[Federal Register Volume 74, Number 65 (Tuesday, April 7, 2009)]
[Notices]
[Pages 15765-15778]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-7494]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[NRC-2009-0148]


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 March 12, 2009 to March 25, 2009. The last 
biweekly notice was published on March 24, 2009 (74 FR 12390).

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 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

[[Page 15766]]

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 
[email protected], or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer \TM\ to 
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer \TM\ is free and is available 
at 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 help 
electronic filing Help Desk can be contacted by telephone at 1-866-672-
7640 or by e-mail at [email protected].
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, in accordance 
with 10 CFR 2.302(g), with their initial paper filing requesting 
authorization to continue to submit documents in paper format. Such 
filings must be submitted by: (1) First class mail addressed to the 
Office of the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff; or (2) courier, express mail, or expedited 
delivery service to the Office of the Secretary, Sixteenth Floor, One 
White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, 
Attention: Rulemaking and Adjudications Staff. Participants filing a 
document in this manner are responsible for serving the document on all 
other participants. Filing is considered complete by first-class mail 
as of the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained

[[Page 15767]]

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 [email protected].

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

    Date of amendment request: September 2, 2008.
    Description of amendment request: The amendments would revise the 
technical specifications to allow manual operation of the containment 
spray system and to revise the upper and lower limits of the refueling 
water storage tank.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Containment Spray System and RWST [refueling water storage 
tank] are accident mitigation equipment. As such, changes in operation 
of these systems cannot have an impact on the probability of an 
accident.
    The RWST will continue to comply with all applicable regulatory 
requirements and design criteria following approval of the proposed 
changes (e.g., train separation, redundancy, and single failure). The 
water level on the containment floor will be higher at the start of 
transfer to the containment sump but will remain below the maximum 
design level analyzed for equipment submergence. The change in the sump 
pH will not result in a significant increase in radiological 
consequences of a LOCA [loss of coolant accident]. Therefore, the 
design functions performed by the equipment are not changed.
    The delay in containment spray operation will result in an increase 
in containment temperature, containment pressure, offsite dose, and 
control room dose during a LOCA or high energy line break inside 
containment. Containment analyses have been performed to demonstrate 
that containment pressure and temperature remain within the design 
limits and there is no significant impact on the environmental 
qualification for equipment inside containment. The impact on piping 
and supports is acceptable without modification. The reduction in 
fission product removal due to delayed containment spray operation does 
not result in exceeding the offsite dose and control room dose limits 
in 10 CFR 50.67 and 10 CFR Part 50, Appendix A, GDC 19. The analysis of 
the change in containment conditions due to a single failure of an 
operating spray pump and the suspension of containment spray determined 
that the pressure remained below the design limits.
    Regarding the proposed change to adopt TSTF-493, Rev. 3 on a 
limited basis, the change clarifies the requirements for 
instrumentation to ensure the instrumentation will actuate as assumed 
in the safety analysis. Instruments are not an assumed initiator of any 
accident previously evaluated. As a result, the proposed change will 
not increase the probability of an accident previously evaluated. The 
proposed change will ensure that the instruments actuate as assumed to 
mitigate the accidents previously evaluated. As a result, the proposed 
change will not increase the consequences of an accident previously 
evaluated.
    Based on this discussion, the proposed amendment does not 
significantly increase the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The modifications to install RWST narrow range level indication 
will be seismically qualified and isolated from the safety related 
portion of the RWST level indication system. As such, the new level 
indication will not create the possibility of a new or different kind 
of accident.
    The modification to the low level setpoint will not install any new 
plant equipment. The setpoint will continue to be included within the 
engineered safeguards features instrumentation and monitored according 
to the applicable surveillance requirements. The evaluation of the new 
level setpoint and the change in the swapover sequence concluded that 
the equipment aligned to the sump will continue to have sufficient 
suction pressure prior to containment sump suction swapover. The design 
of the RWST low level instrumentation-complies with all applicable 
regulatory requirements and design criteria.
    The overall function of the Containment Spray System is not changed 
by this proposed amendment. The proposed change alters the method of 
controlling the safety system following a design basis event so that 
manual actions are substituted for automatic actions. Calculations 
confirm that these actions will be taken within the appropriate 
scenario sequence timing to provide containment cooling and source term 
reduction with no significant increase in radiological consequences and 
without exceeding containment design limits.
    Regarding the proposed change to adopt TSTF-493, Rev. 3 on a 
limited basis, the change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The change 
does not alter assumptions made in the safety analysis but ensures that 
the instruments behave as assumed in the accident analysis. The 
proposed change is consistent with the safety analysis assumptions.
    Therefore, the proposed change does not create the possibility of a 
new or

[[Page 15768]]

different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed change has the potential to increase the radiological 
dose at the site boundary and in the control room. However, the 
calculations demonstrate that the dose consequences at the site 
boundary, low population zone, and control room remain within 
regulatory acceptance limits. Additional analysis concluded:
     Peak containment pressure for analyzed design basis 
accidents will not be significantly increased and containment design 
limits will not be exceeded.
     Assumptions used in the environmental qualification of 
equipment exposed to the containment atmosphere remain bounding.
     Pumps aligned to the RWST and to the containment sump will 
have adequate suction pressure.
    Regarding the proposed change to adopt TSTF-493, Rev. 3 on a 
limited basis, the change clarifies the requirements for 
instrumentation to ensure the instrumentation will actuate as assumed 
in the accident analysis. No change is made to the accident analysis 
assumptions and no margin of safety is reduced as part of this change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Melanie C. Wong.

