[Federal Register Volume 73, Number 204 (Tuesday, October 21, 2008)]
[Notices]
[Pages 62560-62574]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-24896]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses; Involving No Significant Hazards Considerations

I. Background

    Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from September 25, 2008 to October 8, 2008. The 
last biweekly notice was published on October 7, 2008 (73 FR 58669).

[[Page 62561]]

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D44, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
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. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, 
person(s) may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
via electronic submission through the NRC E-Filing system for a hearing 
and a petition for leave to intervene. Requests for a hearing and a 
petition for leave to intervene shall be filed in accordance with the 
Commission's ``Rules of Practice for Domestic Licensing Proceedings'' 
in 10 CFR Part 2. Interested person(s) should consult a current copy of 
10 CFR 2.309, which is available at the Commission's PDR, located at 
One White Flint North, Public File Area 01F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland. Publicly available records will be 
accessible from the Agencywide Documents Access and Management System's 
(ADAMS) Public Electronic Reading Room on the Internet at the NRC Web 
site, 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 
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 supports the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for hearing or a petition for leave to intervene must be 
filed in

[[Page 62562]]

accordance with the NRC E-Filing rule, which the NRC promulgated on 
August 28, 2007 (72 FR 49139). The E-Filing process requires 
participants to submit and serve documents over the Internet or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek a waiver in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
[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 ViewerTM 
to access the Electronic Information Exchange (EIE), a component of the 
E-Filing system. The Workplace Forms ViewerTM 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 technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, in accordance 
with 10 CFR 2.302(g), with their initial paper filing requesting 
authorization to continue to submit documents in paper format. Such 
filings must be submitted by: (1) First class mail addressed to the 
Office of the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff; or (2) courier, express mail, or expedited 
delivery service to the Office of the Secretary, Sixteenth Floor, One 
White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, 
Attention: Rulemaking and Adjudications Staff. Participants filing a 
document in this manner are responsible for serving the document on all 
other participants. Filing is considered complete by first-class mail 
as of the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board that the petition 
and/or request should be granted and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later 
than 11:59 p.m. Eastern Time on the due date.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
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. 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].

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

    Date of amendments request: August 27, 2008.
    Description of amendments request: The amendment would change the 
containment buffering agent from trisodium phosphate (TSP) to sodium 
tetraborate in order to minimize the potential for sump screen blockage 
due to potential adverse chemical interactions between TSP and certain 
insulation materials used in containment under post loss-of-coolant 
accident conditions. This amendment is one of the remaining 
modifications required for Calvert Cliffs Nuclear Power Plant, Unit 
Nos. 1 and 2 to achieve full compliance with the requirements of 
Generic Letter 2004-02, ``Potential Impact of Debris Blockage on 
Emergency Recirculation During Design Basis Accidents at Pressurized-
Water Reactors'' (Agencywide Documents Access and Management System 
(ADAMS) Accession Number ML042360586).
    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?

[[Page 62563]]

    Response-No.
    The proposed amendment does not involve a significant increase 
in the probability of an accident previously evaluated because the 
containment buffering agent is not an initiator of any analyzed 
accident. The proposed change does not impact any failure modes that 
could lead to an accident. The proposed amendment does not involve a 
significant increase in the consequences of an accident previously 
evaluated. The buffering agent in Containment is designed to buffer 
the acids expected to be produced after a loss-of-coolant accident 
(LOCA) and is credited in the radiological analysis for iodine 
retention. Utilizing the required quantity of sodium tetraborate 
decahydrate (STB) as a buffering agent ensures the post-LOCA 
containment sump mixture will have a pH >= 7.0. The proposed change 
of replacing trisodium phosphate (TSP) with STB results in the 
radiological consequences remaining within the limits of 10 CFR 
50.67. There is no dose change with the pH >= 7.0.
    Therefore, operation of the facility in accordance with the 
proposed amendment would 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 amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The STB is a passive component that is proposed to be 
used as a buffering agent to increase the pH of the initially acidic 
post-LOCA containment water to a more neutral pH. Changing the 
proposed buffering agent from TSP to STB does not constitute an 
accident initiator or create a new or different kind of accident 
than previously analyzed. The proposed amendment does not involve 
operation of any required systems, structures, or components in a 
manner or configuration different from those previously recognized 
or evaluated. No new failure mechanisms will be introduced by the 
changes being requested. 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 change involve a significant reduction in a 
margin of safety?
    Response-No.
    The proposed amendment does not involve a significant reduction 
in a margin of safety. The proposed amendment of changing the 
buffering agent from TSP to STB results in equivalent control of 
maintaining sump pH at >= 7.0, thereby controlling containment 
atmosphere iodine and ensuring the radiological consequences of a 
LOCA are within regulatory limits. The change of buffering agent 
from TSP to STB also reduces the amount of calcium phosphate 
precipitate generated thereby reducing the overall amount of 
precipitate that may be formed in a postulated LOCA. The buffer 
change would minimize the potential chemical effects and should 
enhance the ability of the Emergency Core Cooling System to perform 
the post-LOCA mitigating functions.
    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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Generation Group LLC, 750 East Pratt Street, 
17th Floor, Baltimore, MD 21202.
    NRC Branch Chief: Mark G. Kowal.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: July 21, 2008.
    Description of amendment request: The amendment proposes a change 
to the Arkansas Nuclear One, Unit 1 (ANO-1) Technical Specifications 
(TSs) to support adoption of Technical Specification Task Force (TSTF) 
359, ``Increased Flexibility in Mode Restraints.'' The NRC approved 
adoption of TSTF-359 for ANO-1 in TS Amendment 232. The overall intent 
of TSTF-359 was to eliminate exceptions to Limiting Condition for 
Operation (LCO) 3.0.4 within individual specifications and provide 
requirements within LCO 3.0.4 to control mode changes when TS-required 
equipment is inoperable. Following implementation of TS Amendment 232, 
Entergy discovered that one of the marked-up TS pages which contained 
an LCO 3.0.4 exception was not provided to the NRC for review in the 
original submittal.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 2, 2002 (67 FR 50475), as part of the 
Consolidated Line Item Improvement Process (CLIIP), on possible 
amendments to revise the plant-specific TS to modify requirements for 
model change limitations in LCO 3.0.4 and SR 3.0.4.
    The NRC staff subsequently issued a notice of availability of the 
models for Safety Evaluation and No Significant Hazards Consideration 
Determination for referencing in license amendment applications in the 
Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed 
the applicability of the CLIIP, including the model No Significant 
Hazards Consideration Determination, in its application dated October 
22, 2007.
    The proposed TS changes are consistent with NRC-approved Industry 
TSTF STS change, TSTF-359, Revision 8, as modified by 68 FR 16579. 
TSTF-359, Revision 8, was subsequently revised to incorporate the 
modifications discussed in the April 4, 2003, Federal Register notice 
and other minor changes. TSTF-359, Revision 9, was subsequently 
submitted to the NRC on April 28, 2003, and was approved by the NRC on 
May 9, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the NRC staff analysis 
of the issue of no significant hazards consideration is presented 
below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Response: No.
    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 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.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Response: No.
    The proposed change does not involve a 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. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    Response: No.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition

