risk in terms of person-
rem per year within 50 miles resulting from design basis accidents 
was estimated to be acceptably small, and the increase in the large 
early release frequency resulting from the proposed change was 
determined to be within the guidelines published in NRC RG 1.174. 
Additionally, the proposed change maintains defense-in-depth by 
preserving a reasonable balance among prevention of core damage, 
prevention of containment failure, and consequence mitigation. NMPNS 
has determined that the increase in conditional containment failure 
probability due to the proposed change would be very small. 
Therefore, it is concluded that the proposed amendment does not 
significantly increase the consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment adopts the NRC-accepted guidelines of 
NEI-94-01, Revision 2, for development of the NMP2 performance-based 
leakage testing program, and establishes a 15 year interval for the 
performance of the primary containment ILRT. The containment and the 
testing requirements to periodically demonstrate the integrity of 
the containment exist to ensure the plant's ability to mitigate the 
consequences of an accident, and do not involve any accident 
precursors and initiators. The proposed change does not involve a 
physical change to the plant (i.e., no new or different type of 
equipment will be installed) or a change to the manner in which the 
plant is operated or controlled.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment adopts the NRC-accepted guidelines of 
NEI-94-01, Revision 2, for development of the NMP2 performance-based 
leakage testing program, and establishes a 15 year interval for the 
performance of the primary containment ILRT. The amendment does not 
alter the manner in which safety limits, limiting safety system 
setpoints, or limiting conditions for operation are determined. The 
specific requirements and conditions of the 10 CFR 50 Appendix J 
Testing Program Plan, as defined in the TS, ensure that the degree 
of primary containment structural integrity and leak-tightness that 
is considered in the plant safety analyses is maintained. The 
overall containment leakage rate limit specified by the TS is 
maintained, and the Type A, B, and C containment leakage tests will 
continue to be performed at the frequencies established in 
accordance with the NRC-accepted guidelines of NEI 94-01, Revision 
2.
    Containment inspections performed in accordance with other plant 
programs serve to provide a high degree of assurance that the 
containment will not degrade in a manner that is detectable only by 
an ILRT. In addition, the on-line containment monitoring capability 
that is inherent to inerted boiling water reactor containments 
allows for the detection of gross containment leakage that may 
develop during power operation. This combination of factors ensures 
that evidence of containment structural degradation is identified in 
a timely manner. Furthermore, a risk assessment using the current 
NMP2 Probabilistic Risk Assessment model concluded that extending 
the ILRT test interval from 10 years to 15 years results in a very 
small change to the NMP2 risk profile.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Nancy L. Salgado.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendment request: September 15, 2009.
    Description of amendment request: The proposed amendment revises 
Technical Specification (TS) 3.3.2, in Appendix A to Facility Operating 
License Nos. NPF-2 and NPF-8 for the Joseph M. Farley Nuclear Plant 
Units 1 and 2, respectively. P-11 is an engineered safety feature 
actuation system (ESFAS) permissive/interlock which permits normal unit 
cooldown and depressurization without actuation of safety injection 
(SI) from low pressurizer pressure. P-12 is an ESFAS permissive/
interlock which permits normal unit cooldown and depressurization 
without actuation of SI and main steam line isolation on the condition 
of low steam line pressure. Both P-11 and P-12 circuits use input from 
three protection channels. The current wording of Condition K in TS 
3.3.2 states, ``Two channels inoperable.'' As a result, Condition K 
does not explicitly address the possible conditions of one channel or 
three channels inoperable, possibly creating a literal compliance 
issue. The proposed Condition K change from ``Two channels inoperable'' 
to ``One or more channels inoperable'' will resolve the current literal 
compliance issue. The change does not alter the current Condition K 
required action, it simply clarifies that the required action must be 
performed for one, two, or three P-11 or P-12 channels inoperable. In 
addition, an editorial change is proposed for TS 5.6.8 to correct the 
citation of a condition requiring a report for the post-accident 
monitoring instrumentation. The current TS 5.6.8 text states, ``When a 
report is required by Condition B or G of LCO [limiting conditions for 
operation] 3.3.3. * * *'' The citation of Condition B is correct while 
Condition G does not currently exist for LCO 3.3.3; instead TS 5.6.8 
should cite Condition F.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to TS 3.3.2 does not significantly increase 
the probability or consequences of an accident previously evaluated 
in the FSAR. These interlocks do not directly initiate an accident. 
The consequences of accidents previously evaluated in the FSAR are 
not adversely affected by these changes because the changes are made 
to reflect the Improved Standard Technical Specifications and the 
interlocks are verified to be in the required state for the unit 
condition.
    The proposed change to TS 5.6.8 corrects an editorial error and 
therefore does not significantly increase the probability or 
consequences of a previously evaluated accident.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change to TS 3.3.2 does not create the possibility 
of a new or different kind of accident than any accident already 
evaluated in the FSAR. No new accident scenario, failure mechanisms, 
or limiting single failures are introduced as a result of the 
proposed change. The proposed TS 3.3.2 change does not challenge the 
performance or integrity of any safety-related systems. Therefore, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously analyzed.
    The proposed change to TS 5.6.8 corrects an editorial error and 
therefore does not create the possibility of a new or different kind 
of accident from any accident previously analyzed.

