[Federal Register Volume 71, Number 215 (Tuesday, November 7, 2006)]
[Notices]
[Pages 65139-65148]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-18595]


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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 October 13, 2006, to October 26, 2006. The 
last biweekly notice was published on October 24, 2006 (71 FR 62306).

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, Rulemaking, 
Directives and Editing 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

[[Page 65140]]

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

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of amendments request: September 28, 2006.
    Description of amendments request: The proposed amendments would 
revise certain Technical Specification (TS) requirements for mode 
change limitations in Limiting Condition for Operation 3.0.4 and 
Surveillance Requirement 3.0.4. This request is consistent with NRC-
approved Industry/Technical Specification Task Force (TSTF) Traveler 
number TSTF-359, Revision 9, ``Increase Flexibility in Mode 
Restraints.'' In addition, the proposed amendments would correct TS 
Example 1.4-1, ``Surveillance Requirements,'' to accurately reflect the 
changes made by TSTF-359, which is consistent with NRC-approved TSTF-
485, Revision 0.
    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 revises Section 1.4, Frequency, ``Example 
1.4-1,'' to be consistent with Surveillance Requirement (SR) 3.0.4 
and Limiting Condition for Operation (LCO) 3.0.4. This change is 
considered administrative in that it modifies the example to 
demonstrate the proper application of SR 3.0.4 and LCO 3.0.4. The 
requirements of SR 3.0.4 and LCO 3.0.4 are clear and are clearly 
explained in the associated Bases. As a result, modifying the 
example will not result in a change in usage

[[Page 65141]]

of the Technical Specifications (TS). The proposed change does not 
adversely affect accident initiators or precursors, the ability of 
structures, systems, and components (SSCs) to perform their intended 
function to mitigate the consequences of an initiating event within 
the assumed acceptance limits, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Therefore, this change is considered 
administrative and will have no effect on the probability or 
consequences of any accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The change does not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the change does not impose any new or 
different requirements or eliminate any existing requirements. The 
change does not alter assumptions made in the safety analysis. The 
proposed change is consistent with the safety analysis assumptions 
and current plant operating practice.
    Therefore, the 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.
    The proposed change is administrative and will have no effect on 
the application of the Technical Specification requirements. 
Therefore, the margin of safety provided by the Technical 
Specification requirements is unchanged.
    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 
that review, it appears that the three 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: Michael G. Green, Senior Regulatory Counsel, 
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695, 
Phoenix, Arizona 85072-2034.
    NRC Branch Chief: David Terao.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit No. 2 New London County, Connecticut

    Date of amendment request: March 17, 2006.
    Description of amendment request: The proposed amendment would 
revise Millstone Power Station, Unit No. 2 Technical Specification (TS) 
3.4.4 to replace the existing maximum and minimum pressurizer water 
volume and water level limits with a maximum water level limit. The 
associated TS bases will be updated to address the proposed change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change does not change the accident analysis of 
record, maintains the current maximum operating pressurizer level at 
its present value, does not modify any plant equipment and does not 
impact any failure modes that could lead to an accident. Additionally, 
the proposed change has no effect on the consequences of any analyzed 
accident since the change does not affect the function of any equipment 
credited for accident mitigation. Therefore, the proposed amendment 
does not increase the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Since the proposed change does not modify any plant equipment and 
there is no impact on the capability of existing equipment to perform 
its intended functions and no new failure modes are introduced by the 
proposed change, the proposed amendment does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety?
    The proposed change maintains the current maximum operating 
pressurizer level at its present value, and the acceptance criterion 
for the maximum pressurizer level is unchanged. Since there are no 
changes, the proposed change does not involve a reduction in a margin 
of safety.
    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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Branch Chief: Harold K. Chernoff.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: October 16, 2006.
    Description of amendment request: The proposed change will add an 
NRC previously approved topical report to the analytical methods 
referenced in Technical Specification (TS) Section 5.6.5, ``Core 
Operating Limits Report (COLR).'' The current method of performing the 
loss-of-coolant accident (LOCA ) analyses will be replaced by an 
updated method described in AREVA NP (formerly known as Framatome or 
Siemens) topical report, ``EXEM BWR-2000 [Boiling-Water Reactor--2000] 
ECCS [Emergency Core Cooling System] Evaluation Model.''
    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.
    Core operating limits are established each operating cycle in 
accordance with TS 3.2, ``Power Distribution'' and TS 5.6.5, ``Core 
Operating Limits Report (COLR)''. These core operating limits ensure 
that the fuel design limits are not exceeded during any conditions 
of normal operation or in the event of any Anticipated Operational 
Occurrence (AOO). In addition, the Average Planar Linear Heat 
Generation Rate (APLHGR) operating limits imposed by Technical 
Specification 3.2.1 also ensure that the peak cladding temperature 
(PCT) during the postulated design basis LOCA does not exceed the 
2200 [deg]F limit specified in 10 CFR 50.46. The APLHGR is a measure 
of the average linear heat generation rate of all the fuel rods in a 
fuel assembly at any axial location.
    The methods used to determine the operating limits are those 
previously found acceptable by the NRC and listed in TS section 
5.6.5.b. A change to TS section 5.6.5.b is requested to include an 
updated LOCA analysis method, EXEM BWR-2000. The updated method will 
be used to determine the APLHGR operating limits imposed by 
Technical Specification 3.2.1. EXEM BWR-2000 has been reviewed and 
approved by the NRC and is applicable to the RBS [River Bend 
Station] plant design and the AREVA NP fuel being used at RBS. The 
application of the LOCA analytical model will continue to ensure 
that the APLHGR