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

    Date of amendment request: October 2, 2008.
    Description of amendment request: The amendments would revise 
Technical Specifications (TS) associated with the verification of ice 
condenser door operability. The proposed amendment affects the current 
TS surveillance requirements 3.6.13.5 and 3.6.13.6.
    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 only analyzed accidents of possible consideration in regards to 
changes potentially affecting the ice condenser are a loss of coolant 
accident (LOCA) and a high energy line break (HELB) inside Containment. 
However, the ice condenser is not postulated as being the initiator of 
any LOCA or HELB. This is because it is designed to remain functional 
following a design basis earthquake, and the ice condenser does not 
interconnect or interact with any systems that interconnect or interact 
with the Reactor Coolant or Main Steam Systems. Since these proposed 
changes do not result in, or require, any physical change to the ice 
condenser that could introduce an interaction with the Reactor Coolant 
or Main Steam Systems, then there can be no change in the probability 
of an accident previously evaluated. Regarding consequences of analyzed 
accidents, the ice condenser is an engineered safety feature designed, 
in part, to limit the Containment sub-compartment and Containment 
vessel pressure immediately following the initiation of a LOCA or HELB. 
Conservative sub-compartment and Containment pressure analysis shows 
these criteria will be met if the total ice mass within the ice bed is 
maintained in accordance with the DBA analysis; therefore, the proposed 
TS [Technical Specification] SR [surveillance requirement] changes of 
these requirements will not increase the consequences of any accident 
previously evaluated.
    Thus, based on the above, 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.
    As previously described, the ice condenser is not postulated as 
being the initiator of any design basis accident. The proposed changes 
do not impact any plant system, structure or component that is an 
accident initiator. The proposed TSs and TS Bases changes do not 
involve any hardware changes to the ice condenser or other change that 
could create any new accident mechanisms. Therefore, there can be no 
new or different accidents created from those already identified and 
evaluated
    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 and the reactor coolant system will 
not be impacted by the proposed changes. The Application provides a 
description of additional sub-compartment and Containment pressure 
response analysis that has been performed. This analysis demonstrates 
that Containment will remain fully capable of performing its design 
function with implementation of the proposed changes. Therefore, no 
safety margin will be significantly impacted.
    The changes proposed in this LAR [license amendment request] do not 
make any physical alteration to the ice condenser doors, nor does it 
affect the required functional capability of the doors in any way. The 
intent of the proposed changes to the ice condenser door surveillance 
requirements is to eliminate an unnecessary and overly restrictive 
Lower Inlet Door torque surveillance test. There will be no degradation 
in the operable status of the ice condenser doors and the ability to 
confirm operability for the ice condenser doors will be maintained, 
such that the doors will continue to fully perform their safety 
function as assumed in the plant's safety analyses.
    Thus, it can be concluded that the proposed TS and TS Bases 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 C. Wong.

[[Page 15769]]

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

    Date of amendment request: October 8, 2008.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TSs) by removing and updating portions of the 
TSs which are out of date or are obsolete including footnotes and 
references.
    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 are administrative in nature and therefore 
they do not involve any change in the design, configuration, or 
operation of the nuclear units. All Limiting Conditions for Operation, 
Limiting Safety System Settings and Safety Limits specified in the 
Technical Specifications remain unchanged. The Physical Security and 
related plans, Operator Training and Requalification Programs, Quality 
Assurance Programs, and the Emergency Plans will not be materially 
changed by the proposed license amendment due to its administrative 
nature.
    The technical qualifications of the operating licensee will not be 
reduced. Personnel engaged in operation, maintenance, engineering, 
assessment, training, and other related services will not be changed. 
Duke officers and executives currently responsible for the overall safe 
operation of the nuclear plants are expected to continue in the same 
capacity.
    Therefore, the proposed amendment does not involve an increase in 
the probability or consequences of an accident previously analyzed.
    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 are administrative in nature and therefore 
they do not involve any change in the design, configuration, or 
operation of the nuclear plant. The current plant safety analyses, 
therefore, remain complete and accurate in addressing the design basis 
events and in analyzing plant response and consequences.
    The Limiting Conditions for Operations, Limiting Safety System 
Settings and Safety Limits specified in the Technical Specifications 
are not affected by the proposed changes. As such, the plant conditions 
for which the design basis accident analyses were performed remain 
valid.
    The amendment does not introduce a new mode of plant operation or 
new accident precursors, does not involve any physical alterations to 
plant configurations or make changes to system set points that could 
initiate a new or different kind of accident.
    Therefore, the proposed amendment does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes are administrative in nature and therefore 
they do not involve a change in the design, configuration, or operation 
of the nuclear plants. The change does not affect either the way in 
which the plant, structures, systems, and components perform their 
safety function or their design and licensing bases.
    Plant safety margins are established through Limiting Conditions 
for Operation, Limiting Safety System Settings and Safety Limits 
specified in the Technical Specifications. Because there is no change 
to the physical design of the plant, there is no change to any of these 
margins.
    Therefore, the proposed amendment does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Melanie C. Wong.

