[Federal Register Volume 71, Number 39 (Tuesday, February 28, 2006)]
[Notices]
[Pages 10071-10084]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-1737]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

I. Background

    Pursuant to section 189a(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from February 3, 2006, to February 15, 2006. The 
last biweekly notice was published on February 14, 2006 (71 FR 7804).

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

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

[[Page 10072]]

    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
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 email 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].

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

    Date of amendment request: November 30, 2005.
    Description of amendment request: The proposed amendment would 
revise the frequency of the diesel generator automatic trips bypass 
surveillance requirement (SR) 3.8.1.11 from 18 months to 24 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed change decreases the frequency of SR 
3.8.1.11, verification of the DG [diesel generator] automatic trips 
bypass, from 18 months to 24 months. The DG automatic trips bypass 
circuitry is required for DG operability and reliability during 
emergency operation of the DG. The proposed test frequency will 
continue to assure that the DG will perform as required. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated, 
because the factors that are used to determine the probability and 
consequences of accidents are not being affected.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated. There 
are no new or different accident initiators or sequences being 
created by the proposed Technical Specifications change. The 
required surveillance performed at the proposed frequency will 
continue to provide assurance that the trips bypass function is 
operable and is properly supporting operation of the associated DG. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    No. The proposed change does not involve a significant reduction 
in the margin of

[[Page 10073]]

safety. The proposed change will continue to ensure that the DG 
trips bypass function operates as designed. The functionality and 
operability of emergency power system is not being changed. 
Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: January 4, 2006.
    Description of amendment request: The proposed amendment would 
change the Millstone Power Station, Unit No. 2 Technical Specification 
(TS) 3/4.3.3.8, ``Instrumentation, Accident Monitoring,'' to modify the 
description of the pressurizer power operated relief valves (PORVs) and 
pressurizer safety valves position indicators.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment removes the wording ``Acoustic Monitor,'' 
which provides specific details related to system design, from items 
4 and 6 of TS 3/4.3.3.8, Tables 3.3-11 and 4.3-7. The PORVs and 
Pressurizer Safety Valves position indicators (and the associated 
``Acoustic Monitor'') provide only indications of valve position. 
They do not constitute a design feature that is an initial condition 
for a design basis accident or transient analysis. Furthermore, they 
do not affect the function of the system, equipment in the system or 
actuate to mitigate a design basis accident or transient. Therefore, 
the proposed changes do not increase the probability or consequences 
of an accident previously evaluated.
    Additionally, the TS retains the requirement for the total and 
minimum channels required to be OPERABLE and to verify channel 
OPERABILITY at the designated frequencies. The PORVs and Pressurizer 
Safety Valves are equipped with positive position indication that 
meets the requirements of RG [Regulatory Guide] 1.97.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not impact the capability of existing 
equipment to perform its intended functions. No system setpoints are 
being modified and no changes are being made to the method in which 
plant operations are conducted. No new failure modes that would 
impact accident analyses are introduced by the proposed changes. The 
proposed amendment does not introduce accident initiators or 
malfunctions that would cause a new or different kind of accident. 
Therefore, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment removes the wording ``Acoustic Monitor'' 
from items 4 and 6 of TS 3/4.3.3.8, Table[s] 3.3-11 and 4.3-7. The 
proposed changes do not affect any of the assumptions used in the 
accident analysis, nor does it affect any operability requirements 
for equipment important to plant safety. Therefore, the margin of 
safety is not impacted by the proposed amendment.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Branch Chief: Darrell J. Roberts.