[[Page 53781]]

    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to TS 3.3.2 does not involve a significant 
reduction in a margin of safety. The proposed change is made to 
accurately reflect the format of the Improved Standard Technical 
Specifications. The actuation setpoints specified by the Technical 
Specifications and safety analysis limits assumed in the accident 
analysis are unchanged. The margin of safety associated with these 
trip setpoints and the safety analysis acceptance criteria is 
unchanged. Therefore, the proposed change to TS 3.3.2 will not 
significantly reduce the margin of safety as defined in the 
Technical Specifications.
    The proposed change to TS 5.6.8 corrects an editorial error and 
therefore involves no significant reduction in a margin of safety.
    Based on the above, SNC 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 significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Branch Chief: Jon H. Thompson, Acting.

Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260 and 50-
296, Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, 
Alabama

    Date of amendment request: July 27, 2009 (TS-465).
    Description of amendment request: The proposed change is to 
eliminate Technical Specification (TS) surveillance requirement (SR) 
3.6.1.3.11 and the requirement to perform water leak rate testing on 
the listed containment isolation valves. More specifically, the 
proposed change eliminates water local leak rate testing of valves in 
the Containment Leak Rate Program that are being tested to verify the 
combined leakage rate is within the limit that ensures the suppression 
pool level is sufficient to keep lines that terminate below the water 
level for at least 30 days without additional make-up.
    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.