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

    Date of amendment request: October 14, 2008.
    Description of amendment request: The amendments would revise the 
Technical Specification [TS] Administrative Controls, ``Inservice 
Testing Program,'' for consistency with the requirements of Title 10 of 
the Code of Federal Regulations (10 CFR) 50.55a(f)(4) for pumps and 
valves which are classified as American Society of Mechanical Engineers 
[ASME] Code Class 1, Class 2, and Class 3.
    Basis for proposed no significant hazards consideration 
determination: As required by
    10 CFR 50.91(a), the licensee has provided its analysis of the 
issue of no significant hazards consideration, which is presented 
below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise TS 5.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 50.55a(f)(4) 
regarding the inservice testing of pumps and valves which are 
classified as ASME Code Class 1, Class 2, and Class 3. The proposed 
changes incorporate revisions to the ASME Code that result in a net 
improvement in the measures for testing pumps and valves.
    The proposed changes do not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient events. 
The proposed change does not involve the addition or removal of any 
equipment, or any design changes to the facility. Therefore, these 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes revise TS 5.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 50.55a(f)(4) 
regarding the inservice testing of pumps and valves which are 
classified as ASME Code Class 1, Class 2, and Class 3. The proposed 
changes incorporate revisions to the ASME Code that result in a net 
improvement in the measures for testing pumps and valves.
    The proposed changes do not involve a modification to the physical 
configuration of the plant nor does it involve a change in the methods 
governing normal plant operation. The proposed changes will not impose 
any new or different requirements or introduce a new accident 
initiator, accident precursor, or malfunction mechanism. Additionally, 
there is no change in the types or increases in the amounts of any 
effluent that may be released offsite and there is no increase in 
individual or cumulative occupational exposure. Therefore, the

[[Page 15770]]

proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes revise TS 5.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 50.55a(f)(4) 
regarding the inservice testing of pumps and valves which are 
classified as ASME Code Class 1, Class 2, and Class 3. The proposed 
changes do not involve a modification to the physical configuration of 
the plant nor does it change the methods governing normal plant 
operation. The proposed changes incorporate revisions to the ASME Code 
that result in a net improvement in the measures for testing pumps and 
valves. The safety function of the affected pumps and valves will be 
maintained. 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: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Melanie C. Wong.

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

    Date of amendment request: October 2, 2008.
    Description of amendment request: The proposed amendments would 
revise technical specifications (TS) associated with the verification 
of ice condenser door operability. The proposed amendment affects the 
current TS surveillance requirements 3.6.13.5 and 3.6.13.6.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The only analyzed accidents of possible consideration in regards to 
changes potentially affecting the ice condenser are a loss of coolant 
accident (LOCA) and a high energy line break (HELB) inside Containment. 
However, the ice condenser is not postulated as being the initiator of 
any LOCA or HELB. This is because it is designed to remain functional 
following a design basis earthquake, and the ice condenser does not 
interconnect or interact with any systems that interconnect or interact 
with the Reactor Coolant or Main Steam Systems. Since these proposed 
changes do not result in, or require, any physical change to the ice 
condenser that could introduce an interaction with the Reactor Coolant 
or Main Steam Systems, then there can be no change in the probability 
of an accident previously evaluated. Regarding consequences of analyzed 
accidents, the ice condenser is an engineered safety feature designed, 
in part, to limit the Containment sub-compartment and Containment 
vessel pressure immediately following the initiation of a LOCA or HELB. 
Conservative sub-compartment and Containment pressure analysis shows 
these criteria will be met if the total ice mass within the ice bed is 
maintained in accordance with the DBA analysis; therefore, the proposed 
TS [technical specification] SR [surveillance requirement] changes of 
these requirements will not increase the consequences of any accident 
previously evaluated.
    Thus, based on the above, 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.
    As previously described, the ice condenser is not postulated as 
being the initiator of any design basis accident. The proposed changes 
do not impact any plant system, structure or component that is an 
accident initiator. The proposed TSs and TS Bases changes do not 
involve any hardware changes to the ice condenser or other change that 
could create any new accident mechanisms. Therefore, there can be no 
new or different accidents created from those already identified and 
evaluated.
    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 and the reactor coolant system will 
not be impacted by the proposed changes. The Application provides a 
description of additional sub-compartment and Containment pressure 
response analysis that has been performed. This analysis demonstrates 
that Containment will remain fully capable of performing its design 
function with implementation of the proposed changes. Therefore, no 
safety margin will be significantly impacted.
    The changes proposed in this LAR [license amendment request] do not 
make any physical alteration to the ice condenser doors, nor does it 
affect the required functional capability of the doors in any way. The 
intent of the proposed changes to the ice condenser door surveillance 
requirements is to eliminate an unnecessary and overly restrictive 
Lower Inlet Door torque surveillance test. There will be no degradation 
in the operable status of the ice condenser doors and the ability to 
confirm operability for the ice condenser doors will be maintained, 
such that the doors will continue to fully perform their safety 
function as assumed in the plant's safety analyses.
    Thus, it can be concluded that the proposed TS and TS Bases 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 Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: February 24, 2009.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) Surveillance Requirement (SR) 
that governs operability testing of the pressure suppression chamber-
drywell vacuum breakers to incorporate the SR contained within the 
Standard

[[Page 15771]]

