Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 62306-62318 [E6-17546]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 205 (Tuesday, October 24, 2006)]
[Notices]
[Pages 62306-62318]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E6-17546]



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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 to publish notice of any amendments 
issued, or proposed to be issued, and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from September 29, 2006, to October 12, 2006. The 
last biweekly notice was published on October 10, 2006 ( 71 FR 59529).

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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or

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fact. Contentions shall be limited to matters within the scope of the 
amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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: August 16, 2006.
    Description of amendments request: The proposed amendments would 
revise several Surveillance Requirements (SRs) in Technical 
Specification (TS) 3.8.1, ``AC Sources--Operating,'' to allow these SRs 
to be performed, or partially performed, in reactor modes that 
currently are not allowed by the TSs. The proposed changes would also 
require certain SRs to be performed at a power factor of <=0.9 if 
performed with the emergency diesel generators synchronized to the 
grid, unless grid conditions do not permit.
    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 emergency diesel generators (DGs) and their associated 
emergency loads are accident mitigating features, rather than 
accident initiating equipment. Each DG is dedicated to a specific 
vital bus and these buses and DGs are independent of each other. 
There is no common mode failure provided by the testing changes 
proposed in this license amendment request (LAR) that would cause 
multiple bus failures. Therefore, there will be no significant 
impact on any accident probabilities by the approval of the 
requested amendment.
    The design of plant equipment is not being modified by these 
proposed changes.
    The changes include an increase in the online time the DG will 
be paralleled to the grid in Mode[s] 1, 2, 3, and 4. The overall 
time that the DG is paralleled in all modes (outage/non-outage) 
should remain unchanged. As such, the ability of the DGs to respond 
to a design basis accident (DBA) can be adversely impacted by the 
proposed changes. However, the impacts are not considered 
significant based on the DG under test maintaining its ability to 
respond to an auto-start signal were one to be received during 
testing, along with the ability of the remaining DG to mitigate a 
DBA or provide a safe shutdown, and data that shows that the DG 
itself will not perturb the electrical system significantly. 
Furthermore, the proposed amendments for surveillance requirements 
(SR) 3.8.1.10 and SR 3.8.1.14 share the same electrical 
configuration alignment to the current monthly 1-hour loaded 
surveillance.
    SR changes that are consistent with Industry/Technical 
Specification Task Force (TSTF) Standard Technical Specification 
(STS) change TSTF-283, Revision 3 and NUREG-1432, Revision 2 have 
been approved by the NRC, and the on-line tests allowed by the TSTF 
and the NUREG are only to be performed for the purpose of 
establishing operability of the DG being tested. Performance of 
these SRs during previously restricted modes will require an 
assessment to assure plant safety is maintained or enhanced.
    The proposed changes to SRs 3.8.1.10 and 3.8.1.14 to require 
that these SRs be performed at a power factor of <=0.9 if performed 
with the emergency diesel generators synchronized to the grid unless 
grid conditions do not permit are consistent with NRC-approved 
NUREG-1432, Standard Technical Specifications, Combustion 
Engineering Plants, and NRC-approved TSTF-276, Revision 2. This 
requirement ensures that the DG is tested under load conditions that 
are as close to design basis conditions as possible. A power factor 
of <=0.9 is representative of the actual inductive loading a DG 
would see under design basis accident conditions. Under certain 
conditions, however, the proposed change allows the surveillance to 
be conducted at a power factor other than <=0.9. These conditions 
occur when grid voltage is high, and the additional field excitation 
needed to get the power factor to <=0.9 results in voltages on the 
emergency busses that are too high. Under these conditions, the 
power factor should be maintained as close as practicable to 0.9 
while still maintaining acceptable voltage limits on the emergency 
busses. In other circumstances, the grid voltage may be such that 
the DG excitation levels needed to obtain a power factor of 0.9

