[Federal Register Volume 70, Number 224 (Tuesday, November 22, 2005)]
[Notices]
[Pages 70641-70646]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-22795]


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

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 October 28, 2005, to November 9, 2005. The 
last biweekly notice was published on November 8, 2005 (70 FR 67744).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. Within 60 days after the date of publication of this 
notice, the licensee may file a request for a hearing with respect to 
issuance of the amendment to the subject facility operating license and 
any person whose interest may be affected by this proceeding and who 
wishes to participate as a party in the proceeding must file a written 
request for a hearing and a petition for leave to intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the

[[Page 70642]]

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, <a href="/cdn-cgi/l/email-protection#49012c283b20272e0d262a222c3d09273b2a672e263f"><span class="__cf_email__" data-cfemail="1d55787c6f74737a59727e7678695d736f7e337a726b">[email protected]</span></a>; 
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 
<a href="/cdn-cgi/l/email-protection#f3bcb4b0be929a9fb0969d879681b39d8190dd949c85"><span class="__cf_email__" data-cfemail="d39c94909eb2babf90b6bda7b6a193bda1b0fdb4bca5">[email protected]</span></a>. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#ed9d899fad839f8ec38a829b"><span class="__cf_email__" data-cfemail="4d3d293f0d233f2e632a223b">[email protected]</span></a>.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit No. 2, Beaver County, Pennsylvania
    Date of amendment request: October 14, 2005.
    Description of amendment request: The proposed changes would revise 
Technical Specifications (TSs) 3/4.8.2.3, ``D.C. [direct current] 
Distribution--Operating'' and 3/4.8.2.4, ``D.C. Distribution--
Shutdown,'' to permit implementation of design changes associated with 
a battery charger upgrade.
    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.
    Failure of the components associated with the proposed change 
(i.e., battery chargers and Uninterruptible Power Supply [UPS] 
rectifiers) would not initiate any of the accidents described in the 
Updated Final Safety Analysis Report. No design function is being 
changed, and there is no adverse impact on the probability or 
consequences of accidents described in the safety analyses. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Failure modes associated with the rectifiers are reduced by the 
addition of battery chargers. A rectifier failure in the original 
design would result in a loss of battery charging and vital-bus load 
carrying functions. As a result of the modification, a dedicated 
battery charger will eliminate the battery charging design function 
loss upon failure of a rectifier. The failure mode for the new 
battery charger will be limited to a failure mode previously 
associated with the rectifiers. No new failure modes are associated 
with the new battery chargers.
    No new failure modes are created by 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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The modification does not reduce the margin of safety. Vital and 
DC bus support functions are unaffected. The system will still 
consist of rectifier and inverter units to provide AC power to the 
associated vital buses and units that charge the bus batteries and 
carry the associated DC bus loads. The only difference is that the 
Nos. 2-3 and 2-4 UPS unit rectifiers will no longer perform both of 
these functions. Separate battery charger units will charge the bus 
batteries and carry the DC loads. The new arrangement for DC buses 
2-3 and 2-4 will match the existing arrangement of DC buses 2-1 and 
2-2, insofar as the modification will result in all four instrument 
buses having separate UPS rectifier assemblies and battery charger 
devices. Dedicated bus 2-3 and 2-4 charges should increase system 
reliability.

[[Page 70643]]

