[Federal Register Volume 74, Number 95 (Tuesday, May 19, 2009)]
[Notices]
[Pages 23440-23452]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-11268]


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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0204]


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 April 23, 2009, to May 6, 2009. The last 
biweekly notice was published on May 5, 2009 (74 FR 20741).

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 
and Directives Branch, 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. Copies of written comments 
received may be examined at the Commission's Public Document Room 
(PDR), located at One White Flint North, Public File Area O1F21, 11555 
Rockville Pike (first floor), Rockville, Maryland.
    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 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief

[[Page 23441]]

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.
    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 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 
<a href="/cdn-cgi/l/email-protection#b6ded3d7c4dfd8d1d2d9d5ddd3c2f6d8c4d598d1d9c0"><span class="__cf_email__" data-cfemail="e8808d899a81868f8c878b838d9ca8869a8bc68f879e">[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 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 electronic filing 
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, 
Monday through Friday, excluding government holidays. The electronic 
filing Help Desk can be contacted by telephone at 1-866-672-7640 or by 
e-mail at <a href="/cdn-cgi/l/email-protection#773a243f335925120418020514123719051459101801"><span class="__cf_email__" data-cfemail="672a342f234935021408121504022709150449000811">[email protected]</span></a>.
    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:

[[Page 23442]]

Rulemaking and Adjudications Staff. Participants filing a document in 
this manner are responsible for serving the document on all other 
participants. Filing is considered complete by first-class mail as of 
the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the petition 
and/or request should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii).
    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 submission.
    For further details with respect to this amendment action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#3f4f5b4d114d5a4c504a4d5c5a7f514d5c11585049"><span class="__cf_email__" data-cfemail="c8b8acbae6baadbba7bdbaabad88a6baabe6afa7be">[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, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of amendment request: February 19, 2009.
    Description of amendment request: The amendments would relocate the 
reactor coolant system pressure and temperature (P/T) limits and the 
low temperature overpressure protection (LTOP) enable temperatures to a 
licensee-controlled document outside of the Technical Specifications 
(TSs). The P/T limits and LTOP enable temperatures would be specified 
in a Pressure and Temperature Limits Report (PTLR) that would be 
located in the Palo Verde Nuclear Generating Station (PVNGS) Technical 
Requirements Manual and administratively controlled by a new TS 5.6.9.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This proposed change revises the Technical Specifications by 
relocating the reactor coolant system (RCS) pressure and temperature 
limits, heatup and cooldown curves and low temperature overpressure 
protection (LTOP) enable temperatures from the Technical 
Specifications to an [Arizona Public Service] APS-controlled RCS 
Pressure and Temperature Limits Report (PTLR), and requiring that 
the limits in the PTLR be determined using the analytical methods 
described in the NRC-approved Topical Report CE NPSD-683-A. 
Relocation of this information and updating it using NRC-approved 
methodology will not alter the requirement to update the RCS 
pressure and temperature curves and limits in accordance with 10 CFR 
50 Appendices G and H. Updating the P/T curves and LTOP limits 
ensures the reactor coolant system's pressure boundary integrity is 
protected throughout plant life. Consequently, this proposed change 
is determined to not contribute to an increase in the probability 
of, or the initiation of, a design basis accident. Similarly, the 
safety analysis information presented in the Updated Final Safety 
Analysis Report remains unchanged.
    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 revises the Technical Specifications by 
relocating the RCS pressure and temperature limits, heatup and 
cooldown curves and LTOP enable temperatures from the Technical 
Specifications to a PVNGS PTLR, and requiring that the limits in the 
PTLR be determined using the analytical methods described in the 
NRC-approved Topical Report CE NPSD-683-A. The PTLR documents 
removal, testing and analyzing the surveillance capsules, and will 
be updated by APS to reflect the results of testing and analysis of 
surveillance specimens withdrawn in the future. Removal, testing and 
analysis of surveillance specimens may result in a need to implement 
changes to the RCS pressure and temperature limits. Such changes are 
implemented to ensure the integrity of the RCS pressure boundary 
throughout plant lifetime. Updates to the RCS pressure and 
temperature curves and limits will not create a new or different 
kind of accident. Relocating the P/T curves, heatup and cooldown 
rates and LTOP limits to the PTLR has no impact on any safety 
analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Pressure and temperature curves and limits are provided as 
limits to plant operation to ensure RCS pressure boundary integrity 
is maintained throughout the plant's lifetime. Changes to the RCS 
pressure and temperature curves and limits, resulting from the 
removal, testing and analysis of surveillance capsules, are only 
made within the acceptable margin limits thereby maintaining the 
required margin of safety. There is no change to the safety 
analysis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Michael G. Green, Senior Regulatory Counsel, 
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, 
Phoenix, Arizona 85072-2034.
    NRC Branch Chief: Michael T. Markley.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: December 1, 2008.
    Description of amendment request: The proposed amendments would 
correct a non-conservative Technical Specification (TS) Surveillance 
Requirement by revising McGuire TS 3.8.1.4 to increase the minimum 
required amount of fuel oil for the Emergency Diesel Generators fuel 
oil day tank as read on the local fuel gauge used to perform the 
surveillance.

