[Federal Register Volume 71, Number 233 (Tuesday, December 5, 2006)]
[Notices]
[Pages 70553-70569]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-20329]


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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 November 9, 2006, to November 21, 2006. The 
last biweekly notice was published on November 21, 2006 (71 FR 67391).

[[Page 70554]]

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 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

[[Page 70555]]

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

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

    Date of amendment request: April 11, 2006
    Description of amendment request: The proposed amendment would add 
Technical Specification (TS) Limiting Condition for Operation (LCO) 
3.0.8 to allow a delay time for entering a supported system TS when the 
inoperability is due solely to an inoperable snubber. The proposed 
changes are consistent with approval of TS Task Force (TSTF) change 
TSTF-372, Revision 4, ``Addition of LCO 3.0.8, Inoperability of 
Snubbers.''
    The NRC staff issued a notice of availability of a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on November 24, 2004 (69 FR 68412).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an inoperable snubber if risk is assessed and managed. The 
postulated seismic event requiring snubbers is a low-probability 
occurrence and the overall TS system safety function would still be 
available for the vast majority of anticipated challenges. Therefore, 
the probability of an accident previously evaluated is not 
significantly increased, if at all. The consequences of an accident 
while relying on allowance provided by proposed LCO 3.0.8 are no 
different than the consequences of an accident while relying on the TS 
required actions in effect without the allowance provided by proposed 
LCO 3.0.8. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The addition 
of a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Therefore, this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated
    The proposed change does not involve a physical alteration of the 
plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is assessed 
and managed, will not introduce new failure modes or effects and will 
not, in the absence of other unrelated failures, lead to an accident 
whose consequences exceed the consequences of accidents previously 
evaluated. The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
Thus, this change does not create the possibility of a new or different 
kind of accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic event 
requiring snubbers is a low-probability occurrence and the overall TS 
system safety function would still be available for the vast majority 
of anticipated challenges. The risk impact of the proposed TS changes 
was assessed following the three-tiered approach recommended in 
Regulatory Guide 1.177. A bounding risk assessment was performed to 
justify the proposed TS changes. This application of LCO 3.0.8 is 
predicated upon the licensee's performance of a risk assessment and the 
management of plant risk. The net change to the margin of safety is 
insignificant. Therefore, this change does not involve a significant 
reduction in a margin of safety.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Power Company LLC, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: June 5, 2006.

[[Page 70556]]

    Description of amendment request: The amendments would revise the 
Technical Specifications (TSs) to clarify Surveillance Requirement (SR) 
3.8.1.13 and its associated Bases to state that the SR only verifies 
that non-emergency diesel generator (DG) trips are bypassed. It is 
based upon, and consistent with, Industry Technical Specification Task 
Force (TSTF), Standard Technical Specification Traveler, TSTF-400-A, 
Revision 1, ``Clarify Surveillance Requirement on Bypass of DG 
Automatic Trips.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would implementation of the changes proposed in this LAR 
(License Amendment Request) involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. This LAR clarifies the purpose of Surveillance Requirement 
(SR) 3.8.1.13, which is to verify that non-emergency automatic 
diesel generator (DG) trips are bypassed in an accident. The DG 
automatic trips and their bypasses are not initiators of any 
accident that has been previously evaluated. Therefore, the 
probability of any of these accidents is not significantly 
increased. The function of the DG in mitigating accidents is not 
changed. The revised SR continues to ensure that the DG will operate 
as assumed in the accident analyses. Therefore, the consequences of 
any accident previously evaluated are not affected as well.
    2. Would implementation of the changes proposed in this LAR 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    No. The changes proposed in this LAR only clarify the purpose of 
SR 3.8.1.13, which is to verify that non-emergency automatic DG 
trips are bypassed in an accident. The proposed change does not 
involve a physical change to the plant (no new or different type of 
equipment will be installed) or a change in the methods governing 
normal plant operation or testing. Thus, the changes proposed in 
this LAR do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Would implementation of the changes proposed in this LAR 
involve a significant reduction in a margin of safety?
    No. The changes proposed in this LAR only clarify the purpose of 
SR 3.8.1.13, which is to verify that non-emergency automatic DG 
trips are bypassed in an accident. These changes clarify the purpose 
of the SR, which is to verify that the DG is capable of performing 
its assumed safety function. The safety function of the DG is 
unaffected, so the changes do not affect the margin of safety.
    Therefore, this LAR 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: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Power Company LLC, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: April 11, 2006.
    Description of amendment request: The proposed amendment would add 
Technical Specification (TS) Limiting Condition for Operation (LCO) 
3.0.8 to allow a delay time for entering a supported system TS when the 
inoperability is due solely to an inoperable snubber. The proposed 
changes are consistent with approval of TS Task Force (TSTF) Change 
TSTF-372, Revision 4, ``Addition of LCO 3.0.8, Inoperability of 
Snubbers.''
    The NRC staff issued a Notice of Opportunity to Comment of a model 
safety evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on November 24, 2004 (69 FR 68412).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an inoperable snubber if risk is assessed and managed. The 
postulated seismic event requiring snubbers is a low-probability 
occurrence and the overall TS system safety function would still be 
available for the vast majority of anticipated challenges. Therefore, 
the probability of an accident previously evaluated is not 
significantly increased, if at all. The consequences of an accident 
while relying on allowance provided by proposed LCO 3.0.8 are no 
different than the consequences of an accident while relying on the TS 
required actions in effect without the allowance provided by proposed 
LCO 3.0.8. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The addition 
of a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Therefore, this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated
    The proposed change does not involve a physical alteration of the 
plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is assessed 
and managed, will not introduce new failure modes or effects and will 
not, in the absence of other unrelated failures, lead to an accident 
whose consequences exceed the consequences of accidents previously 
evaluated. The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
Thus, this change does not create the possibility of a new or different 
kind of accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic event 
requiring snubbers is a low-probability occurrence and the overall TS 
system safety function would still be available for the vast majority 
of anticipated challenges. The risk impact of the proposed TS changes 
was assessed following the three-tiered approach recommended in 
Regulatory Guide 1.177. A bounding risk assessment was performed to 
justify the proposed TS changes. This application of LCO 3.0.8 is 
predicated upon the licensee's performance of a risk assessment and the 
management of plant risk. The net change to the margin of safety is 
insignificant. Therefore, this change does not involve a significant 
reduction in a margin of safety.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.