[[Page 65142]]

operating limits are established to protect the fuel cladding 
integrity during normal operation, AOOs, and the design basis LOCA.
    The requested TS changes concern the use of analytical methods 
and do not involve any plant modifications or operational changes 
that could affect any postulated accident precursors or accident 
mitigation systems and do not introduce any new accident initiation 
mechanisms.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed TS amendment will not change the design function, 
reliability, performance, or operation of any plant systems, 
components, or structures. It does not create the possibility of a 
new failure mechanism, malfunction, or accident initiators not 
considered in the design and licensing bases. Plant operation will 
continue to be within the core operating limits that are established 
using NRC approved methods that are applicable to the RBS design and 
the RBS fuel.
    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 ECCS performance analysis methods are used to establish the 
APLHGR limits required by Technical Specification 3.2.1. The APLHGR 
limits are specified in the COLR and are the result of fuel design, 
design basis accident (DBA), and transient analyses. Limits on the 
APLHGR are specified to ensure that the fuel design limits are not 
exceeded during anticipated operational occurrences (AOOs) and that 
the peak cladding temperature (PCT) during the postulated design 
basis LOCA does not exceed the 2200 [deg]F limit specified in 10 CFR 
50.46.
    The EXEM BWR-2000 evaluation model is an updated LOCA analytical 
method that has been approved by the NRC and is applicable to the 
RBS plant design and the fuel being used at RBS. A RBS plant 
specific ECCS performance analysis has been performed with the EXEM 
BWR-2000 evaluation model. This evaluation concluded that the 
resulting PCT still afforded adequate margin to the 2200 [deg]F 
limit of 10 CFR 50.46.
    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: Mark Wetterhahn, Esq., Winston & Strawn LLP, 
1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: David Terao.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New 
York

    Date of amendment request: September 25, 2006.
    Description of amendment request: The amendment proposes revisions 
to the Technical Specifications that are editorial in nature and 
consist of typographical corrections, update of references, and 
deletion of obsolete notes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes are editorial in nature and have no affect 
on accident scenarios previously evaluated. The proposed changes 
include typographical corrections, consistent with the current 
version of the Standard Technical Specifications (NUREG 1431, 
Revision 3); updated references, consistent with the current version 
of the Entergy Quality Assurance Program Manual (Revision 13); and 
deletion of notes that provided one-time allowances or are otherwise 
now obsolete. The proposed changes do not affect initiating events 
for accidents previously evaluated and do not affect or modify 
plants systems or procedures used to mitigate the progression or 
outcome of those accident scenarios.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve the installation of new 
plant equipment or modification of existing plant equipment. No 
system or component setpoints are being changed and there are no 
changes being proposed for the way that the plant is operated. There 
are no new accident initiators or equipment failure modes resulting 
from the proposed changes. The proposed changes are editorial in 
nature, consisting of typographical corrections, reference updates, 
and deletion of obsolete notes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes are editorial in nature and do not change 
setpoints or limiting parameters specified in the plant Technical 
Specifications.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Richard J. Laufer.