[[Page 62564]]

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 net 
change to the margin of safety is insignificant. Therefore, this 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff proposes to determine that the request for amendment 
involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Michael T. Markley.

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: December 13, 2007.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) Section 4.3.1, 
``Criticality,'' to add a new requirement to use a blocking device in 
spent fuel storage rack cells that cannot maintain the effective 
neutron multiplication factor, Keff, requirements specified 
in TS Section 4.3.1.1.a. In addition, the proposed change revises TS 
Section 4.3.3 to reflect that the LaSalle County Station, Unit 2 spent 
fuel storage capacity is limited to no more than a combination of 4078 
fuel assemblies and blocking devices.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change adds an additional requirement to the TS to 
ensure that the effective neutron multiplication factor 
Keff, is less than or equal to 0.95, if fully flooded 
with borated water. The additional requirement is to insert a 
blocking device into unusable storage rack cell locations. Since the 
proposed change pertains only to the spent fuel pool (SFP), only 
those accidents that are related to movement and storage of fuel 
assemblies in the SFP could be potentially affected by the proposed 
change.
    The probability that a misplaced fuel assembly would result in 
an inadvertent criticality is unchanged since the process and 
procedural controls governing fuel cell movement in the SFP will not 
be changed. The current criticality analysis for the LSCS Unit 2 SFP 
credits the neutron absorbing properties of the Boraflex neutron 
poison material in the spent fuel storage racks. The current 
analysis demonstrates: (1) Adequate margin to criticality for all 
spent fuel storage cells, (2) adequate margin for fuel assemblies 
inadvertently placed into locations adjacent to the spent fuel 
racks, and (3) adequate margin for assemblies accidentally dropped 
onto the spent fuel racks. The dose consequences of the most 
limiting drop of a fuel assembly in the spent fuel pool is limited 
by the number of the fuel rods damaged and other engineered features 
unaffected by the proposed change, including the fuel design, fuel 
decay time, water level in the spent fuel pool, water temperature of 
the spent fuel pool, and the engineering features of the Reactor 
Building Ventilation System.
    The revised analysis does not result in a significant increase 
in the probability of an accident previously analyzed. The revised 
analysis takes no credit for the Boraflex material. The use of a 
blocking device prevents an inadvertent action to insert a spent 
fuel assembly, and prevents an assembly that is accidentally dropped 
to penetrate into the empty spent fuel cell. In addition to this 
blocking device, administrative controls will be implemented to 
prevent insertion of a bundle into a cell that is blocked. The 
probability that a fuel assembly would be inadvertently placed into 
a location adjacent to the racks is unchanged, and the probability 
that a fuel assembly would be dropped is unchanged by the revised 
analysis. These events involve failures of administrative controls, 
human performance, and equipment failures that are unaffected by the 
presence or absence of Boraflex and the blocking devices.
    The revised analysis does not result in a significant increase 
in the consequence of an accident previously analyzed. The revised 
analysis demonstrates adequate margin to criticality for unblocked 
cells in the LSCS Unit 2 SFP, adequate margin for assemblies 
inadvertently placed into locations adjacent to the spent fuel 
racks, and adequate margin for assemblies accidentally dropped onto 
the spent fuel racks. Placing a spent fuel assembly into a location 
containing a blocking device is not a credible event since there are 
diverse and redundant administrative and physical barriers to 
prevent that.
    The revised analysis does not affect the consequences of a 
dropped fuel assembly. The consequences of dropping a fuel assembly 
onto any other fuel assembly or other structure, other than a 
blocking device, are unaffected by the change. The consequences of 
dropping a fuel assembly onto a blocking device are bounded by the 
event of dropping an assembly onto another assembly, both for 
criticality and for radiological consequences. For criticality, the 
blocking device prevents the dropped assembly from entering the 
blocked cell. For radiological consequences, the number of rods 
damaged when a fuel assembly is accidentally dropped onto a blocking 
device is bounded the by the number of rods damaged by an assembly 
dropped onto another assembly. The change does not affect the 
effectiveness of the other engineered design features to limit the 
offsite dose consequences of the limiting fuel assembly drop 
accident.
    The proposed change to clarify that the capacity of the Unit 2 
SFP is limited to no more than a combination of 4078 fuel assemblies 
and blocking devices does not affect the probability or consequences 
of an accident previously analyzed because no physical modifications 
to the storage racks are proposed. The proposed change will reduce 
the number of allowable fuel assembly storage locations.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Onsite storage of spent fuel assemblies in the SFP is a normal 
activity for which LSCS has been designed and licensed. As part of 
assuring that this normal activity can be performed without 
endangering public health and safety, the ability to safely 
accommodate different possible accidents in the SFP, such as 
dropping a fuel assembly or misloading a fuel assembly, have been 
analyzed. The proposed fuel storage configuration does not change 
the methods of fuel movement or fuel storage. No structural or 
mechanical change to the racks or fuel handling equipment is being 
proposed. The proposed change allows for partial use of storage rack 
locations that have been determined unusable based on the existing 
criticality analysis.
    The blocking devices are passive devices. These devices, when 
inside a spent fuel storage rack cell, perform the same function of 
a spent fuel assembly in that cell. These devices do not add any 
limiting structural loads or affect the removal of decay heat from 
the other assemblies. The devices are resistant to corrosion and 
will maintain their structural integrity over the life of the plant. 
These devices are not under any structural load during normal 
operations. They are only challenged by an accidental fuel assembly 
drop. The existing fuel handling accident, which assumes the drop of 
a fuel bundle, bounds the drop of a blocking device.
    This change does not create the possibility of a misloaded 
assembly into a blocked cell. Placing a spent fuel assembly into a 
location containing a blocking device is not a credible event since 
there are diverse and redundant administrative and physical barriers 
to prevent that.