Technical Specifications (STS), NUREG-1433 and delete the SR that 
requires inspection of the pressure suppression chamber-drywell vacuum 
breakers. Periodic inspections of the pressure suppression chamber-
drywell vacuum breakers are not required by the STS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The operation of Vermont Yankee Nuclear Power Station (VY) in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment does not impact the operability of any 
structure, system or component that affects the probability of an 
accident or that supports mitigation of an accident previously 
evaluated. The proposed amendment does not affect reactor operations or 
accident analysis and has no radiological consequences. The operability 
requirements for accident mitigation systems remain consistent with the 
licensing and design basis. Therefore, the proposed amendment does not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The operation of VY in accordance with the proposed amendment 
will not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed amendment does not change the design or function of 
any component or system. No new modes of failure or initiating events 
are being introduced. Therefore, operation of VY in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The operation of VY in accordance with the proposed amendment 
will not involve a significant reduction in a margin of safety.
    The proposed amendment does not change the design or function of 
any component or system. The proposed amendment does not involve any 
safety limits, safety settings or safety margins. The ability of the 
pressure suppression chamber-drywell vacuum breakers to perform its 
intended function will continue to be required in accordance with the 
VY Technical Specifications.
    Since the proposed controls are adequate to ensure the operability 
of the pressure suppression chamber-drywell vacuum breakers, there will 
still be high assurance that the components are operable and capable of 
performing their respective functions. Therefore, operation of VY in 
accordance with the proposed amendment will not involve a significant 
reduction in the margin to safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Mark G. Kowal.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: October 23, 2008.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to support the application of 
alternative source term (AST) methodology with respect to the loss-of-
coolant accident and the fuel handling accident. The proposed request 
is to support a full-scope application of an AST methodology, with the 
exception that Technical Information Document (TID)-14844, 
``Calculation of Distance Factors for Power and Test Reactor Sites,'' 
will continue to be used as the radiation dose basis for equipment 
qualification.
    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 implementation of AST assumptions has been evaluated in 
revisions to the analyses of the following limiting design basis 
accidents at LSCS [LaSalle County Station]:
     Loss-of-Coolant Accident, and
     Fuel Handling Accident
    Based upon the results of these analyses, it has been demonstrated 
that, with the requested changes, the dose consequences of these 
limiting events are within the regulatory requirements and guidance 
provided by the NRC for use with AST. The regulatory requirements and 
guidance is presented in 10 CFR 50.67, ``Accident source term,'' and 
associated NRC Regulatory Guide 1.183 and Standard Review Plan section 
15.0.1. The AST is an input to calculations used to evaluate the 
consequences of an accident, and does not by itself affect the plant 
response, or the actual pathway of the radiation released from the 
fuel. It does, however, better represent the physical characteristics 
of the release, so that appropriate mitigation techniques may be 
applied. Therefore, the consequences of an accident previously 
evaluated are not significantly increased.
    The equipment affected by the proposed change is mitigative in 
nature, and relied upon after an accident has been initiated. 
Application of the AST does not involve any physical changes to the TS, 
while they revise certain performance requirements, do not involve any 
physical modifications to the plant. As a result, the proposed changes 
do not affect any of the parameters or conditions that could contribute 
to the initiation of any accidents. As such, removal of operability 
requirements during the specified conditions will not significantly 
increase the probability of occurrence for an accident previously 
analyzed. Since plant design basis accidents initiators are not being 
altered by adoption of the AST analyses, the probability of an accident 
previously evaluated is not affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be installed 
and there are no physical modifications to existing equipment 
associated with the proposed change). Similarly, it does not physically 
change any structures, systems, or components involved in the 
mitigation of any accidents. Thus, no new initiators or precursors of a 
new or different kind of accident 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?
    Response: No.

[[Page 15772]]

    Safety margins and analytical conservatisms have been evaluated and 
have been found to be acceptable. The analyzed events have been 
carefully selected and margin has been retained to ensure that the 
analyses adequately bound postulated event scenarios. The dose 
consequences due to design basis accidents comply with the requirements 
of 10 CFR 50.67 and guidance of Regulatory Guide 1.183.
    The proposed change is associated with the implementation of a new 
licensing basis for LSCS design basis accidents. Approval of the change 
from the original source term to a new source term taken from 
Regulatory Guide 1.183 is being requested. The results of the accident 
analyses, revised in support of the proposed license amendment, are 
subject to revised acceptance criteria. The analyses have been 
performed using conservative methodologies, as specified in Regulatory 
Guide 1.183. Safety margins have been evaluated and analytical 
conservatism has been utilized to ensure that the analyses adequately 
bound the postulated limiting event scenario. The dose consequences of 
these design basis accidents remain within the acceptance criteria 
presented in 10 CFR 50.67 and Regulatory Guide 1.183.
    The proposed change continues to ensure that the doses at the 
exclusion area boundary and low population zone boundary, as well as 
the control room, are within corresponding regulatory limits.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 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: February 11, 2009.
    Brief description of amendment: The proposed amendment consists of 
administrative revision to the operating licenses and Technical 
Specifications (TSs) to revise the station name from Comanche Peak 
Steam Electric Station (CPSES) to Comanche Peak Nuclear Power Plant 
(CPNPP); remove the Table of Contents from TSs and maintain and revise 
it in accordance with plant administrative procedures; delete TSs 
3.2.1.1, 3.2.3.1, 5.5.9.1, 5.6.10 and several footnotes from Tables 
3.3.1-1, 3.3.2-1, and TS 3.4.10 since these TSs and footnotes are no 
longer applicable to CPSES, Units 1 and 2 operation; delete several 
topical reports from the list of approved analytical methods used to 
determine core operating limits in TS 5.6.5, no longer in use, since 
these topical reports have been replaced by standard Westinghouse 
methods and Westinghouse methods have been approved for use at CPSES, 
Units 1 and 2, under a separate amendment request; make editorial 
corrections; and reprint and reissue the entire TS due to adoption of 
`FrameMaker' software in place of `Microsoft Word' software.
    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 revises the station name, removes the Table of 
Contents from the Technical Specifications, deletes several Technical 
Specifications and footnotes which are no longer applicable to [CPSES] 
Unit 1 or Unit 2 operation, renumbers subsequent Technical 
Specifications, deletes several topical reports from the list of 
approved analytical methods used to determine core operating limits, 
and corrects various editorial and formatting errors. The Table of 
Contents does not include information required by 10 CFR 50.36 [Title 
10 of the Code of Federal Regulations, Section 50.36] to be reviewed by 
the NRC [U.S. Nuclear Regulatory Commission] staff and is not required 
by the regulation. The Technical Specifications and footnotes which are 
being deleted were only applicable during previous operational cycles 
and are now defunct requirements since both Units have completed the 
applicable operational cycles. The topical reports deleted from 
Technical Specification 5.6.5b are no longer used to determine the core 
operating limits for Comanche Peak Nuclear Power Plant. The remaining 
topical reports listed in Technical Specification 5.6.5b will be used 
to determine the core operating limits for both Comanche Peak Nuclear 
Power Plant units. All other changes proposed are corrections of 
previous inadvertent editorial errors or changes in format to increase 
conformity with the guidelines described in TSTF-RPT-01, ``Writer's 
Guide for Plant-Specific Improved Technical Specifications'', published 
in June, 2005. All of the proposed changes are administrative changes 
which do not change the meaning, intent, interpretation, or application 
of the Technical Specifications. None of the proposed changes affect 
the operation, physical configuration, or function of plant equipment 
or systems. The changes do not affect the initiators or assumptions of 
analyzed events; nor do they impact the mitigation of accidents or 
transient events. 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 revises the station name, removes the Table of 
Contents from the Technical Specifications, deletes several Technical 
Specifications and footnotes which are no longer applicable to [CPSES,] 
Unit 1 or Unit 2 operation, renumbers subsequent Technical 
Specifications, deletes several topical reports from the list of 
approved analytical methods used to determine core operating limits, 
and corrects various editorial and formatting errors. The Table of 
Contents does not include information required by 10 CFR 50.36 to be 
reviewed by the Nuclear Regulatory Commission staff and is not required 
by the regulation. The Technical Specifications and footnotes which are 
being deleted were only applicable during previous operational cycles 
and are now defunct requirements since both Units have completed the 
applicable operational cycles. The topical reports deleted from 
Technical Specification 5.6.5b are no longer used to determine the core 
operating limits for Comanche Peak Nuclear Power Plant. The remaining 
topical reports listed in Technical Specification 5.6.5b will be used 
to determine the core operating limits for both Comanche Peak Nuclear 
Power Plant units. All other changes proposed are corrections of 
previous inadvertent editorial errors or changes in format to increase 
conformity with the guidelines described in TSTF-RPT-01, ``Writer's 
Guide for Plant-Specific Improved Technical Specifications'', published 
in June, 2005. All of the proposed changes are administrative changes 
which do not change the meaning, intent,