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

    Date of amendment request: August 31, 2006.
    Description of amendment request: Entergy Operations, Inc., 
proposes to relocate Technical Specification (TS) 3.8.7 requirements 
associated with 120 Volt Inverter Y-28 and TS 3.8.9 requirements 
associated with 120 VAC electrical power distribution subsystem panel 
C-540 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?
    Response: No.
    The proposed change does not physically alter any plant 
structures, systems, or components and does not affect or create new 
accident initiators or precursors. The loss of Y-28, in itself, has 
no significant impact on station operation because its associated 
instrument panel, C-540, remains energized from an Emergency Diesel 
Generator (EDG) backed vital AC source. A potential loss of vital 
instrument panel C-540 does not prevent the fulfillment of a safety 
function and does not cause Emergency Safeguard Features (ESF) 
systems actuations that could render multiple ESF-related trains 
incapable of performing their intended safety function. Therefore, 
there is no effect on probability of accidents previously evaluated.
    The proposed change relocates operability requirements for Y-28 
and C-540 to the TRM. The TRM is part of the Safety Analysis Report 
(SAR) and is controlled under 10 CFR 50.59. In addition, TS-related 
components powered by C-540 continue to be governed by other TSs 
that limit the time in which the

[[Page 65143]]

components can be out of service or provide compensatory measures 
during the out-of-service period.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not physically alter any structures, 
systems, or components, and does not affect or create new accident 
initiators or precursors. The accident analysis assumptions and 
results are unchanged. No new failures or interactions have been 
created. In addition, the proposed change does not introduce new 
failure modes or mechanisms associated with plant operation and will 
not create a new accident type.
    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 applicable margin of safety is the period of time that 
equipment important to safety is inoperable. There is no increase in 
risk that is a result of the proposed change because (1) affected 
non-TS components are not safety significant, (2) compensatory 
measures are procedurally established for those components governed 
by other regulation (i.e., 10 CFR [Part] 50, Appendix R), and (3) 
TS-related component out-of-service time or related compensatory 
actions are governed by other existing TSs. The proposed change does 
not affect any safety limits, other operational parameters, or 
setpoints in the TS, nor does it affect any margins assumed in the 
accident analyses. In addition, Y-28 and C-540 operability 
requirements will be relocated to the TRM, which is part of the 
Safety Analysis Report (SAR) and controlled by 10 CFR 50.59.
    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

