Involving No Significant Hazards Considerations, 56499-56509 [05-19028]

Agencies

• Biweekly Notice; Applications and Amendments to Facility Operating Licenses
• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 186 (Tuesday, September 27, 2005)]
[Notices]
[Pages 56499-56509]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-19028]


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NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility Operating 
Licenses


Involving No Significant Hazards Considerations

I. Background

    Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from September 1, 2005, to September 15, 2005. 
The last biweekly notice was published on September 13, 2005 (70 FR 
54085).

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 
01F21, 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

[[Page 56500]]

property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    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 email 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 email to pdr@nrc.gov.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina; Docket 
Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, 
and 3, Oconee County, South Carolina; Docket Nos. 50-369 and 50-370, 
McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North 
Carolina

    Date of amendment request: August 5, 2005.
    Description of amendment request: The proposed amendments request 
NRC consent to the indirect transfer of control of the licenses for the 
Oconee Nuclear Station, Units 1, 2, and 3, the McGuire Nuclear Station, 
Units 1 and 2, and the Catawba Nuclear Station, Units 1 and 2. The 
transfers of control will result from the creation of a new holding 
company that will become the parent of the current licensee. The new 
holding company, to be named Duke Energy Corporation, that will result 
from the business combination of Duke Energy with Cinergy Corporation 
(Cinergy). The licensee, current Duke Energy, will convert to a limited 
liability company (LLC) and be re-named Duke Power Company LLC (Duke 
Power).
    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 Conforming Amendments Do Not Involve a Significant Increase in 
the Probability or Consequences of an Accident Previously Evaluated

    The amendments do not involve any change in the design, 
configuration, or operation of the nuclear units. All Limiting 
Conditions for Operation, Limiting Safety System Settings and Safety 
Limits specified in the Technical Specifications remain unchanged. 
Also, the Physical Security Plans and related plans, the Operator 
Training and Requalification Programs, the Quality Assurance 
Programs, and the Emergency Plans will not be materially changed by 
the proposed license transfers and amendments.
    The technical qualifications of the operating licensee will not 
be reduced. Personnel engaged in operation, maintenance, 
engineering, assessment, training, and other related services will 
not be changed. The Duke Energy officers and executives currently 
responsible for the overall safe operation of the nuclear plants are 
expected to continue in the same capacity.

[[Page 56501]]

2. The Conforming Amendments Do Not Create the Possibility of a New or 
Different kind of Accident From Any Accident Previously Evaluated

    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 Operations, Limit Safety System 
Settings and Safety Limits specified in the Technical Specifications 
are not affected by the proposed changes. As such, the plant 
conditions for which the design basis accident analyses were 
performed remain valid.
    The 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.

3. The Conforming Amendments Do Not Involve a Significant Reduction in 
a Margin of Safety

    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 bases.
    Plant safety margins are established through Limiting Conditions 
for Operation, Limiting Safety System Settings and Safety Limits 
specified in the Technical Specifications. Because there is no 
change to the physical design of the plant, there is no change to 
any of these margins.

    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 Section Chief: Evangelos C. Marinos.