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may not cause unacceptable voltages on the emergency busses, but the 
excitation levels are in excess of those recommended for the DG. In 
such cases, the power factor shall be maintained as close as 
practicable to 0.9 without exceeding DG excitation limits.
    As stated above, a power factor <=0.9 should be able to be 
achieved when performing this SR at power and synchronized with 
offsite power by transferring house loads from the auxiliary 
transformer to the startup transformer in order to lower the Class 
1E bus voltage. Transferring house loads from the auxiliary 
transformer to the startup transformer is routinely performed at 
power, in accordance with procedure 40OP-9NA03. The circuit breakers 
supplying the house loads (NAN-S01 and NAN-S02) from the auxiliary 
and startup transformers are interlocked such that one supply 
breaker does not open until the alternate supply breaker is closed. 
This ensures that the bus remains energized during the transfer.
    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 accident from any accident previously evaluated?
    Response: No.
    The proposed changes would create no new accidents since no 
changes are being made to the plant that would introduce any new 
accident causal mechanisms. Equipment will be operated in the same 
configuration currently allowed by other DG SRs that allow testing 
in plant Modes 1, 2, 3, and 4. This license amendment request does 
not impact any plant systems that are accident initiators or 
adversely impact any accident mitigating systems.
    Therefore, the proposed change does not create the possibility 
of a new or different 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 do not involve a significant reduction in a 
margin of safety. The margin of safety is related to the ability of 
the fission product barriers to perform their design safety 
functions during and following an accident situation. These barriers 
include the fuel cladding, the reactor coolant system, and the 
containment system. The proposed changes to the testing requirements 
for the plant DGs do not affect the operability requirements for the 
DGs, as verification of such operability will continue to be 
performed as required (except during different allowed modes). 
Continued verification of operability supports the capability of the 
DGs to perform their required function of providing emergency power 
to plant equipment that supports or constitutes the fission product 
barriers. Only one DG is tested at a time and the remaining DG will 
be available to safely shut down the plant or respond to a DBA, if 
required. Consequently, the performance of these fission product 
barriers will not be impacted by implementation of the proposed 
amendment.
    In addition, the proposed changes involve no changes to safety 
setpoints or limits established or assumed by the accident analysis. 
On this and the above basis, no safety margins will be impacted.
    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 Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin

    Date of amendment request: September 25, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.2.a, ``ASME Code Class 1, 2, 3, 
and MC Components and Supports.'' The revised TS 4.2.a, Item 2, would 
reference the American Society of Mechanical Engineers Code for 
Operation and Maintenance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed change revises the Kewaunee Power Station 
(Kewaunee) Technical Specification (TS) TS 4.2.a.2 regarding in-
service testing of ASME Code Class 1, Class 2 and Class 3 pumps and 
valves. The proposed change revises the TS to be consistent with the 
requirements of 10 CFR [Title 10, Code of Federal Regulations] 
50.55a(f)(4) for pumps and valves which are classified as American 
Society of Mechanical Engineers (ASME) Code Class 1, Class 2 and 
Class 3. The proposed change incorporates revisions to the ASME Code 
that result in a net improvement in the measures for in-service 
testing of pumps and valves.
    As a net improvement in the in-service testing of pumps and 
valves, the proposed change does not negatively impact any accident 
initiators, analyzed events, or assumed mitigation of accident or 
transient events. It does not involve the addition or removal of any 
equipment, or any design changes to the facility. Therefore, this 
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?
    No. The proposed change revises Kewaunee TS 4.2.a.2 regarding 
in-service testing of ASME Code Class 1, Class 2 and Class 3 pumps 
and valves, for consistency with the requirements of 10 CFR 
50.55a(f)(4). The proposed change incorporates revisions to the ASME 
Code that result in a net improvement in the measures for testing 
pumps and valves.
    The proposed change does not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or adversely affect methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. The proposed change does not 
alter existing test criteria or frequencies. Additionally, there is 
no change in the types or increases in the amounts of any effluent 
that may be released off-site and there is no increase in individual 
or cumulative occupational exposure. Therefore, this proposed change 
does not create the possibility of an accident of a different kind 
than previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No. The proposed change revises TS 4.2.a.2 regarding in-service 
testing of ASME Code Class 1, Class 2, and Class 3 pumps and valves, 
for consistency with the requirements of 10 CFR 50.55a(f)(4). The 
proposed change incorporates revisions to the ASME Code that result 
in a net improvement in the measures for testing pumps and valves. 
The safety function of the affected pumps and valves will continue 
to be confirmed through testing. 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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Branch Chief: M. Murphy (Acting).