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

    Date of amendments request: June 30, 2006.
    Description of amendments request: The amendments would relocate 
the movable incore detectors and radioactive gaseous effluent oxygen 
monitoring instrumentation from the Technical Specifications to the 
Updated Final Safety Analysis Report (UFSAR).
    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 change[s] involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would relocate Technical Specification 
(TS) 3/4.3.3.2, ``Movable Incore Detectors,'' and TS 3/4.3.3.9 from 
the TS to the UFSAR. Movable Incore Detectors and Radioactive 
Gaseous Effluent Oxygen Monitoring Instrumentation are not 
initiators to any accident previously evaluated. Consequently, the 
probability of an accident previously evaluated is not significantly 
increased. Movable Incore Detectors and Radioactive Gaseous Effluent 
Oxygen Monitoring Instrumentation are not accident mitigating 
structures, systems, or components. No impact on the plant response 
to accidents will be created. Thus the consequences of accidents 
previously analyzed are unchanged between the existing TS 
requirements and the proposed changes.
    The proposed revision to TS SR [Surveillance Requirement] 
4.11.2.5 is an administrative change to a reference necessitated by 
the proposed relocation of TS Table 3.3-13 from the TS to the UFSAR. 
The proposed revision to the TS Index, page renumbering, and minor 
format changes to improve consistency are also administrative 
changes necessitated by the proposed relocation of TS 3/4.3.3.2 and 
TS 3/4.3.3.9 from the TS to the UFSAR.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or radiological consequences of an 
accident previously evaluated.
    2. Do the proposed change[s] create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated in 
the UFSAR. No new accident scenarios, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
changes. Specifically, no new hardware is being added to the plant 
as part of the proposed changes, no existing equipment is being 
modified, and no significant changes in operations are being 
introduced.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed change[s] involve a significant reduction in 
a margin of safety?
    Response: No.
    The proposed changes will not alter any assumptions, initial 
conditions, or results of any accident analyses. The Movable Incore 
Detectors and oxygen monitoring instrumentation will continue to 
perform as before. The proposed changes relocate TS 3/4.3.3.2 and TS 
3/4.3.3.9 from the TS to the UFSAR consistent with the guidance in 
NRC Generic Letter 95-10 and 10 CFR 50.36, and make conforming 
administrative changes to the TS Index, page renumbering, and minor 
format changes to improve consistency and to TS SR 4.11.2.5 to 
reflect the relocation of TS 3/4.3.3.9 from the TS to the Salem 
UFSAR.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendments request: September 26, 2006.
    Description of amendments request: The amendments would revise 
Technical Specification 6.9.1.9 to remove the revision number and date 
for the topical reports that contain the analytical methods used in the 
Core Operating Limits Report (COLR). The effect of this change is to 
allow the licensee to use current topical reports, as long as they have 
been approved by the NRC. The amendments would also add an NCR-approved 
topical report to the Salem Nuclear Generating Station, Unit No. 2, 
COLR methods.
    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 affect the administrative controls section 
of Technical Specifications (TS) that govern the analytical methods 
used to determine core operating limits. Removal of revision levels 
and dates from NRC-approved methods listed in TS is

[[Page 65144]]

an administrative change that has no impact on the probability or 
consequences of an accident. TS 6.9.1.9.b will still require these 
methods to be reviewed and approved by [the] NRC. The proposed 
change does not affect the required TS actions to be taken in the 
event that any core operating limits are exceeded.
    The proposed use of WCAP-10054-P-A, Addendum 2 for the Salem 
Unit 2 Small Break Loss of Coolant Accident (SBLOCA) analysis is 
consistent with the limitations and conditions of NRC approval. The 
parameters assumed in the analysis are within the design limits of 
the plant equipment. Therefore, there will be no increase in the 
probability of a loss of coolant accident. The consequences of a 
LOCA are not being increased, since it is shown that the Emergency 
Core Cooling System (ECCS) is designed so that its calculated 
cooling performance conforms to the criteria contained in 10 CFR 
50.46, Paragraph b. No other accident is potentially affected by 
this change.
    Therefore, the proposed change does 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.
    No new modes of plant operation are being introduced. The 
parameters assumed in the analysis are within the design limits of 
the plant equipment. TS will continue to require operation within 
the core operating limits determined using NRC-approved analytical 
methods and the proposed change does not affect any actions required 
in the event the core operating limits are exceeded.
    Therefore, the proposed change does not involve an increase in 
the probability or consequences of an accident previously evaluated.
    3. Do the proposed change[s] involve a significant reduction in 
a margin of safety?
    Response: No.
    The proposed changes do not have any impact on plant equipment 
or safety analysis acceptance criteria. Core operating limits will 
continue to be determined using NRC-approved analytical methods. The 
ECCS acceptance criteria of 10 CFR 50.46 will continue to be met 
following the proposed use of WCAP-10054-P-A, Addendum 2 for the 
Salem Unit 2 SBLOCA analysis[.]
    Therefore, the proposed change[s] do[es] 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station, 
Unit No. 2, Salem County, New Jersey