[[Page 15773]]

interpretation, or application of the Technical Specifications. None of 
the changes alter the plant configuration, require installation of new 
equipment, alter assumptions about previously analyzed accidents, or 
impact the operation or function of any plant equipment or systems. 
Therefore, the proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the station name, removes the Table of 
Contents from the Technical Specifications, deletes several Technical 
Specifications and footnotes which are no longer applicable to [CPSES,] 
Unit 1 or Unit 2 operation, renumbers subsequent Technical 
Specifications, deletes several topical reports from the list of 
approved analytical methods used to determine core operating limits, 
and corrects various editorial and formatting errors. The Table of 
Contents does not include information required by 10 CFR 50.36 to be 
reviewed by the Nuclear Regulatory Commission staff and is not required 
by the regulation. The Technical Specifications and footnotes which are 
being deleted were only applicable during previous operational cycles 
and are now defunct requirements since both Units have completed the 
applicable operational cycles. The topical reports deleted from 
Technical Specification 5.6.5b are no longer used to determine the core 
operating limits for Comanche Peak Nuclear Power Plant. The remaining 
topical reports listed in Technical Specification 5.6.5b will be used 
to determine the core operating limits for both Comanche Peak Nuclear 
Power Plant units. All other changes proposed are corrections of 
previous inadvertent editorial errors or changes in format to increase 
conformity with the guidelines described in TSTF-RPT-01, ``Writer's 
Guide for Plant-Specific Improved Technical Specifications'', published 
in June, 2005. All of the proposed changes are administrative changes 
which do not change the meaning, intent, interpretation, or application 
of the Technical Specifications. None of the proposed changes alter the 
effective technical content of the Technical Specifications. Therefore 
the proposed changes do 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.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: October 31, 2008.
    Description of amendment request: The proposed amendment modifies 
the surveillance requirements in Technical Specification (TS) 3.6(3), 
``Containment Recirculating Air Cooling and Filtering System,'' and 
removes the license conditions related to the replacement and testing 
of containment air cooling and filtering (CACF) unit high-efficiency 
particulate air (HEPA) filters and surveillance testing of the CACF 
unit relief ports. These license conditions were committed to by the 
licensee in its letter dated April 10, 2008 (Agencywide Documents 
Access and Management System (ADAMS) Accession No. ML081010122), and 
implemented via TS Amendment No. 255 (ADAMS Accession No. ML081140390), 
dated May 2, 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 amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The containment air cooling and filtering system (CACFS) is not an 
initiator of any accident previously evaluated at the Fort Calhoun 
Station (FCS). The CACFS is an accident mitigation system. The design 
basis function of the CACFS is to limit the containment pressure rise 
by providing a means for cooling the containment following a loss-of-
coolant accident (LOCA) or main steam line break (MSLB). In accordance 
with TS Amendment No. 255, the CACFS high efficiency particulate air 
(HEPA) filters are also credited to reduce post-LOCA radioactive 
leakage from containment.
    The proposed changes provide additional assurance that the CACFS is 
capable of performing its design and licensing basis functions to 
mitigate these design basis accidents (DBAs). The CACFS face and bypass 
dampers are aligned to their accident positions permanently causing the 
CACFS to operate in filtered air mode. Surveillance testing has shown 
that operating the system in this alignment over long periods does not 
jeopardize filter performance. Over the lifetime of the plant, the 
differential pressures measured across the combined HEPA and charcoal 
filter banks have met test acceptance criteria.
    Increasing the number of surveillance requirements will not 
adversely affect the function of the CACFS but rather provides 
additional assurance that the CACFS is capable of responding to a DBA.
    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 CACFS was designed to remove heat released to the containment 
atmosphere during a DBA to the extent necessary to maintain the 
containment structure below its design pressure. The containment 
airflow continually passes through the cooling coils. The proposed 
changes to the surveillance requirements do not affect the active 
function of the CACFS.
    The CACFS will continue to operate in normal and accident 
conditions to remove heat and radioactive particulates and aerosols. 
The proposed changes enhance surveillance testing of the CACFS by 
requiring more frequent exercising of the fans, imposing a more 
stringent pressure drop limit, specifying a HEPA filter replacement 
interval, and instituting a requirement to exercise the relief ports. 
These changes ensure that the CACFS is capable of long-term operation 
in filtered air mode while remaining capable of providing cooling and 
filtering sufficient to mitigate design basis accidents.
    No credible new failure mechanisms, malfunctions, or accident 
initiators not previously considered in the design and licensing basis 
are created and none of the initial condition assumptions of any 
accident evaluated in the safety analysis are impacted.
    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.
    The containment building and associated penetrations are designed 
to