    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 O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Branch Chief: Richard J. Laufer.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Unit 1, Hamilton County, Tennessee
    Date of amendment request: August 31, 2005.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TS) to adopt NRC-approved Revision 4 to 
Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-449, ``Steam Generator Tube 
Integrity.'' The proposed amendment includes 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.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 NRC staff issued a notice of opportunity for comment in the 
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments 
adopting TSTF-449, including a model safety evaluation and model no 
significant hazards consideration (NSHC) determination, using the 
consolidated line item improvement process. The NRC staff subsequently 
issued a notice of availability of the models for referencing in 
license amendment applications in the Federal Register on May 6, 2005 
(70 FR 24126). The licensee affirmed the applicability of the following 
NSHC determination in its application dated August 31, 2005.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change requires a SG Program that includes 
performance criteria that will provide reasonable assurance that the 
SG tubing will retain integrity over the full range of operating 
conditions (including startup, operation in the power range, hot 
standby, cooldown and all anticipated transients included in the 
design specification). The SG performance criteria are based on tube 
structural integrity, accident induced leakage, and operational 
LEAKAGE.
    A steam generator tube rupture (SGTR) event is one of the design 
basis accidents that are analyzed as part of a plant's licensing 
basis. In the analysis of a SGTR event, a bounding primary to 
secondary LEAKAGE rate equal to the operational LEAKAGE rate limits 
in the licensing basis plus the LEAKAGE rate associated with a 
double-ended rupture of a single tube is assumed.
    For other design basis accidents such as a main steamline break 
(MSLB), rod ejection, and reactor coolant pump locked rotor the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). These analyses typically assume that 
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or 
increases to 1 gallon per minute as a result of accident induced 
stresses. The accident induced leakage criterion introduced by the 
proposed changes accounts for tubes that may leak during design 
basis accidents. The accident induced leakage criterion limits this 
leakage to no more than the value assumed in the accident analysis.
    The SG performance criteria proposed change to the TS identify 
the standards against which tube integrity is to be measured. 
Meeting the performance criteria provides reasonable assurance that 
the SG tubing will remain capable of fulfilling its specific safety 
function of maintaining reactor coolant pressure boundary integrity 
throughout each operating cycle and in the unlikely event of a 
design basis accident. The performance criteria are only a part of 
the SG Program required by the proposed change to the TS. The 
program, defined by NEI 97-06, Steam Generator Program Guidelines, 
includes a framework that incorporates a balance of prevention, 
inspection, evaluation, repair, and leakage monitoring. The proposed 
changes do not, therefore, significantly increase the probability of 
an accident previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT I-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The typical analysis of the limiting design basis 
accident assumes that primary to secondary leak rate after the 
accident is 1 gallon per minute with no more than [500 gallons per 
day or 720 gallons per day] in any one SG, and that the reactor 
coolant activity levels of DOSE EQUIVALENT I-131 are at the TS 
values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design basis 
accident and is an improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of a SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.

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

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

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

    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the SG

[[Page 70644]]

Program to manage SG tube inspection, assessment, repair, and 
plugging. The requirements established by the SG Program are 
consistent with those in the applicable design codes and standards 
and are an improvement over the requirements in the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.

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

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to <a href="/cdn-cgi/l/email-protection#7e0e1a0c3e100c1d50191108"><span class="__cf_email__" data-cfemail="2c5c485e6c425e4f024b435a">[email protected]</span></a>.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois
    Date of application for amendment: May 20, 2004, as supplemented 
May 23 and September 30, 2005.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to support the implementation of 24-month fuel 
cycles at Clinton Power Station, Unit 1.
    Date of issuance: October 24, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 169.
    Facility Operating License No. NPF-62: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: July 6, 2004 (69 FR 
40669). The supplements dated May 23 and September 30, 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.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 24, 2005.
    No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois
    Date of application for amendment: August 18, 2004, as supplemented 
May 13 and 25, June 14, August 17, and October 24 and 25, 2005.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 4.3, ``Fuel Storage,'' to reflect the increased 
spent fuel storage capacity at Clinton Power Station, Unit 1.
    Date of issuance: October 31, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 170.
    Facility Operating License No. NPF-62: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: August 29, 2005 (70 FR 
51093). In the supplement dated June 14, 2005, the licensee changed the 
use of the building crane and the temporary crane. This change may have 
impacted the staff''s original proposed no significant hazards 
consideration determination published on December 29, 2004 (69 FR 
78051). Therefore a revised no significant hazards consideration 
determination was published on August 29, 2005. However, the 
supplements dated October 24 and 25, 2005, provided additional 
information that clarified the application, did not expand the scope of 
the application as noticed, and did not change the staff's original 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 31, 2005.
    No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269 and 50-270, Oconee Nuclear 
Station, Units 1 and 2, Oconee County, South Carolina
    Date of application of amendments: August 18, 2005, as supplemented 
by letter dated September 15, 2005.
    Brief description of amendments: The amendments revised the 
Technical Specifications 3.5.2.6 and 3.5.3.6 to accommodate the 
replacement of the reactor building emergency sump suction inlet trash 
racks and screens with strainers.
    Date of issuance: November 1, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 348/350.
    Renewed Facility Operating License Nos. DPR-38 and DPR-47: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2005 (70 FR 
51852). The supplement dated September 15, 2005, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 1, 2005.

