Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 8281-8294 [E9-3515]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 74, Number 35 (Tuesday, February 24, 2009)]
[Notices]
[Pages 8281-8294]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E9-3515]


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NUCLEAR REGULATORY COMMISSION

[NRC-2009-0062]


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 January 29, 2009, to February 11, 2009. The 
last biweekly notice was published on February 10, 2009 (74 FR 6662).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, 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

[[Page 8282]]

Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
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 
hearingdocket@nrc.gov, or by calling (301) 415-1677, to request (1) a 
digital ID certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and/or (2) 
creation of an electronic docket for the proceeding (even in instances 
in which the petitioner/requestor (or its counsel or representative) 
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer 
TM to access the Electronic Information Exchange (EIE), a 
component of the E-Filing system. The Workplace Forms Viewer 
TM is free and is available at https://www.nrc.gov/site-help/
e-submittals/install-viewer.html. Information about applying for a 
digital ID certificate is available on NRC's public Web site at https://
www.nrc.gov/site-help/e-submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at https://www.nrc.gov/
site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitter an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC electronic filing 
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time, 
Monday through Friday, excluding government holidays. The electronic 
filing Help Desk can be contacted by telephone at 1-866-672-7640 or by 
e-mail at MSHD.Resource@nrc.gov.
    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

[[Page 8283]]

submit documents in paper format. Such filings must be submitted by: 
(1) First class mail addressed to the Office of the Secretary of the 
Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, 
express mail, or expedited delivery service to the Office of the 
Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville 
Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff. Participants filing a document in this manner are 
responsible for serving the document on all other participants. Filing 
is considered complete by first-class mail as of the time of deposit in 
the mail, or by courier, express mail, or expedited delivery service 
upon depositing the document with the provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the petition 
and/or request should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii).
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://www.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, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of amendment request: September 29, 2008, as supplemented by 
letter dated January 16, 2009.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) Sections 5.6.1.3.a and 5.6.1.3.b to 
incorporate the results of a new criticality analysis. Specifically the 
TSs would be revised to add new requirements for the Boiling Water 
Reactor (BWR) spent fuel storage racks containing Boraflex in Spent 
Fuel Pools A and B. The requirements for the BWR spent fuel racks as 
currently contained in TS 5.6.1.3 would be revised to specify 
applicability to the spent fuel storage racks containing Boral in Spent 
Fuel Pool B.
    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 activity changes the design basis of the BWR Boraflex 
storage racks, but does not make physical changes to the facility. The 
change to TS Section 5.6.1.3 (BWR Storage Racks in Pools A and B), 
which is an update to the administrative controls for maintaining the 
required boron concentration in the Boraflex BWR spent fuel storage 
racks located in Pools A and B, does not modify the facility.
    The accidents currently analyzed in the FSAR [Final Safety Analysis 
Report] applicable to the proposed activity are fuel handling 
accidents. These accidents include dropping a fuel assembly onto the 
top of a fuel rack or in the space between a rack and the pool wall. 
These events are caused either by personnel error or equipment 
malfunction.
    Based on the new criticality analysis, revised acceptance criteria 
are needed to ensure the criticality safety of fuel storage in BWR 
Boraflex racks in Pools A and B. Similar administrative controls were 
previously placed on fuel stored in the PWR [Pressurized Water Reactor] 
Boraflex racks in Pools A and B. These changes will eliminate the 
dependence on the Boraflex absorber in the BWR storage racks. These 
changes do not impact the probability of having a fuel handling 
accident and do not impact the consequences of a fuel handling 
accident.
    Therefore, this amendment 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.
    These revised acceptance criteria applicable to the irradiated fuel 
stored in the BWR Boraflex racks in Pools A and B are being added to TS 
Section 5.6.1.3.a.
    The proposed change does not result in any credible new failure 
mechanisms, malfunctions or accident initiators not considered in the 
original design and licensing bases.
    Detailed analyses have been performed to ensure a criticality 
accident in Pools A and B is not a credible event. The events that 
could lead to a criticality accident are not new. These events include 
a fuel mispositioning event, a fuel drop event, and a boron dilution 
event. The proposed changes do not impact the probability of any of 
these events.
    The detailed criticality analyses performed demonstrates that 
criticality would not occur following any of these events. Even in a 
more likely event, such as a fuel mispositioning event, the acceptance 
criteria for keff [the effective multiplication factor] 
remains less than or equal to 0.95. In the unlikely event that the 
spent fuel storage pool boron concentration were reduced to zero, 
keff remains less than 1.0. A criticality accident is 
considered ``not credible'' and the proposed action does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Therefore, the proposed change will 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.
    Incorporation of the revised criteria for fuel stored in the BWR 
Boraflex racks in Pools A and B do not involve a reduction in the 
margin of safety. The updated fuel storage condition continues to meet 
keff <0.95 with credit for soluble boron and keff 
< 1.0 when flooded with unborated water.
    The proposed changes for storage of irradiated fuel in BWR Boraflex 
racks in

