[Federal Register Volume 70, Number 147 (Tuesday, August 2, 2005)]
[Notices]
[Pages 44400-44407]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E5-4067]


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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 July 8, 2005, to July 21, 2005. The last 
biweekly notice was published on July 19, 2005 (70 FR 41442).

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

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

[[Page 44401]]

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, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
[email protected]. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

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

    Date of amendment request: July 7, 2005.
    Description of amendment request: The amendments would revise 
Technical Specification 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.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would implementation of the changes proposed in this LAR 
[License Amendment Request] involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. This LAR clarifies Technical Specification [TS] 3.9.1 
regarding the applicability of boron concentration limits when the 
refueling canal and refueling cavity are not connected to the 
reactor coolant system [RCS]. When the refueling canal and the 
refueling cavity are isolated from the RCS, no potential path for 
boron dilution of the RCS exists, thus there is no significant 
increase in the probability of an accident that has been previously 
evaluated, nor would there be a significant increase in the 
consequences of an accident that has been previously evaluated.
    2. Would implementation of the changes proposed in this LAR 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    No. The change proposed in this LAR clarifies the applicability 
of TS 3.9.1 when the refueling canal and refueling cavity are not 
connected to the reactor coolant system. When the refueling canal 
and the refueling cavity are isolated from the RCS, no potential 
path for boron dilution of the RCS exists, thus there is no means to 
initiate an accident that is new or different from any accident that 
has been previously evaluated.
    3. Would implementation of the changes proposed in this LAR 
involve a significant reduction in a margin of safety?
    No. The change proposed in this LAR only clarifies the 
applicability of TS 3.9.1 when the refueling canal and the refueling 
cavity are not connected to the reactor coolant system. [TS 3.9.1 
limits the boron concentrations of the reactor coolant system], the 
refueling canal, and the refueling cavity to ensure that the reactor 
remains subcritical during Mode 6 plant conditions. However, when 
the refueling canal and the refueling cavity are isolated from the 
reactor coolant system, no potential for boron dilution of the RCS 
exists. Therefore, in this condition it is not necessary to place a 
limit on the boron concentration in the refueling canal and the 
refueling cavity, thus there is no significant reduction in a margin 
of safety since no specific boron limits are being changed.

    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.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of amendment request: January 31, 2005.
    Description of amendment request: Entergy Operations, Inc. (EOI) 
has requested a change which would revise

[[Page 44402]]

the requirements associated with the Arkansas Nuclear One, Unit 2 (ANO-
2) containment overcurrent protection devices. EOI proposes to amend 
Operating License NPF-6 to eliminate Technical Specifications (TSs) 
section 3.8.2.5, ELECTRICAL POWER SYSTEMS-Containment Penetration 
Conductor Overcurrent Protection Devices. The proposed change would 
relocate the requirements for containment penetration conductor 
overcurrent protective devices to the Technical Requirements Manual 
(TRM).
    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?
    The proposed changes to relocate the requirements for 
containment penetration conductor overcurrent protective devices 
from Technical Specifications to the TRM will have no adverse effect 
on plant operation, or the availability or operation of any accident 
mitigation equipment. The plant response to the design basis 
accidents will not change. Operation of the containment penetration 
conductor overcurrent protective devices is not an accident 
initiator and can not cause an accident. Whether the requirements 
for the containment penetration conductor overcurrent protective 
devices are located in Technical Specifications or the TRM will have 
no effect on the probability or consequences of any accident 
previously evaluated.
    Therefore, the removal of overcurrent protection devices from 
the TS does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes to relocate the requirements from Technical 
Specifications to the TRM will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. The proposed changes will not 
introduce any new failure modes that could result in a new accident. 
Also, the response of the plant and the operators following the 
design basis accidents is unaffected by the changes.
    Therefore, this 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?
    The proposed changes will relocate the requirements for 
containment penetration conductor overcurrent protective devices 
from Technical Specifications to the TRM. Any future changes to the 
relocated requirements will be in accordance with 10 CFR 50.59 and 
approved station procedures. The proposed changes will have no 
adverse effect on plant operation, or the availability or operation 
of any accident mitigation equipment. The plant response to the 
design basis accidents will not change. In addition, the relocated 
requirements do not meet any of the 10 CFR 50.36c(2)(ii) criteria on 
items for which Technical Specifications must be established.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
    NRC Section Chief: David Terao.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50-
278, Peach Bottom Atomic Power Station, Unit 3, York and Lancaster 
Counties, Pennsylvania