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

    Date of amendment request: August 17, 2005.
    Description of amendment request: The proposed amendment would 
allow a one-time extension of the 72-hour Completion Time for the 
Required Action of Condition B of Technical Specification 3.7.1, 
``Standby Service Water (SW) System and Ultimate Heat Sink (UHS),'' and 
a one-time exemption from Note 1 of part B.1 of this Required Action. 
Specifically, the proposed one-time extension request is for an 
additional 72 hours to the Completion Time and would result in a 144-
hour Completion Time for an inoperable SW subsystem. This would allow 
extensive maintenance to be conducted on the SW train B pump, not 
capable of being completed in the current 72-hour Completion Time.
    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.
    Response: No.
    Since only one subsystem of SW components is affected by the 
condition and an additional failure is not considered while a plant 
is in a Limiting Condition for Operation (LCO) Action, the operable 
SW subsystem is adequate to maintain compliance with the plant's 
design basis. Thus, this condition will not alter assumptions 
relative to the mitigation of an accident or transient event.
    Energy Northwest has determined that there is no significant 
risk associated with the operation of the plant for an additional 3 
days with one SW subsystem out of service. The incremental change in 
risk has been quantitatively evaluated using the guidance of 
Regulatory Guide [RG] 1.174 and 1.177. The incremental risk values 
are within the criteria of Region III (where the increase in risk is 
considered ``very small'') as established in RG 1.174.
    Based on this evaluation, there is no significant increase in 
the probability or consequence 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.
    Response: No.
    This proposed action only extends the CT [Completion Time] and 
will not physically alter the plant. No new or different type of 
equipment will be installed by this action. The changes in methods 
governing normal plant operation are consistent with current safety 
analysis assumptions. No change to the system as evaluated in the 
Columbia Generating Station safety analysis is proposed. Therefore, 
this proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Response: No.
    Columbia is designed with sufficient redundancy such that a SW 
subsystem may be removed from service for maintenance or testing. 
The remaining subsystem is capable of providing water and removing 
heat loads to satisfy the UFSAR [Updated Final Safety Analysis 
Report] requirements for accident mitigation or unit safe shutdown.
    A risk-informed evaluation concluded that the risk contribution 
of the CT extension is non-risk significant.
    There will be no change to the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
change to those plant systems necessary to assure the accomplishment 
of protection functions. For these reasons, the proposed amendment 
does not involve a significant reduction in a margin of safety.
    Based upon the analysis provided herein, the proposed amendments 
do not involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Daniel S. Collins, Acting.

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

    Date of amendment request:
    July 27, 2005.
    Description of amendment request: The proposed change would revise 
technical specification (TS) 3/4.8.1.1, ``A. C. Sources--Operating,'' 
to adopt a more recent standard for diesel fuel oil testing and remove 
the restriction that certain surveillance requirements be performed 
during shutdown.
    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 to TS SR [surveillance requirement] 
4.8.1.1.2.b affects the testing standard for the fuel oil supply for 
the emergency diesel generators (EDGs). The fuel oil supply is not 
an initiator of any accident previously evaluated. The fuel oil 
supply supports the accident mitigation functions of the EDGs, which 
serve as the standby source for A.C. power in the event of a loss of 
offsite power. Adoption of a more recent standard does not affect 
the capability of the diesel fuel oil to perform its required 
function. Therefore, the proposed change to

[[Page 56502]]

SR 4.8.1.1.2.b does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to TS SR 4.8.1.1.2.d affects the performance 
of load rejection testing and the 60-minute loaded run at greater 
than or equal to 2000 kW. Evaluations and operating history have 
demonstrated that performance of these tests online will not impact 
electrical distribution system reliability. No anticipated 
operational occurrence or accident would occur as a result of 
performing these tests online. Although the EDGs are rendered 
inoperable and unavailable during performance of these tests, these 
tests would be performed in conjunction with testing required by 
other specifications; therefore, the accumulated time of EDG 
inoperability and unavailability would not increase. The proposed 
change to SR 4.8.1.1.2.d 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 to TS SR 4.8.1.1.2.b affects the testing 
standard for the fuel oil supply for EDGs. Applying the more recent 
standard for fuel oil testing does not create any new or different 
accident initiators because adoption of a more recent standard does 
not affect the capability of the diesel fuel oil to perform its 
required function.
    The proposed change to TS SR 4.8.1.1.2.d affects the performance 
of load rejection testing and the 60-minute loaded run at greater 
than or equal to 2000 kW. Evaluations and operating experience have 
demonstrated that performance of these tests online will not impact 
electrical distribution system reliability. No new or different 
accidents could occur as a result of performing these tests online. 
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 proposed change to TS SR 4.8.1.1.2.b affects the testing 
standard for the fuel oil supply for EDGs. Adoption of a more recent 
standard does not affect the capability of the diesel fuel oil to 
perform its required function.
    The proposed change to TS SR 4.8.1.1.2.d affects the performance 
of load rejection testing and the 60-minute loaded run at greater 
than or equal to 2000 kW. Evaluations and operating experience have 
demonstrated that performance of these tests online regardless of 
the test outcome will not impact electrical distribution system 
reliability. The required testing will continue to demonstrate 
acceptable EDG performance. 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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Gene Y. Suh.