[[Page 62565]]

    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.
    LSCS TS 4.3.1 .1 requires the spent fuel storage racks to 
maintain the effective neutron multiplication factor, 
Keff, less than or equal to 0.95 when fully flooded with 
unborated water, which includes an allowance for uncertainties. 
Therefore, for criticality, the required safety margin is 5% 
including a conservative margin to account for engineering 
uncertainties.
    The proposed change adds a requirement to use a blocking device 
to ensure that Keff continues to be less than or equal to 
0.95; thus, the required safety margin of 5% is preserved. The 
proposed change also clarifies that the capacity of the Unit 2 SFP 
is limited to no more than a combination of 4078 fuel assemblies and 
blocking devices. This clarification does not impact the required 
safety margin of 5%.
    The current analysis assumes an infinite array of fuel with all 
fuel at the peak reactivity (i.e., the highest combination of 
initial enrichment, gadolinium, and fuel burnup that maximizes the 
reactivity of the fuel). The revised analysis demonstrates the same 
margin to criticality of 5%, including a conservative margin to 
account for engineering uncertainties, is maintained assuming an 
infinite array of fuel with all fuel at the peak reactivity. In 
addition, the margin of safety for radiological consequences of a 
dropped fuel assembly are unchanged because the event involving a 
dropped fuel assembly onto a blocking device is bounded by the 
consequences of a dropped fuel assembly onto another fuel assembly.
    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 Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: May 30, 2008, as supplemented on July 17 
and September 10, 2008.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) Table 3.3.8.1-1, ``Loss of Power 
Instrumentation,'' specifically to change the maximum allowable voltage 
of the 4.16-kV Emergency Bus Undervoltage function from less-than-or-
equal to 3899 V to less-than-or-equal-to 3822 V.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed TS change to the maximum allowable voltage for the 
4160 volt Emergency Bus Undervoltage relays affects when an 
Emergency Bus that is experiencing degraded voltage will disconnect 
from offsite power and transfer to an emergency diesel generator. 
While the maximum allowed voltage that initiates this action will be 
lowered, the function remains the same. The maximum allowed voltage 
has been analyzed to ensure spurious trips will be avoided. The 
proposed change will not affect any accident initiators or 
precursors. As a result, the probability of any accident previously 
evaluated is not significantly increased.
    The consequences of any accident previously evaluated are not 
increased since the 4160 volt Emergency Bus Undervoltage relays will 
continue to meet their required function to transfer the 4160 volt 
Emergency Buses to the emergency diesel generators in the event of a 
degraded voltage condition on the offsite power supply. This 
transfer will ensure that the electrical equipment is capable of 
performing its function to meet the requirements of the accident 
analyses.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The proposed TS change to the maximum allowable voltage for 
the 4160 volt Emergency Bus Undervoltage relays does not affect 
existing or introduce any new accident precursors or modes of 
operation. The relays will continue to detect undervoltage 
conditions and transfer the Emergency Buses to the emergency diesel 
generators at a voltage adequate to ensure proper safety equipment 
performance and to prevent equipment damage. The function of the 
relays remains the same.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed TS change to the maximum allowable voltage for the 
4160 volt Emergency Bus Undervoltage relays will allow all safety 
loads to have sufficient voltage to perform their intended safety 
functions while ensuring spurious trips are avoided. Thus, the 
results of the accident analyses will not be affected as the input 
assumptions are protected.
    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: Mr. R. E. Helfrich, Florida Power & Light 
Company, P. O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Lois M. James.

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

    Date of amendment request: August 19, 2008.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) requirements for mode change 
limitations in accordance with NRC-approved TS Task Force (TSTF) 
traveler TSTF-359, Revision 9, ``Increase Flexibility in MODE 
Restraints,'' and revise TS Section 1.4, ``Frequency,'' in accordance 
with NRC-approved traveler TSTF-485, Revision 0, ``Correct Example 1.4-
1.''
    The NRC staff issued a ``Notice of Availability of Model 
Application Concerning Technical Specification Improvement To Modify 
Requirements Regarding Mode Change Limitations Using the Consolidated 
Line Item Improvement Process'' in the Federal Register on April 4, 
2003 (68 FR 16579). The notice referenced a model safety evaluation and 
a model no significant hazards consideration (NSHC) determination 
published in the Federal Register on August 2, 2002 (67 FR 50475). In 
its application dated August 19, 2008, the licensee affirmed the 
applicability of the model NSHC determination which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee regarding TSTF-359 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

[[Page 62566]]

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 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 a 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 net 
change to the margin of safety is insignificant. Therefore, this 
change does not involve a significant reduction in a margin of 
safety.