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, 
Millstone Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut

    Date of amendment request: September 1, 2006.
    Description of amendment request: The proposed amendment would 
revise the Millstone Power Station, Unit Nos. 2 and 3 (MPS2 and MPS3) 
Technical Specifications (TSs) to replace the terms

[[Page 62309]]

``trash racks and screens'' with the term ``strainers''.
    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:

    Criterion 1:
    Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Although the configurations of the existing sump screen and the 
replacement strainer assemblies are different, they serve the same 
fundamental purpose of passively removing debris from the sump's 
suction supply of the supported system pumps. Replacing trash racks 
with strainers does not adversely impact the adequacy of pump net 
positive suction head assumed in the safety analyses. In fact, it 
will improve it. Likewise, the proposed change does not reduce the 
reliability of any supported systems or introduce any new system 
interactions. A missile evaluation of the new strainer design 
concluded that there is no credible missile that could damage the 
strainer when needed during a loss-of-coolant accident [LOCA]. A jet 
impingement evaluation of the new strainer design concluded that 
there are no credible high energy line break jets that could damage 
the strainer when needed during a LOCA. The greatly increased 
surface area of the new strainer will reduce the approach velocity 
of the strainer face significantly, further decreasing the risk of 
impact from large debris entrained in the sump flow stream. The 
proposed rewording of the SRs [surveillance requirements] will 
continue to ensure that the ECCS [emergency core cooling system] 
sump suction inlet strainers show no evidence of structural distress 
or abnormal corrosion for MPS2 and [MPS]3 with or without the 
strainer modification complete. As such, the proposed change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Criterion 2:
    Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    During the next refueling outage for each unit, DNC [Dominion 
Nuclear Connecticut, Inc.] is replacing the ECCS trash racks and 
screens with strainers in support of the response to Generic Letter 
2004-02 on Millstone Units 2 and 3. The ECCS strainers are passive 
components in standby safety systems used for accident mitigation. 
As such, they are not accident initiators. Therefore, there is no 
possibility that this change could create any accident of any kind. 
A change to TS SRs 4.5.2.j for MPS2 and 4.5.2.d.2 for MPS3 addresses 
differences in nomenclature between the existing and [Generic Safety 
Issue] GSI-191 designs. These changes do not alter the nature of 
events postulated in the Final Safety Analysis Report nor do they 
introduce any unique precursor mechanisms. Therefore, the proposed 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    Criterion 3:
    Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not adversely affect any plant safety 
limits, set points, or design parameters. The changes also do not 
adversely affect the fuel, fuel cladding, reactor coolant system 
(RCS), or containment integrity. Therefore, the proposed TS change, 
which revises the terminology associated with TS SRs, 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, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Acting Branch Chief: Brooke D. Poole.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Units 1 and 2 (ANO-1&2), Pope County, Arkansas