[[Page 15774]]

withstand an internal pressure of 60 psig [pounds per square inch 
gauge] at 305[deg]F, including all thermal loads resulting from the 
temperature associated with this pressure, with a leakage rate of 0.1 
percent by weight or less of the contained volume per 24 hours. [Omaha 
Public Power District] credits the CACFS in the containment pressure 
analysis for a LOCA, and for the containment pressure response to a 
main steam line break (MSLB).
    The proposed changes impose more stringent surveillance test 
requirements. This provides additional assurance that the CACFS will 
perform its design basis and licensing basis functions to be capable of 
long-term post-DBA operation in filtered air mode to limit the 
containment temperature and pressure increase to within design limits 
and to reduce post-LOCA radioactive leakage from containment.
    Neither the design basis nor the licensing basis for post-DBA 
containment heat removal is adversely affected by the proposed changes. 
The ability to maintain design limits for containment peak pressure and 
temperature, as well as long-term containment pressure and temperature, 
are preserved.
    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: David A. Repka, Esq., Winston & Strawn, 1700 
K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: January 30, 2009.
    Description of amendment request: The proposed amendment would 
modify the Fort Calhoun Station (FCS), Unit No. 1, Renewed Operating 
License No. DPR-40, by adding operability and surveillance testing 
requirements to the FCS Technical Specifications (TS) for the steam 
generator (SG) blowdown isolation on a reactor trip. Specifically, the 
proposed changes will revise TS Limiting Conditions for Operation (LCO) 
2.15, Instrumentation and Control Systems, Table 2-4, Instrument 
Operating Conditions for Isolation Functions, to include operability 
requirements for SG blowdown isolation on a reactor trip and to add 
applicable footnotes. In addition, TS 3.1, Instrumentation and Control, 
Table 3-2, Minimum Frequencies for Checks, Calibrations and Testing of 
Engineered Safety Features, Instrumentation and Controls, is being 
revised to include the surveillance test requirements for SG blowdown 
isolation on a reactor trip. An administrative change is also being 
made to TS LCO 2.15(1), to delete the words ``key operated'' as the 
``key'' associated with the bypass switches is not a critical element 
in controlling the use of bypass switches. This amendment will allow 
FCS to credit an automatic SG blowdown isolation interlock being 
installed during the 2009 Refueling Outage (RFO).
    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 change provides Technical Specification (TS) 
operability and surveillance testing requirements for automatic steam 
generator (SG) blowdown isolation on a reactor trip in the event of a 
loss of main feedwater (LMFW). Automatic isolation will ensure that the 
existing 15-minute requirement in the Updated Safety Analysis Report 
(USAR) Chapter 14.10 safety analysis is met without the risk that an 
unanticipated distraction could prevent manual action from occurring at 
the proper time. The installation of this feature will eliminate the 
need for manual isolation of blowdown and thus will eliminate the 
associated operator challenge.
    Automatic isolation of blowdown will reduce the consequences of the 
LMFW event by providing automatic isolation prior to manual isolation 
being initiated by the operators. Automatic isolation at the time of 
reactor trip will reduce the severity of the LMFW event by isolating 
the SGs earlier in the event, thereby conserving SG inventory.
    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.
    No new malfunctions are being introduced by this activity, and 
based on the current redundancy in the design, there are no 
malfunctions of the SG blowdown isolation valves that challenge nuclear 
safety.
    The SG blowdown isolation valves will continue to function as 
currently credited for the LMFW event; thus, this proposed change does 
not alter their ability to function as containment isolation valves to 
maintain containment integrity. The manual isolation capability remains 
unchanged.
    A failure analysis has been prepared which shows that the addition 
of the automatic isolation feature does not introduce a new failure 
mode or malfunction to the valve circuits. An isolation of SG blowdown, 
either through the designed circuit following a reactor trip, or during 
normal operations, does not present a nuclear safety challenge. The 
capability exists for operators to bypass the isolation signal and 
restore blowdown as plant conditions warrant.
    Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    The addition of an automatic isolation interlock to the SG blowdown 
isolation valve circuits that close the valves on a reactor trip 
actually increases the margin of safety by isolating the SG early in 
the event to maintain SG inventories.
    A reactor trip signal is generated in the first seconds of an LMFW 
due to reduced SG inventories. Because it is desirable to isolate 
blowdown as soon as possible following the LMFW event, for maximum 
margin, a reactor trip signal will be used for the SG blowdown 
isolation interlock. Isolating blowdown earlier in an event provides 
greater operating margin in terms of maximizing SG inventories. More 
margin allows operators more time to address operator demands that 
occur during transient events.
    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.