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

    Date of amendment request: July 27, 2005.
    Description of amendment request: The proposed change would revise 
technical specification (TS) 3/4.10.2, ``Special Test Exceptions--
Physics Tests,'' to increase the allowed time between the flux channel 
Channel Functional Tests and the beginning of Mode 2 Physics Tests from 
12 hours to 24 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. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The changes affect the Limiting Condition for Operation for 
``Special Test Exceptions--Physics Tests,'' in particular, the 
neutron flux instrumentation CHANNEL FUNCTIONAL TEST that must 
precede PHYSICS TESTING in MODE 2. The neutron flux instrumentation 
is not an accident initiator, but is credited for two events. These 
events are Uncontrolled Control Rod Assembly Group Withdrawal From a 
Subcritical Condition (Startup Accident), and Uncontrolled Control 
Rod Assembly Group Withdrawal at Power. The proposed change will not 
impact the operation of the neutron flux instrumentation during 
these events. Consequently, the proposed changes will have no impact 
on the probability or consequences of any accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The changes affect the Limiting Condition for Operation for 
Refueling Operations--Instrumentation, in particular, the neutron 
flux instrumentation. The changes are only applicable in MODE 2. 
Under the proposed change, the neutron flux instrumentation will 
continue to operate in the same manner as previously considered. 
Accident initial conditions and assumptions remain as previously 
analyzed.
    The proposed changes do not introduce any new or different 
accident initiators. In addition, the requested increase in the 
allowed time between the flux channel Channel Functional Tests and 
the beginning of Mode 2 Physics Tests from 12 hours to 24 hours will 
not adversely impact the instrumentation's stability or capability. 
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?
    Response: No.
    The changes affect the Limiting Condition for Operation for 
Refueling Operations--Instrumentation; in particular, the neutron 
flux instrumentation. The proposed changes to TS will not result in 
design changes to the neutron flux instrumentation or in changes to 
how the neutron flux instrumentation is used. As discussed in the 
response to question 1 above, channel operability will 
continue to be ensured by the CHANNEL CHECK and CHANNEL CALIBRATION 
requirements of TS 4.3.1.1.1. Therefore, the proposed change will 
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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Gene Y. Suh.

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

    Date of amendment request: August 11, 2005.
    Description of amendment request: The proposed amendment would 
revise Figure 2-3 in Technical Specification (TS) 2.3(4) and related 
technical information to this figure in the Basis of TS 3.6. This 
figure shows the minimum volume of Tri-sodium Phosphate (TSP) required 
for a specified reactor coolant system (RCS) hot zero power (HZP) 
critical boron concentration (CBC) over the operating cycle. 
Maintaining a volume of TSP in the baskets that is within the area of 
acceptable operation of Figure 2-3 ensures that the recirculation water 
in the containment sump attains a pH of 7.0 or greater following a 
loss-of-coolant accident (LOCA). This figure allows the required volume 
of TSP to gradually decrease as HZP CBC decreases during the operating 
cycle. As HZP CBC decreases, less TSP is required to attain a pH of 7.0 
or greater in the containment sump. Also, TS 3.6(2) is being revised to 
remove the term Dodecahydrate to be consistent with Fort Calhoun 
Station TS Amendment No. 232.

[[Page 56503]]