    Date of amendment request: October 25, 2005.
    Description of amendment request: The proposed change modifies 
inventory and inspection requirements associated with the Emergency 
Cooling Pond (ECP), which is a common cooling water source for ANO-1&2 
during conditions that may render the normal cooling water source 
(Dardanelle Reservoir) unavailable.
    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 indicated ECP level is an operator aid for routine 
verification that the required ECP inventory of 70 acre-feet is 
maintained. Relocation of this indication to the TS [technical 
specification] Bases does not change the design basis and, 
therefore, has no impact on any accident described in the SAR 
[safety analysis report]. The relocation of excessive SR 
[surveillance requirement] details to the TS Bases does not reduce 
the level of testing required with regard to ECP operability 
verifications. Actual ECP inspection is more detailed than that 
currently described in the TSs. The relocation of this excessive 
detail to the TS Bases, therefore, has no impact on any accident 
described in the SAR. Finally, the inclusion of a new Action 
associated with the discovery of degradation of the ECP structure is 
more restrictive in that the proposed engineering evaluation must be 
performed within 7 days. Previously, the TS Bases did not require a 
completion time for this action. Actions associated with TS Limiting 
Conditions for Operation (LCO) or SRs are below the level of detail 
described in the SAR and, therefore, have no impact on any accident 
currently described in the SAR.
    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 aforementioned proposed change to the TSs does not require 
any physical alteration to the plant or alter plant design. The ECP 
is not an accident initiator. The proposed change does not adversely 
impact the function of the ECP as credited in any safety analyses 
for the prevention or mitigation of any accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not adversely impact a margin of safety 
analysis for any accident previously evaluated. Relocation of the 
indicated ECP level that corresponds to the required ECP volume of 
70 acre-feet and the relocation of excessive SR details to the TS 
Bases will not result in a credible increase in nuclear safety risk. 
In addition, the TS Bases is part of the SAR and controlled under 10 
CFR 50.59. The inclusion of a new action relocated from the TS Bases 
to the TS with completion time constraint is more conservative than 
currently described in the TS Bases. The proposed change acts to 
correct current TS deficiencies and, therefore, is considered risk 
neutral.
    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.

[[Page 62310]]

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: September 19, 2006.
    Description of amendment request: The proposed change will revise 
River Bend Station, Unit 1, (RBS) Technical Specifications (TS) 
Surveillance Requirement 3.6.1.3.5 to replace the currently specified 
frequency for leak testing containment purge supply and exhaust 
isolation valves with resilient seal materials with a requirement to 
test these valves in accordance with the Containment Leakage Rate 
Testing Program. The RBS Containment Leakage Rate Testing Program is 
implemented in accordance with the Code of Federal Regulations, Part 
50, Appendix J, Option B, and Regulatory Guide (RG) 1.163, 
``Performance-Based Containment Leak Test Program,'' dated September 
1995. RG 1.163 allows a nominal test interval of 30 months for 
containment purge and vent valves.
    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.
    This change deletes the augmented testing requirement for these 
containment isolation valves and allows the surveillance intervals 
to be set in accordance with the Containment Leakage Rate Testing 
Programs. This change does not affect the system function or design. 
The purge valves are not an initiator of any previously analyzed 
accident. Leakage rates do not affect the probability of the 
occurrence of any accident. Operating history has demonstrated that 
the valves do not degrade and cause leakage as previously 
anticipated. Because these valves have been demonstrated to be 
reliable, these valves can be expected to perform the containment 
isolation function as assumed in the accident analyses. Therefore, 
there is no significant increase in the consequences of any 
previously evaluated accident.
    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.
    Extending the test intervals has no influence on, nor does it 
contribute in any way to, the possibility of a new or different kind 
of accident or malfunction from those previously analyzed. No change 
has been made to the design, function or method of performing 
leakage testing. Leakage acceptance criteria have not changed. No 
new accident modes are created by extending the testing intervals. 
No safety-related equipment or safety functions are altered as a 
result of this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The only margin of safety that has the potential of being 
impacted by the proposed changes involves the offsite dose 
consequences of postulated accidents which are directly related to 
the containment leakage rate. The proposed change does not alter the 
method of performing the tests nor does it change the leakage 
acceptance criteria. Sufficient data has been collected to 
demonstrate these resilient seals do not degrade at an accelerated 
rate.
    Because of this demonstrated reliability, this change will 
provide sufficient surveillance to determine an increase in the 
unfiltered leakage prior to the leakage exceeding that assumed in 
the accident analysis.
    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.