    Date of application for amendment: July 6, 2005.
    Description of amendment request: The proposed changes extend the 
use of the Peach Bottom Atomic Power Station, Unit 3, pressure-
temperature (P-T) limits specified in the Technical Specifications 
(TSs) from 22 to 32 effective full power years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No. The proposed changes to the technical 
specifications to extend the use of the existing pressure-
temperature (P-T) limits does not affect the operation or 
configuration of any plant equipment. Thus, no new accident 
initiators are created by this change. The proposed P-T limits are 
based on the projected reactor vessel neutron fluence at 32 
effective full power years (EFPY) of operation. A bounding 
calculation of reactor vessel 32 EFPY fast neutron fluence has been 
completed for Peach Bottom Atomic Power Station (PBAPS), Unit 3, 
using the methodology described in a General Electric (GE) Company 
Licensing Topical Report (LTR), which adheres to the guidance in 
Regulatory Guide 1.190, ``Calculational and Dosimetry Methods for 
Determining Pressure Vessel Neutron Fluence.'' The three-dimensional 
spatial distribution of neutron flux was modeled by combining the 
results of two separate two-dimensional neutron transport 
calculations. The latest available cross section libraries for the 
important components of Boiling Water Reactor (BWR) neutron flux 
calculations, i.e., oxygen, hydrogen and individual iron isotopes, 
were included. The resulting reactor vessel fast neutron fluence 
value was then used in concert with the American Society of 
Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code), 
Section XI, Case -640 and ASME Code, Section XI, Appendix G, 
paragraph G-2214.1 to develop updated P-T curves. A comparison of 
the updated P-T curves with the existing PBAPS, Unit 3 curves 
indicates that the existing curves are bounding through 32 EFPY. 
This provides sufficient assurance that the PBAPS, Unit 3, reactor 
vessel will be operated in a manner that will protect it from 
brittle fracture under all operating conditions.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed changes to the technical 
specifications to extend the use of the existing P-T limits do not 
affect the operation or configuration of any plant equipment. The 
proposed P-T limits will remain valid and conservative throughout 
the proposed extension.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No. The proposed changes extend the use of the 
existing P-T limits. The proposed P-T limits are based on the 
projected reactor vessel neutron fluence at 32 EFPY of operation. A 
bounding calculation of reactor vessel 32 EFPY fast neutron fluence 
has been completed for PBAPS, Unit 3, using the NRC approved 
methodology in a GE LTR, which adheres to the guidance in Regulatory 
Guide 1.190. The three-dimensional spatial distribution of neutron 
flux was modeled by combining the results of two separate two-
dimensional neutron transport calculations. The latest available 
cross section libraries for the important components of BWR neutron 
flux calculations, i.e., oxygen, hydrogen and individual iron 
isotopes, were included. The resulting reactor vessel fast neutron 
fluence value was then used in concert with ASME Code Case -640 and 
ASME Code, Section XI, Appendix G, paragraph G-2214.1 to develop 
updated P-T curves. A comparison of the updated P-T curves with the 
existing PBAPS, Unit 3 curves indicates that the existing curves are 
bounding through 32 EFPY. This provides sufficient margin such that 
the PBAPS, Unit 3, reactor vessel will be operated in a manner that 
will protect it from brittle fracture under all operating 
conditions.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.


[[Page 44403]]


    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 S. O'Neill, Associate and General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Section Chief: Darrell J. Roberts.

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

    Date of amendment request: July 1, 2005.
    Description of amendment request: The proposed change will amend 
the design and licensing basis of the Fort Calhoun Station, Unit 1, by 
revising the updated safety analysis report (USAR) to describe an 
existing Emergency Operating Procedure (EOP) operator action to isolate 
steam generator blowdown within 15 minutes of reactor trip during a 
loss of main feedwater event.
    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 the USAR clarifies reliance on operator 
action which has been utilized since implementation of the EOPs. It 
does not affect an accident initiator previously evaluated in the 
USAR or Technical Specifications and will not prevent safety systems 
from performing their accident mitigating function as discussed in 
the USAR or Technical Specifications.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change provides clarification to the existing USAR 
accident analysis of record. The change does not modify or install 
any safety related equipment. It does not alter any design or 
licensing basis assumptions and does not alter any operating 
procedures other than the explicit specification [of] the time 
constraint of the 15 minutes. Presently the action is included in 
EOP-00 without a time constraint.
    Therefore, this 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 provides clarification to the USAR section 
14.10.1 and has no effect on safety margins.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