    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.
    There are no changes to the design or operation of the plant 
that could affect system, component, or accident functions as a 
result of revising the current volume of active TSP required during 
Operating Modes 1 and 2 with a new figure that reflects the future 
RCS volume change. All systems and components function as designed, 
and the performance requirements have been evaluated and found to be 
acceptable.
    Allowing the required volume of active TSP to decrease over the 
operating cycle as HZP CBC decreases will ensure a pH of 7.0 or 
greater in the containment sump following a LOCA, yet provides [an] 
adequate margin for EEQ [environmental equipment qualification] 
concerns as containment sump pH is less likely to exceed 8.0. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of the proposed change. All 
systems, structures, and components previously required for 
mitigation of an event remain capable of fulfilling their intended 
design function with this change to the TS.
    The proposed change has no adverse effects on any safety-related 
systems or component and does not challenge the performance or 
integrity of any safety-related system. The proposed change has 
evaluated the TSP configuration such that no new accident scenarios 
or single failures are introduced. 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.
    Allowing the required volume of active TSP to decrease as HZP 
CBC decreases still ensures a pH of 7.0 or greater in the 
containment sump following a LOCA and still provides [an] adequate 
margin for EEQ concerns as containment sump pH is less likely to 
exceed 8.0. Therefore, this change does not involve a significant 
reduction in the margin of safety.
    Evaluations were made that indicate that the margin for pH 
control is not altered by the proposed changes. A TSP volume that is 
dependent on HZP CBC has been evaluated with respect to 
neutralization of all borated water and acid sources. These 
evaluations concluded that there would be no impact on pH control, 
and hence, no reduction in the margin of safety related to post LOCA 
conditions.
    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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Daniel S. Collins, Acting.

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

    Date of amendment request: August 11, 2005.
    Description of amendment request: The proposed amendment includes 
various changes to the Technical Specifications (TS). Specifically, 
Omaha Public Power District (OPPD) seeks to delete the surveillance 
requirement (SR) of TS 2.10.2(9)b(iii) to verify the shutdown margin 
every 8-hour shift during low power physics testing. This change will 
make TS 2.10.2(9)b more consistent with SR 3.1.7 of NUREG-1432, 
Standard Technical Specifications--Combustion Engineering Plants, 
Revision 3.
    The Containment Structural Tests Report of TS 5.9.3c is proposed 
for deletion. Amendment No. 216 deleted TS 3.5(5), which required 
submittal of the TS 5.9.3c report. The deletion of the report and the 
remaining changes described in Attachment 1 are considered 
administrative in nature.
    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 [change] involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    This license amendment request (LAR) makes no changes to the 
design or operation of the plant that could affect system, 
component, or accident functions.
    The deletion of Technical Specification (TS) 2.10.2(9)b(iii) 
eliminates the need to verify shutdown margin (SDM) every 8 hours 
during low power physics testing. Reactivity equivalent to at least 
the highest estimated CEA worth is available from the operable CEA 
[control element assembly] groups withdrawn (assuming the most 
reactive CEA of the groups withdrawn is stuck in the fully withdrawn 
position). Each CEA not fully inserted is demonstrated capable of 
full insertion when tripped from at least the 50% withdrawn position 
within 7 days of reducing SDM. Finally, the position of the 
trippable control element assemblies (CEAs) during low power physics 
testing continues to be verified every 2 hours. The SDM provided by 
the CEAs ensures that the operators can respond promptly to 
unexpected increases in core reactivity. Thus, this change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    Deletion of the Containment Structural Tests Report is not an 
initiator of any previously evaluated accidents. OPPD will continue 
to report conditions indicative of containment deterioration or 
degradation in the Inservice Inspection (ISI) Summary Report 
required by 10 CFR 50.55a, ASME [American Society of Mechanical 
Engineers] Section XI, Subsection IWA-6000, and TS 5.9.3a.
    The Commission has provided guidance concerning the application 
of standards for determining whether a significant hazards 
consideration exists by providing certain examples (48 FR 14864) of 
amendments that are considered not likely to involve significant 
hazards considerations. One or more of these examples are cited to 
justify deletion of the Containment Structural Tests Report and for 
each of the remaining administrative changes. Thus, these changes do 
not increase the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment [change] create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This proposed change affects only the TSs and does not involve a 
physical change to the plant. No modifications are made to existing 
components nor will any new or different type of equipment be 
installed. The deletion of the surveillance requirement (SR) to 
verify SDM every 8 hours during low power physics testing does not 
create the possibility of a new or different kind of accident.
    The SRs that remain ensure that the SDM provided by the CEAs is 
adequate and that the CEAs are capable of full insertion. CEA 
positions will continue to be verified at least once per [a] 2-hour 
interval during low power physics testing. The SDM provided by the 
CEAs ensures that the operators can respond promptly to unexpected 
increases in core reactivity.
    The deletion of a report that is redundant to federal 
regulations is an administrative change that does not create the 
possibility of a new or different kind of accident. OPPD will 
continue to report conditions indicative of containment 
deterioration or degradation in the ISI Summary Report.
    The remaining changes proposed by this LAR are administrative in 
nature. These changes do not impose different