[[Page 8284]]

Pools A and B continues to provide the controls necessary to ensure a 
criticality event could not occur in the spent fuel storage pool. The 
acceptance criteria are consistent with the acceptance criteria 
specified in 10 CFR 50.68, which provide an acceptable margin of safety 
with regard to the potential for a criticality event.
    Therefore, this amendment does not involve a significant reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Thomas H. Boyce.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant, Citrus County, Florida

    Date of amendment request: August 28, 2008, as supplemented by 
letter dated January 19, 2009.
    Description of amendment request: The proposed amendment would 
implement the Technical Specification Task Force Standard Technical 
Specification Change Traveler 449, Revision 4 inspection requirements 
for the replacement once through steam generators (OTSGs) that are 
being installed during the Crystal River Unit 3 Nuclear Generating 
Plant fall 2009 refueling outage. The replacement OTSGs differ from the 
existing OTSGs in that the tube material is Alloy 690 thermally treated 
in the replacements versus Alloy 600 in the existing OTSGs. 
Additionally, this amendment would remove inspection requirements that 
are designated for specific damage conditions in the existing OTSGs, 
remove tube repair techniques approved by the license amendment No. 
233, dated May 16, 2007, for the existing OTSGs, and remove inspection 
and reporting requirements specific to those repair techniques.
    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 Proposed Change Does Not Involve a Significant Increase in 
the Probability or Consequences of an Accident Previously Evaluated.
    The proposed change for replacement OTSGs continues to implement 
the current OTSG Program that includes performance criteria which 
provide reasonable assurance that the replacement OTSG tubing will 
retain integrity over the full range of operating conditions (including 
startup, operation in the power range, hot standby, cooldown and all 
anticipated transients included in the design specifications). This 
change removes repair criteria from the OTSG Program that were approved 
by previous License Amendments for the existing Steam Generators which 
are not applicable to the replacement OTSGs. It removes references to 
use of repairs and reporting of repair results in other Technical 
Specification sections. This change removes inspection requirements 
that are designated for specific damage conditions in the existing 
OTSGs.
    The change also revises the inspection interval for 100% 
inspections of OTSG tubes and the maximum interval for inspection of a 
single OTSG consistent with Technical Specification Task Force item 449 
for the Alloy 690 tube material in the replacement OTSGs. The revised 
inspection requirements are based on properties and experience with the 
improved Alloy 690 tube material. The revised inspection requirements 
will result in the same outcome that OTSG tube integrity will continue 
to be maintained.
    This change continues to implement steam generator performance 
criteria for tube structural integrity, accident induced leakage, and 
operational leakage for the replacement OTSGs. Meeting the performance 
criteria provides reasonable assurance that the replacement OTSG tubing 
will remain capable of fulfilling its specific safety function of 
maintaining reactor coolant pressure boundary integrity throughout each 
operating cycle and in the unlikely event of a design basis accident. 
The performance criteria are only a part of the OTSG program required 
by the existing ITS [Improved Technical Specification]. The program, 
defined by NEI [Nuclear Energy Institute] 97-06, Steam Generator 
Program Guidelines, includes a framework that incorporates a balance of 
prevention, inspection, evaluation, repair, and leakage monitoring. 
These features will continue to be implemented as they are currently 
approved. The proposed changes do not, therefore, significantly 
increase the probability of an accident previously evaluated.
    The consequences of design basis accidents are, in part, functions 
of the DOSE EQUIVALENT I-131 in the primary coolant and the primary to 
secondary LEAKAGE rates resulting from an accident. Therefore, limits 
are included in the plant technical specifications for operational 
leakage and for DOSE EQUIVALENT I-131 in the primary coolant to ensure 
the plant is operated within its analyzed condition. The analysis of 
the limiting design basis accident assumes that the primary to 
secondary leak rate, after the accident, is 1 gallon per minute with no 
more than 150 gallons per day in any one SG [steam generator], and that 
the reactor coolant activity levels of DOSE EQUIVALENT I-131 are at the 
TS [technical specification] values before the accident. The proposed 
change to the OTSG inspection program does not affect the design of the 
OTSGs, their method of operation, operational leakage limits, or 
primary coolant chemistry controls. The proposed change does not 
adversely impact any other previously evaluated design basis accident. 
In addition, the proposed changes do not affect the consequences of a 
Main Steam Line Break, rod ejection, or a reactor coolant pump locked 
rotor event, or other previously evaluated accident. Therefore, the 
proposed change does not affect the consequences of a Steam Generator 
Tube Rupture accident and the probability of such an accident is 
unchanged.
    2. The Proposed Change Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed license amendment does not affect the design of the 
OTSGs, their method of operation, or primary or secondary coolant 
chemistry controls. In addition, the proposed amendment does not impact 
any other plant system or component. The change modifies existing OTSG 
inspection requirements for 100% inspection intervals, but establishes 
inspection requirements that are considered equivalent based on 
properties and experience with improved materials. Therefore, the 
proposed change does not create the possibility of a new or different 
type of accident from any accident previously evaluated.
    3. The Proposed Change Does Not Involve a Significant Reduction in 
the Margin of Safety.
    The steam generator tubes in pressurized water reactors are an 
integral part of the reactor coolant pressure boundary and, as such, 
are relied upon to maintain the primary system's pressure and 
inventory. As part of the reactor coolant pressure boundary, the steam 
generator tubes are