[[Page 70557]]

    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Power Company LLC, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: April 11, 2006.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs) related to steam generator 
(SG) tube integrity. The changes are consistent with the consolidated 
line-item improvement process (CLIIP), Nuclear Regulatory Commission-
approved Revision 4 to Technical Specification Task Force (TSTF) 
Standard TS Change Traveler, TSTF-449, ``Steam Generator Tube 
Integrity.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The proposed change requires a SG Program that includes performance 
criteria that will provide reasonable assurance that the SG tubing will 
retain integrity over the full range of operating conditions (including 
startup, operation in the power range, hot standby, cooldown and all 
anticipated transients included in the design specification). The SG 
performance criteria are based on tube structural integrity, accident 
induced leakage, and operational LEAKAGE.
    A (steam generator tube rupture) SGTR event is one of the design 
basis accidents that are analyzed as part of a plant's licensing basis. 
In the analysis of a SGTR event, a bounding primary to secondary 
LEAKAGE rate equal to the operational LEAKAGE rate limits in the 
licensing basis plus the LEAKAGE rate associated with a double-ended 
rupture of a single tube is assumed.
    For other design basis accidents such as MSLB, rod ejection, and 
reactor coolant pump locked rotor the tubes are assumed to retain their 
structural integrity (i.e., they are assumed not to rupture). These 
analyses typically assume that primary to secondary LEAKAGE for all SGs 
is 1 gallon per minute or increases to 1 gallon per minute as a result 
of accident induced stresses. The accident induced leakage criterion 
introduced by the proposed changes accounts for tubes that may leak 
during design basis accidents. The accident induced leakage criterion 
limits this leakage to no more than the value assumed in the accident 
analysis.
    The SG performance criteria proposed change to the TS identify the 
standards against which tube integrity is to be measured. Meeting the 
performance criteria provides reasonable assurance that the SG tubing 
will remain capable of fulfilling its specific safety function of 
maintaining reactor coolant pressure boundary integrity throughout each 
operating cycle and in the unlikely event of a design basis accident. 
The performance criteria are only a part of the SG Program required by 
the proposed change to the TS. The program, defined by NEI 97-06, Steam 
Generator Program Guidelines, includes a framework that incorporates a 
balance of prevention, inspection, evaluation, repair, and leakage 
monitoring. The proposed changes do not, therefore, significantly 
increase the probability of an accident previously evaluated.
    The consequences of design basis accidents are, in part, functions 
of the DOSE EQUIVALENT 1-131 in the primary coolant and the primary to 
secondary LEAKAGE rates resulting from an accident. Therefore, limits 
are included in the plant technical specifications for operational 
leakage and for DOSE EQUIVALENT 1-131 in primary coolant to ensure the 
plant is operated within its analyzed condition. The typical analysis 
of the limiting design basis accident assumes that primary to secondary 
leak rate after the accident is 0.27 gallons per minute with no more 
than 135 gallons per day in any one SG, and that the reactor coolant 
activity levels of DOSE EQUIVALENT 1-131 are at the TS values before 
the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the requirements 
for SG inspections. The proposed change does not adversely impact any 
other previously evaluated design basis accident and is an improvement 
over the current TSs.
    Therefore, the proposed change does not affect the consequences of 
a SGTR accident and the probability of such an accident is reduced. In 
addition, the proposed changes do not affect the consequences of an 
MSLB (main steamline break), rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated
    The proposed performance based requirements are an improvement over 
the requirements imposed by the current technical specifications. 
Implementation of the proposed SG Program will not introduce any 
adverse changes to the plant design basis or postulated accidents 
resulting from potential tube degradation. The result of the 
implementation of the SG Program will be an enhancement of SG tube 
performance. Primary to secondary LEAKAGE that may be experienced 
during all plant conditions will be monitored to ensure it remains 
within current accident analysis assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility of a 
new or different type of accident from any accident previously 
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The SG tubes in pressurized water reactors are an integral part of 
the reactor coolant pressure boundary and, as such, are relied upon to 
maintain the primary system's pressure and inventory. As part of the 
reactor coolant pressure boundary, the SG tubes are unique in that they 
are also relied upon as a heat transfer surface between the primary and 
secondary systems such that residual heat can be removed from the 
primary system. In addition, the SG tubes isolate the radioactive 
fission products in the primary coolant from the secondary system. In 
summary, the safety function of an SG is maintained by ensuring the 
integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the SG Program to manage SG tube inspection, 
assessment, repair, and plugging. The requirements established by the 
SG Program are consistent with those in the applicable design codes and 
standards and are an improvement over the requirements in the current 
TSs.

[[Page 70558]]

    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the TS.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Power Company LLC, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: June 5, 2006.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TSs) to clarify Surveillance Requirement (SR) 
3.8.1.13 and its associated Bases to state that the SR only verifies 
that non-emergency diesel generator (DG) trips are bypassed. It is 
based upon, and consistent with, Industry Technical Specification Task 
Force (TSTF), Standard Technical Specification Traveler, TSTF-400-A, 
Revision 1, ``Clarify Surveillance Requirement on Bypass of DG 
Automatic Trips.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would implementation of the changes proposed in this LAR 
(License Amendment Request) involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. This LAR clarifies the purpose of Surveillance Requirement 
(SR) 3.8.1.13, which is to verify that non-emergency automatic 
diesel generator (DG) trips are bypassed in an accident. The DG 
automatic trips and their bypasses are not initiators of any 
accident that has been previously evaluated. Therefore, the 
probability of any of these accidents is not significantly 
increased. The function of the DG in mitigating accidents is not 
changed. The revised SR continues to ensure that the DG will operate 
as assumed in the accident analyses. Therefore, the consequences of 
any accident previously evaluated are not affected as well.
    2. Would implementation of the changes proposed in this LAR 
create the possibility of a new or different kind of accident from 
any accident previously evaluated?
    No. The changes proposed in this LAR only clarify the purpose of 
SR 3.8.1.13, which is to verify that non-emergency automatic DG 
trips are bypassed in an accident. The proposed change does not 
involve a physical change to the plant (no new or different type of 
equipment will be installed) or a change in the methods governing 
normal plant operation or testing. Thus, the changes proposed in 
this LAR do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Would implementation of the changes proposed in this LAR 
involve a significant reduction in a margin of safety?
    No. The changes proposed in this LAR only clarify the purpose of 
SR 3.8.1.13, which is to verify that non-emergency automatic DG 
trips are bypassed in an accident. These changes clarify the purpose 
of the SR, which is to verify that the DG is capable of performing 
its assumed safety function. The safety function of the DG is 
unaffected, so the changes do not affect the margin of safety. 
Therefore, this LAR 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: Ms. Lisa F. Vaughn, Duke Power Company LLC, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: July 31, 2006.
    Description of amendment request: The proposed amendments would 
revise Technical Specification Section 3.6.3, ``Containment Isolation 
Valves,'' and its associated Bases, by removing the allowance to open 
the upper containment purge isolation valves in the applicable modes 
consistent with the lower containment purge isolation 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 this LAR [License Amendment Request] involve a 
significant increase in the probability or consequences of an 
accident previously evaluated?
    No. The Containment Purge System is not capable of initiating 
any accident by itself so there will be no increase in the 
probability of an accident. Since these containment isolation valves 
will be maintained in the sealed closed position, there can be no 
increase in the consequences of an accident. The design and 
operation of the Containment Purge System is not being modified by 
this LAR. Therefore, approval and implementation of this LAR will 
have no effect on accident probabilities or consequences.
    2. Does this LAR create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. This LAR does not involve any physical changes to the 
Containment Purge System so no new or different accident causal 
mechanisms will be generated. Also, no changes are being made to the 
way in which the Containment Purge System is operated. Some 
surveillance tests will no longer be performed but these tests are 
no longer necessary since the affected components remain in their 
safe, design basis position. Consequently, plant accident analyses 
will not be affected by this LAR.
    3. Does this LAR involve a significant reduction in a margin of 
safety?
    No. Margin of safety is related to the confidence in the ability 
of the fission product barriers to perform their design functions 
during and following accident conditions. These barriers include the 
fuel cladding, the reactor coolant system, and the containment 
system. The performance of these barriers will not be affected by 
the proposed changes. The containment isolation valves in the 
Containment Purge System will continue to perform their design basis 
function after this LAR is implemented.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Power Company LLC, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Branch Chief: Evangelos C. Marinos.