[[Page 15775]]

    Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 
K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: January 30, 2009.
    Description of amendment request: The proposed amendment would 
delete those portions of the Technical Specifications (TS) superseded 
by Title 10 of the Code of Federal Regulations (10 CFR) Part 26, 
Subpart I. The licensee is proposing to adopt the approved Technical 
Specification Task Force (TSTF) change traveler TSTF-511, Revision 0, 
``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, using the consolidated line item 
improvement process. In its application dated January 30, 2009, the 
licensee affirmed the applicability of the model NSHC determination, 
which is presented below.
    Basis for proposed (NSHC) determination: As required by 10 CFR 
50.91(a), an analysis of the issue of NSHC determination 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 Technical Specification restrictions on 
working hours for personnel who perform safety related functions. The 
Technical Specification restrictions are superseded by the worker 
fatigue requirements in 10 CFR Part 26. Removal of the Technical 
Specification requirements will be performed concurrently with the 
implementation of the 10 CFR Part 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 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 Technical Specification restrictions on 
working hours for personnel who perform safety related functions. The 
Technical Specification restrictions are superseded by the worker 
fatigue requirements in 10 CFR Part 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 Technical Specification restrictions on 
working hours for personnel who perform safety related functions. The 
Technical Specification restrictions are superseded by the worker 
fatigue requirements in 10 CFR Part 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 Technical Specification 
administrative requirements will not reduce a margin of safety because 
the requirements in 10 CFR Part 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 analysis adopted by the licensee 
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 NSHC.
    Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 
K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: December 19, 2008.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) to (1) correct an error in TS 
Table 3.3.2-1, ``Engineered Safety Feature Actuation System 
Instrumentation,'' Function 1.a, to reflect the correct CONDITIONS for 
applicable Modes 1, 2, 3, and 4, (2) revise TS Limiting Condition for 
Operation (LCO) 3.3.4 degraded voltage relay and loss of voltage relay 
Limiting Safety System Settings values to reflect the revised analysis, 
and (3) revise the load requirement of Surveillance Requirement 3.8.1.3 
to reflect values supported by the diesel generator accident loading 
analyses.
    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 to LCO 3.3.2 correct an administrative error 
which directed inadequate action in the event that a channel of 
instrumentation is lost for manual safety injection initiation. The 
amendment places the plant in a

[[Page 15776]]

more conservative condition, Mode 5, if the other Required Actions 
cannot be executed within their periodicity.
    The proposed changes to LCO 3.3.4 provide setpoint changes based on 
a revised calculation, which generated new setpoints for the loss of 
voltage relays and degraded voltage relays. The new setpoints ensure 
the protective relays will function when required, will ensure 
protection from thermal damage to loads on the 480V busses, and will 
not cause unintended diesel generator starts even in worst case 
scenarios, with power provided from offsite.
    The proposed changes to LCO 3.8.1 involve an increase in the 
minimum load band value for diesel generator surveillance SR 3.8.1.3. 
This change ensures that the diesel generators are capable of 
synchronizing with the offsite electrical system and accepting loads 
greater than or equal [to] the equivalent of the maximum expected 
accident loads. The new load band value is more conservative than the 
existing value and provides a more thorough test to ensure equipment 
emergency response capability.
    Therefore, the probability or consequences of an accident 
previously evaluated will not be significantly increased.
    2. Do the proposed amendments create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes involve correcting an administrative error and 
revising previously established values associated with the diesel 
generators to increase conservatism. None of these proposed changes 
involve a physical alteration of the plant (i.e., no new or different 
types of equipment will be installed) or a change in methods governing 
normal plant operation. The proposed changes preserve the safety 
analysis assumptions related to accident mitigation. No initiators or 
accident precursors are created by this change. Therefore, the 
possibility of a new or different kind of accident not previously 
evaluated is not created.
    3. Do the proposed amendments involve a significant reduction in a 
margin of safety?
    Response: No.
    The level of safety of facility operation is unaffected by any of 
the proposed changes. The requested administrative change is 
conservative compared to the existing requirement. The response of the 
diesel generators to accident transients reported in the Updated Final 
Safety Analysis Report (UFSAR) is unaffected by these changes. The 
proposed changes preserve the safety analysis assumptions related to 
accident mitigation. Therefore, these 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: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Group, LLC, 750 East Pratt Street, 17 Floor, 
Baltimore, MD 21202.
    NRC Branch Chief: Mark G. Kowal.

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 [email protected].

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

    Date of application for amendments: July 7, 2008, as supplemented 
by letters dated December 17, 2008, and March 9, 2009.
    Brief description of amendments: The amendments revise Surveillance 
Requirement (SR) 3.6.1.6.1 to add a new requirement to verify that each 
vacuum breaker is closed within 6 hours following an operation that 
causes any of the vacuum breakers to open and, also, revise SR 
3.6.1.6.2 by removing the requirement to perform functional testing of 
each vacuum breaker within 12 hours following an operation that causes 
any of the vacuum breakers to open.
    Date of issuance: March 11, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 251 and 279.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: September 23, 2008 (73 
FR 54864). The supplemental letter provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 11, 2009.
    No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

[[Page 15777]]