[[Page 23443]]

    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.
    Implementation of the proposed amendment does not significantly 
increase the probability or the consequences of an accident 
previously evaluated. The Emergency Diesel Generators (EDGs) and 
their associated emergency buses function as accident mitigators. 
The proposed changes do not involve a change in the operational 
limits or the design of the electrical power systems (particularly 
the emergency power systems) or change the function or operation of 
plant equipment or affect the response of that equipment when called 
upon to operate. The proposed change to TS SR 3.8.1.4 confirms the 
minimum supply of fuel oil in the emergency diesel generators (EDG) 
fuel oil day tank. The minimum value for the affected parameter is 
being increased in the conservative direction and further ensures 
the EDGs ability to fulfill their safety related function. Thus, 
based on the above, 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 changes do not involve a change in the operational 
limits or the design capabilities of the emergency electrical power 
systems. The proposed changes do not change the function or 
operation of plant equipment or introduce any new failure 
mechanisms. The evaluation that supports this LAR included a review 
of the EDG fuel oil system to which this parameter applies. The 
proposed changes do not introduce any new or different types of 
failure mechanisms; plant equipment will continue to respond as 
designed and analyzed.
    3. Does the proposed amendment involve a significant reduction 
in the 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, the reactor coolant 
system and the containment system will not be adversely impacted by 
the proposed changes. Thus, it is concluded that the proposed TS and 
TS Basis changes do not involve a significant reduction in the 
margin of safety.

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

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

    Date of amendment request: March 5, 2009.
    Description of amendment request: The proposed amendment will 
revise the Reactor Vessel Heatup, Cooldown, and Low Temperature 
Overpressure Protection curves in Technical Specifications (TSs) 3.4.3 
and 3.4.12 to incorporate the most recent estimates of lifetime neutron 
fluence and the effects of the Stretch Power Uprate (Amendment No. 
241).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated.
    The proposed TS changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
There are no physical changes to the plant being introduced by the 
proposed changes to the heatup and cooldown limitation curves. The 
proposed changes do not modify the RCS [Reactor Coolant System] 
pressure boundary. That is, there are no changes in operating 
pressure, materials, or seismic loading. The proposed changes do not 
adversely affect the integrity of the RCS pressure boundary such 
that its function in the control of radiological consequences is 
affected. The proposed heatup and cooldown limitation curves were 
generated in accordance with the fracture toughness requirements of 
10 CFR 50 [Title 10 of the Code of Federal Regulations Part 50] 
Appendix G, and ASME B&PV code [American Society of Mechanical 
Engineers Boiler and Pressure Vessel Code], Section XI, Appendix G 
edition with 2000 Addenda. The proposed heatup and cooldown 
limitation curves were established in compliance with the 
methodology used to calculate and predict effects of radiation on 
embrittlement of RPV [Reactor Pressure Vessel] beltline materials. 
Use of this methodology provides compliance with the intent of 10 
CFR 50 Appendix G and provides margins of safety that ensure non-
ductile failure of the RPV will not occur. The proposed heatup and 
cooldown limitation curves prohibit operation in regions where it is 
possible for non-ductile failure of carbon and low alloy RCS 
materials to occur. Hence, the primary coolant pressure boundary 
integrity will be maintained throughout the limit of applicability 
of the curves, 29.2 EFPY [Effective Full-Power Years].
    Operation within the proposed LTOPS [Low Temperature 
Overpressure Protection System] limits ensures that 
overpressurization of the RCS at low temperatures will not result in 
component stresses in excess of those allowed by the ASME B&PV Code 
Section XI Appendix G.
    Consequently, the proposed changes do not involve a significant 
increase in the probability or the consequences of an accident 
previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. No new modes of operation are introduced by the proposed 
changes. The proposed changes will not create any failure mode not 
bounded by previously evaluated accidents. Further, the proposed 
changes to the heatup and cooldown limitation curves and the LTOPS 
limits do not affect any activities or equipment other than the RCS 
pressure boundary and do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Consequently, the proposed changes do not involve a significant 
increase in the probability or consequence of a new or different 
kind of accident, from any accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety.
    The proposed TS changes do not involve a significant reduction 
in the margin of safety.
    The revised heatup and cooldown limitation curves and LTOPS 
limits are established in accordance with current regulations and 
the ASME B&PV Code 1998 edition with 2000 Addenda. These proposed 
changes are acceptable because the ASME B&PV Code maintains the 
margin of safety required by 10 CFR 50.55(a). Because operation will 
be within these limits, the RCS materials will continue to behave in 
a non-brittle manner consistent with the original design bases.
    The proposed changes to the allowable operation of charging and 
safety injection pumps when LTOPS is required to be operable is 
consistent with the IP2 licensing bases as established in TS 
Amendment 224.
    Therefore, Entergy has concluded that 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

[[Page 23444]]

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 Acting Branch Chief: Richard V. Guzman.

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

    Date of amendment request: March 25, 2009.
    Description of amendment request: The proposed amendment would add 
two Emergency Core Cooling System (ECCS) valves to Surveillance 
Requirement (SR) 3.5.2.1. The SR is designed to verify that ECCS valves 
whose single failure could cause loss of the ECCS function are in the 
required position with ac power removed so that misalignment or single 
failure cannot prevent completion of the ECCS function. Entergy plans 
to install an alternate source of power during the spring 2010 
refueling outage to provide the required position indication. The 
proposed changes support Entergy's resolution to Generic Letter (GL) 
2004-02 by establishing a licensing basis that supports meeting the 
regulatory requirements of the GL.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response--No.
    The proposed change adds two ECCS valves to SR 3.5.2.1. The 
purpose of the surveillance is to assure that the valves are in 
their required position with normal ac power removed from the valve 
operator so that misalignment or single failure cannot prevent 
completion of the ECCS function. The performance of the SR does not 
involve any actions related to the initiation of an accident and 
therefore the proposed changes cannot increase the probability of an 
accident. Misalignment or single failure of one of the two valves 
being added to TS [Technical Specifications] could cause a loss of 
the ECCS function based on GSI [Generic Safety Issue]-191 
evaluations, so the change will not increase the consequences of an 
accident but rather provide assurance that no such increase can 
occur. Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response--No.
    The proposed change adds two ECCS valves to SR 3.5.2.1. The 
purpose of the surveillance is to assure that the valves are in 
their required position with normal ac power removed from the valve 
operators so that misalignment or single failure cannot prevent 
completion of the ECCS function. The provision of alternate power to 
the existing valve position indication during the upcoming spring 
2010 outage (2R19), will allow the valve operators to be normally 
deenergized. The change assures that the valves will be in their 
correct position and does not introduce any new failure modes or the 
possibility of a different accident. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response--No.
    The proposed change adds two ECCS valves to SR 3.5.2.1. The 
purpose of the surveillance is to assure that the valves are in 
their required position with normal ac power removed so that 
misalignment or single failure cannot prevent completion of the ECCS 
function. The valves will be re-energized 24 hours following a DBA 
[design-basis accident] and therefore will be capable of performing 
their required function of isolating a potential passive failure at 
that time. This ensures that the ECCS function can be performed 
without a 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: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Acting Branch Chief: Richard V. Guzman.