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 amendment request: November 1, 2006.
    Description of amendment request: The proposed amendment would 
modify technical specification (TS) requirements for inoperable 
snubbers by adding Limiting Condition of Operation (LCO) 3.0.8.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on November 24, 2004 (69 FR 68412), on possible 
amendments to revise the plant-specific TS to allow a

[[Page 70559]]

delay time for entering a supported system TS when the inoperability is 
due solely to an inoperable snubber, if risk is assessed and managed 
consistent with the program that is in place for complying with the 
requirements of 10 CFR 50.65(a)(4). LCO 3.0.8 was proposed to be added 
to an individual TS providing this allowance, including a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination, using the consolidated line-item improvement process. 
The NRC staff subsequently issued a notice of availability of the 
models for referencing in license amendment applications in the Federal 
Register on May 4, 2005 (70 FR 23252). The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
November 1, 2006.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated
    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an inoperable snubber if risk is assessed and managed. The 
postulated seismic event requiring snubbers is a low-probability 
occurrence and the overall TS system safety function would still be 
available for the vast majority of anticipated challenges. Therefore, 
the probability of an accident previously evaluated is not 
significantly increased, if at all. The consequences of an accident 
while relying on allowance provided by proposed LCO 3.0.8 are no 
different than the consequences of an accident while relying on the TS 
required actions in effect without the allowance provided by proposed 
LCO 3.0.8. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The addition 
of a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Therefore, this change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated
    The proposed change does not involve a physical alteration of the 
plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is assessed 
and managed, will not introduce new failure modes or effects and will 
not, in the absence of other unrelated failures, lead to an accident 
whose consequences exceed the consequences of accidents previously 
evaluated. The addition of a requirement to assess and manage the risk 
introduced by this change will further minimize possible concerns. 
Thus, this change does not create the possibility of a new or different 
kind of accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic event 
requiring snubbers is a low-probability occurrence and the overall TS 
system safety function would still be available for the vast majority 
of anticipated challenges. The risk impact of the proposed TS changes 
was assessed following the three-tiered approach recommended in RG 
[Regulatory Guide] 1.177. A bounding risk assessment was performed to 
justify the proposed TS changes. This application of LCO 3.0.8 is 
predicated upon the licensee's performance of a risk assessment and the 
management of plant risk. The net change to the margin of safety is 
insignificant. Therefore, this change does not involve a significant 
reduction in a margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a no significant hazards consideration.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: David Terao.

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 amendment request: November 1, 2006.
    Description of amendment request: The proposed change will revise 
the Grand Gulf Nuclear Station (GGNS), Unit 1, Technical Specification 
(TS) Surveillance Requirement 3.3.1.1.7 for the surveillance interval 
of the local power range monitor (LPRM) calibrations from 1,000 
megawatt-days/ton (MWD/T) (approximately every 36 days) to 2,000 MWD/T 
(approximately every 72 days).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The extended surveillance interval continues to ensure that the 
LPRM detectors are adequately calibrated to provide an accurate 
indication of core power distribution and local power changes. The 
change will not alter the basic operation of any process variables, 
structures, systems, or components as described in the safety 
analyses, and no new equipment is introduced. Hence, the probability 
of accidents previously evaluated is unchanged.
    The thermal limits established by safety analysis calculations 
ensure that reactor core operation is maintained within fuel design 
limits during any Anticipated Operational Occurrence (AOO). The 
analytical methods and assumptions used in evaluating these 
transients and establishing the thermal limits assure adequate 
margins to fuel design limits are maintained. These methods account 
for various calculation uncertainties including radial bundle power 
uncertainty which can be affected by LPRM accuracy. Extending the 
LPRM calibration interval does not impact the existing uncertainties 
assumed in the GGNS safety analyses. Plant specific evaluation of 
LPRM sensitivity to exposure has determined that the extended 
calibration interval does not affect the radial bundle power 
distribution uncertainty value currently used in the safety 
analysis. Hence the safety analysis calculations and the associated 
thermal limits are not affected by the extended LPRM calibration 
interval and the consequences of an accident previously evaluated 
are not changed.
    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

[[Page 70560]]

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 GGNS design and the GGNS 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 thermal limits established by safety analysis calculations 
ensure that reactor core operation is maintained within fuel design 
limits during any Anticipated Operational Occurrence (AOO). The 
analytical methods and assumptions used in evaluating these 
transients and establishing the thermal limits assure adequate 
margins to fuel design limits are maintained. These methods account 
for various calculation uncertainties including radial bundle power 
uncertainty which can be affected by LPRM accuracy. Extending the 
LPRM calibration interval does not impact the existing uncertainties 
assumed in the GGNS safety analyses. Plant specific evaluation of 
LPRM sensitivity to exposure has determined that the extended 
calibration interval does not affect the radial bundle power 
distribution uncertainty value currently used in the safety 
analyses. The thermal limits determined by NRC approved analytical 
methods will continue to provide adequate margin to fuel design 
limits.
    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: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213
    NRC Branch Chief: David Terao