    In its application dated August 19, 2008, the licensee also 
affirmed the applicability of the NSHC approved by the NRC in TSTF-485, 
which is presented below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises Section 1.4, Frequency, Example 1.4-
1, to be consistent with Surveillance Requirement (SR) 3.0.4 and 
Limiting Condition for Operation (LCO) 3.0.4. This change is 
considered administrative in that it modifies the example to 
demonstrate the proper application of SR 3.0.4 and LCO 3.0.4. The 
requirements of SR 3.0.4 and LCO 3.0.4 are clear and are clearly 
explained in the associated Bases. As a result, modifying the 
example will not result in a change in usage of the Technical 
Specifications (TS). The proposed change does not adversely affect 
accident initiators or precursors, 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, or radiological release assumptions used in 
evaluating the radiological consequences of an accident previously 
evaluated. Therefore, this change is considered administrative and 
will have no effect on the probability or consequences of any 
accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. 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. In addition, the change does not impose any new or 
different requirements or eliminate any existing requirements. The 
change does not alter assumptions made in the safety analysis. The 
proposed change is consistent with the safety analysis assumptions 
and current plant operating practice.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change is administrative and will have no effect on 
the application of the Technical Specification requirements. 
Therefore, the margin of safety provided by the Technical 
Specification requirements is unchanged.
    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 upon this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendment involves NSHC.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: August 15, 2008.
    Description of amendment request: The proposed amendment would 
revise 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 five (5) years. The proposed amendment 
would allow the next ILRT for NMP1 to be performed within 15 years from 
the last ILRT as opposed to the current 10-year interval.
    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 involves a one-time extension of the primary 
containment ILRT interval from 10 to 15 years. The proposed change 
does not involve a physical change to the plant or a change in the 
manner in which the plant is operated or controlled. The primary 
containment function is to provide an essentially leak tight barrier 
against the uncontrolled release of radioactivity to the environment 
for postulated accidents. As such, the containment itself and the 
testing requirements to periodically demonstrate the integrity of 
the containment exist to ensure the plant's ability to mitigate the 
consequences of an accident, and do not involve any accident 
precursors or initiators. Therefore, the probability of occurrence 
of an accident previously evaluated is not significantly increased 
by the proposed change.
    Continued containment integrity is assured by the established 
programs for local leak rate testing and inservice/containment

[[Page 62567]]

inspections, which are unaffected by the proposed change. As 
documented in NUREG-1493, ``Performance-Based Containment Leak-Test 
Program,'' dated September 1995, industry experience has shown that 
local leak rate tests (Type B and C) have identified the vast 
majority of containment leakage paths, and that ILRTs detect only a 
small fraction of containment leakage pathways.
    The potential consequences of the proposed change have been 
quantified by analyzing the changes in risk that would result from 
extending the ILRT interval from 10 years to 15 years. The increase 
in risk in terms of person-rem per year within 50 miles resulting 
from design basis accidents was estimated to be of a magnitude that 
NUREG-1493 indicates is imperceptible. NMPNS has also analyzed the 
increase in risk in terms of the frequency of large early releases 
from accidents. The increase in the large early release frequency 
resulting from the proposed change was determined to be within the 
guidelines published in NRC Regulatory Guide 1.174. Additionally, 
the proposed change maintains defense-in-depth by preserving a 
reasonable balance among prevention of core damage, prevention of 
containment failure, and consequence mitigation. NMPNS has 
determined that the increase in conditional containment failure 
probability due to the proposed change would be insignificant. 
Therefore, it is concluded that the proposed one-time extension of 
the primary containment ILRT interval from 10 years to 15 years does 
not significantly increase the consequences of an accident 
previously evaluated.
    Based on the above discussion, it is concluded that the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change involves a one-time extension of the primary 
containment ILRT interval. The containment and the testing 
requirements to periodically demonstrate the integrity of the 
containment exist to ensure the plant's ability to mitigate the 
consequences of an accident, and do not involve any accident 
precursors or initiators. The proposed change does not involve a 
physical change to the plant (i.e., no new or different type of 
equipment will be installed) or a change in the manner in which the 
plant is operated or controlled.
    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 proposed one-time extension of the primary containment ILRT 
interval does not alter the manner in which safety limits, limiting 
safety system setpoints, or limiting conditions for operation are 
determined. The specific requirements and conditions of the 10 CFR 
[Part] 50 Appendix J Testing Program Plan, as defined in the TS, 
exist to ensure that the degree of primary containment structural 
integrity and leak-tightness that is considered in the plant safety 
analyses is maintained. The overall containment leakage rate limit 
specified by the TS is maintained, and Type B and C containment 
leakage tests will continue to be performed at the frequency 
currently required by the TS.
    NMP1 and industry experience strongly support the conclusion 
that Type B and C testing detects a large percentage of containment 
leakage paths and that the percentage of containment leakage paths 
that are detected only by the ILRT is small. Containment inspections 
performed in accordance with other plant programs serve to provide a 
high degree of assurance that the containment will not degrade in a 
manner that is detectable only by an ILRT. Additionally, the on-line 
containment monitoring capability that is inherent to inerted 
boiling[-]water reactor containments allows for the detection of 
gross containment leakage that may develop during power operation. 
This combination of factors ensures that the margin of safety that 
is inherent in plant safety analyses is maintained. Furthermore, a 
risk assessment using the current NMP1 Probabilistic Risk Assessment 
interval events model concluded that extending the ILRT test 
interval from 10 to 15 years results in a very small change to the 
NMP1 risk profile.
    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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment request: August 14, 2008.
    Description of amendment request: The proposed amendment would (1) 
revise the NMP2 Technical Specification (TS) Surveillance Requirement 
(SR) frequency in TS 3.1.3, ``Control Rod Operability,'' and (2) revise 
Example 1.4-3 in TS Section 1.4, ``Frequency,'' to clarify the 
applicability of the 1.25 surveillance test interval extension. The 
proposed changes are consistent with Nuclear Regulatory Commission 
(NRC)-approved Revision 1 to TS Task Force (TSTF) Change Traveler, 
TSTF-475, ``Control Rod Notch Testing Frequency and SRM [Source Range 
Monitor] Insert Control Rod Action.'' The availability of this TS 
improvement was announced in the Federal Register on November 13, 2007 
(72 FR 63943) as part of the consolidated line item improvement 
process. The licensee affirmed the applicability of the model no 
significant hazards consideration determination in its application.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which 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 generically implements TSTF-475, Revision 1, 
``Control Rod Notch Testing Frequency and SRM Insert Control Rod 
Action.'' TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and 
NUREG-1434 (BWR/6) STS. The changes: (1) Revise TS testing frequency 
for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod 
OPERABILITY,'' (2) clarify the requirement to fully insert all 
insertable control rods for the limiting condition for operation 
(LCO) in TS 3.3.1.2, Required Action E.2, ``Source Range Monitoring 
Instrumentation'' (NUREG-1434 only), and (3) revise Example 1.4-3 in 
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25 
surveillance test interval extension. The consequences of an 
accident after adopting TSTF-475, Revision 1 are no different than 
the consequences of an accident prior to adoption. 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 Accident Previously 
Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
proposed change 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 analyzed. Thus, 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 [a] Margin of Safety