[[Page 8285]]

unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat can 
be removed from the primary system. In addition, the steam generator 
tubes isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a steam 
generator is maintained by ensuring the integrity of its tubes. Steam 
generator tube integrity is a function of the design, environment, and 
the physical condition of the tube. The proposed change to the OTSG 
inspection program does not affect tube design or operating 
environment. The existing OTSG Program is maintained in this change. 
The repair criteria that are being removed are specific to the existing 
OTSGs and are not applicable to the replacement OTSGs. In the case of 
the roll repair that is being removed, it potentially leads to 
additional cracking over subsequent operating cycles due to tube cold 
working during the re-roll. If tube defects are detected that exceed 
limits in the new generators, then the tube will be removed from 
service. This is considered a more effective means for removing defects 
than repairs. For the above reasons, the margin of safety is not 
changed and overall plant safety will be enhanced by the proposed 
change to the ITS. Based upon the reasoning presented above and the 
previous discussion of the amendment request, the requested change does 
not involve a significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Thomas H. Boyce.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant, Citrus County, Florida

    Date of amendment request: November 6, 2008.
    Description of amendments request: The proposed change would revise 
the Crystal River Unit 3 (CR-3) Improved Technical Specifications 
Surveillance Requirements (SRs); SR 3.8.1.2, SR 3.8.1.6, and SR 
3.8.1.10 to restrict the voltage and frequency limits for all Emergency 
Diesel Generator (EDG) starts. The steady state voltage limits would be 
revised to be more restrictive (plus or minus 2 percent of the nominal 
voltage) to accurately reflect the appropriate calculation and the way 
the plant is operated and tested. The steady state frequency limits 
would be revised to be more restrictive (plus or minus 1 percent for 
all EDG starts) to ensure compliance with the plant design bases and 
the way the plant is operated. These changes would ensure that the EDGs 
are capable of supplying power, with the correct voltage and frequency, 
to the required electrical loads.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The LAR [license amendment request] proposes to provide more 
restrictive steady state voltage and frequency limits for the Emergency 
Diesel Generators (EDGs). The voltage band is going from a range of 
greater than or equal to 3933 V [volts] but less than or equal to 4400 
V, to greater than or equal to 4077 V but less than or equal to 4243 V. 
The proposed limits are +/-2% [percent] around the nominal safety-
related bus voltage of 4160 V. The Frequency Limits are going from a 2% 
tolerance band to a 1% tolerance band around the nominal frequency of 
60 Hz [hertz] (59.4 Hz to 60.6 Hz) for all starts of the EDGs.
    The EDGs are a safety-related system that functions to mitigate the 
impact of an accident with a concurrent loss of offsite power. A loss 
of offsite power is typically a significant contributor to postulated 
plant risk and, as such, onsite AC [alternating current] generators 
have to be maintained available and reliable in the event of a loss of 
offsite power event. The EDGs are not initiators for any analyzed 
accident, therefore; the probability for an accident that was 
previously evaluated is not increased by this change. The revised, 
voltage and frequency limits will ensure the EDGs will remain capable 
of performing their design function.
    The consequences of an accident refer to the impact on both plant 
personnel and the public from any radiological release associated with 
the accident. The EDG supports equipment that is supposed to preclude 
any radiological release. More restrictive voltage and frequency limits 
for the output of the EDG restores design margin, and provides 
assurance that the equipment supplied by the EDG will operate correctly 
and within the assumed timeframe to perform their mitigating functions.
    Until the proposed CR-3 ITS [Improved Technical Specifications] EDG 
voltage and frequency limits are approved by the NRC, administratively 
controlled limits have been established in accordance with NRC 
Administrative Letter 98-10 to ensure all EDG mitigation functions will 
be performed, per design, in the event of a loss of offsite power. 
These administrative limits have been determined as acceptable and have 
been incorporated into the surveillance test procedures under the 
provisions of 10 CFR 50.59. Periodic testing has been performed with 
acceptable results. Since EDGs are mitigating components and are not 
initiators for any analyzed accident, no increased probability of an 
accident can occur. Since administrative limits will ensure the EDGs 
will perform as designed, consequences will not be significantly 
affected.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Administrative voltage limits were established using verified 
design calculations and the guidance of NRC Administrative Letter 98-