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

    Date of amendment request: March 29, 2009.
    Description of amendment request: The proposed amendment will 
establish a more restrictive acceptance criterion for surveillance 
requirement (SR) 3.8.6.6 regarding periodic verification of capacity 
for the affected station batteries.
    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?
    No. The proposed change revises the acceptance criterion applied 
to an existing surveillance test for the Indian Point 2 station 
batteries. Performing a technical specification surveillance test is 
not an accident initiator and does not increase the probability of 
an accident occurring. The proposed revision to the test acceptance 
criterion is based on the design calculation for battery performance 
at the minimum design temperature. The proposed new value for the 
test acceptance criteria is more limiting than the existing value 
which does not account for the minimum environmental design 
temperature assumed for the limiting battery locations. Establishing 
a test acceptance criterion that bounds existing or assumed 
conditions validates the equipment performance assumptions used in 
the accident mitigation safety analyses. Therefore the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed change revises the test acceptance criterion 
for an existing technical specification surveillance test conducted 
on the existing station batteries. The proposed change does not 
involve installation of new equipment or modification of existing 
equipment, so that no new equipment failure modes are introduced. 
Also, the proposed change in test acceptance criterion does not 
result in a change to the way that the equipment or facility is 
operated so that no new accident initiators are created. Therefore 
the proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The conduct of performance tests on safety-related plant 
equipment is a means of assuring that the equipment is capable of 
performing its intended safety function and therefore maintaining 
the margin of safety established in the safety analysis for the 
facility. The proposed change in the acceptance criterion for the 
battery capacity surveillance test is more conservative and more 
restrictive than the value currently in the technical specification 
and is based on the applicable design calculation for these 
components.

    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

[[Page 23445]]

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 Acting Branch Chief: Richard V. Guzman.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois; Docket Nos. 
STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle 
County, Illinois

    Date of amendment request: March 26, 2009.
    Description of amendment request: The proposed amendments would 
revise the fire protection program (FPP) to eliminate the requirement 
for the backup manual carbon dioxide (CO<INF>2</INF>) fire suppression 
system in the upper cable spreading rooms.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the FPP to eliminate the requirement 
to maintain the backup CO<INF>2</INF> fire suppression system for 
the upper cable spreading rooms. With the exception of the 
CO<INF>2</INF> fire suppression system itself, the proposed change 
does not result in any physical changes to safety related 
structures, systems, or components [SSCs], or the manner in which 
they are operated, maintained, modified, tested, or inspected. The 
proposed change does not degrade the performance or increase the 
challenges of any safety related SSCs assumed to function in the 
accident analysis. The proposed change does not change the 
probability of a fire occurring since the fire ignition frequency is 
independent of the method of fire suppression. The proposed change 
does not affect the consequences of an accident previously evaluated 
since the fire safe shutdown analysis assumes fire damage throughout 
the affected fire area. The results of a fire in the upper cable 
spreading room would only affect one engineered safety features 
division. Sufficient redundancy exists in the engineered safety 
features fed from the other division to achieve a reactor shutdown 
and to maintain the reactor in a safe shutdown condition.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the FPP to eliminate the requirement 
to maintain the backup CO<INF>2</INF> fire suppression system for 
the upper cable spreading rooms. With the exception of the 
CO<INF>2</INF> fire suppression system itself, the proposed change 
does not result in any physical changes to safety related 
structures, systems, or components, or the manner in which they are 
operated, maintained, modified, tested, or inspected. The proposed 
change does not degrade the performance or increase the challenges 
of any safety related SSCs assumed to function in the accident 
analysis. As a result, the proposed change does not introduce nor 
increase the number of failure mechanisms of a new or different type 
than those previously evaluated. The fire safe shutdown analysis 
assumes fire damage throughout the area consistent with a complete 
lack of fire suppression capability. Potential habitability hazards 
associated with actuation of the CO<INF>2</INF> system are 
eliminated with the proposed change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the FPP to eliminate the requirement 
to maintain the backup CO<INF>2</INF> fire suppression system for 
the upper cable spreading rooms. With the exception of the 
CO<INF>2</INF> fire suppression system itself, the proposed change 
does not result in any physical changes to safety related 
structures, systems, or components, or the manner in which they are 
operated, maintained, modified, tested, or inspected. The proposed 
change does not degrade the performance or increase the challenges 
of any safety related SSCs assumed to function in the accident 
analysis. Since the backup manual CO<INF>2</INF> fire suppression 
system is not credited in the safe shutdown analysis to protect the 
upper cable spreading rooms, the proposed change does not impact 
plant safety since the conclusions of the fire safe shutdown 
analysis remain unchanged.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The Nuclear Regulatory Commission (NRC) staff has reviewed the 
licensee's analysis and, based on this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