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: July 4, 2005.
    Description of amendment request: The proposed changes would extend 
the allowed outage time for Technical Specification (TS) 3/4.7.4, 
``Essential Cooling Water System,'' and the associated TSs for those 
systems supported by Essential Cooling Water, from 7 days to 14 days.
    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.
    Since only one train of components is affected by the condition 
and single failure is not considered while a plant is in an LCO 
[Limiting Condition for Operation] ACTION, the operable ESF 
[Engineered Safety Feature] trains are adequate to maintain the 
plant's design basis. Thus, this condition will not alter 
assumptions relative to the mitigation of an accident or transient 
event.
    Considering compensatory action and risks involved in a plant 
shutdown, STPNOC [STP Nuclear Operating Company] has determined that 
there is no significant risk associated with extending the Allowed 
Outage Time for the Essential Cooling Water System and the systems 
it supports for an additional 7 days. Additionally, the proposed 
change to remove the one-time note from TS 3.7.4 is considered an 
administrative change and does not impact the probability or 
consequences of any accident previously evaluated.
    Based on this evaluation, there is no significant increase in 
the probability or consequence of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different accident from any accident previously evaluated?
    Response: No.
    This proposed change only extends an Allowed Outage 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[s] as evaluated in the 
South Texas Project safety analysis is proposed. The proposed change 
to remove the one-time note from TS 3.7.4 is considered an 
administrative change and does not create the possibility of a new 
or different kind of accident previously evaluated.
    Therefore, this proposed change[ does not] create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Considering compensatory action and risks involved in a plant 
shutdown, STPNOC has determined that there is no significant risk 
associated with extending the Allowed Outage Time for the Essential 
Cooling Water System and the systems it supports for an additional 7 
days.
    Based on the availability of redundant systems, the compensatory 
actions that will be taken, and the extremely low probability of an 
accident that could not be mitigated by the available systems, 
STPNOC concludes that there is no significant reduction in the 
margin of safety. The proposed change to remove the one-time note 
from TS 3.7.4 is considered an administrative change and does not 
impact any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: David Terao.

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: July 4, 2005.
    Description of amendment request: The proposed change to Technical 
Specification 4.0.5 would add a reference to the NRC-approved

[[Page 44404]]

exemption of selected pumps, valves, and other components from special 
treatment requirements. As an editorial change, references to Title 10, 
Code of Federal Regulations (10 CFR) Part 50, Section 50.55a(f) and
    10 CFR Part 50, Section 50.55a(f)(6)(i) would be added to the 
paragraph for inservice testing, similar to the existing references for 
inservice inspection. In addition, ``inservice testing'' and 
``inservice inspection'' would be reordered for consistency with the 
sequence of the regulations in 10 CFR Part 50, Section 50.55a.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. Including the reference to the exemption in the Technical 
Specifications establishes consistency between the surveillance 
requirements for inservice inspection and testing and the exemption 
as approved by the NRC. There are no changes in the inspection and 
testing procedures as a result of adding the reference because the 
exemption already removes low safety significance and non-risk 
significant components from the requirements for special treatment. 
The proposed changes are administrative in nature and do not have a 
significant adverse effect on plant operation or personnel safety. 
Consequently, the changes will not affect 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?
    No. Including the reference to the exemption in the Technical 
Specifications establishes consistency between the surveillance 
requirements for inservice inspection and testing and the exemption 
as approved by the NRC. There are no changes in the inspection and 
testing procedures as a result of adding the reference because the 
exemption already removes low safety significance and non-risk 
significant components from the requirements for special treatment. 
The proposed changes are administrative in nature and do not have a 
significant adverse effect on plant operation or personnel safety. 
Consequently, the changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. Including the reference to the exemption in the Technical 
Specifications establishes consistency between the surveillance 
requirements for inservice inspection and testing and the exemption 
as approved by the NRC. There are no changes in the inspection and 
testing procedures as a result of adding the reference because the 
exemption already removes low safety significance and non-risk 
significant components from the requirements for special treatment. 
The proposed changes are administrative in nature and do not have a 
significant adverse effect on plant operation or personnel safety. 
Consequently, the changes do not significantly reduce a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: David Terao.