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to the scope of water leak rate testing for 
the subject valves does not affect the probability of the design 
basis accidents. The valves will continue to be maintained in an 
operable state, and in their current design configuration. There is 
no correlation between the scope of the water leak rate testing and 
accident probability.
    TVA reviewed the postulated consequences of design basis events 
on primary containment isolation under the proposed change. The 
primary containment structure, including access openings, 
penetrations and the containment heat removal system, is designed so 
that the containment structure and its internal compartments can 
withstand, without exceeding the design leakage rate (2.0% per day), 
the peak accident pressure and temperature that could occur during 
any postulated LOCA [loss-of-coolant accident].
    For the purposes of considering the consequences of LOCAs under 
the proposed change, a single active failure of a CIV [containment 
isolation valve] or a passive failure of the closed system were 
reviewed, within the limits of the existing licensing basis. Under 
the existing licensing basis, a pipe rupture of seismically 
qualified ECCS [emergency core cooling system] piping does not have 
to be assumed concurrent with the LOCA, except if it is a 
consequence of the LOCA. Consequential failures can be eliminated, 
since a LOCA inside containment is separated from the ECCS piping by 
the containment structure. Consequential failures of the ECCS piping 
from LOCA's outside containment are outside the Appendix J design 
considerations, although they are adequately addressed through the 
redundancy and separation of the ECCS design. A single active 
failure of the CIV, under the LOCA condition, can be accommodated 
since the closed and filled system piping and the suppression pool 
water inventory remain as the leakage barriers. The ECCS passive 
failure criterion does require consideration of system leaks, but 
not pipe breaks, beyond the initiating LOCA. Pipe leakage, 
equivalent to the leakage from a valve or pump seal failure, should 
be considered at 24 hours or greater post-LOCA. The capability to 
make-up inventory to the suppression pool is adequate to ensure that 
postulated seat leakage and pipe leakage does not result in a 
condition that jeopardizes pool level. Make-up capability exists to 
the suppression pool. Actions to make-up to the suppression pool are 
delineated in Emergency Operating Instructions.
    Therefore, the proposal to eliminate the subject water leak rate 
tests 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?
    This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The acceptability of the proposed change to the scope of water 
leak rate testing for the subject valves is based on maintaining the 
existing barriers to primary containment leakage, and ensuring that 
the suppression pool level is assured for 30 days during all design 
basis, post-accident modes of operation. By meeting these dual 
objectives, the plant response to the design basis events will be 
unchanged, and no new accident scenarios will be encountered. These 
two objectives are related, in that, the suppression pool inventory 
creates a passive barrier to primary containment atmospheric leakage 
for valves associated with penetrations which are located below the 
minimum water level of the pool.
    The proposed Technical Specification change does not alter the 
configuration of the subject containment isolation valves or their 
associated systems. The valves will continue to be tested and 
maintained to ensure their operability. The subject valves are all 
isolation valves associated with lines that penetrate the primary 
containment. For closed system valves, the redundant isolation 
boundary for each of the affected valves is the closed system 
associated with the valve. The closed system piping is verified via 
a 10 CFR 50 Appendix J Type A test. The integrity of the closed 
systems is also monitored and controlled via Technical Specification 
5.5.2, ``Primary Coolant Sources Outside Containment.''
    The subject valves may be open, or change state, post-accident 
to support the design function of their associated ECCS systems 
(HPCI [high-pressure coolant injection], Core Spray, RHR [residual 
heat removal]), RCIC [reactor core isolation coolant] or RHR 
Sampling using the Post Accident Sampling System. The subject valves 
function as system valves during the periods when they are open or 
in an intermediate state, not as containment isolation valves. 
Reliance is placed on the suppression pool seal and the closed 
system piping to maintain the barrier between primary and secondary 
containment atmospheres.
    Therefore, with the valve configuration and closed systems 
configuration unaffected by the proposed change, the existing 
barriers to primary containment atmospheric leakage are maintained, 
so long as the suppression pool level is ensured.
    The suppression pool is designed and operated so that it is 
filled with water in accordance with Technical Specifications 
3.6.2.2, ``Suppression Pool Water Level,'' and the associated Bases. 
As such, the supply of water in the suppression pool is assured for 
30 days during all design basis, post-accident modes of operation. 
Water leak rate testing has historically been performed on valves 
associated with lines that connect to the suppression pool. The 
acceptance criteria for combined leakage from these penetrations is 
72.79 cfh [cubic feet per hour]. This leakage rate is at a level 
which ensures the 30-day post-accident suppression pool level.