Nuclear Management Company, LLC, Docket No. 50-255, Palisades Nuclear 
Plant (PNP), Van Buren County, Michigan

    Date of amendment request: May 30, 2006.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS), Section 5.5.8, ``Steam Generator 
Program,'' to modify the steam generator (SG) provisions for tube 
inspections, as contained in the PNP TS Surveillance Requirements, 
Section 5.5.8.d. The purpose of these changes is to define the depth of 
the required tube inspections. WCAP-16208-P, ``NDE Inspection Length 
for CE [Combustion Engineering] Steam Generator Tubesheet Region 
Explosive Expansions,'' Revision 1, provided recommended tubesheet 
region inspection lengths for plants with CE-supplied steam generators 
with explosive expansions. This inspection length is referred to as C* 
(``C-Star''). Nuclear Management Company (NMC) intends to implement the 
C* inspection methodology for PNP.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not involve a significant increase 
in the probability of an accident previously evaluated because the 
modification to TS Section 5.5.8.d maintains the existing design 
limits and would not increase the probability or consequences of an 
accident involving tube burst or primary to secondary accident-
induced leakage, as previously analyzed in the UFSAR [Updated Final 
Safety Analysis Report]. Also, the tube burst and collapse criteria 
of NRC [Nuclear Regulatory Commission] Regulatory Guide 1.121, 
``Basis for Plugging Degraded PWR Steam Generator Tubes,'' would 
continue to be satisfied.
    Tube burst is precluded for a tube with defects within the 
tubesheet region because of the constraint provided by the 
tubesheet. As such, tube pullout resulting from the axial forces 
induced by primary to secondary differential pressures would be a 
prerequisite for tube burst to occur. A joint industry test program, 
WCAP-16208-P, has defined the nondegraded tube to tubesheet joint 
length required to preclude tube pullout C [deg]) and maintain 
acceptable primary to secondary accident-induced leakage, assuming a 
360[deg] circumferential through wall crack existed immediately 
below this length. For PNP, C [deg] is 12.5 inches. Any degradation 
below C [deg] is shown by empirical test results and analyses to be 
acceptable, thereby precluding an event with consequences similar to 
a postulated tube rupture event.
    WCAP-1 6208-P incorporates an assumed primary to secondary 
accident-induced leakage value of 0.1 gpm/SG. The NMC TSTF 
[Technical Specifications Task Force]-449 submittal to the NRC 
provided the PNP SG tube integrity related TS. LCO [Limiting 
Condition for Operation] 3.4.13, item d., ``PCS Operational 
Leakage,'' states that operational leakage through any one SG shall 
be limited to 150 gallons per day. The UFSAR Chapter 14.14-6 
accident-induced leakage limit assumption based on MSLB [main steam-
line break] is 0.3 gallons per minute (432 gallons per day). 
Therefore, the LCO leakage limit is conservatively less than the 
design basis accident induced leakage limit.
    In summary, the proposed modifications to the PNP Technical 
Specifications maintain existing design limits and do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated in the UFSAR.
    Therefore, operation of the facility in accordance with the 
proposed amendment

[[Page 62311]]