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

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

Tennessee Valley Authority, Docket No. 50-259 , Browns Ferry Nuclear 
Plant, Unit 1, Limestone County, Alabama

    Date of application for amendments: June 28, 2004, as supplemented 
February 23 and April 25, 2005.
    Description of amendments request: The proposed amendment would 
change the operating license to increase the maximum authorized power 
level from 3293 megawatts thermal (MWt) to 3952 MWt; an increase of 
approximately 20 percent. The amendment would also change the licensing 
bases and any associated Technical Specifications for containment 
overpressure, the maximum ultimate heat sink temperature, and the upper 
bound peak cladding temperature.
    Date of publication of individual notice in the Federal Register: 
July 11, 2005 (70 FR 39803).
    Expiration date of individual notice: August 10, 2005 (Public 
comments) and September 9, 2005 (Hearing requests).

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

    Date of application for amendments: June 25, 2004, as supplemented 
February 23 and April 25, 2005.
    Description of amendments request: The proposed amendments would 
change the operating licences to increase the maximum authorized power 
level from 3458 megawatts thermal (MWt) to 3952 MWt; an increase of 
approximately 15 percent. The amendment would also change the licensing 
bases and any associated Technical Specifications for containment 
overpressure.
    Date of publication of individual notice in the Federal Register: 
July 12, 2005 (70 FR 40064).
    Expiration date of individual notice: August 11, 2005 (Public 
comments) and September 12, 2005 (Hearing requests).

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

[[Page 44405]]

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

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

    Date of application for amendment: March 25, 2005, as supplemented 
on June 10, 2005.
    Brief description of amendment: The amendment revised Section 3.7, 
``Auxiliary Electrical Power,'' of the Technical Specifications to 
reflect the capability upgrade of one of the offsite power supply lines 
from 69 kilovolts (KV) to 230 KV.
    Date of Issuance: July 14, 2005.
    Effective date: July 14, 2005 and shall be implemented as soon as 
the upgraded offsite supply line is placed in service.
    Amendment No.: 256.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 12, 2005 (70 FR 
19113).
    The June 10, 2005, letter provided clarifying information within 
the scope of the original application and did not change the staff's 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of this amendment is contained in a 
Safety Evaluation dated July 14, 2005.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 15, 2004.
    Brief description of amendment: This amendment revises Technical 
Specifications by extending the inspection interval for reactor coolant 
pump flywheels to 20 years.
    Date of issuance: June 21, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 119.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 2005 (70 FR 
9988).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 21, 2005.
    No significant hazards consideration comments received: No.

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 application of amendments: February 14, 2005.
    Brief description of amendments: The amendments revised the 
Technical Specification Surveillance Requirement 3.3.7.1 to extend the 
frequency of the channel functional test for the Engineered Safeguards 
Protective System digital actuation logic channels from once every 31 
days to once every 92 days.
    Date of Issuance: May 19, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 345, 347 and 346.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 15, 2005 (70 FR 
12745).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 19, 2005.
    No significant hazards consideration comments received: No.

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 application of amendments: March 14, 2005.
    Brief description of amendments: The amendments deleted Technical 
Specification 5.5.4, ``Post Accident Sampling.''
    Date of Issuance: July 12, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment Nos.: 346, 348, and 347.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: May 10, 2005 (70 FR 
24649)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 12, 2005.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of application for amendment: December 17, 2004.
    Brief description of amendment: The proposed change revises the air 
lock surveillance test acceptance criteria to be consistent with the 
NRC approved Industry Technical Specification Task Force (TSTF) change 
to the Standard Technical Specifications TSTF-52, entitled, ``Implement 
10 CFR [Part] 50, Appendix J, Option B.'' By letter dated April 6, 
1998, the NRC Staff issued amendment number 135 to the Grand Gulf 
Nuclear Station license permitting the implementation of the 
containment leak rate testing provisions of 10 CFR Part 50, Appendix J, 
Option B.
    Date of issuance: July 12, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 168.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 1, 2005 (70 FR 
5242).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 12, 2005.
    No significant hazards consideration comments received: No.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: February 4, 2004, as supplemented by 
letter dated March 16, 2005.
    Description of amendment request: The amendment modified the 
Seabrook Station Technical Specification (TS) Index; TS Table 3.3-10, 
``Accident Monitoring Instrumentation''; TS Table 4.4-2, ``Steam 
Generator Tube Inspection''; TS 6.0, ``Administrative Controls''; and 
Appendix B to Facility Operating License (FOL) No. NPF-86, 
``Environmental Protection Plan''.
    Date of issuance: July 18, 2005.