[[Page 56504]]

requirements and do not alter assumptions made in the safety 
analysis and licensing basis. Therefore, they do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment [change] involve a significant 
reduction in a margin of safety?
    Response: No.
    The proposed changes do not affect any safety analysis 
assumptions. During low power physics testing, the position of the 
trippable CEAs will continue to be verified at 2-hour intervals. The 
deleted 8-hour SDM surveillance requirement is performed less 
frequently, is redundant and unnecessary. The SDM provided by the 
CEAs ensures that the operators can respond promptly to unexpected 
increases in core reactivity. The Containment Structural Tests 
Report can be deleted since OPPD will continue to report conditions 
indicative of containment deterioration or degradation in accordance 
with 10 CFR 50.55a in the ISI Summary Report required by TS 5.9.3a.

    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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Daniel S. Collins, Acting.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station (VCSNS), 
Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: June 22, 2005.
    Description of amendment request: The proposed changes would revise 
the reactor coolant system heatup and cooldown curves located in 
Technical Specification (TS) section 3/4.4.9 to reflect the results of 
the last reactor vessel surveillance specimen that was removed from the 
reactor vessel and analyzed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise the P/T [pressure/temperature] limit 
curves to provide figures that reflect the results of the analysis 
performed on reactor vessel surveillance specimen Z. This analysis 
was performed using NRC approved methodology as documented in WCAP 
14040-NP-A, Revision 4, utilizing the 1998 ASME [American Society of 
Mechanical Engineers] Code, Section XI through the 2000 addenda, 
Appendix G requirements. These curves provide the limits for 
operation of the Reactor Coolant System during heatup, cooldown, 
criticality, and hydrostatic testing. These curves are provided 
without instrument uncertainties included, however, the 
uncertainties are included in the curves provided in the plant 
operating procedures. The limits protect the reactor vessel from 
brittle fracture by separating the region of acceptable operation 
from the region where brittle fracture is postulated to occur. 
Failure of the reactor vessel is not a VCSNS design basis accident, 
and, in general, reactor vessel failure has a low probability of 
occurrence and is not considered in the safety analysis. Therefore, 
the 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 changes revises the P/T limits curves, Section 3/
4.4.9, to incorporate the results fo the analysis performed on 
reactor vessel specimen Z. There are no physical plant design 
changes or significant changes in any operating procedures. This 
change adjusts the heatup and cooldown curves to reflect the shift 
in nil-ductility reference temperature of the reactor vessel as a 
result of neutron embrittlement. Therefore, the change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed changes revise the P/T limits curves, Section 3/
4.4.9, to incorporate the results of the analysis performed on 
reactor vessel specimen Z. The new P/T curves ensure that the 10 CFR 
50 Appendix G, requirements are not exceeded during normal operation 
including Reactor Coolant System transients during heatup, cooldown, 
criticality and hydrostatic testing. The new P/T curves were 
prepared, using approved industry methodology, for a projected 
reactor vessel neutron exposure of 56 EFPY [effective full-power 
year]. The proposed P/T limit curves reflect a shift of the limits 
in a conservative direction from the current requirements. 
Therefore, the 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: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: Evangelos C. Marinos.

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

    Date of amendment request: July 29, 2005.
    Brief description of amendments: The proposed amendments revised 
the technical specification (TS) testing frequency for the surveillance 
requirement (SR) 3.1.4.2, control rod scram time testing, from 120 days 
cumulative operation in MODE 1 to 200 days cumulative operation in MODE 
1.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in licensing amendment applications in the Federal Register on August 
23, 2004 (69 FR 51864). The licensee affirmed the applicability of the 
model NSHC determination in its application dated July 29, 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:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The frequency 
of surveillance testing is not an initiator of any accident 
previously evaluated. The frequency of surveillance testing does not 
affect the ability to mitigate any accident previously evaluated, as 
the tested component is still required to be operable. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change does not result in any new or different modes of plant 
operation. 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?