    TSTF-475, Revision 1 will: (1) [revise the TS SR 3.1.3.2 
frequency in TS 3.1.3, ``Control Rod OPERABILITY,'' (2) clarify the 
requirement to fully insert all insertable control rods for the 
limiting condition for operation (LCO) in TS 3.3.1.2, ``Source Range 
Monitoring Instrumentation,'' and (3)] revise

[[Page 62568]]

Example 1.4-3 in Section 1.4 ``Frequency'' to clarify the 
applicability of the 1.25 surveillance test interval extension. [The 
GE Nuclear Energy Report, ``CRD Notching Surveillance Testing for 
Limerick Generating Station,'' dated November 2006, concludes that 
extending the control rod notch test interval from weekly to monthly 
is not expected to impact the reliability of the scram system and 
that the analysis supports the decision to change the surveillance 
frequency.] Therefore, the proposed changes in TSTF-475, Revision 1 
are acceptable and do not involve a significant reduction in a 
margin of safety.

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

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

    Date of amendment request: August 18, 2008.
    Description of amendment request: The proposed amendment would 
revise the NMP1 Technical Specification (TS) Section 3/4.1.1, ``Control 
Rod System,'' to increase the Surveillance Requirement (SR) frequency 
associated with control rod exercising. The proposed change would 
revise the required SR frequency from once each week to once every 31 
days. The proposed change is consistent with Nuclear Regulatory 
Commission (NRC)-approved Revision 1 to TS Task Force (TSTF) Change 
Traveler, TSTF-475, ``Control Rod Notch Testing Frequency and SRM 
[Source Range Monitor] Insert Control Rod Action,'' and NUREG-1433, 
``Standard Technical Specifications General Electric Plants, BWR/4,'' 
Revision 3.1. The availability of the TS improvement was announced in 
the Federal Register on November 13, 2007 (72 FR 63943) as part of the 
consolidated line item improvement process. The licensee affirmed the 
applicability of the model no significant hazards consideration 
determination in its application.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which 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 generically implements TSTF-475, Revision 1, 
``Control Rod Notch Testing Frequency and SRM Insert Control Rod 
Action.'' TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and 
NUREG-1434 (BWR/6) STS. The changes: (1) revise TS testing frequency 
for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod 
OPERABILITY,'' (2) clarify the requirement to fully insert all 
insertable control rods for the limiting condition for operation 
(LCO) in TS 3.3.1.2, Required Action E.2, ``Source Range Monitoring 
Instrumentation'' (NUREG-1434 only), and (3) revise Example 1.4-3 in 
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25 
surveillance test interval extension. The consequences of an 
accident after adopting TSTF-475, Revision 1 are no different than 
the consequences of an accident prior to adoption. 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 Accident Previously 
Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
proposed change 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 analyzed. Thus, 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 [a] Margin of Safety

    TSTF-475, Revision 1 will: (1) [revise the TS SR 3.1.3.2 
frequency in TS 3.1.3, ``Control Rod OPERABILITY,'' (2) clarify the 
requirement to fully insert all insertable control rods for the 
limiting condition for operation (LCO) in TS 3.3.1.2, ``Source Range 
Monitoring Instrumentation,'' and (3)] revise Example 1.4-3 in 
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25 
surveillance test interval extension. [The GE Nuclear Energy Report, 
``CRD Notching Surveillance Testing for Limerick Generating 
Station,'' dated November 2006, concludes that extending the control 
rod notch test interval from weekly to monthly is not expected to 
impact the reliability of the scram system and that the analysis 
supports the decision to change the surveillance frequency.] 
Therefore, the proposed changes in TSTF-475, Revision 1 are 
acceptable and do not involve a significant reduction in a margin of 
safety.

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

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment request: July 11, 2008.
    Description of amendment request: The proposed amendments would 
establish Conditions, Required Actions, and Completion Times in the 
Prairie Island Nuclear Generating Plant, Units 1 and 2, Technical 
Specifications (TSs) for the condition where one steam supply to the 
turbine-driven auxiliary feedwater (AFW) pump is inoperable concurrent 
with an inoperable motor-driven AFW train. The proposed amendments 
would also make changes to the TSs that establish specific Actions for 
when the turbine-driven AFW train is inoperable either (a) due solely 
to one inoperable steam supply, or (b) due to reasons other than the 
one inoperable steam supply.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on March 19, 2007 (72 FR 12845), on possible 
amendments concerning the consolidated line item improvement process 
(CLIIP), including a model safety evaluation and a model no significant 
hazards consideration (NSHC) determination. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on July 17, 2007 
(72 FR 39089), as part of the CLIIP. In its application dated July 11, 
2008, the licensee affirmed the applicability of the following 
determination.
    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:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The Auxiliary/Emergency Feedwater (AFW/EFW) System is not an 
initiator of any design basis accident or event, and therefore the 
proposed changes do not increase the probability of any accident 
previously evaluated. The proposed changes to address the condition 
of one or two motor driven AFW/EFW trains inoperable and the turbine

[[Page 62569]]

driven AFW/EFW train inoperable due to one steam supply inoperable 
do not change the response of the plant to any accidents.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not adversely 
affect the ability of structures, systems, and components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated. 
Further, the proposed changes do not increase the types and amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not result in a change in the manner in 
which the AFW/EFW System provides plant protection. The AFW/EFW 
System will continue to supply water to the steam generators to 
remove decay heat and other residual heat by delivering at least the 
minimum required flow rate to the steam generators. There are no 
design changes associated with the proposed changes. The changes to 
the Conditions and Required Actions do not change any existing 
accident scenarios, nor create any new or different accident 
scenarios.
    The changes do 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. In addition, 
the changes do not impose any new or different requirements or 
eliminate any existing requirements. The changes do not alter 
assumptions made in the safety analysis. The proposed changes are 
consistent with the safety analysis assumptions and current plant 
operating practice.
    Therefore, the 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 do 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 impacted by these changes. The proposed changes will not 
result in plant operation in a configuration outside the design 
basis.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment requests 
involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Lois M. James.