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: March 1, 2006
    Description of amendment request: The proposed amendment would 
modify the Special Operations Limiting Condition for Operation (LCO) 
3.10.1, ``System Leakage and Hydrostatic Testing Operation,'' allowance 
for operation with the average reactor coolant temperature greater than 
212 [deg]F while considering operational conditions to be in MODE 4, to 
include operations where temperature exceeds 212 [deg]F as a 
consequence of maintaining reactor pressure for a system leakage or 
hydrostatic test, or as a consequence of maintaining reactor pressure 
for control rod scram time testing initiated in conjunction with a 
system leakage or hydrostatic test. This change would allow more 
efficient testing during a refueling outage.
    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.
    Technical Specifications currently allow for operation at >212 
[deg]F while imposing MODE 4 requirements in addition to the 
secondary containment requirements required to be met. Extending the 
activities that can apply this allowance will not adversely impact 
the probability or consequences of an 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Technical Specifications currently allow for operation at >212 
[deg]F while imposing MODE 4 requirements in addition to the 
secondary containment requirements required to be met. No new 
operational conditions beyond those currently allowed by LCO 3.10.1 
are introduced. The extended allowances would result from operations 
that commence at reduced temperatures, but approach the normal MODE 
4 limit of 212 [deg]F prior to completion of the inspections or 
testing. The changes do 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 changes do not impose any new or 
different requirements or eliminate any existing requirements. The 
changes do not alter assumptions made in the safety analysis. The 
proposed changes are 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 previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Technical Specifications currently allow for operation at >212 
[deg]F while imposing MODE 4 requirements in addition to the 
secondary containment requirements required to be met. Extending the 
activities that can apply this allowance will not adversely impact 
any margin of safety. Allowing completion of inspections and testing 
and supporting completion of scram time testing initiated in 
conjunction with a system leakage or hydrostatic test prior to power 
operation, results in enhanced safe operations by eliminating 
unnecessary maneuvers to control reactor temperature and pressure.
    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: Mr. R. E. Helfrich, Florida Power & Light 
Company, P. O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: L. Raghavan.

GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear 
Station, Unit 2, Dauphin County, Pennsylvania

    Date of amendment request: October 10, 2006.
    Description of amendment requests: The amendment application 
proposes a revision to the Technical Specification Surveillance 
Requirement 4.1.1.3 to extend the containment airlock surveillance 
frequency from once per year to once every five years.
    Basis for proposed no significant hazards consideration 
determination:
    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated? 
No.
    The proposed change does not introduce any new degradation or 
failure mechanism. The failure mechanism in this case would be a 
failure of an airlock door to open, thus no new release path to the 
environment is created. As no release path is created, there is not 
the possibility of a significant increase in the probability or 
consequences of an accident.
    (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 does not introduce any new degradation or 
failure mechanism.
    The failure mechanism in this case would be a failure of an 
airlock door to open, thus no new release path to the environment is 
created. As no release path is created, there is not the possibility 
of a new or different kind of accident from any accident previously 
evaluated being created.

[[Page 70561]]

    (3) Does the proposed change involve a significant reduction in 
a margin of safety? No.
    The proposed change does not introduce any new degradation or 
failure mechanism. The failure mechanism in this case would be a 
failure of an airlock door to open, thus no new release path to the 
environment is created. Thus, 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.
    NRC Branch Chief: Claudia Craig.

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

    Date of amendment request: October 17, 2006.
    Description of amendment request: The proposed amendment would 
revise the Cooper Nuclear Station (CNS) Technical Specifications (TS) 
4.3.1.1.c by adding a new nominal center-to-center distance between 
fuel assemblies for the new storage racks, and would revise TS 4.3.3 by 
increasing the capacity of the spent fuel storage pool from 2366 
assemblies to 2651 assemblies.
    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 probability of a seismic event, and the resulting loss of 
spent fuel pool cooling flow, is not influenced by the proposed 
changes. In addition, the probability of an accidental fuel assembly 
drop or misloading is primarily influenced by the methods used to 
lift and move these loads. The method of handling fuel will not be 
changed since the same equipment and procedures will be used. 
Shipping cask movements in the SFP [spent fuel pool] will not be 
performed during installation of the new racks. There is no change 
to the methods or equipment to be used in moving fuel casks. 
Expanding the spent fuel storage capacity does not have a 
significant impact on the frequency of occurrence for any accident 
previously evaluated.
    Therefore, this change will not significantly increase the 
probability of occurrence of any accident previously analyzed.
    The consequences of a dropped spent fuel assembly in the SFP 
have been re-evaluated for the proposed change by analyzing a 
potential impact onto the new racks. The results show that the 
postulated accident of a fuel assembly striking the new storage 
racks will not distort the racks sufficiently to impair their 
functionality. The minimum subcriticality margin required by the 
current TS (i.e., neutron multiplication factor [keff] less than or 
equal to 0.95) will be maintained. The structural damage to the 
Reactor Building, pool liner, and fuel assembly resulting from a 
dropped fuel assembly striking the pool floor or another assembly 
located in the racks is primarily dependent on the mass of the 
falling object and the drop height. Since these two parameters are 
not changed by the proposed modification, the postulated structural 
damage to these items remains unchanged. The radiological dose at 
the exclusion area boundary will not be increased since no changes 
are being made to in-core hold time or burnup as a result of the 
proposed amendment.
    Loss of SFP cooling was evaluated. The concern with this event 
is a reduction of spent fuel pool water inventory as a result of 
boiling in the fuel pool, with the inventory reduction resulting in 
an unacceptable increase in dose rates. Loss of spent fuel pool 
cooling at CNS is mitigated procedurally by supplying makeup water 
to the pool prior to the time that the temperature of the pool 
reaches boiling. The thermal-hydraulic analysis for the proposed 
license amendment determined, for a complete loss of forced cooling 
and a full core discharge, that the minimum time to boil is 4.19 
hours. This has been determined to be sufficient time for the 
operators to provide alternate means of makeup water to the SFP 
before the water begins to boil. Based on this the consequences of a 
loss of SFP cooling are not significantly increased.
    The consequences of a design basis seismic event are evaluated 
on the basis of subsequent fuel damage or compromise of the fuel 
storage or building configurations leading to radiological or 
criticality concerns. The new racks have been analyzed in their new 
configuration and were found to be safe during seismic motion. Fuel 
has been determined to remain intact and the storage racks maintain 
the fuel and fixed poison configurations subsequent to a seismic 
event. The structural capability of the pool and liner will not be 
exceeded under the anticipated combinations of dead weight, thermal, 
and seismic loads. The Reactor Building structure will remain intact 
during a seismic event and will continue to adequately support and 
protect the fuel racks, storage array, and pool moderator/coolant. 
Therefore, the consequences of a design basis seismic event are not 
increased.
    The consequence of a fuel misloading accident has been analyzed 
for the worst possible storage configuration subsequent to the 
proposed modification. It has been determined that the consequences 
remain acceptable with respect to the same criteria used previously.
    Therefore, the proposed change does not result in a significant 
increase in the consequences of a previously evaluated accident.
    In summary, the proposed change does 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.
    A drop of a fuel assembly onto fuel assemblies stored in the SFP 
has been previously analyzed for CNS and is not a new or different 
kind of accident. The only event which would represent a new or 
different kind of accident is an accidental drop of a rack during 
movement in the pool.
    Dropping a rack onto stored spent fuel or the pool floor liner, 
commonly referred to as a ``heavy load drop,'' is not postulated due 
to the defense-in-depth approach to be taken. A lifting rig designed 
to meet the requirements of NUREG 0612 [Nuclear Regulatory 
Commission technical report designation 0612] and ANSI N 14.6 
[American National Standards Institute N 14.6] will be used to 
install the new racks. Dropping a new rack onto fuel is precluded by 
not allowing the new racks being placed into the SFP to travel over 
racks containing fuel assemblies. A rack drop to the pool liner is 
not postulated since the lifting components either provide 
redundancy in supporting the racks or are designed with safety 
margins greater than a factor of ten. Movements of heavy loads over 
the pool will comply with the applicable administrative controls and 
guidelines (i.e. plant procedures, NUREG 0612, etc.). Therefore, the 
rack drop does not represent a new or different kind of accident.
    The proposed change does not alter the operation of the plant or 
equipment credited for the mitigation of the design basis accidents. 
The proposed change does not affect the important parameters 
required to ensure safe fuel storage.
    In summary, 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The function of the spent fuel pool is to store the fuel 
assemblies in a subcritical and coolable configuration under 
postulated environmental and abnormal loadings, such as an 
earthquake or fuel assembly drop. The new rack design meets the 
applicable requirements for safe storage and is functionally 
compatible with the SFP.
    The Holtec Licensing Report was prepared using the guidance of 
the applicable provisions of the NRC Guidance entitled, ``OT 
Position for Review and Acceptance of Spent Fuel Storage and 
Handling Applications.'' The rack materials used are compatible with 
the spent fuel assemblies and the SFP environment. The design of the 
new racks preserves the proper margin of safety during abnormal 
loads, e.g., loads from a seismic event, a dropped assembly, and 
tensile loads from a stuck fuel assembly. It has been shown that 
such loads will not invalidate the mechanical design and material 
selection to safely store fuel in a coolable and subcritical 
configuration.
    The methodology used in the criticality analysis of the expanded 
spent fuel pool

