[Federal Register Volume 71, Number 59 (Tuesday, March 28, 2006)]
[Notices]
[Pages 15479-15494]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-2908]


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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

[[Page 15480]]

Commission publish notice of any amendments issued, or proposed to be 
issued and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license upon a 
determination by the Commission that such amendment involves no 
significant hazards consideration, notwithstanding the pendency before 
the Commission of a request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 3, 2006 to March 16, 2006. The last 
biweekly notice was published on March 14, 2006 (71 FR 13169).

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. Within 60 days after the date of publication of this 
notice, the licensee 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 for a hearing and a petition for leave to intervene.
    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, Rules and 
Directives 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 6D22, 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, the 
licensee 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 
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 persons 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 support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to

[[Page 15481]]

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 a hearing or a petition for leave to intervene must 
be filed 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; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this 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].

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

    Date of amendment request: December 1, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.6.4.1, ``Secondary Containment.'' 
Specifically, the change would modify Surveillance Requirements (SRs) 
3.6.4.1.4 and 3.6.4.1.5 to clarify their intent with respect to 
secondary containment boundary integrity. The change is submitted in 
accordance with the TS Task Force Traveler 322-A, Revision 2, 
``Secondary Containment and Shield Building Boundary Integrity SRs.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No. This change involves an administrative clarification 
to reflect the original intent of the Technical Specifications. There 
is no impact on the availability or capability of the secondary 
containment or Standby Gas Treatment (SGT) system as a result of the 
proposed change. Both the secondary containment and SGT system are 
considered accident-mitigating equipment and are not initiators of any 
previously evaluated accidents. Therefore, the proposed change does not 
involve an increase in the probability of an accident previously 
evaluated. Additionally, the proposed change does not alter the 
secondary containment or SGT systems' performance measures or their 
ability to perform their accident mitigation functions.
    Therefore, the proposed change does not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed changes to the wording of TS SRs 
3.6.4.1.4 and 3.6.4.1.5 clarify that only one SGT subsystem is required 
to ensure the requirements of TS 3.6.4.1 are met. The proposed change 
does not alter the parameters within which the plant is operated. There 
are no new system operating conditions or performance measures 
introduced by this proposed change that will affect the secondary 
containment and SGT systems' protective or mitigative functions. The 
proposed changes will not alter the methods in which equipment is 
operated or tested. No new accident scenarios or assumptions, failure 
mechanisms, or limiting single failures are introduced as a result of 
the proposed change.
    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. Margins of safety are established in the design of 
components, the configuration of components to meet certain performance 
parameters, and in the establishment of setpoints to initiate alarms or 
actions. The proposed change does not impact any of these margins of 
safety parameters. This change involves an administrative clarification 
to reflect the original intent of the TS. There is no adverse effect on 
the operability or design requirements of the secondary containment or 
SGT system. The equipment will continue to be tested in a manner and at 
a frequency necessary to provide confidence that the equipment can 
perform its intended safety function. There is no impact on the plant 
safety analyses.
    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. Bradley J. Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Daniel S. Collins.

[[Page 15482]]

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: February 27, 2006.
    Description of amendments request: The amendment would revise 
Technical Specification 4.2.1, ``Fuel Assemblies,'' to allow fuel with 
advanced cladding material to be installed in the core for Cycle 19 
only at Unit No. 1 or Cycle 17 only at Unit No. 2. Advanced cladding 
material from Framatome-ANP may be used in up to 2 lead test 
assemblies, and advanced cladding material from Westinghouse may be 
used in up to 2 lead test assemblies.
    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. Would not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Calvert Cliffs Technical Specification 4.2.1, Fuel Assemblies, 
states that fuel rods are clad with either Zircaloy or 
ZIRLOTM. Calvert Cliffs Nuclear Power Plant, Inc. proposes 
to re-insert up to four fuel assemblies into Calvert Cliffs Unit 1 or 
Unit 2 that have some fuel rods clad in zirconium alloys that do not 
meet the definition of Zircaloy or ZIRLOTM. A temporary 
exemption to the regulations has also been requested to allow these 
fuel assemblies to be re-inserted into Unit 1 or Unit 2. The proposed 
change to the Calvert Cliffs Technical Specifications will allow the 
use of cladding materials that are not Zircaloy or ZIRLOTM 
for one fuel cycle once the temporary exemption is approved. The 
proposed change to the Technical Specification is effective only as 
long as the temporary exemption is effective. The addition of what will 
be an approved temporary exemption for Unit 1 or Unit 2 to Technical 
Specification 4.2.1 does not change the probability or consequences of 
an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Would not create the possibility of a new or different [kind] of 
accident from any accident previously evaluated.
    The proposed change does not add any new equipment, modify any 
interfaces with existing equipment, change the equipment's function, or 
change the method of operating the equipment. The proposed change does 
not affect normal plant operations or configuration. Since the proposed 
change does not change the design, configuration, or operation, it 
could not become an accident initiator.
    Therefore, the proposed change does not create the possibility of a 
new or different [kind] of accident from any previously evaluated.
    3. Would not involve a significant reduction in [a] margin of 
safety.
    The proposed change will add an approved temporary exemption to the 
Calvert Cliffs Technical Specifications allowing the installation of up 
to four lead fuel assemblies. The assemblies use advanced cladding 
materials that are not specifically permitted by existing regulations 
or Calvert Cliffs' Technical Specifications. A temporary exemption to 
allow the installation of these assemblies has been requested. The 
addition of an approved temporary exemption to Technical Specification 
4.2.1 is an administrative change to allow the installation of the lead 
fuel assemblies under the provisions of the temporary exemption. The 
license amendment is effective only as long as the exemption is 
effective. This amendment does not change the margin of safety since it 
only adds a reference to an approved, temporary exemption to the 
Technical Specifications.
    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 
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: Richard J. Laufer.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: January 5, 2005, supplemented November 
21, 2005.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification (TS) 5.5.19.b, TS 5.5.19.c, and TS 
Surveillance Requirement (SR) 3.8.1.9. TS 5.5.19.b currently requires 
verification that a Lee Combustion Turbine (LCT) can supply the 
equivalent of one Unit's maximum safeguard loads, plus two Units' Mode 
3 loads, when connected to the system grid every 12 months. In the 
proposed amendments, this requirement would be more clearly specified 
as, ``Verify an LCT can supply equivalent of one Unit's Loss of Coolant 
Accident (LOCA) loads plus two Unit's Loss of Offsite Power (LOOP) 
loads when connected to system grid every 12 months.'' TS 5.5.19.b and 
SR 3.8.1.9 would be revised for consistency.
    This notice supersedes the notice published in the Federal Register 
on February 15, 2005 (70 FR 7764).
    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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    Duke proposes to revise TS 5.5.19.b to clarify the Lee Combustion 
Turbine (LCT) testing requirements. Duke proposes to revise TS 5.5.19.c 
and TS 3.8.1 Surveillance Requirement (SR) 3.8.1.19 to be consistent 
with the proposed change to TS 5.5.19.b. The proposed change makes the 
wording of the test requirement consistent with the UFSAR [Updated 
Final Safety Analysis Report]. LCT testing has no impact on the 
probability of an accident analyzed in the UFSAR. The LCT can be 
credited to mitigate the consequences of an accident analyzed in the 
UFSAR. However, this clarification of LCT testing requirements has no 
impact on its ability to mitigate the consequences of an accident. As 
such, the proposed LAR [license amendment request] does not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    (2) Create the possibility of a new or different kind of accident 
from any kind of accident previously evaluated:
    Duke proposes to revise TS 5.5.19.b to clarify the Lee Combustion 
Turbine (LCT) testing requirements. Duke proposes to revise TS 5.5.19.c 
and TS 3.8.1 SR 3.8.1.9 to be consistent with the proposed change to TS 
5.5.19.b. The proposed change makes the wording of the test requirement 
consistent with the UFSAR. These changes do not alter the nature of 
events postulated in the Safety Analysis Report nor do they introduce 
any unique precursor mechanisms. Therefore, the proposed amendment will 
not create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (3) Involve a significant reduction in a margin of safety:

