Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 15479-15494 [06-2908]

Agencies

• NUCLEAR REGULATORY COMMISSION

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


-----------------------------------------------------------------------

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, 
https://www.nrc.gov/reading-rm/doc-collections/cfr/. 
If a request 
for a hearing or petition for leave to intervene is filed within 
60 days, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic 
Safety and Licensing Board Panel, will rule on the request 
and/or petition; and the Secretary or the Chief Administrative 
Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to

[[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, HearingDocket@nrc.gov; 
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 
OGCMailCenter@nrc.gov. 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, 
https://www.nrc.gov/reading-rm/adams.html. 
If you do not have access 
to ADAMS or if there are problems in accessing the documents 
located in ADAMS, contact the PDR Reference staff at 
1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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