[[Page 70562]]

complies with the appropriate NRC guidelines and the ANSI standards 
(Draft GDC 66 [General Design Criterion 66], NUREG 0800, Section 
9.1.2, the OT Position for Review and Acceptance of Spent Fuel 
Storage and Handling Applications, Reg. Guide 1.13, and ANSI ANS 
8.17 [American Nuclear Society 8.17]).
    The subcriticality margin (keff) for spent fuel 
stored in the SFP is required to be less than or equal to 0.95 under 
normal storage, fuel handling, and accident conditions, including 
uncertainties. This margin will be maintained with the proposed 
increased capacity.
    The thermal-hydraulic and cooling evaluation of the pool 
determined that the pool can be maintained below the specified 
thermal limits under the conditions of the maximum heat load. The 
pool temperature will not exceed the design temperature of 150[deg]F 
during operation of the cooling systems. The maximum local water 
temperature in the hot channel will remain below the boiling point. 
The maximum cladding temperature after a loss of cooling remains 
less than the current licensing basis value of 350 [deg]F with bulk 
boiling in the pool. The stored fuel will not undergo any 
significant heat up with blockage of a dropped fuel assembly lying 
horizontally on top of the racks. The thermal limits specified for 
the evaluations performed to support the proposed change are the 
same as those which were used in the previous evaluations.
    The time to boiling, in the event of a complete loss of SFP 
cooling with a full core discharge, has been reduced from 5 hours to 
4.19 hours. However, this has been determined to be sufficient time 
for providing makeup to the SFP.
    Based on the above it is concluded that 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: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: October 19, 2006.
    Description of amendment request: The proposed amendment would 
revise the surveillance requirements in Technical Specification (TS) 
4.1.1, ``Control Rod System,'' to modify the conditions under which 
scram time testing (STT) of control rods is required, and add a 
requirement to perform STT on a defined portion of control rods, at a 
specified frequency, during the operating cycle. The requirement to 
test ``eight selected [control] rods'' after a reactor scram or other 
outage would be replaced by a requirement to periodically test at least 
20 control rods, on a rotating basis, every 180 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change adds new surveillance requirements (SR) to 
the MCPR [minimum critical power ratio] Technical Specification (TS) 
which requires determination of the MCPR operating limit following 
the completion of scram time testing (STT) of the control rods. Use 
of the scram speed in determining the MCPR operating limit (i.e., 
Option B) is an alternative to the current method for determining 
the operating limit (i.e., Option A). The probability of an accident 
previously evaluated is unrelated to the MCPR operating limit that 
is provided to ensure no fuel damage results during anticipated 
operational occurrences. This is an operational limit to ensure 
conditions following an assumed accident do not result in fuel 
failure and therefore do not contribute to the occurrence of an 
accident.
    The proposed change revises allowable conditions for the STT of 
non-maintenance affected control rods and eliminates the requirement 
to test ``eight [selected] rods'' after a reactor scram or other 
outage. The requirement to test ``eight selected rods'' is replaced 
by a new SR to perform periodic STT. No active or passive failure 
mechanisms that could lead to an accident are affected by this 
proposed change and the STT acceptance criteria are not being 
revised. Therefore, the proposed change in STT requirements does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    The proposed change ensures that the appropriate MCPR operating 
limit is in place. By implementing the correct MCPR operating limit, 
the MCPR SL [safety limit] will continue to be ensured. Ensuring the 
MCPR SL is not exceeded will result in prevention of fuel failure. 
Therefore, since there is no increase in the potential for fuel 
failure, there is no increase in the consequences of any accidents 
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 adds a new SR to the MCPR TS which requires 
determination of the MCPR operating limit following the completion 
of the [STT] of the control rods. The proposed change revises 
allowable conditions for the STT of non-maintenance affected control 
rods and eliminates the requirement to test ``eight [selected] 
rods'' after a reactor scram or other outage. The requirement to 
test ``eight selected rods'' is replaced by a new SR to perform 
periodic STT. The proposed change does not involve the use or 
installation of new equipment. Installed equipment is not operated 
in a new or different manner. No new or different system 
interactions are created, and no new processes are introduced. No 
new failures have been created by the addition of the proposed SR 
and the use of the alternate method for determining the MCPR 
operating limit. 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.
    Use of Option B for determining the MCPR operating limit will 
result in a reduced operating limit in comparison to the use of 
Option A. However, a reduction in the operating limit margin does 
not result in a reduction in the safety margin. The MCPR SL remains 
the same regardless of the method used for determining the operating 
limit. The proposed change revises allowable conditions for the STT 
of non-maintenance affected control rods and eliminates the 
requirement to test ``eight [selected] rods'' after a reactor scram 
or other outage. The requirement to test ``eight selected rods'' is 
replaced by a new SR to perform periodic STT. No active or passive 
failure mechanisms that could adversely impact the consequences of 
an accident are affected by this proposed change. All analyzed 
transient results remain within the design values for structures, 
systems and components. 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 J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Richard J. Laufer.

Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: October 23, 2006.
    Description of amendment request: The proposed changes to the 
technical specifications (TSs) would eliminate the use of the defined 
term CORE ALTERATIONS in the TSs.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 70563]]

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. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change eliminates the use of the defined term CORE 
ALTERATIONS from the Technical Specifications. CORE ALTERATIONS are 
not an initiator of any accident previously evaluated except a fuel 
handling accident. The revised Technical Specifications that protect 
the initial conditions of a fuel handling accident also require the 
suspension of movement of irradiated fuel assemblies, which protects 
the initial condition of a fuel handling accident.
    Therefore, suspension of CORE ALTERATIONS do not affect the 
initiators of the accidents previously evaluated and suspension of 
CORE ALTERATIONS does not affect the mitigation of the accidents 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical modification of the 
plant (i.e., no new or different type of equipment will be 
installed) or a significant change in the methods governing normal 
plant operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    Only two accidents are postulated to occur during plant 
conditions where CORE ALTERATIONS may be made: A fuel handling 
accident and a boron dilution accident. Suspending movement of 
irradiated fuel assemblies prevents a fuel handling accident. Also, 
requiring the suspension of CORE ALTERATIONS is redundant to 
suspending movement of irradiated fuel assemblies and does not 
increase the margin of safety. CORE ALTERATIONS have no effect on a 
boron dilution accident. Core components are not involved in the 
initiation or mitigation of a boron dilution accident. Therefore, 
CORE ALTERATIONS have no effect on the margin of safety related to a 
boron dilution accident.
    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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Acting Branch Chief: L. Raghavan.

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 amendment request: August 4, 2006.
    Description of amendment request: The amendments would allow the 
use of blind flanges for containment isolation in the containment purge 
system supply and exhaust lines, and make corresponding changes to the 
Technical Specifications (TSs). The amendments would also consolidate 
the containment isolation requirements by moving the requirements of TS 
3/4 6.1.7, ``Containment Ventilation System,'' to TS 3/4 6.3.1 (TS 3/4 
6.3 for Unit No. 2), ``Containment Isolation 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.
    The proposed change to the Containment purge supply and exhaust 
penetrations presents no change in the probability or the 
consequence of an accident, since the penetrations continue to 
conform to the TS requirements for containment integrity, and will 
be appropriately tested as required by 10 CFR 50 Appendix J. The 
blind flanges are passive devices not susceptible to an active 
failure or malfunction that could result in a loss of isolation or 
leakage that exceeds limits assumed in the safety analysis. The 
blind flanges are leak rate tested in accordance with the 
containment leakage rate testing program. Containment integrity is 
not lessened by this change.
    The change to the Containment Purge System does not affect the 
design basis limit for any fission product barrier.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to the Containment purge supply and exhaust 
penetrations does not change the function of the system and does not 
alter containment integrity. The penetrations continue to conform to 
the TS requirements for containment integrity and will be 
appropriately tested as required by 10 CFR 50 Appendix J. No new 
accident scenarios, failure mechanisms, or limiting single failures 
are introduced as a result of the proposed changes.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed change will not alter any assumptions, initial 
conditions or results specified in any accident analysis. The 
Containment purge supply and exhaust penetrations will continue to 
conform to the TS requirements for containment integrity, and will 
be appropriately tested as required by 10 CFR 50 Appendix J. The 
blind flanges are passive devices not susceptible to an active 
failure or malfunction that could result in a loss of isolation or 
leakage that exceeds limits assumed in the safety analysis. The 
blind flanges are leak rate tested in accordance with the 
containment leakage rate testing program. Containment integrity is 
not lessened by this change. Therefore, there is no 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: Harold K. Chernoff.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: October 3, 2006.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) and licensing basis to support 
the resolution of the Nuclear Regulatory Commission's (NRC's) Generic 
Safety Issue (GSI) 191, assessment of debris accumulation on 
containment sump performance and its impact on emergency recirculation 
during an accident, and NRC 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:


[[Page 70564]]


    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed changes include a physical alteration to the RS 
system to start the inside and outside [Recirculation Spray] RS 
pumps on [Refueling Water Storage Tank] RWST Level Low coincident 
with High High containment pressure. The RS system is used for 
accident mitigation only, and changes in the operation of the RS 
system cannot have an impact on the probability of an accident. The 
other changes do not affect equipment and are not accident 
initiators. The RWST Level Low instrumentation will comply with all 
applicable regulatory requirements and design criteria (e.g., train 
separation, redundancy, and single failure). Therefore, the design 
functions performed by the RS system are not changed.
    Delaying the start of the RS pumps creates more challenging 
long-term containment pressure and temperature profiles. The 
environmental qualification of safety-related equipment inside 
containment was confirmed to be acceptable, and accident mitigation 
systems will continue to operate within design temperatures and 
pressures. Delaying the RS pump start reduces the emergency diesel 
generator loading early during a design basis accident, and 
staggering the RS pump start avoids overloading on each emergency 
bus. The reduction in iodine removal efficiency during the delay 
period is offset by changes to other assumptions in the [loss-of-
coolant accident] LOCA dose analysis. The predicted offsite doses 
and control room doses following a design basis LOCA remain within 
regulatory limits.
    The [Updated Final Safety Analysis Report] UFSAR safety analysis 
acceptance criteria continue to be met for the proposed changes to 
the RS pump start method, the proposed TS containment air partial 
pressure limits, the proposed TS containment temperature limit, the 
implementation of the GOTHIC containment analysis methodology, the 
proposed change to the [safety injection] SI [recirculation mode 
transfer] RMT allowable values, and the changes to the LOCA dose 
consequences analyses. Based on this discussion, the proposed 
amendments do not increase the probability or consequence of an 
accident previously evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
identified?
    Response: No.
    The proposed change alters the RS pump circuitry by initiating 
the start sequence with a new RWST Level Low signal instead of a 
timer after the High High containment pressure setpoint is reached. 
The timers for the inside RS pumps will be used to sequence pump 
starts and preclude diesel generator overloading. The RS pump 
function is not changed. The RWST Level Low instrumentation will be 
included as part of the Engineered Safety Features Actuation System 
(ESFAS) instrumentation in the North Anna TS and will be subject to 
the ESFAS surveillance requirements. The design of the RWST Level 
Low instrumentation complies with all applicable regulatory 
requirements and design criteria. The failure modes have been 
analyzed to ensure that the RWST Level Low circuitry can withstand a 
single active failure without affecting the RS system design 
functions. The RS system is an accident mitigation system only, so 
no new accident initiators are created.
    The remaining changes to the containment analysis methodology, 
the containment air partial pressures, the maximum containment 
temperature operating limit, the TS allowable values for SI RMT, and 
the LOCA [alternate source term] AST analysis basis do not impact 
plant equipment design or function. Together, the changes assure 
that there is adequate margin available to meet the safety analysis 
criteria and that dose consequences are within regulatory limits. 
The proposed changes do not introduce failure modes, accident 
initiators, or malfunctions that would cause a new or different kind 
of accident. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously identified.
    3. Does the proposed license amendment involve a significant 
reduction in a margin of safety?
    Response: No.
    The changes to the actuation of the RS pumps and the increased 
containment air partial pressure have created an adverse effect on 
the containment response analyses and the LOCA dose analysis. 
Analyses have been performed that show the containment design basis 
limits are satisfied and the post-LOCA offsite and control room 
doses meet the required criteria for the proposed changes to the 
containment analysis methodology, the RS pump start method, the TS 
containment air partial pressure limits, the TS containment 
temperature maximum limit, the TS allowable values for SI RMT, and 
the LOCA AST bases. Therefore, the proposed amendment 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 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.