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

    Date of application for amendments: September 4, 2008.
    Brief description of amendments: The proposed amendment will delete 
the Technical specification (TS) requirements related to hydrogen 
recombiners and hydrogen monitors. Licensees were generally required to 
implement upgrades as described in NUREG-0737, ``Clarification of TMI 
[Three Mile Island] Action Plan Requirements,'' and Regulatory Guide 
(RG) 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.'' Implementation of these upgrades was an outcome of the 
lessons learned from the accident that occurred at TMI, Unit 2.
    Requirements related to combustible gas control were imposed by 
Order for many facilities and were added to or included in the TSs for 
nuclear power reactors currently licensed to operate. The revised 10 
CFR 50.44, ``Standards for Combustible Gas Control System in Light-
Water-Cooled Power Reactors,'' eliminated the requirements for hydrogen 
recombiners and relaxed safety classifications and licensee commitments 
to certain design and qualification criteria for hydrogen and oxygen 
monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on September 
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated September 4, 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to 17 approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
RG 1.97 Category 1, is intended for key variables that most directly 
indicate the accomplishment of a safety function for design-basis 
accident events. The hydrogen monitors no longer meet the definition 
of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 
50.44, the Commission found that Category 3, as defined in RG 1.97, 
is an appropriate categorization or the hydrogen monitors because 
the monitors are required to diagnose the course of beyond design-
basis accidents. The regulatory requirements for the hydrogen 
monitors can be relaxed without degrading the plant emergency 
response. The emergency response, in this sense, refers to the 
methodologies used in ascertaining the condition of the reactor 
core, mitigating the consequences of an accident, assessing and 
projecting offsite releases of radioactivity, and establishing 
protective action recommendations to be communicated to offsite 
authorities. Classification of the hydrogen monitors as Category 3, 
and removal of the hydrogen monitors from TS will not prevent an 
accident management strategy through the use of the SAMGs, the 
emergency plan (EP), the emergency operating procedures (EOP), and 
site survey monitoring that support modification of emergency plan 
protective action recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TS, does not involve a 
significant increase in the probability or the consequences of any 
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 elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner

[[Page 62570]]

and hydrogen monitor equipment are not considered accident 
precursors, nor does their existence or elimination have any adverse 
impact on the pre-accident state of the reactor core or post 
accident confinement of radionuclides within the containment 
building.
    Therefore, this 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 the Margin of Safety

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, in light of existing plant equipment, 
instrumentation, procedures, and programs that provide effective 
mitigation of and recovery from reactor accidents, results in a 
neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: January 14, 2008.
    Description of amendment request: The proposed amendment would 
modify the Technical Specification (TS) requirements related to control 
room envelope habitability in accordance with TS Task Force (TSTF) 
traveler TSTF-448-A, ``Control Room Habitability,'' Revision 3.
    The NRC staff issued a ``Notice of Availability of Technical 
Specification Improvement to Modify Requirements Regarding Control Room 
Envelope Habitability Using the Consolidated Line Item Improvement 
Process'' in the Federal Register on January 17, 2007 (72 FR 2022). The 
notice referenced a model safety evaluation, a model no significant 
hazards consideration (NSHC) determination, and a model license 
amendment request published in the Federal Register on October 17, 2006 
(71 FR 61075). In its application dated January 14, 2008, the licensee 
affirmed the applicability of the model NSHC determination which is 
presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee is presented below:

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

    The proposed change 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 has reviewed the analysis adopted by the licensee 
and, based upon this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendment involves NSHC.
    Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

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

[[Page 62571]]

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.

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

    Date of amendment request: July 30, 2008.
    Description of amendment request: This amendment revises the Indian 
Point Nuclear Generating Unit No. 2 Technical Specification 3.8.1, 
Required Action A.4, to allow a one time extension to the completion 
time for the loss of one offsite power circuit from 72 hours to 144 
hours. This change will ensure that there is enough time for the failed 
oil cooling pump on the station auxiliary transformer to be removed, 
and for the new oil cooling pump to be installed and tested.
    Date of publication of individual notice in Federal Register: 
August 27, 2008.
    Expiration date of individual notice: October 27, 2008.

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].

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

    Date of amendment request: March 10, 2008, as supplemented by 
letters dated June 30, 2008, and September 29, 2008.
    Description of amendment request: The amendment revised the Oyster 
Creek Technical Specifications (TSs) 3.3, ``Reactor Coolant.'' 
Specifically, the amendment relocated the pressure and temperature 
limit curves to the licensee controlled document, ``Pressure and 
Temperature Limits Report'' (PTLR). Additionally, the amendment 
introduced supporting definitions and adds controls regarding the PTLR 
to Section 6.0, ``Administrative Controls.''
    Date of issuance: September 30, 2008.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 269.
    Facility Operating License No. DPR-16: The amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: June 17, 2008 (73 FR 
34339). The supplemental letters 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 initial proposed 
no significant hazards determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
September 30, 2008.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, 
Millstone Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut