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

    Date of amendment request: April 1, 2009.
    Brief description of amendments: The proposed amendment would 
delete Technical Specification (TS) 5.2.2.d, in TS 5.2.2, ``Unit 
Staff,'' regarding the requirement to develop and implement 
administrative procedures to limit the working hours of personnel who 
perform safety-related functions. The requirements of TS 5.2.2.d have 
been superseded by Title 10 of the Code of Federal Regulations (10 CFR) 
Part 26, Subpart I. The change is consistent with U.S. Nuclear 
Regulatory Commission (NRC)-approved Revision 0 to Technical 
Specification 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 NRC staff issued a ``Notice of Availability of Model Safety 
Evaluation, Model No Significant Hazards Determination, and Model 
Application for Licensees That Wish to Adopt TSTF-511, Revision 0, 
`Eliminate Working Hour Restrictions from TS 5.2.2 to Support 
Compliance with 10 CFR Part 26,' '' in the Federal Register on December 
30, 2008 (73 FR 79923). The notice included a model safety evaluation, 
a model no significant hazards consideration (NSHC) determination, and 
a model license amendment request. In its application dated April 1, 
2009, the licensee affirmed the applicability of the model NSHC 
determination, which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee, is presented below:

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

    The proposed change removes TS restrictions on working hours for 
personnel who perform safety related functions. The TS restrictions 
are superseded by the worker fatigue requirements in 10 CFR 26. 
Removal of the TS requirements will be performed concurrently with 
the implementation of the 10 CFR 26, Subpart I, requirements. The 
proposed change does not impact the physical configuration or 
function of plant structures, systems, or components (SSCs) or the 
manner in which SSCs are operated, maintained, modified, tested, or 
inspected. Worker fatigue is not an initiator of any accident 
previously evaluated. Worker fatigue is not an assumption in the

[[Page 23446]]

consequence mitigation of any accident previously evaluated.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

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

    The proposed change removes TS restrictions on working hours for 
personnel who perform safety related functions. The TS restrictions 
are superseded by the worker fatigue requirements in 10 CFR 26. 
Working hours will continue to be controlled in accordance with NRC 
requirements. The new rule allows for deviations from controls to 
mitigate or prevent a condition adverse to safety or as necessary to 
maintain the security of the facility. This ensures that the new 
rule will not unnecessarily restrict working hours and thereby 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    The proposed change does not alter the plant configuration, 
require new plant equipment to be installed, alter accident analysis 
assumptions, add any initiators, or effect the function of plant 
systems or the manner in which systems are operated, maintained, 
modified, tested, or inspected.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

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

    The proposed change removes TS restrictions on working hours for 
personnel who perform safety related functions. The TS restrictions 
are superseded by the worker fatigue requirements in 10 CFR 26. The 
proposed change does not involve any physical changes to plant or 
alter the manner in which plant systems are operated, maintained, 
modified, tested, or inspected. The proposed change does not alter 
the manner in which safety limits, limiting safety system settings 
or limiting conditions for operation are determined. The safety 
analysis acceptance criteria are not affected by this change. The 
proposed change will not result in plant operation in a 
configuration outside the design basis. The proposed change does not 
adversely affect systems that respond to safely shutdown the plant 
and to maintain the plant in a safe shutdown condition.
    Removal of plant-specific TS administrative requirements will 
not reduce a margin of safety because the requirements in 10 CFR 26 
are adequate to ensure that worker fatigue is managed.
    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: Timothy P. Matthews, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: April 2, 2009.
    Brief description of amendments: The amendment revises Technical 
Specification (TS) 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' to add Surveillance Requirement (SR) 3.3.1.16 to 
Function 3 of TS Table 3.3.1-1. SR 3.3.1.16 requires that RTS RESPONSE 
TIMES be verified to be within limits every 18 months on a STAGGERED 
TEST BASIS. Function 3 is the power range neutron flux--high positive 
rate reactor trip function (hereafter referred to as the positive flux 
rate trip (PFRT) function). This change is based on a reanalysis of the 
Rod Cluster Control Assembly Bank Withdrawal at Power event.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change imposes additional surveillance requirements 
to assure safety related structures, systems, and components are 
verified to be consistent with the safety analysis and licensing 
basis. In this specific case, a response time verification 
requirement will be added to the positive flux rate trip (PFRT) 
function.
    Overall protection system performance will remain within the 
bounds of the accident analysis since there are no hardware changes. 
The design of the Reactor Trip System (RTS) instrumentation, 
specifically the positive flux rate trip (PFRT) function, will be 
unaffected. The reactor protection system will continue to function 
in a manner consistent with the plant design basis. All design, 
material, and construction standards that were applicable prior to 
the request are maintained.
    The proposed changes will not modify any system interface. The 
proposed changes will not affect the probability of any event 
initiators. There will be no degradation in the performance of or an 
increase in the number of challenges imposed on safety-related 
equipment assumed to function during an accident situation. There 
will be no change to normal plant operating parameters or accident 
mitigation performance. The proposed changes will not alter any 
assumptions or change any mitigation actions in the radiological 
consequences evaluations in the updated Final Safety Analysis Report 
(FSAR).
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) to 
perform their intended function to mitigate the consequences of an 
initiating event within the assumed acceptance limits. The proposed 
changes do not affect the source term, containment isolation, or 
radiological release assumptions used in evaluating the radiological 
consequences of an accident previously evaluated. The proposed 
changes are consistent with safety analysis assumptions and 
resultant consequences.
    Therefore, the proposed changes 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.
    The proposed change imposes additional surveillance requirements 
to assure safety related structures, systems, and components are 
verified to be consistent with the safety analysis and licensing 
basis.
    There are no hardware changes nor are there any changes in the 
method by which any safety related plant system performs its safety 
function. This change will not affect the normal method of plant 
operation or change any operating parameters. No performance 
requirements will be affected; however, the proposed change does 
impose additional surveillance requirements. The additional 
requirements are consistent with assumptions made in the safety 
analysis and licensing basis.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of these changes.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change imposes additional surveillance requirements 
to assure safety related structures, systems, and components are 
verified to be consistent with the safety analysis and licensing 
basis.
    The proposed changes do not affect the acceptance criteria for 
any analyzed event. The margin of safety is affected in that in the 
new analyses of the Rod (Bank) Withdrawal at Power analyses, it is 
necessary to credit a previously uncredited reactor trip function in 
an analysis. However, that reactor trip function is described in the 
plant Technical