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: November 16, 2006.
    Description of amendment request: The proposed amendments would add 
a reference in Technical Specification (TS) 6.2.C, ``Core Operating 
Limits Report (COLR),'' to permit the use of the Westinghouse Best-
Estimate Large Break Loss of Coolant Accident (BE-LBLOCA) analysis 
methodology using the Automated Statistical Treatment of Uncertainty 
Method (ASTRUM) for the analysis of LBLOCA.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The probability of occurrence or the consequences of an 
accident previously evaluated are not significantly increased.No 
physical plant changes are being made as a result of using the 
Westinghouse Best Estimate Large Break LOCA (BE-LBLOCA) analysis 
methodology. The proposed TS change simply involves updating the 
references in TS 6.2.C, Core Operating Limits Report (COLR), to 
reference the Westinghouse BE-LBLOCA analysis methodology. The 
consequences of a LOCA are not being increased, since the analysis 
has shown that the Emergency Core Cooling System (ECCS) is designed 
such that its calculated cooling performance conforms to the 
criteria contained in 10 CFR 50.46, ``Acceptance criteria for 
emergency core cooling systems for light-water nuclear power 
reactors.'' No other accident consequence is potentially affected by 
this change.
    All systems will continue to be operated in accordance with 
current design requirements under the new analysis, therefore no new 
components or system interactions have been identified that could 
lead to an increase in the probability of any accident previously 
evaluated in the Updated Final Safety Analysis Report (UFSAR). No 
changes were required to the Reactor Protection System (RPS) or 
Engineering Safety Features (ESF) setpoints because of the new 
analysis methodology.
    An analysis of the LBLOCA accident for Surry Units 1 and 2 has 
been performed with the Westinghouse BE-LBLOCA analysis methodology 
using ASTRUM. The analysis was performed in compliance with all the 
NRC conditions and limitations as identified in WCAP-16009-P-A. 
Based on the analysis results, it is concluded that the Surry Units 
1 and 2 continue to maintain a margin of safety to the limits 
prescribed by 10 CFR 50.46.
    There are no changes to assumptions of the radiological dose 
calculations. Hence, there is no increase in the predicted 
radiological consequences of accidents postulated in the UFSAR.
    Therefore, neither the probability of occurrence nor the 
consequences of an accident previously evaluated is significantly 
increased.
    2. The possibility for a new or different type of accident from 
any accident previously evaluated is not created.
    The use of the Westinghouse BE-LBLOCA analysis methodology with 
ASTRUM does

[[Page 70565]]

not impact any of the applicable design criteria and all pertinent 
licensing basis criteria will continue to be met. Demonstrated 
adherence to the criteria in 10 CFR 50.46 precludes new challenges 
to components and systems that could introduce a new type of 
accident. Safety analysis evaluations have demonstrated that the use 
of Westinghouse BE-LBLOCA analysis methodology with ASTRUM is 
acceptable. All design and performance criteria will continue to be 
met and no new single failure mechanisms will be created. The use of 
the Westinghouse BE-LBLOCA analysis methodology with ASTRUM does not 
involve any alteration to plant equipment or procedures that would 
introduce any new or unique operational modes or accident 
precursors. Furthermore, no changes have been made to any RPS or ESF 
actuation setpoints. Based on this review, it is concluded that no 
new accident scenarios, failure mechanisms, or limiting single 
failures are introduced as a result of the proposed changes.
    Therefore, the possibility for a new or different kind of 
accident from any accident previously evaluated is not created.
    3. The margin of safety is not significantly reduced.
    It has been shown that the analytical technique used in the 
Westinghouse BE-LBLOCA analysis methodology using ASTRUM 
realistically describes the expected behavior of the reactor system 
during a postulated LOCA. Uncertainties have been accounted for as 
required by 10 CFR 50.46. A sufficient number of LOCAs with 
different break sizes, different locations, and other variations in 
properties have been considered to provide assurance that the most 
severe postulated LOCAs have been evaluated. The analysis has 
demonstrated that all acceptance criteria contained in 10 CFR 50.46 
continue to be satisfied.
    Therefore, it is concluded that this change does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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].

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of application for amendment: December 1, 2005.
    Brief description of amendment: The amendment revised Technical 
Specification 3.6.4.1, ``Secondary Containment.'' Specifically, the 
amendment revised Surveillance Requirement (SR) 3.6.4.1.4 and SR 
3.6.4.1.5 to clarify their intent with respect to secondary containment 
boundary integrity.
    Date of issuance: November 17, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 175.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specification Surveillance Requirements and License.
    Date of initial notice in Federal Register: March 28, 2006 (71 FR 
15481).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 17, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 26, 2006, as 
supplemented by the letter dated November 3, 2006.
    Brief description of amendments: The amendments revise TS 3.7.2, 
``Main Steam Isolation Valves (MSIVs),'' to include specific 
requirements for the MSIV actuator trains.
    Date of issuance: November 17, 2006.
    Effective date: Effective as of the date of issuance to be 
implemented within 10 days from the date of issuance.
    Amendment Nos.: Unit 1--163, Unit 2--163, Unit 3--163.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 5, 2006 (71 FR 
58879). The supplemental letter dated November 3, 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 November 17, 2006.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of application for amendments: February 27, 2006.
    Brief description of amendments: The amendments revise Technical

[[Page 70566]]

Specification 4.2.1, ``Fuel Assemblies,'' to permit up to four lead 
fuel assemblies (LFAs) with advanced cladding material to be re-
inserted into either the Unit 1 or Unit 2 core for the next operating 
cycle, which is Cycle 19 for Unit 1 and Cycle 17 for Unit 2. Two of 
these LFAs were manufactured by Westinghouse Electric Company and 
contain a limited number of fuel rods with advanced zirconium-based 
alloys. The other two LFAs were manufactured by Framatome ANP, Inc. 
with fuel rod cladding material classified as M5TM alloy. 
These LFAs were originally inserted into the Unit 2 core in April 2003 
(Operating Cycles 15 and 16) and are scheduled to be discharged during 
the 2007 refueling outage.
    Date of issuance: November 16, 2006.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 280 and 257.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the License and Technical Specifications.
    Date of initial notice in Federal Register: March 28, 2006 (71 FR 
15482).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated November 16, 2006.
    No significant hazards consideration comments received: No

Exelon Generation Company, LLC, Docket No. 50-249, Dresden Nuclear 
Power Station, Unit 3, Grundy County, Illinois