10. These administrative limits will ensure the EDGs will perform as 
designed. No new configuration is established by this change. The 
administrative limits for the EDG frequency were determined to be 
sufficient to account for measurement and other uncertainties.
    The proposed amendment will place the administrative limits into 
the CR-3 ITS. The more restrictive voltage and frequency limits will 
provide additional assurance that the EDG can provide the necessary 
power to supply the required safety-related loads during an analyzed 
accident.
    The proposed ITS voltage and frequency limits restore the EDG 
capability to those analyzed by engineering calculation. No new 
configuration is established. Therefore, no new or different kind of 
accident from any previously evaluated can be created.
    3. Does not involve a significant reduction in a margin of safety.
    The LAR proposes to provide more restrictive steady state voltage 
and frequency limits for the EDGs. The change in the acceptance 
criteria for specific surveillance testing provides assurance that the 
EDGs will be capable of performing their design function. Previous test 
history has shown that the new limits are well within the

[[Page 8286]]

capability of the EDGs and are repeatable. The ``as-left'' settings for 
voltage and frequency will be adjusted such that they remain within a 
tight band and this ensures that the ``as-found'' settings will be in 
an acceptable tolerance band.
    The proposed ITS limits on voltage and frequency will ensure that 
the EDG will be able to perform all design functions assumed in the 
accident analyses. Administrative limits are in place to ensure these 
parameters remain within analyzed limits. As such, 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, NC 27602.
    NRC Branch Chief: Thomas H. Boyce.

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

    Date of amendment request: September 26, 2008.
    Description of amendment request: The amendments would revise the 
Technical Specifications to adopt Nuclear Regulatory Commission (NRC)-
approved Revision 3 to Technical Specification Task Force (TSTF) 
Improved Standard Technical Specification Change Traveler, TSTF-448, 
``Control Room Envelope Habitability.'' The proposed amendments include 
changes to the TS requirements related to control room envelope (CRE) 
habitability in TS 3/4.7.5, ``Control Room Emergency Ventilation System 
(CREVS),'' and TS Section 6.8, ``Administrative Controls--Procedures 
and Programs.'' In addition, the improvements to TSTF-448, Revision 3 
as recommended in TSTF-508, Revision 0, ``Revise Control Room Envelope 
Habitability Actions to Address Lessons Learned from TSTF-448 
Implementation,'' have been incorporated as appropriate.
    The NRC staff published a notice of opportunity for comment in the 
Federal Register on October 17, 2006 (71 FR 61075), on possible 
amendments adopting TSTF-448, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line-item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on January 17, 2007 (72 FR 2022). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated September 26, 2008.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility. The proposed change 
does not alter or prevent the ability of structures, systems, and 
components (SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change revises the TS for the CRE emergency 
ventilation system, which is a mitigation system designed to 
minimize unfiltered air leakage into the CRE and to filter the CRE 
atmosphere to protect the CRE occupants in the event of accidents 
previously analyzed. An important part of the CRE emergency 
ventilation system is the CRE boundary. The CRE emergency 
ventilation system is not an initiator or precursor to any accident 
previously evaluated.
    Therefore, the probability of any accident previously evaluated 
is not increased. Performing tests to verify the operability of the 
CRE boundary and implementing a program to assess and maintain CRE 
habitability ensure that the CRE emergency ventilation system is 
capable of adequately mitigating radiological consequences to CRE 
occupants during accident conditions, and that the CRE emergency 
ventilation system will perform as assumed in the consequence 
analyses of design basis accidents. Thus, the consequences of any 
accident previously evaluated are not increased. Therefore, the 
proposed 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 does not impact the accident analysis. The 
proposed change does not alter the required mitigation capability of 
the CRE emergency ventilation system, or its functioning during 
accident conditions as assumed in the licensing basis analyses of 
design basis accident radiological consequences to CRE occupants. No 
new or different accidents result from performing the new 
surveillance or following the new program. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a significant 
change in the methods governing normal plant operation. The proposed 
change does not alter any safety analysis assumptions and is 
consistent with current plant operating practice. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.