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

    Date of amendment request: December 30, 2005.
    Description of amendment request: The proposed amendment 
establishes a combined leakage rate limit for the sum of the four Main 
Steam line leakage rates that is equal to four times the current 
individual Main Steam Isolation Valve (MSIV) leakage rate limit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not involve a change to structures, 
systems, or components that would affect the probability of an 
accident previously evaluated in the Cooper Nuclear Station (CNS) 
Updated Safety Analysis Report (USAR). The proposed amendment 
results in no change in the radiological consequences of the design 
basis Loss-of-Coolant Accident (LOCA) as currently analyzed for CNS. 
That analysis was calculated for a combined Main Steam Isolation 
Valve (MSIV) leakage for determining acceptance to the regulatory 
limits for the offsite and Control Room radiation doses, as 
contained in 10 CFR 100 [Part 100 of Title 10 of the Code of Federal 
Regulations] and 10 CFR 50[,] Appendix A, General Design Criterion 
(GDC) 19. The aggregate Main Steam line leakage rate limit has no 
adverse effect on the environmental qualification of equipment 
important to safety, as provided for in 10 CFR 50.49.
    Based on the above conclusions, this proposed amendment does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not modify the MSIVs or any other plant 
system or structure associated with this amendment and therefore, 
will not affect their capability to perform their design function. 
The combined total Main Steam line leakage rate is included in the 
current radiological analyses for the assessment of radiation 
exposure following an accident. This License Amendment Request 
revises the allowable leakage rate from a per valve limit to a total 
combined leakage rate limit for all four Main Steam lines but does 
not change the cumulative limit.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously analyzed.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    The leakage rate limit specified for the MSIVs is used to 
quantify the maximum amount of Secondary Containment bypass leakage 
assumed in the LOCA radiological analysis. Results of the analysis 
are evaluated against the dose limits contained in 10 CFR 50[,] 
Appendix A[,] GDC 19 and 10 CFR 100. The margin of safety in this 
context is considered to be the difference between the calculated 
dose exposures and the limits provided by GDC 19 and 10 CFR 100.
    Therefore, since the proposed combined Main Steam line leakage 
rate limit is unchanged from the assumed maximum leakage rate for 
MSIVs, for the purpose of calculating [a] potential radiation dose, 
the margin of safety is not affected because the postulated 
radiation doses remain the same.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 10074]]

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.

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

    Date of amendment request: January 30, 2006.
    Description of amendment request: 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 consistent with the program in place for 
complying with the requirements of 10 CFR 50.65(a)(4). Limiting 
Condition for Operation (LCO) 3.0.8 is added to the TS to provide this 
allowance and define the requirements and limitations for its use.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff 
issued a notice of opportunity for comment in the Federal Register on 
November 24, 2004 (69 FR 68412), on possible amendments concerning 
TSTF-372, 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 following NSHC determination 
in its application dated January 30, 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 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 a 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. The proposed LCO 
3.0.8 defines limitations on the use of the provision and includes a 
requirement for the licensee to assess and manage the risk 
associated with operation with an inoperable snubber. 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: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: November 11, 2005.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.6.5, ``Containment Spray and 
Cooling Systems''; an existing Condition, two Surveillance 
Requirements, and add a new Condition which will allow continued plant 
operation with TS limitations when two Containment Cooling System fan 
coil units (FCUs), one in each train, are inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment proposes to revise the Technical 
Specifications to allow plant operation to continue for a limited 
time period under Technical Specification controls with two fan coil 
units, one fan coil unit from each containment cooling train, 
providing the required cooling function. Analyses demonstrate that 
any two fan coil units, whether they are in the same train or from 
opposite trains, are sufficient to supply the required containment 
cooling following a design basis accident when the plant in the 
proper configuration as required by the proposed Technical 
Specifications.
    The containment cooling system is required for accident 
mitigation and is not an accident initiator, thus revising the 
equipment required to provide the safety function does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    Since the proposed change continues to provide the post-accident 
containment cooling function under Technical Specification controls, 
this change does not involve an increase in the consequences of an 
accident. Thus this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This license amendment proposes to revise the Technical 
Specifications to allow plant operation to continue for a limited 
time period under Technical Specification controls with two fan coil 
units, one fan coil unit from each containment cooling train, 
providing the required cooling function. Analyses demonstrate that 
any two fan coil units, whether they are in the same train or

[[Page 10075]]