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois

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

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

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

Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

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

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

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

    Date of application for amendments: April 21, 2008, as supplemented 
on March 11, 2009.
    Brief description of amendments: The proposed amendment removes 
references to and limits provided by Nuclear Regulatory Commission 
Generic Letter (GL) 82-12, ``Nuclear Power Plant Staff Working Hours,'' 
from the subject plants' technical specifications (TS). The references 
and limitations have been superseded by the requirements of Title 10 of 
the Code of Federal Regulations, Part 26 (10 CFR 26), Subpart I, 
``Managing Fatigue.''
    Date of issuance: March 23, 2009.
    Effective date: As of the date of issuance and shall be implemented 
by October 1, 2009.
    Amendment Nos.: 157, 157, 162, 162, 185, 231, 224, 192, 179, 198, 
159, 274, 271, 275, 243, 238, 270.
    Facility Operating License Nos. NPF-72, NPF-77, NPF-37, NPF-66, 
NPF-62, DPR-19, DPR-25, NPF-11, NPF-18, NPF-39, NPF-85, DPR-16, DPR-44, 
DPR-56, DPR-29, DPR-30, DPR-50: The amendments revised the Technical 
Specifications/Licenses.
    Date of initial notice in Federal Register: June 3, 2008 (73 FR 
31721). The March 11, 2009, supplement contained clarifying information 
and did not change the NRC staff's initial proposed finding of no 
significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 23, 2009.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 24, 2008, as supplemented by 
letters dated September 11 and 19, 2008, November 6, 2008, and February 
26, 2009.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Section 3.7.3, ``Reactor Equipment Cooling (REC) 
System,'' to allow credit for the ability to align the service water 
system to the REC system.
    Date of issuance: March 20, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 232.
    Facility Operating License No. DPR-46: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 22, 2008 (73 FR 
21660). The supplemental letters dated September 11 and 19, 2008, 
November 6, 2008, and February 26, 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 March 20, 2009.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station, Unit No. 1 (NMP1), Oswego County, New York

    Date of application for amendment: August 15, 2008, as supplemented 
on December 4, 2008.
    Brief description of amendments: The amendment revises NMP1 
Technical Specification (TS) 6.5.7, ``10 CFR 50 [Part 50 of Title 10 of 
the Code of Federal Regulations Appendix J Testing Program Plan,'' to 
allow a one-time extension of the Integrated Leak Rate Test (ILRT) 
interval for no more than 5 years. The amendment allows the next ILRT 
for NMP1 to be performed within 15 years from the last ILRT as opposed 
to the current 10-year interval.
    Date of issuance: March 11, 2009.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 202.
    Renewed Facility Operating License No. DPR-063: The amendment 
revises the License and TSs.
    Date of initial notice in Federal Register: October 21, 2008 (73 FR 
62566). The supplement dated December 4, 2008, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the Nuclear 
Regulatory Commission staff's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 11, 2009.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 3, 2008, as supplemented 
on February 23, 2009.
    Brief description of amendment: The amendment adopted the proposed 
requirements regarding control room envelope habitability set forth in 
Technical Specifications Task Force (TSTF) change traveler TSTF-448, 
Revision 3. Specifically, the amendment revised the requirements in TS 
Section 3.7.4, ``Control Room Emergency Filtration (CREF) System,'' 
adds a new TS Section 5.5.13, ``Control Room Envelope Habitability 
Program,'' and added a license condition to the operating license to 
implement the TS changes.
    Date of issuance: March 17, 2009.
    Effective date: As of the date of issuance and shall be implemented 
by November 1, 2009.
    Amendment No.: 160.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 6, 2008 (73 FR 
25043). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 17, 2009.
    No significant hazards consideration comments received: No.

[[Page 15778]]

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of application for amendments: July 31, 2008.
    Brief description of amendments: The amendments changed the PPL 
Susquehanna, LLC (PPL) Units 1 and 2 Technical Specification 3.6.1.3 
``Primary Containment Isolation Valves (PCIVs).'' It revised the 
Secondary Containment Bypass Leakage limit in Surveillance Requirement 
3.6.1.3.11 from ``less than or equal to 9 standard cubic foot/feet per 
hour (scfh)'' to ``less than or equal to 15 scfh when pressurized to 
greater than or equal to Pa.''
    Date of issuance: March 18, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 251 for Unit 1 and 231 for Unit 2.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the License and Technical Specifications.
    Date of initial notice in Federal Register: November 18, 2008 (73 
FR 68455). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation (SE) dated March 18, 2009.
    No significant hazards consideration comments received: No. 
However, comments have been received from the Commonwealth of 
Pennsylvania and have been addressed in the SE.

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

    Date of application for amendment: March 19, 2008, as supplemented 
October 7, 2008, November 17, 2008, and December 10, 2008.
    Brief description of amendment: The amendments revise the technical 
specifications (TSs) to (1) delete TS 3.7.13, ``MCR/ESGR Bottled Air 
System,'' (2) create TS 3.3.6, ``Main Control Room/Emergency Switchgear 
Room (MCR/ESGR) Envelope Isolation Actuation Instrumentation,'' to 
establish the operability requirements for the MCR/ESGR envelope 
isolation function, and (3) incorporate TS 3.7.14, ``MCR/ESGR Emergency 
Ventilation During Movement of Recently Irradiated Fuel Assemblies,'' 
into TS 3.7.10, ``MCR/ESGR Emergency Ventilation System.'' The changes 
revise the TSs to be consistent with the assumptions of the current 
dose analysis of record, performed in accordance with Title 10 of the 
Code of Federal Regulations, Section 50.67, ``Accident Source Term,'' 
and the results of the nonpressurized MCR/ESGR envelope tracer gas 
testing.
    Date of issuance: March 25, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 255/236.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
change the licenses and the technical specifications.
    Date of initial notice in Federal Register: April 22, 2008 (73 FR 
21661). The supplements dated October 7, 2008, November 17, 2008, and 
December 10, 2008, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated March 25, 2009.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 30th of March, 2009.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
 [FR Doc. E9-7494 Filed 4-6-09; 8:45 am]
BILLING CODE 7590-01-P