    Date of amendment request: April 6, 2006.
    Description of amendment request: The amendment would change the 
Technical Specifications to reduce the maximum allowable reactor power 
when two main steam safety valves (MSSVs) are inoperable.
    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[es] the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change reduces the power level at which Salem Unit 
2 may be operated with a maximum of two inoperable MSSVs in any 
steam generator. This change is consistent with analyses of the 
limiting transients for secondary system pressure (loss of load/
turbine trip and rod withdrawal at power), performed to demonstrate 
the acceptance criterion of 110% of design pressure will continue to 
be met following steam generator replacement. The proposed change 
does not involve any changes to the MSSV actuation setpoints; they 
remain well above the Main Steam System operating pressures. The 
proposed change does not challenge the relief capacity of the MSSVs. 
Therefore, the probability of an event associated with mis-operation 
of the MSSVs (e.g., inadvertent depressurization of the Main Steam 
System) is not impacted by the proposed change. The proposed 
reduction in allowable power level establishes initial conditions 
consistent with the safety analyses, and does not affect the 
probability of any previously evaluated accident.
    The proposed change is necessitated by analyses of limiting 
secondary system pressure transients, whose acceptance criteria 
continue to be met provided that plant operation is restricted to 
58% RTP [rated thermal power] with a maximum of two inoperable MSSVs 
in any steam generator. There is no impact on any radiological 
consequences of an accident associated with the proposed reduction 
in maximum power level.
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Do[es] the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Reducing the allowable power level per the proposed change does 
not introduce any new accident scenarios or malfunctions. The 
proposed change establishes an operating restriction consistent with 
current safety analysis methodology. It represents a change to an 
input assumption used in analyses of limiting secondary 
pressurization transients to ensure plant operation does not 
challenge any design limits.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do[es] the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    Acceptable margins of safety are inherent in the safety analysis 
acceptance criteria, including the limit on secondary system 
pressure to 110% of design pressure during a loss of load/turbine 
trip (LOL/TT) or rod withdrawal at power (RWAP) transient. The 
purpose of the proposed change is to limit operation with a maximum 
of two inoperable MSSVs for any steam generator, such that the 
acceptance criterion for secondary pressure continues to be met. The 
proposed change does not modify any acceptance criteria, nor would 
it cause any design limit to be exceeded.
    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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: September 29, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.7.8, ``Service Water (SW) System,'' to 
change the limiting conditions for operation (LCOs), Actions, 
Completion Times, and Surveillance Requirements (SRs). Specifically, 
the proposed amendment would change the LCO to require a specific 
number of SW pumps to be operable rather than the current SW train 
operability. The LCO Actions, Completion Times, and SRs would also be 
revised based on pump operability.
    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:


[[Page 65145]]


    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The safety related function of the Service Water (SW) System is 
to provide cooling for safety related equipment, mitigate the 
containment response effects of a Main Steam Line Break (MSLB) and 
design basis Loss of Coolant Accident (LOCA), and provide long term 
containment and core cooling in the event of a LOCA. The operation 
of the SW system, including the number of pumps operating or 
available, has no affect on the probability of these accidents.
    The probability of a loss of SW event is not increased. The 
proposed TS provides for more restrictive actions for pump 
inoperability than the existing TS, thereby reducing the probability 
of this event.
    The consequences of a[n] MSLB or LOCA or other design basis 
accidents are not increased beyond that assumed in the accident 
analysis. Two service water pumps are sufficient for all accident 
mitigation functions. The change provides for adequate service water 
supply (2 pumps) for both normal and accident conditions. The 
availability of associated power supplies is also considered. For a 
reduction in the total number of available pumps, appropriate LCO 
action statements ensure that the pumps are returned to service 
within a time limit commensurate with an acceptable level of plant 
safety and risk, or the plant is placed in a safe mode.
    The loss of SW has been previously evaluated and measures 
implemented to mitigate the event. Since a loss of SW assumes no SW 
pumps are operating, the proposed amendment has no affect on 
consequences of this event.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The only accidents directly initiated from this system are the 
loss of SW or flooding concerns. Both of these accidents have been 
previously evaluated with acceptable results. Therefore, this change 
does not create the possibility of a new or different [kind] of 
accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change will ensure that sufficient SW pumps are available 
for accident mitigation at any one time while still providing the 
appropriate operational flexibility. A risk determination 
demonstrates that any increase in risk associated with this change 
is within the established regulatory guidelines. The technical 
analysis shows that appropriate action statements exist to ensure 
adequate SW is available for accident mitigation, considering 
emergency power supply availability. Therefore, this 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: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Branch Chief: Richard J. Laufer.