[[Page 53782]]

    As mentioned above, the integrity of the closed system piping is 
verified via a 10 CFR 50 Appendix J Type A test and is monitored and 
controlled via Technical Specification 5.5.2. TS 5.5.2 establishes a 
program to monitor and control leakage from systems located outside 
containment that could contain highly radioactive fluids during a 
serious transient or accident. This program applies to the ECCS and 
RCIC systems affected by the proposed change and ensures that 
leakage into secondary containment via packing, flanges, seals, 
etc., is controlled. Leakage from these systems has been found to be 
very low, and well below the 20 gpm [gallons per minute] limit 
established for these systems. The proposed change is not expected 
to contribute to higher levels of system leakage. Normal operational 
monitoring of suppression pool level, operator rounds, housekeeping 
inspections, and system pressure testing further ensure external 
leakage is identified and minimized while suppression pool level is 
being maintained.
    A review of water leak rate test data for the subject CIVs 
showed that the valves have had leakage rates within the acceptance 
criteria. Testing of the valves in accordance with ASME [American 
Society of Mechanical Engineers] Code requirements ensure valve 
operability.
    Therefore, leakage past the CIVs is expected to be low and in 
keeping with the design basis for the suppression pool. However, the 
capability does exist to make-up water to the suppression pool if 
necessary. Existing Emergency Operating Instructions require actions 
if suppression pool level is less than the required level. Thus, the 
level of the suppression pool is ensured, independent of the current 
CIV water leak rate testing requirement.
    The proposed change to the scope of water leak rate testing for 
the subject valves maintains the existing barriers to primary 
containment leakage, and ensures that the suppression pool level is 
assured for 30 days during all design basis, post-accident modes of 
operation. Therefore, the plant response to the design basis events 
is unchanged, and the proposal 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?
    This change does not involve a significant reduction in a margin 
of safety.
    As discussed in the responses to questions 1 and 2, the proposed 
change does not alter the plant response to existing accident 
scenarios, and does not introduce new or different scenarios. So the 
margin of safety from a design basis accident standpoint is 
maintained.
    Historically, the leakage rate through the subject valves has 
been determined in accordance with TS SR 3.6.1.3.11. This leakage 
rate has always been within the acceptance criteria. Quantifying 
leakage past the CIVs has been used to ensure that the suppression 
pool level is assured for 30 days post-accident. Under the proposed 
change, this leakage rate will not be quantified. In addition, 
closed system leakage is monitored and controlled by an existing 
Technical Specification program. Closed system leakage has been 
found to be very low on each of the units, and is currently well 
below the 20 gpm allowable. Therefore, leakage past the CIVs is 
expected to be low and in keeping with the design basis for the 
suppression pool. However, the capability does exist, and is 
proceduralized, to make-up water to the suppression pool if 
necessary. Thus the current capability to maintain adequate 
suppression pool level for 30 days post-accident is assured under 
the proposed change.
    Therefore the proposed change to the scope of water leak rate 
testing for the subject valves 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Thomas H. Boyce.

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.

Florida Power and Light Company, Docket Nos. 50-250 and 251, Turkey 
Point Plant Units 3, and 4, Miami-Dade County, Florida

    Date of application for amendment: September 1, 2009.
    Description of amendment request: Delay the date specified in 
License Amendments 234 and 229 for the implementation of the Boraflex 
Remedy in the spent fuel pools.
    Date of publication of individual notice in the  Federal Register: 
September 15, 2009 (74 FR 47278).
    Expiration date of individual notice: October 15, 2009 (Public 
comments) and November 16, 2009 (Hearing requests).

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of application for amendments: July 23, 2009.
    Description of amendments request: Revise the scope of the 
inservice inspections required in the tubesheet regions of the steam 
generators.
    Date of publication of individual notice in the1 Federal Register: 
August 28, 2009 (74 FR 44405).
    Expiration date of individual notice: September 28, 2009 (Public 
comments) and October 27, 2009 (Hearing requests).

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

[[Page 53783]]

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 (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#59293d2b772b3c2a362c2b3a3c19372b3a773e362f"><span class="__cf_email__" data-cfemail="ddadb9aff3afb8aeb2a8afbeb89db3afbef3bab2ab">[email protected]</span></a>.