[[Page 15483]]

    The proposed TS change does not unfavorably affect any plant safety 
limits, set points, or design parameters. The changes also do not 
unfavorably affect the fuel, fuel cladding, RCS [reactor coolant 
system], or containment integrity. Therefore, the proposed TS change, 
which clarifies TS requirements associated with the LCT testing 
program, does not involve a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Branch Chief: Evangelos C. Marinos.

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: March 13, 2006.
    Description of amendment request: The proposed amendments would 
make changes to the technical specifications (TS) for LaSalle County 
Station (LSCS), Units 1 and 2. Surveillance Requirement (SR) 3.7.3.1 
verifies the cooling water temperature supplied to the plant from the 
core standby cooling system (CSCS) pond (i.e., the ultimate heat sink 
(UHS)) is <= 100 [deg]F. Currently, if the temperature of the cooling 
water supplied to the plant from the CSCS pond is > 100 [deg]F, the UHS 
must be declared inoperable in accordance with TS 3.7.3. TS 3.7.3, 
Required Action B.1, requires that both units be placed in Mode 3 
within 12 hours and Required Action B.2 requires that both units be 
placed in Mode 4 within 36 hours.
    Prolonged hot weather in the area during the summer months, in 
conjunction with high humidity during the daytime, minimal cooling at 
night and little precipitation, has resulted in sustained elevated 
cooling water temperature supplied to the plant from the CSCS pond. 
This license amendment is being requested to increase the temperature 
limit of the cooling water supplied to the plant from the CSCS pond to 
<= 101.5 [deg]F by reducing the temperature measurement uncertainty by 
replacing the existing thermocouples with higher precision temperature 
measuring equipment. Should the UHS indicated temperature exceed 101.5 
[deg]F, Required Action B.1 would be entered and both units would be 
placed in Mode 3 within 12 hours and Mode 4 within 36 hours.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated? 
The proposed change will allow the indicated temperature of the cooling 
water supplied to the plant from the CSCS pond to be increased to <= 
101.5 [deg]F based on reducing the temperature measurement uncertainty 
by replacing the existing thermocouples with higher precision 
temperature measuring equipment.
    Analyzed accidents are assumed to be initiated by the failure of 
plant structures, systems, or components. An inoperable UHS is not 
considered as an initiator of any analyzed events. As such, there is 
not a significant increase in the probability of a previously evaluated 
accident. Allowing the UHS to operate at a higher allowable indicated 
temperature, but still within the design limits of the equipment it 
supplies, will not affect the failure probability of that equipment. 
The current heat analyses calculations of record for LSCS, Units 1 and 
2, assume a UHS temperature of 100 [deg]F and post-accident peak inlet 
temperature of 104 [deg]F. The proposed temperature increase is based 
solely on a reduction of the existing instrument loop uncertainty 
value. The current analysis bounds the proposed change. This higher 
allowable indicated temperature does not impact the LOCA [loss-of-
coolant accident] Peak Clad Temperature Analysis, LOCA Containment 
Analysis or the non-LOCA analyses; therefore, continued operation with 
a UHS temperature > 100 [deg]F but <= 101.5 [deg]F will not increase 
the consequences of an accident previously evaluated in the UFSAR.
    Based on the above information, the increase in the allowable 
indicated temperature of the cooling water supplied to the plant from 
the UHS to <= 101.5 [deg]F by reducing the existing instrument loop 
uncertainty value has no effect on the result of the design basis event 
and will continue to allow each required heat exchanger to perform its 
safety function. The heat exchangers will continue to provide 
sufficient cooling for the heat loads during the most severe 30-day 
period.
    Based on the above information, increasing the allowable indicated 
temperature of the cooling water supplied to the plant from the CSCS 
pond from <= 100 [deg]F to <= 101.5 [deg]F by reducing the instrument 
uncertainty value has no impact on any analyzed accident; therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change create the possibility of a new or different 
kind of accident from any previously evaluated?
    The proposed change involves replacing the presently installed 
thermocouples with higher accuracy temperature measurement equipment. 
This proposed action will not alter the manner in which equipment is 
operated, nor will the functional demands on credited equipment be 
changed. No alteration in the procedures that ensure the units remain 
within analyzed limits is proposed, and no change is being made to 
procedures relied upon to respond to an off-normal event. Raising the 
UHS temperature limit does not introduce any new or different modes of 
plant operation, nor does it affect the operational characteristics of 
any safety-related equipment or systems; as such, no new failure modes 
are being introduced. The proposed action reduces the instrument 
uncertainty value but does not alter assumptions made in the safety 
analysis.
    Increasing the allowable indicated temperature of the cooling water 
supplied to the plant from the CSCS pond from <= 100 [deg]F to <= 101.5 
[deg]F has no impact on safety related systems. The plant is designed 
such that the RHR [residual heat removal] pumps on the unit undergoing 
the LOCA/LOOP [loss of offsite power] conditions would start upon the 
receipt of a signal, and would load onto their respective Emergency 
Diesel Generators emergency bus during the LOOP event. The increase in 
the allowable indicated temperature of the cooling water supplied to 
the plant from the CSCS pond will not require operation of additional 
RHR pumps; therefore, system operation is unaffected by the proposed 
change in the UHS temperature limit.
    Based on the above information, the proposed change does not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed change allows an increase in the allowable indicated 
temperature of the cooling water supplied to the plant from the CSCS

[[Page 15484]]

pond to <= 101.5 [deg]F. The margin of safety is determined by the 
design and qualification of the plant equipment, the operation of the 
plant within analyzed limits, and the point at which protective or 
mitigative actions are initiated. The proposed action does not impact 
these factors as the analyzed peak inlet temperature of the UHS is 
unaffected based on the improved instrument uncertainty of the new high 
precision temperature measurement instrumentation. No setpoints are 
affected, and no other change is being proposed in the plant 
operational limits as a result of this change. All accident analysis 
assumptions and conditions will continue to be met. Adequate design 
margin is available to ensure that the required margin of safety is not 
significantly reduced.
    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, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Daniel S. Collins.