    Date of application for amendment: August 15, 2007, as supplemented 
on May 27, 2008, July 24, 2008, and September 3, 2008.
    Brief description of amendment: The proposed amendment modified 
Technical Specification (TS) 3.3.3.1, ``Radiation Monitoring,'' TS 
3.4.6.1, ``Reactor Coolant System Leakage Detection Systems,'' and 
Surveillance Requirements 4.4.6.1, ``Reactor Coolant System Leakage 
Detection Systems.'' Specifically, the proposed amendment removed 
credit for the gaseous radiation monitor for Reactor Coolant System 
leakage detection. Improvements in nuclear fuel reliability over time 
have resulted in the reduction of effectiveness of the monitors in 
detecting very small leaks and very small changes in the leak rate. The 
proposed change also addressed the condition when the remaining 
monitoring systems are all inoperable.
    Date of issuance: September 30, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 306 and 244.
    Renewed Facility Operating License Nos. DPR-65 and NPF-49: 
Amendment revised the License and Technical Specifications.
    Date of initial notice in Federal Register: June 17, 2008 (73 FR 
34341). The supplements dated May 27, 2008, July 24, 2008, and 
September 3, 2008, clarified the application, did not expand the scope 
of the application as originally noticed, and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 30, 2008.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423, 
Millstone Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: May 8, 2008, as supplemented by 
letter dated August 14, 2008.
    Brief description of amendment: This amendment request contains 
sensitive unclassified non-safeguards

[[Page 62572]]

information. The changes allow for interim alternate steam generator 
tube repair criterion, as specified in the Millstone Power Station, 
Unit 3 (MPS3) technical specifications. The interim alternate repair 
criterion is for the upcoming refueling outage and the subsequent 
operating cycle. The amendment also adds three reporting criteria to 
the MPS3 technical specifications for steam generator tube inspections.
    Date of issuance: September 30, 2008.
    Effective date: As of the date of issuance and shall be implemented 
prior to Mode 5 startup.
    Amendment No.: 245.
    Renewed Facility Operating License No. NPF-49: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2008 (73 FR 
39054). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 30, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 20, 2007.
    Brief description of amendment: The amendment reflects the direct 
transfer of the undivided ownership interest of the Saluda River 
Electric Cooperation, Inc., in Catawba Nuclear Station, Unit 1, to Duke 
Energy Carolinas, LLC, a current owner and operator, and the North 
Carolina Electric Membership Corporation, a current owner.
    Date of issuance: September 30, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 245.
    Facility Operating License Nos. NPF-35: Amendment revised the 
license.
    Date of initial notice in Federal Register: July 21, 2008 (73 FR 
42375). The supplement dated May 29, 2008, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated September 25, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 26, 2007, as superseded by 
application dated August 8, 2007, and as supplemented by letters dated 
November 19, 2007, and June 5 and July 21, 2008.
    Brief description of amendment: The amendment revises the 
requirements of Technical Specification (TS) 3.3.5.2, ``Reactor Core 
Isolation Cooling (RCIC) System Instrumentation,'' and TS 3.5.2, ``ECCS 
[Emergency Core Cooling System]-Shutdown,'' to increase the Condensate 
Storage Tank level.
    Date of issuance: September 30, 2008.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 210.
    Facility Operating License No. NPF-21: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 28, 2007 (72 FR 
49572).
    The supplements dated November 19, 2007, and June 5 and July 21, 
2008, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 30, 2008.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 22, 2008, as supplemented 
by letters date July 2, July 22, and September 24, 2008.
    Brief description of amendment: The amendment modified Technical 
Specification (TS) 1.0, ``Definitions,'' Limiting Conditions for 
Operation and Surveillance Requirement Applicability Section 3.4.9, 
``RCS [Reactor Coolant System] Pressure and Temperature (P-T) Limits,'' 
and Section 5.0, ``Administrative Controls,'' to delete reference to 
the pressure and temperature curves, and include reference to the 
Pressure and Temperature Limits Report (PTLR). This change adopted the 
methodology of SIR-05-044-A, ``Pressure-Temperature Limits Report 
Methodology for Boiling Water Reactors,'' for preparation of the 
pressure and temperature curves, and incorporated the guidance of TSTF-
419-A, ``Revise PTLR Definition and References in ISTS [Improved 
Standard Technical Specifications] 5.6.6, RCS PTLR.''
    Date of issuance: October 3, 2008.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 292.
    Facility Operating License No. DPR-59: The amendment revised the 
License and the Technical Specifications.
    Date of initial notice in Federal Register: July 1, 2008 (73 FR 
37503). The supplemental submissions dated July 2, July 22, and 
September 24, 2008, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the NRC staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register. The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated October 3, 2008.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station (Byron), Unit Nos. 1 and 2, Ogle County, Illinois

    Date of application for amendment: June 17, 2008.
    Brief description of amendment: The amendments revise Technical 
Specification (TS) 5.5.9, ``Steam Generator (SG) Program,'' and TS 
5.6.9, ``Steam Generator (SG) Tube Inspection Report.'' For TS 5.5.9, 
the amendments incorporate a one-cycle interim alternate repair 
criteria in the provisions for SG tube repair criteria during Byron, 
Unit No. 2, refueling outage 14 and the subsequent operating cycle. For 
TS 5.6.9, the amendments revise the current reporting requirements. 
These changes only affect Byron, Unit No. 2; however, this action is 
docketed for both Byron units because the TS are common to both units.
    Date of issuance: October 1, 2008.
    Effective date: As of the date of issuance and shall be implemented 
prior to the return to service from Byron, Unit No. 2, fall 2008 
Refueling Outage 14.
    Amendment Nos.: Unit 1--158; Unit 2--158.
    Facility Operating License Nos. NPF-37 and NPF-66: The amendment 
revised the TSs and License.