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

    Date of amendment request: October 12, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.3.3, ``Capacity,'' to change the 
limit on the number of fuel assemblies in the spent fuel pool. The 
proposed amendment would also revise TS 3.7.13, ``Spent Fuel Pool 
Storage,'' to remove the references to Type 4 spent fuel pool storage 
racks, which are not currently installed.
    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 reduces the total number of fuel assemblies 
that can be stored in the current spent fuel pool storage locations 
and reduces the number of available locations. This will limit the 
potential inventory of spent fuel in the pool. The probability of an 
accident has not changed since the number of stored fuel assemblies 
is not a precursor for a spent fuel handling accident. A comparison 
of the criticality analysis of fuel assemblies to be used in 
subsequent Extended Power Uprate core reloads to the current 
criticality analysis has been performed. The design parameter 
assumptions used in the licensing basis criticality analyses are 
bounding.
    There are no new components or new functions associated with the 
spent fuel cooling system so the probability of an accident has not 
changed. The effect of a single failure on the spent fuel pool 
system's capability to provide for heat removal from the fuel pool 
has been analyzed. The analysis concluded that the system remains 
within the parameters previously evaluated. The implementation of 
the Extended Power Uprate does not affect the capability of the 
system to perform its function.
    The Extended Power Uprate conditions do not add any new or 
previously unevaluated materials to the spent fuel pool storage 
system and do not include any reductions in the boron concentration 
requirements so the probability of an accident has not changed. The 
total soluble boron concentration required to maintain the spent 
fuel pool in a subcritical condition with the transition to the new 
fuel has not changed. The conclusions in the Ginna UFSAR [Updated 
Final Safety Analysis Report], assuming the most limiting accident, 
remain valid.
    Therefore, the consequences of a fuel handling accident, a loss 
of spent fuel cooling, and a boron reduction concentration event 
previously evaluated have not increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not alter the function of the spent fuel 
pool or any related equipment, nor cause it to operate differently 
than it was designed to operate. All equipment required to mitigate 
the consequences of an accident would continue to operate as before. 
The proposed changes reduce the maximum number of fuel assemblies 
that can be stored in the spent fuel pool and the number of storage 
locations. Therefore, this change does not create the possibility of 
a new or different [kind] of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes reduce the maximum number of fuel 
assemblies that can be stored in the spent fuel pool and the number 
of storage locations. The changes are in accordance with conclusions 
supporting Extended Power Uprate and have been determined to be 
acceptable. The design parameter assumptions used in the licensing 
basis criticality analysis bound those of the new fuel assemblies. 
Although the individual heat load per assembly has increased due to 
the changed fuel design, the maximum spent fuel pool heat load has 
decreased due to the reduction in the number of fuel assemblies that 
will be stored based on future plans to use dry cask storage. 
Therefore, this proposed change does not reduce 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: Daniel F. Stenger, Ballard Spahr Andrews & 
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 
20005.
    NRC Branch Chief: Richard J. Laufer.

[[Page 65146]]