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

    Date of amendment request: February 14, 2006.
    Description of amendment request: The proposed amendment would 
revise the frequency of the Mode 5 Intermediate Range Monitoring (IRM) 
Instrumentation CHANNEL FUNCTIONAL TEST contained in Technical 
Specification (TS) 3.3.1.1 from 7 days to 31 days. The methodology used 
for the IRM drift analysis is based upon guidance contained in Generic 
Letter 91-04, ``Changes in Technical Specification Surveillance 
Intervals to Accommodate a 24-month Fuel Cycle,'' and Electric Power 
Institute Report TI-103335, ``Guidance for Instrument Calibration 
Extension/Reduction Programs.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed Technical Specifications (TS) change involves an 
increase in the Mode 5 CHANNEL FUNCTIONAL TEST interval for RPS 
[Reactor Protection System] IRM channels from 7 days to 31 days. The 
IRM system is used for event mitigation. The failure of an IRM does not 
initiate an accident or transient event. The proposed TS change does 
not alter the design or function of the IRM system for no physical 
changes are being made to the plant. Evaluation of the proposed testing 
interval change demonstrated that the availability of IRMs to mitigate 
the consequences of a control rod withdrawal event at low power levels 
are not significantly affected based on the effectiveness of other, 
required TS surveillance testing that is performed, the availability of 
redundant systems and equipment, and the high reliability of the IRM 
equipment.
    Therefore, the proposed 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 accident previously evaluated.
    The proposed TS change involves an increase in the Mode 5 IRM 
CHANNEL FUNCTIONAL TEST interval from 7 clays [days] to 31 days. 
Existing TS testing requirements ensure the operability of the IRMs. 
The proposed TS change does not introduce any failure mechanisms of a 
different type than those previously evaluated, since no physical 
changes to the plant are being made. No new or different equipment is 
being installed, and no installed equipment is being operated in a 
different manner. As a result, no new failure modes are introduced. In 
addition, the manner in which surveillance tests are performed remains 
unchanged.
    Therefore, the proposed TS change does not create the possibility 
of a new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed TS change involves an increase in the Mode 5 CHANNEL 
FUNCTIONAL TEST interval for RPS IRM channels from 7 days to 31 days. 
There is expected to be no impact on system operability, based upon the 
performance of the more frequent Channel Checks, Control Room 
monitoring when the IRMs are in use, and the overall IRM reliability.
    Furthermore, a historical review of surveillance test results and 
associated maintenance records did not indicate evidence of any failure 
that would invalidate the above conclusions.
    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: David W. Jenkins, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Mindy S. Landau, Acting.

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

    Date of amendment request: December 16, 2005.
    Description of amendment request: The proposed change to Technical 
Specification (TS) Surveillance Requirement (SR) 4.1.4d relocates the 
SR for testing the core spray header differential pressure ([Delta]P) 
instrumentation to licensee-controlled documents. TS SR 4.1.4d 
currently requires that the core spray header [Delta]P instrumentation 
be periodically tested such that a check of each sensor is performed at 
least once each day and each channel is calibrated and tested at least 
once every 3 months. The proposed change will allow these SRs to be 
placed in licensee-controlled documents where future changes will be 
made pursuant to Title 10 of the Code of Federal Regulations (10 CFR), 
Section 50.59. The functional description of the core spray header 
[Delta]P instrumentation will also be relocated from the TS Bases to 
licensee-controlled documents consistent with the proposed TS change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes are limited to the relocation of selected 
instrumentation requirements. The proposed relocated requirements were

[[Page 15485]]

determined not to meet the 10 CFR 50.36 screening criteria for 
retention in the TSs and will be maintained in licensee-controlled 
documents in accordance with the provisions of 10 CFR 50.59. The 
proposed changes do not introduce any new modes of plant operation, 
make any physical changes to the plant, or alter any operational 
setpoints which could degrade the performance of any safety system 
assumed to function in the accident analysis. Therefore, the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed changes do not introduce any new modes 
of plant operation, make any physical changes to the plant, or alter 
any operational setpoints which could create new accident initiators or 
failure mechanisms. The proposed changes are limited to the relocation 
of selected instrumentation requirements, and will have no impact on 
the accident assumptions and initial conditions as previously analyzed 
in the UFSAR [Updated Final Safety Analysis Report]. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed changes are consistent with the Improved 
Standard TSs (NUREG-1433, Rev. 3) and will have no impact on the 
instrumentation setpoints, logic, or functional requirements as 
described in the TSs, TS Bases, and UFSAR. The proposed relocated 
requirements were determined to not meet the 10 CFR 50.36 screening 
criteria for retention in the TSs. Thus, the relocated requirements 
will be maintained in accordance with 10 CFR 50.59 as required. 
Accordingly, the proposed relocated requirements will not degrade the 
quality or performance of any safety system assumed to mitigate an 
accident or assure operation within the safety limits. Therefore, the 
proposed changes do not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Richard J. Laufer.

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

    Date of amendment request: February 28, 2006.
    Description of amendment request: The proposed amendments would 
change the SSES 1 and 2 Technical Specification (TS) Surveillance 
Requirements (SRs) 3.8.4.7 and 3.8.4.8 to clarify that diesel generator 
``E'' (DG E) electrical power subsystem testing does not require a mode 
restriction when the DG E diesel is not required to be OPERABLE.
    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. Performance of TS required SRs are not initiators to 
any accident sequences analyzed in the Final Safety Analysis Report 
(FSAR). The changes do not involve any physical change to structures, 
systems, or components, (SSCs) and do not alter the method of operation 
or control of SSCs. The current assumptions in the safety analysis 
regarding accident initiators and mitigation of accidents are 
unaffected by these changes. No additional failure modes or mechanisms 
are being introduced and the likelihood of previously analyzed failures 
remains unchanged.
    Operation in accordance with the proposed Technical Specification 
(TS) ensures that the DC [direct current] distribution system and 
supported equipment functions remain capable of performing the function 
as described in the FSAR. Therefore, the mitigative functions supported 
by the system will continue to provide the protection assumed by the 
analysis.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed change does not involve a physical 
alteration of the plant. No new equipment is being introduced, and 
installed equipment is not being operated in a new or different manner. 
There are no setpoints, at which protective or mitigative actions are 
initiated, affected by this change. This change will not alter the 
manner in which equipment operation is initiated, nor will the function 
demands on credited equipment be changed. No alterations in the 
procedures that ensure the plant remains within analyzed limits are 
being proposed, and no changes are being made to the procedures relied 
upon to respond to an off-normal event as described in the FSAR. As 
such, no new failure modes are being introduced. The change does not 
alter assumptions made in the safety analysis and licensing basis.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The margin of safety is established through equipment 
design, operating parameters, and the setpoints at which automatic 
actions are initiated. The proposed change is acceptable because 
performance of SRs on equipment not require[d] to be OPERABLE and 
isolated from the OPERABLE plant equipment cannot affect any margin of 
safety. Therefore, the plant response to analyzed events will continue 
to provide the margin of safety assumed by the analysis.
    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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Richard J. Laufer

Southern California Edison Company (SCE), et al., Docket Nos. 50-361 
and 50-362, San Onofre Nuclear Generating Station, Units 2 and 3 (SONGS 
2 and 3), San Diego County, California

    Date of amendment requests: March 10, 2006.
    Description of amendment requests: The licensee requests the 
Nuclear Regulatory Commission consent to the transfer of the City of 
Anaheim's 3.16 percent undivided ownership interest in SONGS 2 and 3 to 
Southern California Edison, excluding Anaheim's interest in its spent 
fuel and the SONGS 2 and 3 independent spent fuel storage installation.