[[Page 56505]]

    Response: No.
    The proposed change extends the frequency for testing control 
rod scram time testing from every 120 days of cumulative Mode 1 
operation to 200 days of cumulative Mode 1 operation. The proposed 
change continues to test the control rod scram time to ensure the 
assumptions in the safety analysis are protected. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Section Chief: Michael L. Marshall, Jr.

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

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

Duke 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: August 18, 2005.
    Brief description of amendment request: The proposed Technical 
Specification changes are needed to accommodate the replacement of the 
Reactor Building Emergency Sump suction inlet trash racks and screens 
with strainers.
    Date of publication of individual notice in Federal Register: 
August 31, 2005 (70 FR 51852).
    Expiration date of individual notice: September 30, 2005.

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

    Date of amendment request: January 27, 2005, revised by letter 
dated August 12, 2005.
    Brief description of amendment request: The proposed amendment 
would allow the licensee to utilize a probabilistic methodology to 
determine the contribution to main steamline break leakage rates for 
the once-through steam generator (OTSG) from the tube end crack (TEC) 
alternate repair criteria described in Improved Technical Specification 
(ITS) 5.6.2.10.2.f and also involves a change to ITS 5.6.2.10.2.f to 
incorporate the basis of the proposed probabilistic methodology and the 
method and technical justification for projecting the TEC leakage that 
may develop during the next operating cycle following the inservice 
inspection of each OTSG.
    Date of publication of individual notice in Federal Register: 
August 26, 2005 (70 FR 50424).
    Expiration date of individual notice: September 26, 2005.

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, 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.

Duke Energy Corporation, et al., Docket Nos. 50-413, 50-414, 50-369, 
and 50-370, Catawba Nuclear Station, Units 1 and 2, York County, South 
Carolina and McGuire Nuclear Station, Units 1 and 2, Mecklenburg 
County, North Carolina.

    Date of application for amendments: July 7, 2005.
    Brief description of amendments: The amendments revised the 
Technical Specifications TS 3.9.1, ``Boron Concentration,'' to clarify 
the technical requirements for boron concentration when the refueling 
canal and the refueling cavity are not connected to the reactor coolant 
system.
    Date of issuance: September 1, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 231/213 and 226/221.
    Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9 and 
NPF-17: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 2, 2005 (70 FR 
44401).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 1, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 20, 2004, as supplemented by 
letters dated June 6 and August 10, 2005.

[[Page 56506]]

    Brief description of amendment: The Amendment revised the safety 
analysis report (SAR) to allow the licensee the use of a lifting tripod 
(a special lifting device) to remove and install the reactor vessel 
(RV) head and certain RV internals during refueling outages, using the 
reactor building polar crane.
    Date of issuance: August 30, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance. The SAR changes shall be 
implemented in the next periodic update to the SAR in accordance with 
Paragraph 50.71(e) of Title 10 of the Code of Federal Regulations.
    Amendment No.: 225.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the SAR.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5242)
    The supplements dated June 6 and August 10, 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 amendment is contained 
in a Safety Evaluation dated August 30, 2005.
    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 11, 2005.
    Brief description of amendment: This amendment allows a one-time 
extension of the surveillance interval for the reactor vessel internals 
vent valves from September 2005 to March 2006.
    Date of issuance: September 6, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 14 days.
    Amendment No.: 268.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications/License.
    Date of initial notice in Federal Register: July 5, 2005 (70 FR 
38719).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 6, 2005.
    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: December 20, 2004, as 
supplemented by letter dated April 6, 2005.
    Brief description of amendment: This amendment revised Technical 
Specification (TS) 3/4.9.2, ``Refueling Operations--Instrumentation.'' 
Specifically, the changes revised TS 3/4.9.2 concerning source range 
flux monitors to be more consistent with improved Standard Technical 
Specifications. The changes achieve consistency with corresponding 
requirements in NUREG-1430, ``Standard Technical Specifications Babcock 
and Wilcox Plants,'' Revision 3, dated June 2004, with exceptions to 
account for plant-specific design differences and retention of current 
licensing basis requirements and commitments.
    Date of issuance: September 12, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 269.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 2005 (70 
FR 7765). The supplement dated April 6, 2005, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally notice, and did not change the NRC staff 
original proposed no significant hazards consideration determination as 
published in the Federal Register on February 15, 2005 (70 FR 7765).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 12, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 8, 2004.
    Brief description of amendment: The amendment deletes one-time use 
footnotes that have expired or have already been used from the Crystal 
River Unit 3 (CR-3) Improved Technical Specifications (ITS). 
Specifically, ITS 3.7.9, ``Nuclear Services Seawater System'' and ITS 
3.8.1, ``AC Sources--Operating (Emergency Diesel Generator)'' notes are 
removed. These changes are administrative in nature and do not alter 
any operating license requirements.
    Date of issuance: September 6, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 220.
    Facility Operating License No. DPR-72: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 3, 2004 (69 FR 
46585).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 6, 2005.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: September 21, 2004, as 
supplemented by letters dated March 18, April 7, May 6, and August 10, 
2005.
    Brief description of amendments: The amendments the 69 kV offsite 
power circuit limiting conditions for operation action statements. Add 
a license condition to extend the required action completion time for 
an inoperable alternate offsite power source (69 kV circuit) from the 
current 72 hours to 14 days on a one-time basis.
    Date of issuance: September 9, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment Nos.: 289, 271.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 26, 2004 (69 FR 
62476)
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 9, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 27, 2004, as 
supplemented by letter dated August 2, 2005.
    Brief description of amendments: The amendments revised Technical