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

    Date of amendment request: October 3, 2006.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications surveillance requirements related 
to inspection of the containment sump trash racks and screens, inside 
recirculation spray (RS) pump wells, and outside RS and low head safety 
injection pump suction inlets resulting from Nuclear Regulatory 
Commission's (NRC's) Generic Safety Issue (GSI) 191 and Generic Letter 
(GL) 2004-02.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not impact the condition or performance 
of any plant structure, system or component. Furthermore, the 
proposed change does not affect the initiators of any previously 
analyzed event or the assumed mitigation of accident or transient 
events since the plant will be operated in the same manner and 
within the same operating limits that are currently in place. The 
proposed TS change is administrative in nature given that inspection 
of containment sump components for debris accumulation and 
structural integrity will continue to be performed. The revised TS 
surveillance wording is being implemented as a clarification to 
facilitate inspection of the containment sump in its current 
configuration, as well as after containment sump modifications have 
been implemented in response to GSI-191 and GL 2004-002. As a 
result, the proposed change to the Surry TS does not involve any 
increase in the probability or the consequences of any accident or 
malfunction of equipment important to safety previously evaluated 
since neither accident probabilities nor consequences are being 
affected by this proposed change.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change is administrative in nature and, as such, 
does not involve any changes in station operation or physical 
modifications to the plant. In addition, no changes are being made 
in the methods used to respond to plant transients that have been 
previously analyzed. No changes are being made to plant parameters 
within which the plant is normally operated or in the setpoints, 
that initiate protective or mitigative actions, since the plant will 
be operated in the same manner and within the same operating limits 
that are currently in place. Since plant operation will not be 
affected by this change, no new failure modes are being introduced. 
Therefore, the proposed change to the Surry TS does not create the 
possibility of a new or different kind of accident or malfunction of 
equipment important to safety from any previously evaluated.
    3. Does the change involve a significant reduction in the margin 
of safety?
    The proposed change is administrative in nature given that 
inspection of the containment sump components for debris 
accumulation and structural integrity will continue to be performed 
on an established frequency. The more general nature of the TS 
surveillance wording is being implemented as a clarification to 
facilitate inspection of the containment sump in its current 
configuration, as well as after containment sump modifications have 
been implemented in response to GSI-191 and GL 2004-002. The 
proposed change does not impact station operation or any plant 
structure, system or component that is relied upon for accident 
mitigation. Furthermore, the margin of safety assumed in the plant 
safety analysis is not affected in any way by the proposed change 
since the plant will be operated in the same manner and within the 
same operating limits and setpoints that are currently in place. 
Therefore, the proposed change to the Surry Technical Specifications 
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: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Branch Chief: Evangelos C. Marinos.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to [email protected].

Carolina Power & Light Company, Docket No. 50-261, H.B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: November 30, 2005.
    Brief description of amendment: The amendment revises the 
surveillance requirements (SR) for the emergency diesel generator 
automatic trips bypass of SR 3.8.1.11 from 18 months to 24 months.
    Date of issuance: October, 4, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No. 208.
    Renewed Facility Operating License No. DPR-23. Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: February 28, 2006 (71 
FR 10072).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 4, 2006.

[[Page 65147]]

    No significant hazards consideration comments received: No.

Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

Duke Power Company LLC, et al., Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: July 27, 2005, as supplemented 
May 4, 2006, and August 8, 2006.
    Brief description of amendments: The amendments revise the Catawba 
and McGuire Technical Specification 3.4.15, ``RCS Leakage Detection 
Instrumentation.''
    Date of issuance: September 30, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 234/230 and 235/217.
    Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9 and 
NPF-17: Amendments revised the licenses and the technical 
specifications.
    Date of initial notice in Federal Register: August 30, 2006 (71 FR 
51644).
    The supplement dated August 8, 2006, provided clarifying 
information that did not expand the scope of the July 27, 2005, 
application as modified May 4, 2006.
    The Commission's related evaluation, Final No Significant Hazards 
Finding, and State consultation of the amendments are contained in a 
Safety Evaluation dated September 30, 2006.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. STN 50-457, Braidwood 
Station, Unit No. 2, Will County, Illinois