[[Page 15486]]

    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 amendments do not involve any change in the 
design, configuration, or operation of the nuclear plant. All Limiting 
Conditions for Operation, Limiting Safety System Settings, and Safety 
Limits specified in the Technical Specifications remain unchanged. SCE 
will continue to be the licensed operator of the units.
    The technical qualifications of SCE to carry out its exclusive 
responsibilities under the operating licenses, as amended, will remain 
unchanged. Personnel engaged in operation, maintenance, engineering, 
assessment, training, and other related services are not changed. The 
SCE officers and executives currently responsible for the overall safe 
operation of the nuclear plants will continue in the same capacity.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The amendments do not involve any change in the 
design, configuration, or operation of the nuclear plant. The current 
plant design and design bases will remain the same. The current plant 
safety analyses, therefore, remain complete and accurate in addressing 
the design basis events and in analyzing plant response and 
consequences.
    The Limiting Conditions for Operation, Limiting Safety System 
Settings, and Safety Limits specified in the Technical Specifications 
are not affected by the change. As such, the plant conditions for which 
the design basis accident analyses were performed remain valid.
    The amendments do not introduce a new mode of plant operation or 
new accident precursors, do not involve any physical alterations to 
plant configurations, or make changes to system set points that could 
initiate a new or different kind of accident.
    Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The amendments do not involve a change in the design, 
configuration, or operation of the nuclear plants. The change does not 
affect either the way in which the plant structures, systems, and 
components perform their safety function, or their design and licensing 
basis.
    Plant safety margins are established through Limiting Conditions 
for Operation, Limiting Safety System Settings, and Safety Limits 
specified in the Technical Specifications. Because there is no change 
to the physical design of the plant, there is no change to any of these 
margins.
    Therefore, the proposed 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 requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: David Terao.

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 amendment request: September 19, 2005.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Limiting Conditions for Operation 
(LCO) 3.3.1, ``Reactor Trip system (RTS) Instrumentation'' and TS 
Surveillance Requirements (SR) 3.2.4.2, ``Quadrant Power Tilt Ration 
(QPTR)'' to avoid confusion as to when a flux map for QPTR is required.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes do not adversely affect accident 
initiators or precursors nor alter the design assumptions, conditions, 
or configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or prevent 
the ability of structures, systems, and components (SSCs) from 
performing their intended function to mitigate the consequences of an 
initiating event within the assumed acceptance limits. The proposed 
changes do not affect the source term, containment isolation, or 
radiological release assumptions used in evaluating the radiological 
consequences of an accident previously evaluated. Further, the proposed 
changes do not increase the types or amounts of radioactive effluent 
that may be release offsite, nor significantly increase individual or 
cumulative occupational/public radiation exposures. The proposed 
changes are consistent with safety analysis assumptions and resultant 
consequences.
    Therefore, the proposed changes do not increase 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 previously evaluated?
    No. The proposed changes do not result in a change in the manner in 
which the RTS and ESFAS provide plant protection. The RTS and ESFAS 
will continue to have the same set points after the proposed changes 
are implemented. There are no design changes associated with the 
license amendment.
    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 proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    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. Redundant RTS and ESFAS trains are 
maintained, and diversity with regard to the signals that provide 
reactor trip and engineered safety features actuation is also

[[Page 15487]]

maintained. All signals credited as primary or secondary, and all 
operator actions credited in the accident analyses will remain the 
same. The proposed changes will not result in plant operation in a 
configuration outside the design basis.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: January 10, 2006 (TS-453).
    Description of amendment request: The proposed amendment would 
specify the methodology used for determining, setting, and evaluating 
as-found setpoints for those drift susceptible instruments, which are 
either necessary to ensure compliance with a Safety Limit or critical 
in ensuring the fuel peak cladding temperature acceptance criteria of 
10 CFR 50.46 are met.
    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. Including references to TVA's methodology for 
determining, setting, and evaluating as-found instrument setpoints in 
the TS is an administrative change. There will be no change to the 
manner in which Safety Limits, Analytical Limits, or Allowable Values 
are determined. No changes are proposed in the manner in which the 
Reactor Protection System (RPS), Emergency Core Cooling System (ECCS), 
Reactor Core Isolation Cooling (RCIC), or Primary Containment Isolation 
systems provide plant protection or which create new modes of plant 
operation.
    The proposed request will not affect the probability of any event 
initiators. There will be no degradation in the performance of, or an 
increase in the number of challenges imposed on, safety-related 
equipment assumed to function during an accident situation. There will 
be no change to normal plant operating parameters or accident 
mitigation performance.
    Therefore, the proposed amendment 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. There are no hardware changes nor are there any 
changes in the method by which any plant system performs a safety 
function. This request does not affect the normal method of plant 
operation. The proposed amendment does not introduce new equipment, 
which could create a new or different kind of accident.
    No new external threats, release pathways, or equipment failure 
modes are created. No new accident scenarios, transient precursors, 
failure mechanisms, or limiting single failures are introduced as a 
result of this request. Therefore, the implementation of the proposed 
amendment will not create a possibility for an accident of a new or 
different type than those previously evaluated.
    3. Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No. Including references to TVA's methodology for 
determining, setting, and evaluating as-found instrument setpoints in 
the TS is an administrative change. No changes are proposed in the 
manner in which the RPS, ECCS, RCIC, or Primary Containment Isolation 
systems satisfy the Updated Final Safety Analysis Report requirements 
for accident mitigation or unit safe shutdown. There will be no change 
to Safety Limits, Analytical Limits, Allowable Values, or post-Loss Of 
Coolant Accident peak clad temperatures. For these reasons, the 
proposed amendment does not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Michael L. Marshall, Jr.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: February 6, 2006.
    Description of amendment request: The proposed amendment would 
modify technical specification (TS) requirements for inoperable 
snubbers by adding Limiting Condition for Operation 3.0.7. The changes 
are consistent with Nuclear Regulatory Commission approved Industry/
Technical Specification Task Force (TSTF) standard TS change TSTF-373, 
Revision 4. The availability of this TS improvement was published in 
the Federal Register on May 4, 2005 (70 FR 23252), as part of the 
consolidated line item improvement process.
    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 allows a delay time before 
declaring supported TS systems inoperable when the associated 
snubber(s) cannot perform its required safety function. Entrance into 
Actions or delaying entrance into Actions is not an initiator of any 
accident previously evaluated. Consequently, the probability of an 
accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on the delay time allowed 
before declaring a TS supported system inoperable and taking its 
Conditions and Required Actions are no different than the consequences 
of an accident under the same plant conditions while relying on the 
existing TS supported system Conditions and Required Actions. 
Therefore, the consequences of an accident previously evaluated are not 
significantly increased by this change. Therefore, this 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