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

    Date of application for amendment: May 28, 2009, as supplemented by 
letter dated August 3, 2009.
    Brief description of amendment: The amendments eliminated working 
hour restrictions from Technical Specification (TS) 5.2.2 for Palo 
Verde Nuclear Generating Station, Units 1, 2, and 3, to support 
compliance with the revisions to Title 10 of the Code of Federal 
Regulations (10 CFR), Part 26, ``Fitness for Duty Programs,'' that 
became effective on March 31, 2008. The changes are consistent with the 
NRC-approved Technical Specification Task Force (TSTF) Standard 
Technical Specification change traveler, TSTF-511, Revision 0, 
``Eliminate Working Hour Restrictions from TS 5.2.2 to Support 
Compliance with 10 CFR part 26.''
    Date of issuance: September 30, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: Unit 1-175; Unit 2-175; Unit 3-175.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: July 28, 2009 (74 FR 
37247). The supplemental letter dated August 3, 2009, 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.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 30, 2009.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: May 13, 2009.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 5.5.8, ``Inservice Testing Program,'' by 
incorporating TS Task Force Traveler (TSTF) 479, ``Changes to Reflect 
Revision of 10 CFR [Title 10 of the Code of Federal Regulations] 
50.55a,'' and TSTF-497, ``Limit Inservice Testing Program SR 
[Surveillance Requirement] 3.0.2 Application to Frequencies of 2 Years 
or Less.'' Specifically, the amendments (1) replace references to the 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code, Section XI with the ASME Code for Operation and 
Maintenance of Nuclear Power Plants for inservice testing activities, 
and (2) applies the extension allowance of SR 3.0.2 to other normal and 
accelerated inservice testing frequencies of 2 years or less that were 
not included in the frequencies of the table listed in TS 5.5.8.a.
    Date of issuance: September 28, 2009.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 294 and 270.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the License and Technical Specifications.
    Date of initial notice in Federal Register: July 14, 2009 (74 FR 
34046). The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated September 28, 2009.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 20, 2008, as supplemented by 
letter dated August 12, 2009.
    Brief description of amendment: The amendment revised Technical 
Specification 5.6.5, ``Core Operating Limits Report (COLR),'' to add a 
reference to an analytical method that will be used to determine the 
core operating limits. The change is needed to support the use of GE14 
fuel during refueling outage 15 scheduled for the fall of 2009.
    Date of issuance: September 29, 2009.
    Effective date: As of the date of issuance and shall be implemented 
prior to Cycle 16 operation.
    Amendment No.: 166.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 23, 2009 (74 FR 
4249). The supplemental letter dated August 12, 2009, 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 September 29, 2009.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 17, 2008, as supplemented by 
letters dated February 26, June 30, and September 24, 2009.
    Brief description of amendment: The amendment revised the Waterford 
3 Technical Specifications (TSs) to take credit for soluble boron in 
Region 1 (cask storage pit) and Region 2 (spent fuel pool and refueling 
canal) fuel storage racks for the storage of both Standard and Next 
Generation Fuel assemblies. Two new TSs were added which included a 
surveillance that ensures the required boron concentration is 
maintained in the spent fuel storage racks and to reflect the results 
of the new criticality analysis.
    Date of issuance: September 30, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 223.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 14, 2009 (74 FR 
17228). The application dated September 17, 2008, contained an 
evaluation of the TS change in accordance with 10 CFR 50.91(a)(1) using 
criteria in 10 CFR 50.92(c), and the licensee determined that the 
change involved no significant hazards consideration (NSHC). However, 
based on the discussions between the staff and the licensee, the 
licensee provided a revised NSHC in its supplemental letter dated 
February 26, 2009. Based on the February 26, 2009, revised NSHC, the 
staff's proposed

[[Page 53784]]

NSHC determination was published in the Federal Register on April 14, 
2009. The supplemental letters dated June 30 and September 24, 2009, 
provided additional information that clarified the application, did not 
expand the scope of the application as noticed, and did not change the 
staff's 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 September 30, 2009.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit No. 2, Beaver County, Pennsylvania