from opposite trains, are sufficient to supply the required 
containment cooling following a design basis accident when the plant 
in the proper configuration as required by the proposed Technical 
Specifications.
    The proposed licensing basis changes do not involve a change in 
the function or use of the containment cooling system. It does 
assure that the containment cooling function is provided during 
plant operations for post-accident mitigation. There are no new 
failure modes or mechanisms created through allowing different 
combinations of fan coil units to provide the cooling function as 
proposed by this Technical Specification change. There are no new 
accident precursors generated by providing the required cooling 
function with an operable fan coil unit from each train.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    This license amendment proposes to revise the Technical 
Specifications to allow plant operation to continue for a limited 
time period under Technical Specification controls with two fan coil 
units, one fan coil unit from each containment cooling train, 
providing the required cooling function. Analyses demonstrate that 
any two fan coil units, whether they are in the same train or from 
opposite trains, are sufficient to supply the required containment 
cooling following a design basis accident when the plant in the 
proper configuration as required by the proposed Technical 
Specifications.
    Current plant Technical Specifications allow plant operation to 
continue for 7 days with the containment cooling function provided 
by the two operable fan coil units of a single operable containment 
cooling train. This is acceptable because engineering analyses 
demonstrate that the two fan coil units of a single train can 
provide the required post-accident containment cooling.
    Likewise, engineering analyses demonstrate that any two fan coil 
units from opposite containment cooling trains can also provide the 
required post-accident containment cooling if the cooling water flow 
to the other fan coil unit in each train is isolated. This license 
amendment request proposes Technical Specifications which will allow 
plant operation to continue for 7 days with the containment cooling 
function provided by two fan coils from opposite trains provided the 
cooling water flow to the other fan coil unit in each train is 
isolated. Thus, from a cooling capacity perspective, this proposed 
Technical Specification change does not involve a reduction in a 
margin of safety.
    When inoperable plant systems are under Technical Specification 
controls that limit the time for inoperability, a single failure in 
addition to the inoperable equipment is not postulated. Therefore, 
whether two inoperable fan coil units are in the same train or 
opposite trains does not change the availability of the two 
remaining operable fan coil units. Thus from a Technical 
Specification perspective, this proposed Technical Specification 
change does not involve a reduction in a margin of safety.
    Therefore, based on the considerations given above, the proposed 
changes do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve 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: Timothy J. Kobetz.

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

    Date of amendment request: December 19, 2005.
    Description of amendment request: The proposed change will revise 
Fort Calhoun Station, (FCS) Technical Specification 2.4, ``Containment 
Cooling,'' (and associated Bases) to reduce the required number of 
operable Containment Spray (CS) pumps from three to two in order to 
enhance net positive suction head (NPSH) margins. This change will be 
accomplished by disabling the containment spray actuation signal (CSAS) 
automatic start feature of CS pump SI-3C. This change will reduce the 
head loss across the containment sump strainers during the 
recirculation phase of a design-basis accident (DBA) by reducing flow 
rates, and will improve NPSH available (NPSHA).
    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 Containment Spray (CS) system is not an initiator of any 
accident previously evaluated at the Fort Calhoun Station (FCS); the 
CS system is an accident mitigation system. The CS system's 
licensing basis functions are to limit the containment pressure rise 
and reduce the leakage of airborne radioactivity from the 
containment by providing a means for cooling the containment 
following a loss-of-coolant accident (LOCA) or main steam line break 
(MSLB) inside containment. The proposed change disables the CSAS 
automatic start feature of one of the three CS pumps.
    The only FCS safety analysis that currently assumes three CS 
pumps operating to mitigate an accident is the Containment Pressure 
Analysis for a[n] MSLB inside containment. Even though this analysis 
assumes operation of all three CS pumps, it also shows that peak 
containment pressure occurs prior to the CS system starting, 
therefore, the CS system does not mitigate the peak pressure for 
a[n] MSLB. The reviews evaluated both existing AORs [analyses of 
record] and those analyses developed for the Steam Generator 
Replacement (RSG) project. The analysis developed for the RSG 
project that evaluates the Containment Pressure Analysis for MSLB 
inside containment was reviewed for the impact of reducing the 
number of operating CS pumps from three to two. This review 
determined that the RSG MSLB analysis will be acceptable and will 
continue to be bounded by the analysis currently documented in USAR. 
AOR peak pressure is unaffected by implementation of this proposed 
change. Therefore, the combination of the RSG project and this 
containment spray modification will not result in an increase in the 
currently documented peak containment pressure for an MSLB. 
Therefore, the evaluation for the MSLB event has determined that the 
containment pressure response is acceptable with less than three CS 
pumps operating.
    The LOCA analysis source term is based on operation of minimum 
safeguards due to a worst-case single failure. The minimum 
safeguards configuration is unchanged by this modification. 
Following implementation of the proposed change at least one CS pump 
will be available to mitigate a LOCA as currently assumed in the 
analysis, therefore, the proposed change will have no adverse effect 
on the radiological consequences following a LOCA. The analyses that 
establish the radiological consequences for the site are based on a 
Large Break LOCA with a single CS pump in operation, therefore, 
single CS pump operation during a[n] MSLB inside containment is 
bounded by the LOCA analysis.
    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.
    The proposed change will reduce the number of operable CS pumps 
from three to two; however, previous accident analyses will remain 
valid. No credible new failure mechanisms, malfunctions, or accident 
initiators not considered in the design and licensing basis have 
been created and none of the initial condition assumptions of any 
accident evaluated in the safety analysis are impacted.
    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?