[[Page 23447]]

Specifications with well-defined operability requirements. An 
additional attribute, specifically the channel response time 
verification on, a periodic frequency, provides additional assurance 
that the trip function performs as credited in the accident 
analysis. With the credit for this reactor trip function, all 
relevant event acceptance criteria continue to be met. None of the 
event acceptance limits are exceeded, and none of the event 
acceptance limits are revised by the proposed activity. There is no 
effect on the manner in which safety limits, limiting safety system 
settings, or limiting conditions for operation are determined nor is 
there any effect on those plant systems necessary to assure the 
accomplishment of protection functions. There is no impact on the 
overpower limit, the minimum departure from nucleate boiling ratio 
limit, the radial and axial peaking factor limits, the loss of 
coolant accident (LOCA) peak clad temperature limit, nor any other 
limit which, in whole or in part, defines a margin of safety. The 
radiological dose consequence acceptance criteria listed in the 
Standard Review Plan will continue to be met.
    Therefore the proposed change does not involve a 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: Timothy P. Matthews, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: Michael T. Markley.

Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-220, Nine 
Mile Point Nuclear Station Unit No. 1 (NMP 1), Oswego County, New York

    Date of amendment request: March 3, 2009.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) Section 3.2.1, ``Reactor Vessel 
Heatup and Cooldown Rates,'' and Section 3.2.2, ``Minimum Reactor 
Vessel Temperature for Pressurization,'' by replacing the existing 
reactor vessel heatup and cooldown rate limits and the pressure and 
temperature limit curves with references to the Pressure and 
Temperature Limits Report (PTLR). In addition, a new definition for the 
PTLR would be added to TS Section 1.0, ``Definitions,'' and a new 
section addressing administrative requirements for the PTLR would be 
added to TS Section 6.0, ``Administrative Controls.'' The proposed 
changes are consistent with the guidance in Generic Letter 96-03, 
``Relocation of the Pressure Temperature Limit Curves and Low 
Temperature Overpressure Protection System Limits,'' as supplemented by 
TS Task Force (TSTF) traveler TSTF-419-A, ``Revise PTLR Definition and 
References in ISTS 5.6.6, RCS [Reactor Coolant System] PTLR.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes modify the TS by replacing references to 
existing reactor vessel heatup and cooldown rate limits and P-T 
[pressure-temperature] limit curves with references to the PTLR. The 
proposed amendment also adopts the NRC-approved methodology of SIR-
05-044-A for the preparation of NMP1 P-T limit curves. In 10 CFR 50 
Appendix G, requirements are established to protect the integrity of 
the reactor coolant pressure boundary (RCPB) in nuclear power 
plants. Implementing the NRC-approved methodology for calculating P-
T limit curves and relocating those curves to the PTLR provide an 
equivalent level of assurance that RCPB integrity will be 
maintained, as specified in 10 CFR 50 Appendix G.
    The proposed changes do not adversely affect accident initiators 
or precursors, and do not alter the design assumptions, conditions, 
or configuration of the plant or the manner in which the plant is 
operated and maintained. The ability of structures, systems, and 
components to perform their intended safety function is not altered 
or prevented by the proposed changes, and the assumptions used in 
determining the radiological consequences of previously evaluated 
accidents are not affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The change in methodology for calculating P-T limits and the 
relocation of those limits to the PTLR do not alter or involve any 
design basis accident initiators. RCPB integrity will continue to be 
maintained in accordance with 10 CFR 50 Appendix G, and the assumed 
accident performance of plant structures, systems and components 
will not be affected. These changes do not involve any physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed), and installed equipment is not being operated in 
a new or different manner. Thus, no new failure modes are 
introduced.
    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?
    Response: No.
    The proposed changes do not affect the function of the RCPB or 
its response during plant transients. By calculating the P-T limits 
using NRC-approved methodology, adequate margins of safety relating 
to RCPB integrity are maintained. The proposed changes do not alter 
the manner in which safety limits, limiting safety system settings, 
or limiting conditions for operation are determined, there are no 
changes to the setpoints at which actions are initiated, and the 
operability requirements for equipment assumed to operate for 
accident mitigation are not affected.
    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 Acting Branch Chief: John P. Boska.

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: March 9, 2009.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) testing frequency for the 
surveillance requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.'' 
Specifically, the proposed change is based on TS Task Force (TSTF) 
change traveler TSTF-460-A, Revision 0, and extends the frequency for 
testing control rod scram time testing in SR 3.1.4.2 from every 120 
days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 
operation. A notice of availability of this proposed TS change using 
the consolidated line item improvement process was published in the 
Federal Register on August 23, 2004 (69 FR 51864). The licensee 
affirmed the applicability of the model no significant hazards 
consideration determination in its application.
    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

[[Page 23448]]

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 extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The frequency 
of surveillance testing is not an initiator of any accident 
previously evaluated. The frequency of surveillance testing does not 
affect the ability to mitigate any accident previously evaluated, as 
the tested component is still required to be operable.
    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 extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change does not result in any new or different modes of plant 
operation.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    4. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change continues to test the control rod scram time to ensure the 
assumptions in the safety analysis are protected.

    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 Acting Branch Chief: John P. Boska.