[[Page 62573]]

    Date of initial notice in Federal Register: August 5, 2008 (73 FR 
45485).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 1, 2008.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: July 16, 2007, as supplemented 
May 20 and August 26, 2008.
    Brief description of amendments: Amendments modified the technical 
specification requirements related to control room envelope 
habitability in accordance with Technical Specification Task Force 
(TSTF) Traveler TSTF-448, Revision 3, ``Control Room Habitability.''
    Date of Issuance: September 30, 2008.
    Effective Date: Unit 1--Amendment is effective as of the date of 
its issuance and shall be implemented following implementation of the 
Amendment No. 152, regarding Alternative Source Term and with the 
completion of the installation and testing of the plant modifications 
described in the licensee's application, including letters dated July 
16, 2007, February 14, March 18, April 14, June 2, July 11, and August 
13, 2008. Unit 2--This license amendment is effective as of the date of 
its issuance and shall be implemented following implementation of 
License Amendment No. 152.
    Amendment Nos.: 205 and 153.
    Renewed Facility Operating License Nos. DPR-67 and NPF-16: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 28, 2007 (72 FR 
49578). The supplements dated May 20 and August 26, 2008, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 30, 2008.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: July 16, 2007, as supplemented 
by letters dated February 14, March 18, April 14, June 2, July 11, and 
August 13, 2008.
    Brief description of amendment: The amendment modifies the 
facility's operating licensing bases to adopt the alternative source 
term as allowed in 10 CFR 50.67, and as described in Regulatory Guide 
1.183. The licensee revised the plant licensing basis through 
reanalysis of the radiological consequences of the following Updated 
Final Safety Analysis Report Chapter 15 accidents: Loss-of-Coolant 
Accident, Fuel-Handling Accident, Main Steam Line Break, Steam 
Generator Tube Rupture, Reactor Coolant Pump Shaft Seizure, Control 
Element Assembly Ejection, Letdown Line Break, and Feedwater Line 
Break.
    Date of issuance: September 29, 2008.
    Effective date: Effective as of the date of issuance and shall be 
implemented within 180 days.
    Amendment No.: 152.
    Renewed Facility Operating License No. NPF-16: The amendment 
revises the Technical Specifications and the Renewed Facility Operating 
License.
    Date of initial notice in Federal Register: June 12, 2008 (73 FR 
33460). The supplements dated February 14, March 18, April 14, June 2, 
July 11, and August 13, 2008, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.
    Public comments received as to proposed no significant hazards 
consideration (NSHC): No.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 29, 2008.
    Attorney for licensee: M. S. Ross, Managing Attorney, Florida Power 
and Light Company, P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Thomas H. Boyce.

Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220 and 50-410, 
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2 (NMP1 and NMP2), 
Oswego County, New York

    Date of application for amendment: December 20, 2007.
    Brief description of amendments: The amendments revise NMP1 
Technical Specification (TS) Section 6.3, ``Unit Staff 
Qualifications,'' and NMP2 TS Section 5.3, ``Unit Staff 
Qualifications,'' to update requirements that have been superseded due 
to the accreditation of the NMPNS licensed operator training program 
and due to promulgation of the revised Title 10 of the Code of Federal 
Regulations (10 CFR), Part 55, ``Operators' Licenses,'' which became 
effective on May 26, 1987 (52 FR 9453). Additionally, the amendment for 
NMP1 revises the TSs by eliminating the qualification requirement 
exceptions listed for the position of Manager Operations which were 
previously approved by the NRC staff. The position of Manager 
Operations would meet the minimum qualification requirements as 
required in American National Standard Institute Standard NI8.1-1971, 
``American National Standard for Selection and Training of Nuclear 
Power Plant Personnel.''
    Date of issuance: September 29, 2008.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment Nos.: 198 and 127.
    Renewed Facility Operating License No. DPR-63 and NPF-069: 
Amendments revise the License and TSs.
    Date of initial notice in Federal Register: January 28, 2008 (73 FR 
5225).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 29, 2008.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: October 3, 2007.
    Brief description of amendments: The amendments revised a footnote 
in Technical Specifications Table 3.3.2.1-1, ``Control Rod Block 
Instrumentation,'' such that a new banked position withdrawal sequence 
shutdown sequence could be utilized. Associated changes are made to the 
TS Bases. This operating license improvement was made available by the 
NRC staff on May 23, 2007, as part of the consolidated line item 
improvement process.
    Date of issuance: October 1, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--258, Unit 2--202.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the licenses and the technical specifications.

[[Page 62574]]

    Date of initial notice in Federal Register: November 6, 2007 (72 FR 
62691).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 1, 2008.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: October 5, 2007.
    Brief description of amendments: The amendments revise the TSs 
completion times (CTs) for TS Limiting Condition of Operation (LCO) 
3.8.1, Conditions B and C, by specifying when maintenance restrictions 
need to be met and by adding a 72-hour CT for the swing DG 1B.
    Date of issuance: October 2, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days from the date of issuance.
    Amendment Nos.: Unit 1--259, Unit 2--203.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: November 6, 2007, (72 
FR 62691).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 2, 2008.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama Southern Nuclear Operating Company, Inc., Georgia Power 
Company, Oglethorpe Power Corporation, Municipal Electric Authority of 
Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin 
I. Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia Southern 
Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle 
Electric Generating Plant, Units 1 and 2, Burke County, Georgia

    Date of application for amendments: June 12, 2008.
    Brief description of amendments: The amendments revised the 
Technical Specifications requirement for the Plant Manager or the 
Operations Manager regarding the holding of a Senior Reactor Operator 
license.
    Date of issuance: October 7, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Farley Unit 1--179; Unit 2--171; Hatch Unit 1--260; 
Unit 2--204; Vogtle Unit 1--153; Unit 2--134.
    Facility Operating License Nos. NPF-2 and NPF-8; DPR-57 and NPF-5; 
NPF-68 and NPF-81: Amendments revised the licenses and the technical 
specifications.
    Date of initial notice in Federal Register: July 1, 2008, 73 FR 
37505.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 7, 2008.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Unit 1, Hamilton County, Tennessee

    Date of application for amendment: April 14, 2008.
    Brief description of amendment: The amendment revises the list of 
topical reports referenced in Technical Specification Section 
6.9.1.14.a for use in preparing the core operating limits report by 
adding EMF-2103P-A, ``Realistic Large Break LOCA Methodology for 
Pressurized Water Reactors.'' The change will be utilized in core 
loading designs for Unit 1 fuel-load configurations in future operating 
cycles.
    Date of issuance: September 24, 2008.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 320.
    Facility Operating License No. DPR-77: Amendment revises the 
technical specifications.
    Date of initial notice in Federal Register: June 10, 2008 (73 FR 
32746). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 24, 2008.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 10th day of October 2008.

    For the Nuclear Regulatory Commission.
Joseph Gitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E8-24896 Filed 10-20-08; 8:45 am]
BILLING CODE 7590-01-P