[[Page 10076]]

    Response: No.
    The containment building and associated penetrations are 
designed to withstand an internal pressure of 60 psig at 305 [deg]F, 
including all thermal loads resulting from the temperature 
associated with this pressure, with a leakage rate of 0.1 percent by 
weight or less of the contained volume per 24 hours. The CS System 
and the Containment Fan Coolers are credited for maintaining 
containment pressure and temperatures within design limitations, and 
assure that the release of fission products to the environment 
following a design[-]basis accident will not exceed regulatory 
guidelines. The FCS licensing basis credits only one of the three CS 
pumps to limit the containment pressure to below the design value 
for a LOCA. Currently, the FCS licensing basis credits three CS 
pumps for a[n] MSLB, however, the CS system is not credited for 
limiting peak containment pressure for a[n] MSLB.
    The EEQ [electrical equipment qualification] profile developed 
for the current plant configuration bounds those associated with the 
upcoming RSG modification. Both the proposed CS system changes and 
the RSG projects are scheduled for the same refueling outage. The 
thermal lag analysis of equipment performed using the current plant 
configuration demonstrated a large margin between the equipment 
evaluated during the accident versus the conditions under which it 
was tested. The RSG modification will further increase this margin. 
As part of the RSG effort the EEQ analysis will be revised to 
address RSG issues and will include the changes to containment 
spray. When the margins associated with the current analysis as well 
as increases in margin when the new analysis is implemented it is 
expected that the changes to the containment spray system will not 
produce an adverse result. All equipment will remain qualified to 
operate in the accident environment.
    Additionally, the CFCs [containment fan coolers] operate 
independently of the CS system to remove heat from the containment 
atmosphere. The CFCs consist of two redundant trains; each train 
with one air cooling and filtering unit and one air cooling unit, 
for a total of four cooling units. Operation of the CFCs is credited 
in the MSLB containment pressure analysis. The CFCs are not impacted 
by this proposed change. During the MSLB containment spray takes 
place after the peak containment pressure occurs. Therefore, the 
licensing basis capabilities of the Containment Cooling System, 
which consists of the CS and CFCs, is not adversely affected by the 
proposed change; the ability to maintain containment peak pressure 
and temperature and long[-]term containment pressure and temperature 
will be maintained.
    Particulate fission products that are released into the 
containment following a DBA are removed by the CS system for those 
events that result in CS actuation. The water spray strips 
radioactive particles from the atmosphere where they fall to the 
floor and are washed into the containment sump. The radiological 
consequences analysis credits CS system operation for removal of 
particulates from the containment atmosphere during a LOCA. The LOCA 
analysis source term is based on operation of minimum safeguards due 
to a worst-case single failure, and a presumption of core damage. 
Minimum safeguards corresponds to one CS pump and one CS header 
operation and take into account pump degradation, and instrument 
uncertainties. The analyses that establish the radiological 
consequences for the site are not impacted by the proposed 
modification. These analyses are based on a Large Break LOCA with a 
single CS pump in operation. Therefore, single CS pump operation 
bounds the plant configuration following the proposed modification.
    The Large Break LOCA assumes that there will be three CS pumps 
operating when evaluating the effects of containment pressure on 
ECCS [emergency core cooling system] performance. The analysis 
assumes three CS pumps, which minimizes containment pressure, to 
conservatively evaluate ECCS performance in response to a LOCA. The 
use of two CS pumps versus three improves ECCS performance and thus 
increases margin to 10 CFR 50.46 limits on peak clad temperature.
    In summary, following implementation of the proposed change:
     Peak containment pressure for analyzed DBAs will not be 
increased;
     The assumptions used in the environmental qualification 
of equipment exposed to the containment atmosphere following a DBA 
remaining bounding; and
     The radiological consequences for the bounding DBA 
remains unchanged.
     The currently calculated peak clad temperature 
following a LOCA remains bounded by existing analysis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon 
Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California