[[Page 15488]]

of accident from any accident previously evaluated?
    Response: No. The proposed change allows a delay time before 
declaring supported TS systems inoperable when the associated 
snubber(s) cannot perform its required safety function. 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 operations. Thus, this change does not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No. The proposed change allows a delay time before 
declaring supported TS systems inoperable when the associated 
snubber(s) cannot perform its required safety function. The proposed 
change restores an allowance in the pre-ISTS conversion TS that was 
unintentionally eliminated by the conversion. The pre-ISTS TS were 
considered to provide an adequate margin of safety for plant operation, 
as does the post-ISTS conversion TS. Therefore, this change does not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Michael L. Marshall, Jr.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant, 
Unit 2, Hamilton County, Tennessee

    Date of amendment request: February 15, 2006.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TS) to adopt NRC-approved Revision 4 to 
Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-449, ``Steam Generator Tube 
Integrity.'' The proposed amendment includes changes to the TS 
definition of Leakage, TS 3.4.6.2, ``Reactor Coolant System, 
Operational Leakage,'' TS 3.4.5, ``Steam Generator (SG) Tube 
Integrity,'' and adds TS 6.8.4.k, ``Steam Generator (SG) Program,'' and 
TS 6.9.1.16, ``Steam Generator Tube Inspection Report.'' The proposed 
changes are necessary in order to implement the guidance for the 
industry initiative on NEI 97-06, ``Steam Generator Program 
Guidelines.''
    The amendment would also delete License Condition 2.C.8 Item b. 
This License Condition references the licensee's letters from 1997 that 
contain commitments associated with NRC Generic Letter 95-05, 
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes 
Affected by Outside Diameter Stress Corrosion Cracking,'' and the 
application of voltage-based alternate repair criteria to the steam 
generators. The licensee has concluded that the provisions and 
requirements of the proposed TS changes bound the commitments 
identified in the existing License Condition.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments 
adopting TSTF-449, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on May 6, 2005 
(70 FR 24126). The licensee affirmed the applicability of the following 
NSHC determination in its application dated August 31, 2005.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change requires a SG Program that includes performance 
criteria that will provide reasonable assurance that the SG tubing will 
retain integrity over the full range of operating conditions (including 
startup, operation in the power range, hot standby, cooldown and all 
anticipated transients included in the design specification). The SG 
performance criteria are based on tube structural integrity, accident 
induced leakage, and operational LEAKAGE.
    A steam generator tube rupture (SGTR) event is one of the design 
basis accidents that are analyzed as part of a plant's licensing basis. 
In the analysis of a SGTR event, a bounding primary to secondary 
LEAKAGE rate equal to the operational LEAKAGE rate limits in the 
licensing basis plus the LEAKAGE rate associated with a double-ended 
rupture of a single tube is assumed.
    For other design basis accidents such as a main steamline break 
(MSLB), rod ejection, and reactor coolant pump locked rotor the tubes 
are assumed to retain their structural integrity (i.e., they are 
assumed not to rupture). These analyses typically assume that primary 
to secondary LEAKAGE for all SGs is 1 gallon per minute or increases to 
1 gallon per minute as a result of accident induced stresses. The 
accident induced leakage criterion introduced by the proposed changes 
accounts for tubes that may leak during design basis accidents. The 
accident induced leakage criterion limits this leakage to no more than 
the value assumed in the accident analysis.
    The SG performance criteria proposed change to the TS identify the 
standards against which tube integrity is to be measured. Meeting the 
performance criteria provides reasonable assurance that the SG tubing 
will remain capable of fulfilling its specific safety function of 
maintaining reactor coolant pressure boundary integrity throughout each 
operating cycle and in the unlikely event of a design basis accident. 
The performance criteria are only a part of the SG Program required by 
the proposed change to the TS. The program, defined by NEI 97-06, Steam 
Generator Program Guidelines, includes a framework that incorporates a 
balance of prevention, inspection, evaluation, repair, and leakage 
monitoring. The proposed changes do not, therefore, significantly 
increase the probability of an accident previously evaluated.
    The consequences of design basis accidents are, in part, functions 
of the DOSE EQUIVALENT I-131 in the primary coolant and the primary to 
secondary LEAKAGE rates resulting from an accident. Therefore, limits 
are included in the plant technical specifications for operational 
leakage and for DOSE EQUIVALENT I-131 in primary coolant to ensure the 
plant is operated within its analyzed condition. The typical analysis 
of the limiting design basis accident assumes that primary to secondary 
leak rate after the accident is 1 gallon per minute and that the 
reactor coolant activity levels of DOSE EQUIVALENT I-131 are at the TS 
values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the requirements 
for SG inspections. The proposed change does not adversely impact any 
other previously evaluated

[[Page 15489]]

design basis accident and is an improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences of 
a SGTR accident and the probability of such an accident is reduced. In 
addition, the proposed changes do not affect the consequences of an 
MSLB, rod ejection, or a reactor coolant pump locked rotor event, or 
other previously evaluated accident.

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

    The proposed performance based requirements are an improvement over 
the requirements imposed by the current technical specifications. 
Implementation of the proposed SG Program will not introduce any 
adverse changes to the plant design basis or postulated accidents 
resulting from potential tube degradation. The result of the 
implementation of the SG Program will be an enhancement of SG tube 
performance. Primary to secondary LEAKAGE that may be experienced 
during all plant conditions will be monitored to ensure it remains 
within current accident analysis assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility of a 
new or different type of accident from any accident previously 
evaluated.