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

    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The proposed change does not affect safety 
analysis acceptance criteria. The proposed change will not result in 
plant operation in a configuration outside the design basis for an 
unacceptable period of time without compensatory measures. The 
proposed change does not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Thomas H. Boyce.

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

    Date of amendment request: January 5, 2009.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications (TS) requirements for mode change 
limitations in accordance with Revision 9 of Nuclear Regulatory 
Commission (NRC)-approved TS Task Force (TSTF) change TSTF-359, 
``Increase Flexibility in Mode Restraints.''
    In a Federal Register notice dated August 2, 2002 (67 FR 50475), 
the NRC staff issued a notice of opportunity to comment on a model 
safety evaluation and model no significant hazards consideration (NSHC) 
determination for proposed license amendments adopting TSTF-359 using 
the consolidated line item improvement process (CLIIP).
    In a Federal Register notice dated April 4, 2003 (68 FR 16579), the 
NRC staff issued a notice of availability of a model application for 
proposed license amendments adopting TSTF-359 using the CLIIP. The 
notice also included a revised model safety evaluation and a

[[Page 8287]]

model NSHC determination. In its application dated January 5, 2009, the 
licensee affirmed the applicability of the model NSHC determination 
which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC 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 allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO [Limiting Condition for Operation] 3.0.4, 
are no different than the consequences of an accident while entering 
and relying on the required actions while starting in a condition of 
applicability of the TS. Therefore, the consequences of an accident 
previously evaluated are not significantly affected by this change. 
The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
Therefore, 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 Previously Evaluated

    The proposed change does not involve the physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS Limiting 
Conditions for Operation (LCO). The risk associated with this 
allowance is managed by the imposition of required actions that must 
be performed within the prescribed completion times. The net effect 
of being in a TS condition on the margin of safety is not considered 
significant. The proposed change does not alter the required actions 
or completion times of the TS. The proposed change allows TS 
conditions to be entered, and the associated required actions and 
completion times to be used in new circumstances. This use is 
predicated upon the licensee's performance of a risk assessment and 
the management of plant risk. The change also eliminates current 
allowances for utilizing required actions and completion times in 
similar circumstances, without assessing and managing risk. The new 
change to the margin of safety is insignificant. Therefore, this 
change does not involve a significant reduction in a margin of 
safety.

    Based upon the reasoning presented above 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