[[Page 44406]]

    Effective date: As of its date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 104.
    Facility Operating License No. NPF-86: The amendment revised the 
TSs and Appendix B to the FOL.
    Date of initial notice in Federal Register: March 2, 2004 (69 FR 
9861). The March 16, 2005, supplement provided clarifying information 
that 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 amendment is contained in a Safety Evaluation dated 
July 18, 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: June 29, 2004, as supplemented 
by letter dated June 14, 2005.
    Brief description of amendments: The proposed changes revise the 
Technical Specifications (TSs) to implement the following miscellaneous 
TS changes: Revise TS 2.2.5 Safety Limit Violations Licensee Event 
Report reporting period from 30 days to 60 days; revise 3.4.3.1.2 
Pressurizer Heatup/Cooldown Limits Surveillance Requirements frequency 
to reflect pressurizer spray cyclic limits being governed by the 
temperature differentials between the spray nozzle and the spray line; 
revise TS 5.5.2.11 Steam Generator Tube Surveillance requirements to 
correct typographical errors; remove TS 5.5.2.14 Configuration Risk 
Management Program in accordance with Federal Register Notice Vol. 64, 
No. 137 (64 FR 38551, July 19, 1999); and revise TS 5.7.1.5 Core 
Operating Limits Report (COLR) to delete revision numbers and dates 
from the referenced documents in this section, consistent with the NRC 
approved industry Technical Specifications Task Force (TSTF) Standard 
Technical Specifications Traveler number TSTF-363, ``Revise Topical 
Report References in ITS (Improved Technical Specifications) 5.6.5 
COLR.''
    Date of issuance: July 19, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 197, 188.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 2004 (69 FR 
46588). The supplemental letter dated June 14, 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 July 19, 2005.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 13, 2004.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 5.6.5b by adding two topical reports (TRs) into the 
list of approved analytical methods used to determine the core 
operating limits, deleting four TRs for analytical methods no longer 
used to determine the core operating limits, and sequentially 
renumbering the remaining approved analytical methods in TS 5.6.5b.
    Date of issuance: July 13, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 119, 119.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 21, 2004 (69 
FR 76495).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 13, 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 1, 2004, as supplemented by 
letters dated and October 28, 2004, and November 16, 2004.
    Brief description of amendment: These amendments revise the reactor 
coolant pressure and temperature limits, low-temperature overpressure 
protection system (LTOPS) setpoint values, and LTOPS enable 
temperatures that are valid for 50.3 effective full-power years (EFPY) 
and 52.3 EFPY of operation for North Anna, Units 1 and 2, respectively.
    Date of issuance: July 8, 2005.
    Effective date: As of the date of issuance and shall be implemented 
within 6 months from the date of issuance.
    Amendment Nos.: 242 and 223.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2004 (69 FR 
53114). The supplements dated October 28, 2004, and November 16, 2004, 
contained clarifying information only and did not change the initial no 
significant hazards consideration determination or expand the scope of 
the initial application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 8, 2005.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of application for amendments: November 4, 2004, as 
supplemented on February 21 and June 2, 2005.
    Brief Description of amendments: These amendments revise the 
Technical Specifications (TS) to delete the Inservice Inspection (ISI) 
and Inservice Testing (IST) requirements in TS 4.0.5; relocate the IST 
requirements to the administrative section of the TS as a program; 
revise the TS to reference the IST program instead of TS 4.0.5; delete 
the individual TS references to the ISI program; and add a TS Bases 
Control Program to the TS Administrative Controls section.
    Date of issuance: July 15, 2005.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment Nos.: 243 and 242.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments change the Technical Specifications.
    Date of initial notice in Federal Register: February 15, 2005 (70 
FR 7771). The February 21 and June 2, 2005, supplements contained 
clarifying information only and did not change the initial proposed no 
significant hazards consideration determination or expand the scope of 
the initial application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 15, 2005.
    No significant hazards consideration comments received: No.


[[Page 44407]]


    Dated at Rockville, Maryland, this 25th day of July 2005.

For the Nuclear Regulatory Commission
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. E5-4067 Filed 8-1-05; 8:45 am]
BILLING CODE 7590-01-P