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

    Date of amendment request: March 5, 2009, as supplemented by letter 
dated April 13, 2009.
    Description of amendment request: The proposed amendments would 
make changes to the PINGP Technical Specifications (TSs) to revise TS 
3.8.1, ``AC Sources--Operating,'' Surveillance Requirement (SR) 3.8.1.8 
Frequency to allow use of the SR 3.0.2 interval extension (1.25 times 
the specified 24 month Frequency). This would be an exception to the SR 
3.0.2 limitations in the PINGP TS, which do not allow use of the 
interval extension for SRs with a 24 month Frequency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment request proposes to add a Frequency Note 
to Surveillance Requirement 3.8.1.8 which will allow application of 
the Surveillance Requirement 3.0.2 interval extension (1.25 times 
the specified 24 month Frequency) for performance of this 
surveillance. This would be an exception to the limitations 
specified in the Prairie Island Nuclear Generating Plant Technical 
Specification Surveillance Requirement 3.0.2 for Surveillance 
Requirements with a 24 month Frequency and would allow an interval 
up to 30 months for performance of the surveillance.
    The emergency diesel generators are not accident initiators and 
therefore, these changes do not involve a significant increase [in] 
the probability of an accident.
    Failure of the bypass relay, by itself, does not prevent an 
emergency diesel generator from performing its safety related 
functions. Since the accident analyses only require one of the two 
trains of onsite emergency AC to be operable, the changes proposed 
in the license amendment request do not involve a significant 
increase in the consequences of an accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This license amendment request proposes to add a Frequency Note 
to Surveillance Requirement 3.8.1.8 which will allow application of 
the Surveillance Requirement 3.0.2 interval extension (1.25 times 
the specified 24 month Frequency) for performance of this 
surveillance. This would be an exception to the limitations 
specified in the Prairie Island Nuclear Generating Plant Technical 
Specification Surveillance Requirement 3.0.2 for Surveillance 
Requirements with a 24 month Frequency and would allow an interval 
up to 30 months for performance of the surveillance.
    The changes proposed for the emergency diesel generators do not 
change any system operations or maintenance activities. Testing 
requirements will be revised and will continue to demonstrate that 
the Limiting Conditions for Operation are met and the system 
components are functional. The revised test Frequency does not 
create new failure modes or mechanisms and no new accident 
precursors are generated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This license amendment request proposes to add a Frequency Note 
to Surveillance Requirement 3.8.1.8 which will allow application of 
the Surveillance Requirement 3.0.2 interval extension (1.25 times 
the specified 24 month Frequency) for performance of this 
surveillance. This would be an exception to the limitations 
specified in the Prairie Island Nuclear Generating Plant Technical 
Specification Surveillance Requirement 3.0.2 for Surveillance 
Requirements with a 24 month Frequency and would allow an interval 
up to 30 months for performance of the surveillance.
    The proposed change will continue to ensure that the DG trips 
bypass function operates as designed. The functionality and 
operability of the emergency power system is not being changed. 
Since the requested change only allows extension of the relay 
testing interval and failure of the relay by itself does not prevent 
the diesel from performing its safety function, this change does not 
involve a significant reduction in a margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Lois M. James.

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: March 30, 2009.
    Description of amendment request: The proposed amendment revises 
the Technical Specifications (TS), Appendix A to Facility Operating 
License Nos. NPF-2 and NPF-8 for the Joseph M. Farley Nuclear Plant, 
Units 1

[[Page 23449]]

and 2, respectively. The changes would eliminate the Reactor Coolant 
Pump (RCP) Breaker Position reactor trip. The changes will allow the 
elimination of a trip circuitry that is susceptible to single failure 
vulnerabilities which can result in unwarranted reactor trips.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated in 
the Final Safety Analysis Report (FSAR). All of the safety analyses 
have been evaluated for impact. The elimination of Reactor Coolant 
Pump Breaker Position reactor trip will not initiate any accident; 
therefore, the probability of an accident has not been increased. An 
evaluation of dose consequences, with respect to the proposed 
changes, indicates there is no impact due to the proposed changes 
and all acceptance criteria continue to be met. Therefore, these 
changes 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 previously evaluated?
    Response: No.
    The proposed changes do not create the possibility of a new or 
different kind of accident than any accident already evaluated in 
the FSAR. No new accident scenarios, failure mechanisms or limiting 
single failures are introduced as a result of the proposed changes. 
The changes have no adverse effects on any safety-related system. 
Therefore, all accident analyses criteria continue to be met and 
these changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not involve a significant reduction in a 
margin of safety. All analyses that credit the Reactor Coolant 
System Low Flow reactor trip function have been reviewed and no 
changes to any inputs are required. The evaluation demonstrated that 
all applicable acceptance criteria are met. Therefore, the proposed 
changes do not involve a significant reduction in the margin of 
safety.
    Based on the preceding evaluation, SNC has determined that the 
proposed changes meet the requirements of 10 CFR 50.92(c) and do not 
involve a significant hazards consideration.

    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: Melanie C. Wong.

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

    Date of amendment request: April 30, 2009.
    Description of amendment request: The proposed amendment would 
revise technical specification (TS) Section 5.7, ``Procedures, 
Programs, and Manuals,'' to correct typographical errors introduced in 
Amendment No. 70, dated October 8, 2008.
    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. This change is limited only to correcting a 
typographical error in a section number (5.7.2.20 versus 5.2.7.20) 
contained in Technical Specification Section 5.0, which will not 
change the intent of the added section previously approved in 
License Amendment 70. Therefore, no increase in the probability or 
consequences of an accident previously evaluated has been created.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. This change is limited only to correcting a 
typographical error in a section number (5.7.2.20 versus 5.2.7.20) 
contained in Technical Specification Section 5.0, which will not 
change the intent of the added section previously approved in 
License Amendment 70. Therefore, the possibility of a new or 
different kind of accident from those previously analyzed has not 
been created.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. This change is limited only to correcting a 
typographical error in a section number (5.7.2.20 versus 5.2.7.20) 
contained in Technical Specification Section 5.0, which will not 
change the intent of the added section previously approved in 
License Amendment 70. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.
    Based on the above, TVA concludes that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no 
significant hazards consideration'' is justified.