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

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of amendment request: January 5, 2009.
    Description of amendment request: The proposed amendments would 
delete Section 2.F of the Facility Operating License (FOL) for Hope 
Creek Generating Station (Hope Creek) and Section 2.I of the FOL for 
Salem Nuclear Generating Station (Salem) Unit No. 2. The FOL sections 
being deleted require reporting of violations of the requirements in 
Section 2.C of the respective FOLs. The proposed amendments would also 
delete Technical Specification (TS) 6.9.3 for Hope Creek, Salem Unit 
No. 1 and Salem Unit No. 2. These TSs contain a reporting requirement 
that is duplicative of Nuclear Regulatory Commission (NRC) regulations.
    The NRC staff issued a ``Notice of Opportunity to Comment on Model 
Safety Evaluation on Elimination of Typical License Condition Requiring 
Reporting of Violations of Section 2.C of Operating Licensing Using the 
Consolidated Line Item Improvement Process,'' in the Federal Register 
on August 29, 2005 (70 FR 51098). The notice included a model safety 
evaluation (SE) and a model no significant hazards consideration (NSHC) 
determination. On November 4, 2005, the NRC staff issued a notice in 
the Federal Register (70 FR 67202) announcing that the model SE and 
model NSHC determination may be referenced in plant-specific 
applications to adopt the changes. In its application dated January 5, 
2009, the licensee affirmed the applicability of the model NSHC 
determination which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change involves the deletion of a reporting 
requirement. The change does not affect plant equipment or operating 
practices and therefore does not significantly increase the probability 
or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change is administrative in that it deletes a 
reporting requirement. The change does not add new plant equipment, 
change existing plant equipment, or affect the operating practices of 
the facility. Therefore, the change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change deletes a reporting requirement. The change 
does not affect plant equipment or operating practices and therefore 
does not involve a significant reduction in a margin of safety.
    Based on the above, the NRC staff proposes that the change presents 
no significant hazards consideration under the standards set forth in 
10 CFR 50.92(c).
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

[[Page 8288]]

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

    Date of amendment request: October 30 and November 20, 2008 (TS-
463-T).
    Description of amendment request: The BFN requests adoption of an 
approved change to the Standard Technical Specifications (TSs) for 
General Electric Plants (NUREG-1433, BWR/4) and plant-specific TSs, 
that allows: (1) Revising the frequency of Surveillance Requirement 
(SR) 3.1.3.2, notch testing of fully withdrawn control rod, from ``7 
days after the control rod is withdrawn and THERMAL POWER is greater 
than the low-power set point (LPSP) of rod worth minimizer (RWM)'' to 
``31 days after the control rod is withdrawn and THERMAL POWER is 
greater than the LPSP of the RWM,'' (2) adding the word ``fully'' to 
Limiting Condition for Operation LCO 3.3.1.2, Required Action E.2 to 
clarify the requirement to fully insert all insertable control rods in 
core cells containing one or more fuel assemblies when the associated 
source range monitor instrument is inoperable, and (3) revising Example 
1.4-3 in Section 1.4 ``Frequency'' to clarify that the 1.25 
surveillance test interval extension in SR 3.0.2 is applicable to time 
periods discussed in NOTES in the ``SURVEILLANCE'' column in addition 
to the time periods in the ``FREQUENCY'' column.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No
    This change does not affect either the design or operation of the 
Control Rod Drive Mechanism (CRDM). The affected surveillance and 
Required Action is not considered to be an initiator of any analyzed 
event. Revising the frequency for notch testing fully withdrawn control 
rods will not affect the ability of the control rods to shutdown the 
reactor if required. Given the extremely reliable nature of the CRDM, 
as demonstrated through industry operating experience, the proposed 
monthly notch testing of all withdrawn control rods continues to 
provide a high level of confidence in control rod operability. Hence, 
the overall intent of the notch testing surveillances, which is to 
detect either random stuck control rods or identify generic concerns 
affecting control rod operability, is not significantly affected by the 
proposed change. Requiring control rods to be fully inserted when the 
associated SRM is inoperable is consistent with other similar 
requirements and will increase the shutdown margin. The clarification 
of Example 1.4-3 in Section 1.4 ``Frequency'' is an editorial change 
made to provide consistency with other TSTF-475, Rev. 1 discussions in 
Section 1.4. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No
    Revising the frequency for notch testing fully withdrawn control 
rods does not involve physical modification to the plant and does not 
introduce a new mode of operation. Requiring control rods to be fully 
inserted will make this action consistent with other similar actions. 
The clarification of Example 1.4-3 in Section 1.4 ``Frequency'' is an 
editorial change made to provide consistency with other discussions in 
Section 1.4. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No
    The CRDs and CRDMs are extremely reliable systems and, as such, 
reducing the number of control rod notch tests will not significantly 
impact the likelihood of detecting a stuck control rod. If a stuck 
control rod is detected, existing action requirements will ensure 
prompt action is taken to ensure there is not a generic problem. Other 
surveillances are routinely performed to ensure that the performance of 
the control rods in the event of a DBA [design-basis accident] or 
transient meets the assumptions used in the safety analyses. As such, 
potential effects of reducing the number of notch tests are far 
outweighed by the benefit of reducing undue burden on reactor operators 
and reducing the potential for mispositioning events which accompanies 
any control rod manipulation. Requiring control rods to be fully 
inserted instead of partially inserted when the associated SRM is 
inoperable will increase the margin of safety. The clarification of 
Example 1.4-3 in Section 1.4 ``Frequency'' is an editorial change made 
to provide consistency with other discussions in Section 1.4. 
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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Thomas H. Boyce.