    Date of application for amendment: November 18, 2005, as 
supplemented by letters dated August 18 and September 28, 2006.
    Brief description of amendment: The amendment revised TS 5.5.9, 
``Steam Generator (SG) Tube Surveillance Program,'' regarding the 
required SG inspection scope for Braidwood Station, Unit No. 2, during 
refueling outage 12 and the subsequent operating cycle.
    Date of issuance: October 24, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 141.
    Facility Operating License No. NPF-77: The amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: (71 FR 29676; May 23, 
2006).
    The August 18 and September 28, 2006, supplements contained 
clarifying information and did not change the NRC staff's initial 
proposed finding of no significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 24, 2006.
    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: March 23, 2006.
    Description of amendment request: The amendment deleted License 
Condition 2.G, ``Reporting to the Commission,'' as described in the 
Notice of Availability published in the Federal Register on April 25, 
2006 (71 FR 23955). The change was requested as part of the 
consolidated line item improvement process and consistent with the 
model safety evaluation published in the Federal Register on November 
4, 2005 (70 FR 67202).
    Date of issuance: October 17, 2006.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 113.
    Facility Operating License No. NPF-86: The amendment revised 
Facility Operating License No. NPF-86 and the Technical Specifications.
    Date of initial notice in Federal Register: April 25, 2006 (71 FR 
23955).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 17, 2006.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 16, 2006.
    Brief description of amendment: The amendment revised the Technical 
Specification 3.10.1, ``Inservice Leak and Hydrostatic Testing 
Operation,'' to extend the scope to include provisions for temperature 
increases above 212 [deg]F as a consequence of inservice leak or 
hydrostatic testing, and as a consequence of control rod scram time 
testing initiated in conjunction with the inservice leak test or 
hydrostatic test, when initial test conditions are below 212 [deg]F.
    Date of issuance: October 23, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 225.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 1, 2006 (71 FR 
43535)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 23, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 7, 2005, as 
supplemented by letters dated July 20 and September 5, 2006.
    Brief description of amendments: These amendments revised the 
Technical Specifications to delete Surveillance Requirement (SR) 
4.9.2.b, which requires performance of a channel functional test (CFT) 
of each source range neutron flux monitor within 8 hours prior to the 
initial start of core alterations. An associated administrative change 
would renumber current SR 4.9.2.c as SR 4.9.2.b. The amendments would 
also eliminate the restriction in SRs 4.10.3.2 and 4.10.4.2 that the 
CFTs of the intermediate and power range monitors be performed within 
12 hours prior to initiating physics tests.
    Date of issuance: October 13, 2006.
    Effective date: As of the date of issuance, to be implemented in 60 
days.
    Amendment Nos.: 275, 257.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: August 2, 2006 (71 FR 
43819). The supplements provided clarifying information that did not 
change the initial proposed no significant hazards consideration 
determination or expand the application beyond the scope of the 
original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 13, 2006.
    No significant hazards consideration comments received: No.

[[Page 65148]]

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

    Date of application for amendment: November 15, 2005, as 
supplemented May 31, August 31, and September 29, 2006.
    Brief description of amendment: The amendment revises the Virgil C. 
Summer Nuclear Station Technical Specifications (TS) 3/4.3 for the 
reactor trip instrumentation and the engineered safety feature 
actuation system instrumentation to implement the allowed outage time 
and bypass test time changes approved in WCAP-14333-P-A, Revision 1, 
``Probabilistic Risk Analysis of the RPS and ESFAS Test Times and 
Completion Times,'' and makes several additional changes to TS outside 
of the scope of WCAP-14333.
    Date of issuance: October 24, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No. 177.
    Renewed Facility Operating License No. NPF-12: Amendment revises 
the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 2005 (70 
FR 75496).
    The supplemental letters provided clarifying information that was 
within the scope of the initial notice and did not change the initial 
proposed no significant hazards consideration. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
October 24, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 10, 2003 (TS-430), as 
supplemented by letter dated November 8, 2004.
    Brief description of amendment: The amendment incorporates the 
necessary Technical Specification (TS) changes for the planned 
replacement of the power range monitoring portion of the existing 
Neutron Monitoring System with a digital upgrade. These changes expand 
the current allowable operating domain to the Maximum Extended Load 
Line Limit region of the power/flow chart.
    Date of issuance: September 27, 2006.
    Effective date: Date of issuance, to be implemented within 30 days.
    Amendment No.: 262.
    Facility Operating License No. DPR-33: Amendment revised the TSs.
    Date of initial notice in Federal Register: February 3, 2004 (69 FR 
5208). The November 8, 2004, supplement, contained clarifying 
information and did not change the NRC staff's initial proposed finding 
of no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 27, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 6, 2006.
    Brief description of amendments: The amendments modify Technical 
Specification (TS) requirements for inoperable snubbers by adding 
Limiting Condition for Operation 3.0.7. This operating license 
improvement was made available by the Nuclear Regulatory Commission 
(NRC) on May 4, 2005 (70 FR 23252) as part of the consolidated line 
item improvement process and is consistent with NRC approved Technical 
Specification Task Force (TSTF) standard TS change TSTF-372, Revision 
4.
    Date of issuance: October 4, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos. 312/301.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the technical specifications.
    Date of initial notice in Federal Register: March 28, 2006 (71 FR 
15487).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 4, 2006.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 30th day of October 2006.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E6-18595 Filed 11-6-06; 8:45 am]
BILLING CODE 7590-01-P