[[Page 56507]]

Specifications (TSs) related to the reactor coolant pump flywheel 
inspection program by increasing the inspection interval to 20 years.
    Date of issuance: September 9, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 265 and 247.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: March 29, 2005 (70 FR 
15945). The licensee's supplement dated August 2, 2005, did not change 
the scope of the proposed amendment as described in the original notice 
of proposed action published in the Federal Register, and did not 
change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 9, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 26, 2004, as 
supplemented by letter dated July 18, 2005.
    Brief description of amendments: The amendments revise the 
following Technical Specifications (TSs): TS 4.2.1, ``Fuel 
Assemblies,'' adds reference to ZIRLO\TM\ clad fuel and filler rods; 
and TS 5.7.1.5, ``Core Operating Limits Report (COLR),'' adds the 
following references to the list of analytical methods used to 
determine the core operating limits: ``Calculative Methods for the CE 
Nuclear Power Large Break LOCA [loss-of-coolant accident] Evaluation 
Model,'' CENPD-132, Supplement 4-P-A, August 2000, and ``Implementation 
of ZIRLO\TM\ Cladding Material in CE Nuclear Power Fuel Assembly 
Designs,'' CENPD-404-P-A, November 2001. These changes were requested 
to implement ZIRLO\TM\ fuel rod cladding material into the fuel design 
for San Onofre Nuclear Generating Station, Units 2 and 3.
    Date of issuance: September 14, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days from the date of issuance.
    Amendment Nos.: Unit 2-199; Unit 3-190.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 28, 2004 (69 
FR 57991).
    The supplemental letter dated July 18, 2005, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 14, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 27, 2005.
    Brief description of amendments: The amendments revised the 
applicability of Technical Specification (TS) 3.4.3, Functional Unit 
18.A, ``Turbine Trip, Low Fluid Oil Pressure,'' and TS Functional Unit 
18.B, ``Turbine Trip, Turbine Stop Valve Closure,'' by altering Table 
3.3-1, ``Reactor Trip System Instrumentation,'' and Table 4.3-1, 
``Reactor Trip System Instrumentation Surveillance Requirements.'' The 
change adds a footnote that indicates that the Mode 1 applicability is 
limited to operation above the P-9 (50 percent rated thermal power) 
interlock setpoint value. Additionally, the action for an inoperable 
turbine stop valve closure channel is revised to be consistent with the 
design of this function. Finally, an option is added to permit a 
reduction in thermal power to below the P-9 interlock within 10 hours 
for an inoperable turbine stop valve closure channel.
    Date of issuance: September 2, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 304 and 294.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the technical specifications.
    Date of initial notice in Federal Register: July 5, 2005 (70 FR 
38722).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 2, 2005.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    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 for the 
amendment complies with the standards and requirements of