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

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

Duke Power Company LLC, Docket Nos. 50-414, Catawba Nuclear Station, 
Unit 2, York County, South Carolina

    Date of application for amendments: January 20, 2009.
    Brief description of amendments: The proposed amendment would allow 
a one-time limited duration extension of the Technical Specification 
(TS) Surveillance (SR) 3.3.1.4 frequency. SR 3.3.1.4 is a Trip 
Actuating Device Operational Test (TADOT) of the reactor trip breakers 
(RTBs) and reactor trip bypass breakers.
    Date of publication of individual notice in Federal Register: 
January 28, 2009 (74 FR 4986).
    Expiration date of individual notice: 30 days February 27, 2009; 60 
days March 30, 2009.

[[Page 8289]]

Notice of Issuance of Amendments to Facility Operating Licenses

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

Carolina Power & Light Company, et. al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: January 4, 2008.
    Brief description of amendment: The amendment establishes more 
effective and appropriate action, surveillance, and administrative 
requirements related to ensuring the habitability of the control room 
envelope in accordance with the NRC-approved Technical Specification 
Task Force (TSTF) Standard Technical Specification change traveler 
TSTF-448, Revision 3, ``Control Room Habitability.'' This technical 
specification improvement was initially made available in the Federal 
Registerby the NRC on January 17, 2007 (72 FR 2022).
    Date of issuance: January 29, 2009.
    Effective date: Effective as of the date of issuance and shall be 
implemented within 180 days.
    Amendment No: 128.
    Renewed Facility Operating License No. NPF-63: The amendment 
revises the Technical Specifications and Facility Operating License.
    Date of initial notice in Federal Register: May 20, 2008 (73 FR 
29161).
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated January 29, 2009.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, et. al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: April 3, 2008, as supplemented 
by letters dated December 9, 2008, and January 9, 2009.
    Brief description of amendment: The amendment revises Technical 
Specification Section 5.6.3.b to allow a reconfiguration of the fuel 
racks in Spent Fuel Pool (SFP) C and allow the use of Metamic as an 
alternate neutron poison material in the new storage racks for SFP C 
and D. The amendment: (1) Revises the rack configuration in SFP C to 
allow the substitution of four previously approved (13 x 13 cell) 
Boiling Water Reactor racks with an equal number of (9 x 9 cell) 
Pressurized Water Reactor racks, and (2) authorizes the use of Metamic 
as an alternate spent fuel rack poison material.
    Date of issuance: January 29, 2009.
    Effective date: Effective as of the date of issuance and shall be 
implemented within 60 days.
    Amendment No: 129.
    Renewed Facility Operating License No. NPF-63: The amendment 
revises the Technical Specifications and Facility Operating License.
    Date of initial notice in Federal Register: June 10, 2008 (73 FR 
32744). The supplemental letters provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated January 29, 2009.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 22, 2008.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) requirements related to control room 
envelope habitability in accordance with TS Task Force (TSTF) traveler 
TSTF-448, ``Control Room Habitability,'' Revision 3. This TS 
improvement was made available by the Commission on January 17, 2007 
(72 FR 2022) as part of the consolidated line item improvement process 
(CLIIP).
    Date of issuance: January 30, 2009.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 249 and 229.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: March 25, 2008 (73 FR 
15784).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 30, 2009.
    No significant hazards consideration comments received: No

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

    Date of application for amendment: September 22, 2008.
    Brief description of amendment: The amendment revised the Technical 
Specification (TS) to change requirements related to Battery Systems 
specified in TS Section 3.10 resulting in

[[Page 8290]]

removing the Limiting Condition for Operation pertaining to 345 kV 
switchyard batteries, chargers and associated direct current 
distribution panel.
    Date of Issuance: February 11, 2009.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 234.
    Facility Operating License No. DPR-28: Amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: November 18, 2008 (73 
FR 68454).
    The Commission's related evaluation of t