    Date of application for amendment: July 21, 2006, as supplemented 
by letter dated October 19, 2006.
    Brief description of amendment: The amendment revised the values of 
the safety limit minimum critical power ratio in Technical 
Specification Section 2.1.1, ``Reactor Core SLs [Safety Limits].''
    Date of issuance: November 7, 2006.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup for cycle 20.
    Amendment Nos.: 213.
    Renewed Facility Operating License Nos. DPR-19 and DPR-25: The 
amendment revised the Technical Specifications and License.
    Date of initial notice in Federal Register: August 29, 2006 (71 FR 
51228). The October 19, 2006 supplement provided additional clarifying 
information that did not expand the scope of the application as 
originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination published 
in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 7, 2006.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 21, 2006, as supplemented on 
September 6 and October 10, 2006.
    Brief description of amendment: The amendment changed the Technical 
Specifications (TSs) to: (1) Revise TS Section 2.3(4) to change the 
reactor containment building sump buffering agent from trisodium 
phosphate to sodium tetraborate and change the TS section title to 
``Containment Sump Buffering Agent Specification and Volume 
Requirement,'' (2) revise TS 3.6(2)d to require a volume of sodium 
tetraborate that is within an area of acceptable operation, as shown in 
TS Figure 2-3, and (3) an administrative correction to TS 3.6(2)d(i). 
The amendment allows OPPD to replace the trisodium phosphate in the 
containment with sodium tetraborate. Changes were also made to the 
corresponding TS Bases. The TS changes are approved for Cycle 24 only, 
ending in the spring 2008 refueling outage.
    Date of issuance: November 13, 2006.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 247.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 30, 2006 (71 FR 
51646). The September 6 and October 10, 2006, supplemental letters 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated November 13, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 3, 2005, as 
supplemented by letters dated May 1, August 15, and October 5, 2006.
    Brief description of amendments: The amendments revised Technical 
Specification Section 5.5.2.11 to modify the definitions of steam 
generator tube ``Repair Limit'' and ``Tube Inspection.'' The changes 
define the extent of the required tube inspections and repair criteria 
within the tubesheet regions.
    Date of issuance: November 9, 2006.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 2--206; Unit 3--198.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 6, 2005 (70 FR 
72676). The May 1, August 15, and October 5, 2006, supplemental letters 
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 no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 9, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 14, 2006.
    Brief description of amendments: The amendments deleted duplicative 
notifications, reporting, and restart requirements if a safety limit 
was violated; replaced plant-specific position titles with generic 
position titles; and additional administrative changes.
    Date of issuance: November 15, 2006.
    Effective date: As of date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 2--207; Unit 3--199.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.

[[Page 70567]]

    Date of initial notice in Federal Register: September 12, 2006 (71 
FR 53720).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 15, 2006.
    No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: May 11, 2006.
    Brief description of amendment: The amendment revised Surveillance 
Requirements (SRs) 3.7.2.1, 3.7.3.1, and 3.7.3.3 on verifying the 
closure time of the main steam isolation valves (MSIVs), main feedwater 
regulating valves (MFRVs), main feedwater regulating valve bypass 
valves (MFRVBVs), and main feedwater isolation valve (MFIVs) in the 
Technical Specifications (TS). These valves are the Main Steam and Main 
Feedwater System isolation valves. The revisions replace (1) the 
specified maximum acceptable valve closure time for the MSIVs, MFRVs, 
and MFRVBVs, and (2) TS Figure 3.7.3-1, which shows acceptable valve 
closure times for the MFIVs, by the reference to the valve closure time 
is verified to be ``within limits.'' The maximum acceptable valve 
closure times for the MFRVs and MFRVBVs, and TS Figure 3.7.3-1 are now 
located in the TS Bases. The maximum acceptable valve closure time for 
the MSIV is already in the TS Bases.
    Date of issuance: November 15, 2006.
    Effective date: Effective as of its date of issuance, and shall be 
implemented within 90 days of the date of issuance.
    Amendment No.: 176.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 20, 2006 (71 FR 
35461).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 15, 2006.
    No significant hazards consideration comments received: No.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, 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 System's (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].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. 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

[[Page 70568]]

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, and electronically on the Internet at the 
NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If 
there are problems in accessing the document, contact the PDR Reference 
staff at 1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. 
If a request for a hearing or petition for leave to intervene is filed 
by the above date, 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 identify 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 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
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.\1\ 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 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.
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    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly avaialble because they 
are asserted to contain safeguards or proprietary information, 
petitioners desiring access to this information should contact the 
applicant or applicant's counsel and discuss the need for a 
protective order.
---------------------------------------------------------------------------

    Each contention shall be given a separate numeric or alpha 
designation within one of the following groups:
    1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the 
applications.
    2. Environmental--primarily concerns/issues relating to matters 
discussed or referenced in the environmental analysis for the 
applications.
    3. Miscellaneous--does not fall into one of the categories outlined 
above.
    As specified in 10 CFR 2.309, if two or more petitioners/requestors 
seek to co-sponsor a contention, the petitioners/requestors shall 
jointly designate a representative who shall have the authority to act 
for the petitioners/requestors with respect to that contention. If a 
petitioner/requestor seeks to adopt the contention of another 
sponsoring petitioner/requestor, the petitioner/requestor who seeks to 
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that 
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the 
petitioners/requestors with respect to that contention.
    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. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    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 or 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).

Pacific Gas and Electric Company, Docket No. 50-275, Diablo Canyon 
Nuclear Power Plant, Unit No. 1, San Luis Obispo County, California

    Date of application for amendment: October 18, 2006, as 
supplemented on November 2, 2006.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Section 3.8.4, ``DC Sources--Operating,'' Condition 
B to extend the completion time (CT) to restore an inoperable vital 
battery from 2 hours to 4 hours for the current operating Cycle 14, 
provided certain required actions are taken. The extended CT would 
allow sufficient time to correct a degraded condition on the station 
Vital Battery 1-1.
    Date of issuance: November 15, 2006
    Effective date: As of its date of issuance and shall be implemented 
within 7 days of the date of issuance.
    Amendment No.: 190
    Facility Operating License No. DPR-80: The amendment revised the 
Technical Specifications and license.
    Public comments requested as to proposed no significant hazards

[[Page 70569]]

consideration (NSHC): Yes. An individual 14-day Notice of Consideration 
of Issuance of Amendment to Facility Operating License was published on 
October 27, 2006 (71 FR 63040) in the Federal Register. The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided an opportunity to request a hearing by December 26, 2006, but 
indicated that if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment.
    The November 2, 2006, supplemental letter 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 amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a safety evaluation dated November 15, 2006.
    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Branch Chief: David Terao

    Dated at Rockville, Maryland, this 22nd day of November 2006.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. E6-20329 Filed 12-4-06; 8:45 am]
BILLING CODE 7590-01-P