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

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

    Date of amendment request: March 26, 2009.
    Description of amendment request: The proposed amendments would 
increase each unit's rated thermal power (RTP) level from 2893 
megawatts thermal (MWt) to 2940 MWt, and make technical specification 
changes as necessary to support operation at the uprated power level. 
The proposed change is an increase in RTP of approximately 1.6 percent.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change will increase the North Anna Power Station 
(NAPS) Units 1 and 2 rated thermal power (RTP) from 2893 megawatts 
thermal (MWt) to 2940 MWt. Nuclear steam supply systems and balance-
of-plant systems, components and analyses that could be affected by 
the proposed change to the RTP were evaluated using revised design 
parameters. The evaluations determined that these structures, 
systems and components are capable of performing their design 
function at the proposed uprated RTP of 2940 MWt. An evaluation of 
the accident analyses demonstrates that the applicable analysis 
acceptance criteria are still met with the proposed changes. Power 
level is an input assumption to equipment design and accident 
analyses, but it is not a transient or accident initiator. Accident 
initiators are not affected by the power uprate, and plant safety 
barrier challenges are not created by the proposed changes.
    The radiological consequences of operation at the uprated power 
conditions have been assessed. The proposed change to RTP does not 
affect release paths, frequency of release, or the analyzed source 
term for any accidents

[[Page 23450]]

previously evaluated in the NAPS Updated Final Safety Analysis 
Report. Structures, systems and components required to mitigate 
transients are capable of performing their design functions with the 
proposed changes, and are thus acceptable. Analyses performed to 
assess the effects of mass and energy releases remain valid. The 
source term used to assess radiological consequences was reviewed 
and determined to bound operation at the proposed power level.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of any proposed changes. The UFM 
has been analyzed, and system failures will not adversely affect any 
safety-related system or any structures, systems or components 
required for transient mitigation. Structures, systems and 
components previously required for transient mitigation are still 
capable of fulfilling their intended design functions. The proposed 
changes have no significant adverse affect on any safety-related 
structures, systems or components and do not significantly change 
the performance or integrity of any safety-related system.
    The proposed changes do 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 previously 
evaluated. Operating at RTP of 2940 MWt does not create any new 
accident initiators or precursors. Credible malfunctions are bounded 
by the current accident analyses of record or recent evaluations 
demonstrating that applicable criteria are still met with the 
proposed changes.
    Therefore, this 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 margins of safety associated with the power uprate are those 
pertaining to core thermal power. These include fuel cladding, 
reactor coolant system pressure boundary, and containment barriers. 
Core analyses demonstrate that power uprate implementation will 
continue to meet the current nuclear design basis. Impacts to 
components associated with the reactor coolant system pressure 
boundary structural integrity, and factors such as pressure-
temperature limits, vessel fluence, and pressurized thermal shock 
were determined to be bounded by the current analyses.
    Systems will continue to operate within their design parameters 
and remain capable of performing their intended safety functions 
following implementation of the proposed change. The current NAPS 
safety analyses, including the design basis radiological accident 
dose calculations, bound the power uprate.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, 
VA 23219.
    NRC Branch Chief: Melanie C. Wong.

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, 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#d0a0b4a2fea2b5a3bfa5a2b3b590bea2b3feb7bfa6"><span class="__cf_email__" data-cfemail="8afaeef8a4f8eff9e5fff8e9efcae4f8e9a4ede5fc">[email protected]</span></a>.

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

    Date of application for amendment: September 11, 2008, as 
supplemented by letter dated December 17, 2008, and January 20, 2009.
    Brief Description of amendment: The amendment revised the Technical 
Specifications, extending the 15-year interval between containment Type 
A tests specified by Specification 4.4.a, ``Integrated Leak Rate 
Test,'' by 6 months. The current Type A test interval expires at the 
end of April 2009. The amendment extends this interval, on a one-time 
basis, to October 2009 to coincide with completion of the next 
scheduled refueling outage.
    Date of issuance: April 27, 2009.
    Effective date: As of the date of issuance and should be 
implemented within 60 days.
    Amendment No.: 204.
    Facility Operating License No. DPR-43: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 4, 2008 (73 FR 
65689). The commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 27, 2009.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 7, 2008, as supplemented on 
September 19, 2008, and March 17, 2009.
    Brief description of amendment: The amendment revised the licensing 
basis, authorizing the licensee to use the methodology conveyed in the 
licensee's letters cited above to determine the seismic loads on the 
recently upgraded Auxiliary Building crane. The authorization is 
conveyed by addition of a new License Condition 2.C.(11) to Facility 
Operating License DPR-43.
    Date of issuance: April 30, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 205.

[[Page 23451]]