would not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because SG tube leakage and structural integrity will continue to be 
maintained during all plant conditions upon implementation of the 
proposed inspection scope to the PNP TSs. The revised inspection 
scope does not introduce any new mechanisms that might result in a 
different kind of accident from those previously evaluated. Even 
with the limiting circumstances of a complete circumferential 
separation (360-degree through wall crack) of a tube below the C* 
length, tube pullout is precluded and leakage is predicted to be 
maintained within the TS limits during all plant conditions.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment does not involve a significant reduction 
in a margin of safety. The requirements for the inspection of SG 
tubes are intended to ensure that this portion of the primary 
coolant system maintains its integrity. Tube integrity means that 
the tubes are capable of performing these functions in accordance 
with the plant design and licensing basis. Tube integrity includes 
both structural and leakage integrity. The proposed tubesheet 
inspection depth of 12.5 inches will ensure tube integrity is 
maintained because any degradation below C* is shown by empirical 
test results and analyses to be acceptable. In addition, operation 
with potential tube degradation below the C* inspection length 
continues to meet the margin of safety as defined by RG [Regulatory 
Guide] 1.121, ``Basis for Plugging Degraded PWR Steam Generator 
Tubes,'' and RG 1.83, ``Inservice Inspection of Pressurized Water 
Reactor Steam Generator Tubes.'' Therefore, the proposed 
modifications 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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Branch Chief: Martin C. Murphy, Acting Branch Chief.

Notice of Issuance of Amendments to Facility Operating Licenses

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

Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power 
Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: March 28, 2006.
    Brief description of amendment: The amendment revised Facility 
Operating License No. NPF-49 by deleting Section 2.F, which specifies 
reporting of violations of the requirements of Section 2.C of the 
renewed operating license.
    Date of issuance: October 4, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 234.
    Facility Operating License No. NPF-49: The amendment revised the 
License.
    Date of initial notice in Federal Register: May 9, 2006 (71 FR 
26997).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 4, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 19, 2005, as 
supplemented by letters dated February 28, May 31, and September 26, 
2006.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.6.2.1, ``Containment Spray System.'' Specifically, 
the change revised the allowable outage time (AOT) for TS 3.6.2.1 from 
72 hours to 7 days during fuel cycles 19 and 20. Per the license 
amendment request, the AOT extension may only be invoked twice (i.e., 
once for each train or twice for one train). The requested changes are 
sought to provide needed flexibility in the performance of selected 
corrective and preventative maintenance activities during power 
operations. Currently, the licensee's maintenance activities on 
containment spray system components are performed during the refueling 
outages; taking several days of ``around the clock'' effort.
    Date of issuance: September 28, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 268.
    Renewed Facility Operating License No. NPF-6: The amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: January 3, 2006 (71 FR 
148). The supplements dated February 28, May 31, and September 26, 
2006,

[[Page 62312]]

provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 28, 2006.
    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: August 17, 2005, as supplemented 
by letter dated May 19, 2006.
    Brief description of amendment: The proposed changes revised the 
Operating License Condition (OLC) 2.C.(41) to add reference to a 
Nuclear Regulatory Commission (NRC) Safety Evaluation that allows the 
application of certain risk-informed, performance-based fire protection 
methods and tools.
    Date of issuance: September 29, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No: 170.
    Facility Operating License No. NPF-29: The amendment revised the 
OLC 2.C.(41).
    Date of initial notice in Federal Register: October 25, 2005 (70 FR 
61658). The supplement dated May 19, 2006, provided additional 
information that clarified the change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 29, 2006
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: November 5, 2004.
    Brief description of amendment: The amendment modified Waterford 3 
Technical Specification (TS) 3.7.4, ``Ultimate Heat Sink,'' to provide 
clarification that the ambient temperature monitoring requirement that 
is specified in TS 3.7.4.d only applies when the affected ultimate heat 
sink train is considered to be operable. The NRC is not approving the 
request to delete TS 3.7.4.c, which would allow the plant to take 
credit for the dry cooling tower fans that are not protected from 
tornado missiles when a tornado warning is in effect.
    Date of issuance: September 28, 2006.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 208.
    Facility Operating License No. NPF-38: The amendment revised the 
Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: December 7, 2004 (69 FR 
70717).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 28, 2006.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: October 27, 2005.
    Brief description of amendment: The amendment modified Surveillance 
Requirement (SR) 4.5.2e of Technical Specification (TS) 3.5.2, ``ECCS 
[Emergency Core Cooling Systems] Subsystems--Modes 1, 2 and 3,'' SR 
4.6.2.1d of TS 3.6.2, ``Containment Spray System,'' and SR 4.7.3b of TS 
3.7.3, ``Component Cooling Water and Auxiliary Component Cooling Water 
Systems,'' to remove the words ``during shutdown.'' This will provide 
flexibility allowing components required to be tested by these SRs to 
be tested online. Additionally, a revision to delete SR 4.7.12.1c of TS 
3.7.12, ``Essential Services Chilled Water system,'' is approved. A 
modification permanently separating the safety and non-safety portions 
of the Essential Services Chilled Water system has eliminated the need 
for automatic isolation valves and thus this SR.
    Date of issuance: October 6, 2006.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 209.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications and the Facility Operating License.
    Date of initial notice in Federal Register: December 20, 2005 (70 
FR 75491).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 6, 2006.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: October 25, 2005.
    Brief description of amendment: The amendment modifies Waterford 3 
Technical Specification 6.9.1.11, ``Core Operating Limits Report 
COLR,'' to add a methodology that will allow the use of zirconium 
diboride burnable absorber coating on fuel pellets.
    Date of issuance: October 6, 2006.
    Effective date: As of the date of issuance and shall be implemented 
30 days from the date of issuance.
    Amendment No.: 210.
    Facility Operating License No. NPF-38: The amendment revised the 
Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: December 6, 2005 (70 FR 
72673).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 6, 2006.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: June 11, 2004, as supplemented 
by letters dated December 12, 2005, April 4, 2006, and July 28, 2006.
    Brief description of amendments: This amendment incorporated a 
revision to the Technical Specifications (TSs) and licensing and design 
bases that relocates surveillance test intervals of various TS 
surveillance requirements to a new program, the Surveillance Frequency 
Control Program, which will be located in the Administrative Controls 
Section of the TSs. These amendments are pilot submittals in support of 
the Boiling Water Reactor Owners' Group Risk-Informed Initiative 5b, 
``Relocate Surveillance Test Intervals to Licensee Control.''
    Date of issuance: September 28, 2006.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos. 186, 147.
    Facility Operating License Nos. NPF-39 and NPF-85. This amendment 
revised the facility operating licenses and the TSs.
    Date of initial notice in Federal Register: May 24, 2005 (70 FR 
29793). The supplements provided clarifying information that did not 
expand the scope of the application as originally

[[Page 62313]]

noticed, and did not change the NRC staff's original proposed no 
significant hazards consideration determination as originally published 
in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 28, 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: September 29, 2005, as supplemented on 
August 8, September 18, and September 28, 2006.
    Description of amendment request: The amendment revised the 
Seabrook Station, Unit No. 1 Technical Specifications (TSs) to permit a 
one-time change in the steam generator tube inspection requirements to 
include a sampling of the bulges and over-expansions for portions of 
the steam generator tubes within the hot-leg tubesheet region.
    Date of issuance: September 29, 2006.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 112.
    Facility Operating License No. NPF-86: The amendment revised the 
License and the Tss.
    Date of initial notice in Federal Register: November 8, 2005 (70 FR 
67749). The licensee's August 8 and September 28, 2006, supplements 
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 supplement dated September 18, 2006, modified the requested 
amendment to request a one-time change in lieu of a permanent one. This 
narrowing of scope did not alter the validity of the NRC staff's 
proposed no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated September 29, 2006.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of application for amendment: May 31, 2006.
    Brief description of amendment: The amendment approved elimination 
of the resistance temperature detector (RTD) bypass piping and 
installing fast response thermowell-mounted RTDs in the reactor coolant 
system loop piping. The amendment also revised Surveillance Requirement 
3.3.1.15 of the Technical Specifications, deleting the requirement to 
perform surveillance on the reactor coolant system RTD bypass loop flow 
rate.
    Date of issuance: October 6, 2006.
    Effective date: As of the date of issuance and shall be implemented 
prior to entry into Mode 2 from the fall 2006 refueling outage.
    Amendment No.: 296.
    Facility Operating License No. DPR-58: Amendment revise the 
Technical Specifications.
    Date of initial notice in Federal Register: July 5, 2006 (71 FR 
38182). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 6, 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: January 30, 2006, as supplement by May 
17 and August 29, 2006.
    Brief description of amendment: The amendment revised the Cooper 
Nuclear Station Technical Specification Section 5.5.12, ``Primary 
Containment Leakage Rate Testing Program,'' to allow a one-time 
extension of no more than 5 years for the Type A, Integrated Leakage 
Rate Test (ILRT) interval. This revision is a one-time exception to the 
10-year frequency of the performance-based leakage rate testing program 
for Type A tests as defined in Nuclear Energy Institute (NEI) document, 
NEI 94-01, Revision 0, ``Industry Guideline for Implementing 
Performance-Based Option of 10 CFR Part 50, Appendix J,'' pursuant to 
10 CFR Part 50, Appendix J, Option B. The requested exception is to 
allow the ILRT to be performed within 15 years from the last ILRT, last 
performed on December 7, 1998.
    Date of issuance: October 3, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 224.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 25, 2006 (71 FR 
23957). The supplement dated May 17 and August 29, 2006, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated October 3, 2006.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station, Unit No. 1, Oswego County, New York