    Date of amendment requests: January 13, 2006.
    Description of amendment requests: The proposed amendment would 
revise Technical Specification 5.6.5, ``Core Operating Limits Report 
(COLR),'' by adding WCAP-16009-P-A, ``Realistic Large-Break LOCA [Loss-
of-Coolant Accident] Evaluation Methodology Using the Automated 
Statistical Treatment of Uncertainty Method (ASTRUM),'' dated January 
2005, as an approved analytical method for determining core operating 
limits for Unit 2.
    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 allow the use of the best estimate loss-
of-coolant accident (LOCA) analysis methodology using the automated 
statistical treatment of uncertainty methodology (ASTRUM) does not 
involve a physical alteration of any plant equipment or change 
operating practice at Unit 2 of Diablo Canyon Power Plant (DCPP). 
Therefore, there will be no increase in the probability of a LOCA. 
The consequences of a LOCA are not being increased.
    The plant conditions assumed in the analysis are bounded by the 
design conditions for all equipment in Unit 2. That is, it is shown 
that the emergency core cooling system is designed so that its 
calculated cooling performance conforms to the criteria contained in 
10 CFR [Title 10 of the Code of Federal Regulations, Section] 50.46, 
paragraph b. No other accident is potentially affected by this 
change.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different accident from any accident previously evaluated?
    Response: No.
    The proposed change would not result in any physical alteration 
to any Unit 2 system, and there would not be a change in the method 
by which any safety [-]related system performs its function. 
Analyses of transient events have confirmed that no transient event 
results in a new sequence of events that could lead to a new 
accident scenario. The parameters assumed in the analysis are within 
the design limits of existing plant equipment.
    In addition, employing the ASTRUM methodology does not create 
any new failure modes that could lead to a different kind of 
accident. The design of all systems remains unchanged and no changes 
are being made to any reactor protection system or engineered 
safeguard features 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 proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.

[[Page 10077]]

    It has been shown that the analytic technique used in the 
analysis realistically describes the expected behavior of the DCPP 
Unit 2 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 analyzed to provide assurance 
that the most severe postulated LOCAs were analyzed. The analysis 
has demonstrated that all acceptance criteria contained in 10 CFR 
50.46[,] paragraph b continue to be satisfied.
    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 requests involve no significant hazards consideration.
    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.

Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power 
Plant (HBPP), Unit 3 Humboldt County, California