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

    The SG tubes in pressurized water reactors are an integral part of 
the reactor coolant pressure boundary and, as such, are relied upon to 
maintain the primary system's pressure and inventory. As part of the 
reactor coolant pressure boundary, the SG tubes are unique in that they 
are also relied upon as a heat transfer surface between the primary and 
secondary systems such that residual heat can be removed from the 
primary system. In addition, the SG tubes isolate the radioactive 
fission products in the primary coolant from the secondary system. In 
summary, the safety function of an SG is maintained by ensuring the 
integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the SG Program to manage SG tube inspection, 
assessment, repair, and plugging. The requirements established by the 
SG Program are consistent with those in the applicable design codes and 
standards and are an improvement over the requirements in the current 
TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the TS.
    The NRC staff proposes to determine that the amendments request 
involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: December 15, 2005.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TS) to adopt NRC-approved Revision 4 to 
Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-449, ``Steam Generator Tube 
Integrity.'' The proposed amendment includes:

--Revised TS definition of Leakage,
--Revised TS 3.4.13, ``RCS [Reactor Coolant System] Operational 
Leakage,''
--Added new TS 3.4.17, ``Steam Generator Tube Integrity,''
--Revised TS 5.7.2.12, ``Steam Generator (SG) Tube Surveillance 
Program,'' and
--Revised TS 5.9.9, ``SG Tube Inspection Report.''

The proposed changes are necessary in order to implement the guidance 
for the industry initiative on NEI 97-06, ``Steam Generator Program 
Guidelines.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments 
adopting TSTF-449, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on May 6, 2005 
(70 FR 24126). The licensee affirmed the applicability of the following 
NSHC determination in its application dated December 15, 2005.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change requires a SG Program that includes performance 
criteria that will provide reasonable assurance that the SG tubing will 
retain integrity over the full range of operating conditions (including 
startup, operation in the power range, hot standby, cooldown and all 
anticipated transients included in the design specification). The SG 
performance criteria are based on tube structural integrity, accident 
induced leakage, and operational LEAKAGE.
    A steam generator tube rupture (SGTR) event is one of the design 
basis accidents that are analyzed as part of a plant's licensing basis. 
In the analysis of a SGTR event, a bounding primary to secondary 
LEAKAGE rate equal to the operational LEAKAGE rate limits in the 
licensing basis plus the LEAKAGE rate associated with a double-ended 
rupture of a single tube is assumed.
    For other design basis accidents such as a main steamline break 
(MSLB), rod ejection, and reactor coolant pump locked rotor the tubes 
are assumed to retain their structural integrity (i.e., they are 
assumed not to rupture). These analyses typically assume that primary 
to secondary LEAKAGE for all SGs is 1 gallon per minute or increases to 
1 gallon per minute as a result of accident induced stresses. The 
accident induced leakage criterion introduced by the proposed changes 
accounts for tubes that may leak during design basis accidents. The 
accident induced leakage criterion limits this leakage to no more than 
the value assumed in the accident analysis.
    The SG performance criteria proposed change to the TS identify the 
standards against which tube integrity is to be measured. Meeting the 
performance criteria provides reasonable assurance that the SG tubing 
will remain capable of fulfilling its specific safety function of 
maintaining reactor coolant pressure boundary integrity throughout each 
operating cycle and in the unlikely event of a design basis accident. 
The performance criteria are only a part of the SG Program required by 
the

[[Page 15490]]

proposed change to the TS. The program, defined by NEI 97-06, Steam 
Generator Program Guidelines, includes a framework that incorporates a 
balance of prevention, inspection, evaluation, repair, and leakage 
monitoring. The proposed changes do not, therefore, significantly 
increase the probability of an accident previously evaluated.
    The consequences of design basis accidents are, in part, functions 
of the DOSE EQUIVALENT I-131 in the primary coolant and the primary to 
secondary LEAKAGE rates resulting from an accident. Therefore, limits 
are included in the plant technical specifications for operational 
leakage and for DOSE EQUIVALENT I-131 in primary coolant to ensure the 
plant is operated within its analyzed condition. The typical analysis 
of the limiting design basis accident assumes that primary to secondary 
leak rate after the accident is 1 gallon per minute and that the 
reactor coolant activity levels of DOSE EQUIVALENT I-131 are at the TS 
values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the requirements 
for SG inspections. The proposed change does not adversely impact any 
other previously evaluated design basis accident and is an improvement 
over the current TSs.
    Therefore, the proposed change does not affect the consequences of 
a SGTR accident and the probability of such an accident is reduced. In 
addition, the proposed changes do not affect the consequences of an 
MSLB, rod ejection, or a reactor coolant pump locked rotor event, or 
other previously evaluated accident.

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

    The proposed performance based requirements are an improvement over 
the requirements imposed by the current technical specifications. 
Implementation of the proposed SG Program will not introduce any 
adverse changes to the plant design basis or postulated accidents 
resulting from potential tube degradation. The result of the 
implementation of the SG Program will be an enhancement of SG tube 
performance. Primary to secondary LEAKAGE that may be experienced 
during all plant conditions will be monitored to ensure it remains 
within current accident analysis assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility of a 
new or different type of accident from any accident previously 
evaluated.

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

    The SG tubes in pressurized water reactors are an integral part of 
the reactor coolant pressure boundary and, as such, are relied upon to 
maintain the primary system's pressure and inventory. As part of the 
reactor coolant pressure boundary, the SG tubes are unique in that they 
are also relied upon as a heat transfer surface between the primary and 
secondary systems such that residual heat can be removed from the 
primary system. In addition, the SG tubes isolate the radioactive 
fission products in the primary coolant from the secondary system. In 
summary, the safety function of an SG is maintained by ensuring the 
integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the SG Program to manage SG tube inspection, 
assessment, repair, and plugging. The requirements established by the 
SG Program are consistent with those in the applicable design codes and 
standards and are an improvement over the requirements in the current 
TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the TS.
    The NRC staff proposes to determine that the amendments request 
involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: December 12, 2005.
    Brief description of amendments: The amendments requested would 
revise Technical Specification (TS) 3.3.1, ``RTS [Reactor Trip System] 
Instrumentation,'' Surveillance Requirements (SRs) 3.3.1.2 and SR 
3.3.1.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No. Overall protection system performance will remain 
within the bounds of the previously performed accident analyses since 
there are no hardware changes. The Reactor Trip System (RTS) 
Instrumentation will be unaffected. Protection systems will continue to 
function in a manner consistent with the plant design basis. All 
design, material, and construction standards that were applicable prior 
to the request are maintained.
    The probability and consequences of accidents previously evaluated 
in the Final Safety Analysis Report (FSAR) are not adversely affected 
because the change to the daily surveillance for the normalization of 
the Nuclear Instrumentation System (NIS) Power Range and Nitrogen-16 
(N-16) Power Monitor indications assures the conservative response of 
the channel even at reduced power levels.
    The proposed changes will not affect the probability of any event 
initiators. There will be no degradation in the performance of, or an 
increase in the number of challenges imposed on, safety-related 
equipment assumed to function during an accident situation. There will 
be no change to normal plant operating parameters or accident 
mitigation performance.
    The proposed changes will not alter any assumptions or change any 
mitigation actions in the radiological consequence evaluations in the 
FSAR.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. There are no hardware changes nor are there any 
changes in the method by which any safety-related plant system performs 
its safety function. This amendment will not affect the normal method 
of plant operation or change any operating parameters. No performance 
requirements or response time limits will be affected.