    Date of application for amendment: October 10, 2008, as 
supplemented by letters dated. June 16 and July 14, 2009.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 5.5.5 to allow an additional method of repair for 
steam generator (SG) tubes by installation of leak limiting Alloy 800 
sleeves developed by Westinghouse and clarifies an existing reporting 
requirement in TS 5.6.6.2.4 concerning SG tube inspections.
    Date of issuance: September 30, 2009.
    Effective date: As of the date of issuance and shall be implemented 
prior to achieving Mode 4 during startup from the fall 2009 refueling 
outage.
    Amendment No: 170
    Facility Operating License No. NPF-73. Amendment revised the 
License and TSs.
    Date of initial notice in Federal Register: February 17, 2009 (74 
FR 7482). The June 16 and July 14, 2009, supplemental letters 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 September 30, 2009.
    No significant hazards consideration comments received: No.

Northern States Power Company--Minnesota, LLC, Docket No. 50-263, 
Monticello Nuclear Generating Plant, Wright County, Minnesota

    Date of application for amendment: May 29, 2009.
    Brief description of amendment: The amendment changes the Technical 
Specifications, revising the applicability for isolation of the Reactor 
Water Cleanup System on a Standby Liquid Control system initiation to 
align with the modes stated in Specification 3.1.7.
    Date of issuance: September 28, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 164.
    Facility Operating License No. DPR-22. Amendment revised the 
Facility Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: July 28, 2009 (74 FR 
37248). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 28, 2009.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-424 and 50-425, Vogtle Electric 
Generating Plant, Units 1 and 2, Burke County, Georgia

    Date of application for amendments: May 19, 2009, as supplemented 
August 28, 2009 (three submittals) and September 11, 2009.
    Brief description of amendments: The amendments revised TS 5.5.9, 
``Steam Generator (SG) Program,'' to exclude portions of the tubes 
within the tubesheet from periodic SG inspections (establish alternate 
repair criteria). The amendments also revised TS 5.6.10, ``Steam 
Generator Tube Inspection Report,'' to remove reference to previous 
interim alternate repair criteria and provide specific reporting 
requirements for Unit 1 during refueling outage (RFO) 15 and the 
subsequent operating cycle, and for Unit 2 during RFO 14 and the 
subsequent operating cycle.
    Date of issuance: September 24, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 157 and 138.
    Renewed Facility Operating License Nos. NPF-68 and NPF-81: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: June 18, 2009 (74 FR 
28962). The supplements dated August 28, 2009, and September 11, 2009, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 24, 2009.
    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: May 4, 2009.
    Brief description of amendment: The amendment revised the Callaway 
Plant Technical Specification (TS) 5.2.2, ``Unit Staff,'' to eliminate 
working hour restrictions in paragraph d of TS 5.2.2 to support 
compliance with Title 10 of the Code of Federal Regulations (10 CFR) 
Part 26. The change is consistent with U.S. Nuclear Regulatory 
Commission (NRC)-approved Revision 0 to TS Task Force (TSTF) Improved 
Technical Specification change traveler, TSTF-511, ``Eliminate Working 
Hour Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 
26.'' The availability of this TS improvement was announced in the 
Federal Register on December 30, 2008 (73 FR 79923), as part of the 
consolidated line item improvement process.
    Date of issuance: September 29, 2009.
    Effective date: As of its date of issuance and shall be implemented 
by October 1, 2009.
    Amendment No.: 193.
    Facility Operating License No. NPF-30: The amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: July 28, 2009 (74 FR 
37250).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 29, 2009.
    Date of initial notice in Federal Register: July 28, 2009 (74 FR 
37250).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 29, 2009.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 8th day of October 2009.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E9-24915 Filed 10-19-09; 8:45 am]
BILLING CODE 7590-01-P


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