[[Page 70645]]

    No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
    Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
Units 1 and 2, Rock Island County, Illinois
    Date of application for amendments: November 4, 2004, as 
supplemented by letters dated March 8, May 25 and July 8, 2005.
    Brief description of amendments: The amendment revises TS Section 
3.4.9, ``Reactor Coolant System Pressure and Temperature (P/T) 
Limits.'' The changes revise the P/T limit curves for 54 effective full 
power years (EFPY) to support an additional 20 years of operation under 
the renewed license and resolve a non-conservative condition for TS 
3.4.9, Figure 3.4.9-2, ``Non-Nuclear Heatup/Cooldown Curve,'' for Quad 
Cities Nuclear Power Station.
    Date of issuance: October 17, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 217, 209, 228, 223.
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5244). The supplements dated March 8, May 25 and July 8, 2005, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 17, 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: January 27, 2005, as 
supplemented by letters dated August 12, September 9, and October 21, 
2005.
    Brief description of amendment: The amendment allows 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 issuance: October 31, 2005.
    Effective date: October 31, 2005.
    Amendment No.: 222.
    Facility Operating License No. DPR-72: Amendment revises the 
Technical Specifications.
    Date of notice in Federal Register: August 26, 2005 (70 FR 50424) 
and Repeat Notice dated September 27, 2005 (70 FR 56505). The August 
26, 2005, Notice revised the previous notice dated March 15, 2005 (70 
FR 12746). The licensee's supplement dated August 12, 2005, revised the 
proposed no significant hazards consideration determination and the 
licensee's supplements dated September 9, and October 21, 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 proposed no significant hazards consideration 
determination as published August 26, 2005, in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 31, 2005.
    No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California
    Date of application for amendments: March 11, 2005, and its 
supplement dated August 25, 2005.
    Brief description of amendments: The amendments allow use of the 
steam generator tube W* (W-star) alternate repair criteria for 
indications in the Westinghouse explosive tube expansion region on a 
permanent basis.
    Date of issuance: October 28, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented prior to startup of Cycle 14.
    Amendment Nos.: Unit 1-182; Unit 2-184.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 26, 2005 (70 FR 
21462).
    The August 25, 2005, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 28, 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: February 3, 2005.
    Brief description of amendments: The amendments revised Technical 
Specification 3.6.3, ``Containment Isolation Valves,'' Surveillance 
Requirements 3.6.3.3 and 3.6.3.4 for Containment Isolation Valves and 
Blind Flanges (CIVs), by adding a provision to exempt CIVs that are 
locked, sealed, or otherwise secured from the position verification 
requirements.
    Date of issuance: November 3, 2005.
    Effective date: November 3, 2005, to be implemented within 60 days 
of issuance.
    Amendment Nos.: Unit 2-201; Unit 3-192.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 1, 2005 (70 FR 
9996).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 3, 2005.
    No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas
    Date of amendment request: June 2, 2005.
    Brief description of amendments: The amendments change Technical 
Specification (TS) 3.4.6.1, ``Reactor Coolant System Leakage Detection 
Systems,'' to specifically require only one containment radioactivity 
monitor (particulate channel) to be operable in Modes 1, 2, 3, and 4. 
Additionally, corresponding changes to the Surveillance Requirement 
(SR) 4.4.6.1 and 4.4.6.2.1, ``Reactor Coolant System Operational 
Leakage,'' were also made.
    Date of issuance: October 17, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.

[[Page 70646]]

    Amendment Nos.: Unit 1-174; Unit 2-162.
    Facility Operating License Nos. NPF-76 and NPF-80: The 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 October 17, 2005.
    No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
    Date of application for amendment: March 1, 2005, as supplemented 
by letters dated June 16 and September 23, 2005.
    Brief description of amendment: These amendments revise the 
frequency for the trip actuating device operational test (TADOT) of the 
P-4 interlock function. The proposed changes would revise the 
surveillance requirement frequency in Technical Specification 3.3.2 
from ``once per reactor trip breaker cycle'' to ``18 months'' for North 
Anna, Units 1 and 2.
    Date of issuance: October 24, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 244/225.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: April 26, 2005 (70 FR 
21465).
    The supplements dated June 16 and September 23, 2005, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 24, 2005.
    No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
    Date of application for amendment: July 14, 2005.
    Brief description of amendment: These amendments correct two errors 
in the units of measure used to determine the Overtemperature [delta]T 
Function Allowable Value.
    Date of issuance: October 25, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 245/226.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: August 16, 2005 (70 FR 
48208)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 25, 2005.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 9th day of November, 2005.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing Office of Nuclear 
Reactor Regulation.

[FR Doc. 05-22795 Filed 11-21-05; 8:45 am]
BILLING CODE 7590-01-P


</pre><script data-cfasync="false" src="/cdn-cgi/scripts/5c5dd728/cloudflare-static/email-decode.min.js"></script></body>
</html>