[[Page 15491]]

    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result of 
this amendment. There will be no adverse effect or challenges imposed 
on any safety-related system as a result of this amendment.
    Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No. The proposed changes require a revision to the 
criteria for implementation of NIS Power Range and N-16 Power Monitor 
indication adjustments based on secondary power calorimetric 
calculations; however, the changes do not eliminate any RTS 
surveillances or alter the frequency of surveillances required by the 
TS. The revision to the criteria for implementation of the daily 
surveillance will remove a requirement for normalization of the NIS 
Power Range and N-16 Power Monitor indications at reduced power 
conditions that could result in safety performance outside the bounds 
of the safety analyses. Therefore, the Nominal Trip Setpoints and 
Allowable Values for the Reactor Trip System functions, as specified in 
the TS and related Bases, as well as the safety analysis limits assumed 
in the transient and accident analyses, are unchanged. None of the 
acceptance criteria for any accident analysis is changed.
    There will be no effect on the manner in which safety limits or 
limiting safety systems settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment of 
protection functions. There will be no impact on the overpower limit, 
departure from nucleate boiling ratio (DNBR) limits, heat flux hot 
channel factor (FQ), nuclear enthalpy rise hot channel 
factor (F[Delta]H), loss of coolant accident peak cladding temperature 
(LOCA PCT), peak local power density, or any other margin of safety. 
The radiological dose consequences are unaffected by this proposed 
change.
    The imposition of appropriate surveillance testing requirements 
will not reduce any margin of safety since the changes will assure that 
safety analysis assumptions on equipment operability are verified on a 
periodic frequency.
    Therefore the proposed change does not involve a reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: David Terao.

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

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

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

    Date of amendment request: February 14, 2006.
    Description of amendment request: Revise the Technical 
Specifications regarding the Containment Ventilation System to allow 
additional corrective actions for inoperable containment purge supply 
and exhaust valves.
    Date of publication of individual notice in the Federal Register: 
March 1, 2006 (71 FR 10566).
    Expiration date of individual notice: March 15, 2006.

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 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, 
http://www.nrc.gov/reading-rm/adams.html. 
If you do not have access 
to ADAMS or if there are problems in accessing the documents 
located in ADAMS, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

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

    Date of application for amendment: June 20, 2005, as supplemented 
by letter dated November 2, 2005.
    Brief description of amendment: This amendment revises the 
footnotes in Tables 3.4-2 and 4.4-3 of Technical Specification (TS) 3/
4.4.7 by increasing the temperature limit above which (1) reactor 
coolant sampling and analysis for dissolved oxygen is required, and (2) 
when limit for dissolved oxygen, specified in TS 4.4.7, applies. This 
temperature limit will be increased from 180 [deg]F to 250 [deg]F.
    Date of issuance: March 8, 2006.
    Effective date: March 8, 2006.
    Amendment No. 120.

[[Page 15492]]

    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 11, 2005 (70 FR 
59084). The supplemental letter provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 8, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 1, 2005, as 
supplemented by letters dated December 22, 2005, and January 23, 2006.
    Brief description of amendment: This amendment revises the 
technical specification (TS) requirements for pressurized-water reactor 
Boraflex fuel storage racks and adds TS requirements for fuel storage 
pool boron concentration. Specifically, the amendment (1) adds a new TS 
3/4.7.14, ``Fuel Storage Pool Boron Concentration,'' with a Limiting 
Condition for Operation that requires a fuel pool boron concentration 
of at least 2000 ppm at all times, (2) revises and reformats TS 5.6.1 
to specify the design features and fuel storage limitations in 
accordance with the categorization of spent fuel storage racks in 
various spent fuel pools, and (3) revises TS 5.3.1 to remove the cross-
reference to TS 5.6.1.b.
    Date of issuance: March 10, 2006.
    Effective date: March 10, 2006.
    Amendment No. 121.
    Facility Operating License No. NPF-63: Amendment revises the TS.
    Date of initial notice in Federal Register: November 8, 2005 (70 FR 
67745). The supplemental letters provided additional information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 10, 2006.
    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 amendments: February 25, 2005.
    Brief description of amendments: The amendments made various 
administrative changes to the Technical Specifications (TSs).
    Date of issuance: March 16, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 291 and 229.
    Facility Operating License Nos. DPR-65 and NPF-49: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: March 29, 2005 (70 FR 
15942).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 16, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 17, 2004.
    Brief description of amendments: The amendments revised Appendix B, 
Environmental Protection Plan (non-radiological), of the LaSalle County 
Facility Operating Licenses.
    Date of issuance: March 8, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 176/162.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Environmental Protection Plan.
    Date of initial notice in Federal Register: April 12, 2005 (70 FR 
19115).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 8, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 15, 2005.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.3.2.2, ``Feedwater System and Main Turbine High 
Water Level Trip Instrumentation,'' to reflect a design change in the 
instrumentation logic that trips the three feedwater pumps and main 
turbine.
    Date of issuance: March 9, 2006.
    Effective date: As of the date of issuance and shall be implemented 
prior to start-up from the spring 2006 refueling outage for Unit 2 and 
prior to start-up from the spring 2007 refueling outage for Unit 1.
    Amendment Nos.: 330/225.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications and Surveillance Requirements.
    Date of initial notice in Federal Register: August 30, 2005 (70 FR 
51381).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 9, 2006.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al. (FENOC), Docket No. 50-
346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: August, 20, 2004, as 
supplemented by letters dated June 16 and December 6, 2005.
    Brief description of amendment: The amendment revised TS 3/
4.8.1.1,``A.C. Sources--Operating,'' by deleting Surveillance 
Requirement (SR) 4.8.1.1.2.d.4, which requires verification that the 
emergency diesel generator auto-connected loads do not exceed the 2000-
hour load limit. In addition, the amendment revised TS 4/3.8.1.2, 
``A.C. Sources--Shutdown,'' to add exceptions to SR 4.8.1.2 when 
performed in Modes 5 and 6. As a result of discussions held on October 
20, 2005, FENOC decided to withdraw the portion of the amendment 
request (LAR 01-0009) that requested clarification of SR 4.8.1.1.b.
    Date of issuance: March 2, 2006
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 273.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 28, 2004 (69 
FR 57989).
    The June 16 and December 6, 2005, supplements, contained clarifying 
information and did not change the NRC staff's initial proposed finding 
of no significant hazards consideration or expand the scope of the 
original Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 2, 2006.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: May 2, 2005, as supplemented by 
letters