    Facility Operating License No. DPR-43: The amendment revised 
Facility Operating License No. DPR-43.
    Date of initial notice in Federal Register: August 26, 2008 (73 FR 
50358). The Commission's related evaluation of the amendment is 
contained in a safety evaluation dated April 30, 2009.
    No Significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: July 16, 2008, as supplemented 
by letters dated January 2 and March 19, 2009.
    Brief description of amendment: The amendment revised Technical 
Specifications 3.1.4, ``Control Rod Scram Times,'' 3.2.2, ``Minimum 
Critical Power Ratio (MCPR),'' and 5.6.3, ``Core Operating Limits 
Report (COLR),'' to allow incorporation of the analytical methodologies 
associated with operation of Global Nuclear Fuel-Americas (GNF) fuel 
into the licensing basis to support transition to GNF GE14 fuel.
    Date of issuance: May 5, 2009.
    Effective date: As of its date of issuance and shall be implemented 
prior to beginning operating cycle 20.
    Amendment No.: 211.
    Facility Operating License No. NPF-21: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 14, 2008 (73 FR 
60729).
    The supplements dated January 2 and March 19, 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 May 5, 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 18, 2008, as supplemented by 
letter dated February 4, 2009.
    Brief description of amendment: The amendment modified Technical 
Specification (TS) requirements for inoperable snubbers by relocating 
the current TS 3.7.8, ``Snubbers,'' to the Technical Requirements 
Manual and adding Limiting Condition for Operation (LCO) 3.0.8. The 
amendment also made conforming changes to TS LCO 3.0.1. The proposed 
amendment is consistent with U.S. Nuclear Regulatory Commission (NRC)-
approved Technical Specification Task Force (TSTF) Improved Standard 
Technical Specifications Change Traveler, TSTF-372, Revision 4, 
``Addition of LCO 3.0.8, Inoperability of Snubbers,'' as part of the 
consolidated line item improvement process.
    Date of issuance: May 1, 2009.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 219.
    Facility Operating License No. NPF-38: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 16, 2008 (73 
FR 76410). The supplemental letter dated February 4, 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 May 1, 2009.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of application for amendment: November 18, 2008.
    Brief description of amendment: This amendment modifies Technical 
Specification 5.5.6 to incorporate Technical Specification Task Force 
(TSTF) Travelers TSTF-479, ``Changes to Reflect Revision of 10 CFR 
[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.''
    Date of issuance: May 1, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 151.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: January 27, 2009 (74 FR 
4772). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 1, 2009.
    No significant hazards consideration comments received: No.

GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear 
Station, Unit 2, Dauphin County, Pennsylvania

    Date of amendment request: June 11, 2008, as supplemented by 
letters dated September 15, 2008, December 10, 2008, and March 16, 
2009.
    Brief description of amendment: The amendment deletes Technical 
Specification 6.5, which provided the requirements related to review 
and audit functions.
    Date of issuance: May 1, 2009.
    Effective date: May 1, 2009.
    Amendment No.: 63.
    Possession Only License No. DPR-73: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 26, 2008 (73 FR 
50356) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation Report, dated May 1, 2009.
    No significant hazards consideration comments received: No.

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: June 27, 2007, as supplemented 
on April 28, September 4, and December 17, 2008.
    Brief description of amendment: The amendment revises surveillance 
requirements in Technical Specifications (TS) Section 3.8.1, ``AC 
Sources--Operating,'' associated with the diesel generator (DG) steady-
state frequency and voltage. The amendment corrects non-conservative TS 
frequency and voltage values, which the licensee states have the 
potential to result in undesirable effects such as centrifugal charging 
pump motor brake horsepower exceeding its nameplate maximum horsepower, 
and subsequently overloading the DGs.
    Date of issuance: April 30, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days from the date of issuance April 30, 2009.
    Amendment Nos.: 309 (CNP-1), 291 (CNP-2).
    Facility Operating License Nos. DPR-58 and DPR-74: Amendment 
revises the Renewed Operating License and Technical Specifications.
    Date of initial notice in Federal Register: November 4, 2008 (73 FR

[[Page 23452]]

65696). The April 28 and December 17, 2008 supplements provided 
additional information that clarified the application, but did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed significant hazards consideration 
published in the Federal Register on August 14, 2007.
    The September 4, 2008 supplement provided additional information 
which expanded the scope of the application as originally noticed. The 
NRC staff identified that the specified DG voltage of 3,740 volts at 10 
seconds after the DG start was non-conservative and inconsistent with 
the 3,910 volt minimum steady-state voltage provided in other parts of 
TS Section 3.8.1. The licensee proposed additional changes to TS 
Section 3.8.1 in its September 4, 2008 letter. The NRC staff determined 
that the proposed expanded scope of the amendment involved a proposed 
no significant hazards consideration as published in the Federal 
Register on November 4, 2008 (73 FR 65696).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 30, 2009.
    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: February 24, 2009.
    Brief description of amendments: The amendments deleted the 
requirement for the power range neutron flux rate-high negative rate 
trip (Function 3.b) in Technical Specification (TS) Table 3.3.1-1, 
``Reactor Trip System Instrumentation.'' The changes are consistent 
with the NRC-approved methodology presented in Westinghouse Topical 
Report, WCAP-11394-P-A, ``Methodology for the Analysis of the Dropped 
Rod Event,'' dated January 1990. The amendments also incorporated 
editorial changes to reflect the deletion of Function 3.b in TS Table 
3.3.1-1.
    Date of issuance: April 29, 2009.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: Unit 1--205; Unit 2--206.
    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: March 24, 2009 (74 FR 
12394).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 29, 2009.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 31, 2008 superseded the 
application dated August 1, 2008, as supplemented by letters dated 
November 25 and December 31, 2008.
    Brief description of amendment: The amendment revised WBN Unit 1 
Technical Specification (TS) 4.2.1, ``Fuel Assemblies,'' and TS 
surveillance requirements (SRs) 3.5.1.4, ``Accumulators,'' and 3.5.4.3, 
``RWST [Refueling Water Storage Tank],'' to increase the maximum number 
of tritium producing burnable absorber rods from 400 to 704.
    Date of issuance: April 30, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment No.: 77.
    Facility Operating License No. NPF-90: Amendment revises the TS 
4.2.1 and TS SRs 3.5.1.4 and 3.5.4.3.
    Date of initial notice in Federal Register: Originally November 12, 
2008 (73 FR 66946) was superseded by a notice on January 27, 2009 (74 
FR 4776).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 30, 2009.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 7th day of May 2009.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E9-11268 Filed 5-18-09; 8:45 am]
BILLING CODE 7590-01-P


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