    Date of application for amendment: January 18, 2006.
    Brief description of amendment: The amendment deletes the reference 
to the hydrogen monitors in Technical Specification 3.6.11, ``Accident 
Monitoring Instrumentation.''
    Date of issuance: October 2, 2006.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 191.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications and License.
    Date of initial notice in Federal Register: July 18, 2006 (71 FR 
40749) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 2, 2006.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: December 13, 2005, supplemented 
by letters dated June 7, and July 21, 2006.
    Brief description of amendments: The amendments revise technical 
specification 5.5.14 ``Containment Leakage Rate Testing Program'' for 
Prairie Island Nuclear Generating Plant Units 1 and 2, to allow a one-
time interval extension of no more than 5 years for the Appendix J Type 
A, Integrated Leakage Rate Test.
    Date of issuance: October 2, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 174 and 164.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 31, 2006 (71 FR 
5081) The supplemental information provided by letters dated June 7, 
and July 21, 2006, did not change the no significant hazards 
determination.

[[Page 62314]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 2, 2006.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, 
California

    Date of application for amendments: January 19, 2006, as 
supplemented by letter dated June 20, 2006.
    Brief description of amendments: The amendments deleted the 
antitrust conditions from the facility operating licenses.
    Date of issuance: October 2, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: Unit 1-189; Unit 2-191.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: April 14, 2006 (71 FR 
19551) The supplemental letter dated June 20, 2006, provided additional 
information that clarified the application, and did not expand the 
scope of the application as originally noticed. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated October 2, 2006.
    No significant hazards consideration comments received: No.

PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, 
Pennsylvania

    Date of application for amendments: November 9, 2004, as 
supplemented on December 15, 2005, June 30, 2006, August 18, 2006, and 
September 28, 2006.
    Brief description of amendments: The amendments revise the SSES 1 
and 2 Technical Specifications (TSs) 3.8.4, ``DC Sources--Operating,'' 
3.8.5, ``DC Sources-Shutdown,'' 3.8.6, ``Battery Cell Parameters,'' and 
add a new TS Section, 5.5.13, ``Battery Monitoring and Maintenance 
Program.'' These changes are consistent with TS Task Force (TSTF) 360, 
Revision 1.
    Date of issuance: September 28, 2006.
    Effective date: As of the date of issuance and to be implemented 
within 60 days.
    Amendment Nos.: 238 and 215.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the TSs and license.
    Date of initial notice in Federal Register: January 17, 2006 (71 FR 
2596). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 28, 2006.
    The supplements dated December 15, 2005, June 30, 2006, August 18, 
2006, and September 28, 2006, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original