[[Page 15493]]

dated August 28, September 15, 2005, and January 12, 2006, and January 
13, February 9, and February 28, 2006.
    Brief description of amendment: This amendment revised the 
Technical Specifications (TSs) Section 2.1.1, ``Safety Limits--Reactor 
Core,'' and TS Section 2.2.1, ``Limiting Safety Settings--Reactor 
Protection System Setpoints.'' The amendment supports the use of the 
Framatome Mark B-HTP fuel design for Cycle 15, which is scheduled to 
begin following the refueling outage in March 2006.
    Date of issuance: March 2, 2006
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 274.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 24, 2005 (70 FR 
29796).
    The August 28, September 15, 2005, and January 12, January 13, 
February 9, and February 28, 2006, supplements, contained clarifying 
information and did not change the NRC staff's initial proposed finding 
of no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 2, 2006.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 25, 2005.
    Brief description of amendment: The amendment revised the 
definitions of Channel Calibration, Channel Function Test, and Logic 
System Functional Test in accordance with the Technical Specification 
Task Force Traveler (TSTF)-205-A.
    Date of issuance: March 10, 2006
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 217.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 11, 2005 (70 FR 
59086).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 10, 2006
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 11, 2005.
    Brief description of amendment: The amendment deleted the 
surveillance requirement (SR) of TS 2.10.2(9)b(iii) to verify shutdown 
margin every 8-hour shift during low power physics testing. This change 
made TS 2.10.2(9)b more consistent with SR 3.1.7 of NUREG-1432, 
``Standard Technical Specifications Combustion Engineering Plants, 
Revision 3.'' In addition, the Containment Structural Tests Report has 
been deleted from TS 5.9.3c and several administrative and editorial 
changes were made.
    Date of issuance: February 1, 2006.
    Effective date: February 1, 2006 and shall be implemented within 60 
days from the date of issuance.
    Amendment No.: 237.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 27, 2005 (70 
FR 56503)
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated February 1, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 27, 2005, as supplemented 
on December 1, 2005, and February 28, 2006.
    Brief description of amendments: These amendments change the SSES 1 
and 2 technical specifications for reactor protection system and 
control rod block instrumentation, oscillation power range monitor 
instrumentation, recirculation loops operating, shutdown margin test--
refueling, and the core operating limits report. The amendments modify 
the power range neutron monitor system (PRNMS) by installation of the 
General Electric Nuclear Measurement Analysis and Control PRNMS. The 
modification of the PRNMS replaces analog technology with a digital 
upgrade.
    Date of issuance: March 3, 2006
    Effective date: As of the date of issuance and to be implemented 
prior to startup following the Cycle 14 refueling outage for Unit 1 and 
the Cycle 13 refueling outage for Unit 2.
    Amendment Nos.: 230 and 207.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 13, 2005 (70 
FR 54088).
    The supplements dated December 1, 2005, and February 28, 2006, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 3, 2006.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric 
Station, Unit 2 (SSES 2), Luzerne County, Pennsylvania

    Date of application for amendment: March 18, 2005, as supplemented 
on February 28, 2006.
    Brief description of amendment: The amendment revises the SSES 2 
Technical Specification 3.3.8.1, ``Loss of Power (LOP) 
Instrumentation,'' to (1) clarify that Condition A applies to the LOP 
instrumentation associated with both the Unit 1 and Unit 2 4.16 
Kilovolt (kV) Engineered Safeguards System (ESS) buses since both the 
Unit 1 and Unit 2 buses are required to support Unit 2 operation, (2) 
add a new Condition B to allow the LOP instrumentation for two Unit 1 
4.16kV ESS buses in the same division to be inoperable for up to 8 
hours for the performance of Surveillance Requirement 3.8.1.19 on Unit 
1. In addition, the amendment revises the SSES 2 TS 3.8.7, 
``Distribution Systems--Operating,'' to (1) eliminate ``or more'' and 
the plural to ``subsystems'' such that the condition will read ``one 
Unit 1 AC [alternating current] electrical power distribution subsystem 
inoperable,'' and (2) add a new Condition D for two Unit 1 AC 
electrical power distribution subsystems inoperable.
    Date of issuance: March 16, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 208.
    Facility Operating License No. NPF-22: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 24, 2005 (70 FR 
29800).
    The supplement dated February 28, 2006, provided additional 
information that clarified the application, did not expand the scope of 
the application as

[[Page 15494]]

originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 16, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 4, 2005.
    Brief description of amendment: The amendment relocated the 
Transversing In-Core Probe (TIP) system Technical Specification (TS) to 
the Hope Creek Generating Station Updated Final Safety Analysis Report, 
as well as removed the note on the TIP system from the Reactor 
Protection System Instrumentation Surveillance Requirements table.
    Date of issuance: March 8, 2006.
    Effective date: As of the date of issuance, to be implemented 
within 60 days from date of issuance.
    Amendment No.: 164.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: March 15, 2005 (70 FR 
12750).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 8, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 29, 2005, as supplemented 
on July 1 and November 21, 2005.
    Brief description of amendment: The amendment revised Technical 
Specification 3.7.3, ``Main Feedwater Regulating Valves (MFRVs), 
Associated Bypass Valves, and Main Feedwater Pump Discharge Valves 
(MFPDVs),'' to allow the use of the main feedwater isolation valves in 
lieu of the MFPDVs to provide isolation capability to the steam 
generators in the event of a steam line break.
    Date of issuance: March 16, 2006
    Effective date: As of the date of issuance to be implemented prior 
to startup from the fall 2006 refueling outage.
    Amendment No.: 95.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: June 7, 2005 (70 FR 
33218).
    The July 1 and November 21, 2005, 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 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 16, 2006.
    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: August 31, 2005.
    Brief description of amendment: The amendment revises the Technical 
Specifications associated with steam generator tube integrity 
consistent with Revision 4 to Technical Specification Task Force (TSTF) 
Standard Technical Specification Change Traveler, TSTF-449, ``Steam 
Generator Tube Integrity.''
    Date of issuance: February 23, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 306.
    Facility Operating License No. DPR-77: Amendment revises the 
technical specifications.
    Date of initial notice in Federal Register: November 22, 2005 (70 
FR 70643).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 23, 2006.
    No significant hazards consideration comments received: No.

TXU Generation Company LP, Docket No. 50-446, Comanche Peak Steam 
Electric Station, Unit No. 2, Somervell County, Texas

    Date of amendment request: April 27, 2005, as supplemented by 
letter dated July 20, 2005.
    Description of amendment: The amendment revises the Technical 
Specifications to add Topical Report WCAP-13060-P-A to the list of NRC 
approved methodologies to be used at Comanche Peak Steam Electric 
Station, Unit 2.
    Date of issuance: March 15, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 123.
    Facility Operating License No. NPF-89: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 8, 2005 (70 FR 
67753).
    No significant hazards consideration comments received: No.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: October 6, 2004, as supplemented by 
letters dated September 16 and November 22, 2005.
    Brief description of amendments: The amendments revised the 
Technical Specification 3.8.1, ``AC Sources--Operating,'' to remove 
mode restrictions on surveillance requirements.
    Date of issuance: March 15, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 124.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 15, 2005 (70 FR 
12751).
    The supplements dated September 16 and November 22, 2005, 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 March 15, 2006.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 20th day of March 2006.

    For the Nuclear Regulatory Commission.
Edwin M. Hackett,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 06-2908 Filed 3-27-06; 8:45 am]
BILLING CODE 7590-01-P