    Date of amendment request: January 19, 2006.
    Description of amendment request: The licensee has proposed to 
revise the Technical Specifications (TS) to correct an editorial error 
in TS 3.1.2, ``Spent Fuel Pool Load Restrictions,'' and to change TS 
5.2.2, ``Facility Staff,'' to allow the Unit 3 control room to be 
temporarily unmanned during emergency conditions that require personnel 
to evacuate buildings for their safety.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed editorial change has no impact on probability or 
consequences of accidents. The following discussion applies to the 
proposed change related to control room evacuation.
    Allowing plant personnel to not continuously man the control 
room has no impact on the probability of an accident from occurring, 
especially acts of nature such as earthquakes and tsunamis.
    The HBPP DSAR, Appendix A, and NRC SER, Section 10, dated April 
29, 1987, evaluate various accidents at HBPP. Because all fuel has 
been removed from the reactor vessel and stored in the spent fuel 
pool, the majority of accidents analyzed pertain to events that 
could only affect spent fuel or the spent fuel pool. All accidents 
affecting spent fuel or the spent fuel pool do not require operator 
action to protect the public health and safety or to maintain 
offsite radiological doses well within regulatory limits. In 
addition, NRC SER, Section 10.7, ``Impact of Tsunami Flooding,'' 
analyzes the impact of tsunami flooding. That analysis identifies a 
likely impact of the tsunami to be a release of the radwaste tank 
radionuclide contents to the bay and some damage to the reactor 
building. For both situations, no operator action is required to 
maintain offsite radiological doses well within regulatory limits.
    Allowing the control room to be temporarily unmanned under 
emergency conditions does not create problems that could increase 
the consequences of an accident. The primary function of manning the 
control room is for an operator to observe and acknowledge alarms. 
Recovery actions to respond to damage to spent fuel, the spent fuel 
pool, or radwaste tanks are taken by personnel outside the control 
room. No recovery actions are required to be taken by the control 
room operator to respond to damage to spent fuel, the spent fuel 
pool, or radwaste tanks.
    Evacuating occupied buildings, including the control room, 
during a tsunami, allows the control room operator to return to the 
control room after the tsunami and assess damage by observing 
indicators and alarms. Upon returning to the control room, the 
operator would be able to direct and monitor recovery efforts from 
the control room that may be necessary to bring plant parameters 
within required specifications.
    If an operator remains in the control room during a tsunami and 
becomes injured, that operator would be unable to direct and monitor 
recovery efforts. Under this scenario, other plant personnel who 
evacuated to higher ground onsite within the OCA would eventually 
return to the plant, including the control room, and perform any 
required recovery functions. Therefore, consequences of a tsunami 
are not increased by not continually manning the control room during 
the event.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident evaluated?
    Response: No.
    The proposed editorial change has no impact on accidents. The 
following discussion applies to the proposed change related to 
control room evacuation.
    As discussed in the response to question 1 above, none of the 
analyzed accidents require operator action to keep offsite 
radiological doses well within regulatory limits. In addition, 
allowing plant personnel to not continuously man the control room 
after an emergency situation has occurred, has no impact on the 
possibility of a new or different kind of accident from occurring. 
If the plant is evacuated, no work activities will be performed in 
the plant. With the plant in SAFSTOR and no work being performed, 
there are no actions required to be taken by personnel manning the 
control room.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: The proposed editorial change has no impact on margin 
of safety. The following discussion applies to the proposed change 
related to control room evacuation.
    NRC SER Section 10.8, ``Accident Analysis Conclusions,'' 
summarizes the consequences from accidents in terms of offsite 
radiological doses. SER Section 10.8 includes the statement, ``The 
(NRC) staff has determined that offsite radiological consequences 
due to a tsunami are within acceptable dose guideline values.'' As 
discussed in the response to question 1 above, none of the analyzed 
accidents require operator action to keep offsite radiological doses 
well within regulatory limits. Therefore, temporarily not manning 
the control room during an emergency will have no impact on the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based upon 
the staff's review of the licensee's analyses as well as the staff's 
own evaluation, the staff concludes 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: Richard F. Locke, Esquire, Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Claudia Craig.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: January 31, 2006.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 3.8.3.1, ``Onsite Power Distribution-
Operating,'' to extend the allowed outage time (AOT) for an inoperable 
Class 1E vital 120-volt alternating current inverter. The TS currently 
provides an AOT of 24 hours to restore an inoperable inverter. Based on 
risk-informed assessment, the amendments would extend the AOT to 7 
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 formatting changes to TS 3.8.3.1 Action b and the 
change to the AOT

[[Page 10078]]

for an inoperable inverter to be extended from 24 hours to 7 days do 
not alter any plant equipment or operating practices in such a 
manner that the probability of an accident is increased. The 
proposed changes will not alter assumptions relative to the 
mitigation of an accident or transient event.
    An evaluation was performed to determine the risk significance 
of the proposed change to the AOT. The risk evaluation concludes 
that the [Delta]CDF [core damage frequency] and [Delta]LERF [large 
early release frequency] associated with the proposed changes are 
1.88E-07 and 2.05E-09, respectively, which are characterized as 
``very small changes'' by RG [Regulatory Guide] 1.174. The ICCDP 
[incremental conditional core damage probability] and ICLERP 
[incremental conditional large early release probability] associated 
with the proposed change are 3.63E-07 and 1.08E-08, respectively, 
which are within the acceptance criteria in RG 1.177. Therefore, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. Therefore, 
the proposed changes do 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.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel and fuel cladding, reactor 
coolant pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed change to TS 
3.8.3.1 to allow the AOT for an inoperable inverter to be extended 
from 24 hours to 7 days has been evaluated for its effect on plant 
safety. The risk-informed evaluation concludes that the [Delta]CDF 
and [Delta]LERF associated with the proposed change are 1.88E-07 and 
2.05E-09, respectively, which are characterized as ``very small 
changes'' by RG 1.174. The ICCDP and ICLERP associated with the 
proposed change are 3.63E-07 and 1.08E-08, respectively, which are 
within the acceptance criteria in RG 1.177. The proposed changes to 
the formatting of TS 3.8.3.1 Action b are administrative only and 
have no impact on margin of safety. Therefore, the proposed changes 
do 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Branch Chief: David Terao.

Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear 
Plant, Unit 1 (WBN) Rhea County, Tennessee

    Date of amendment request: December 14, 2005 (TS-05-07).
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 5.7.2.19, ``Containment Leakage 
Rate Testing Program,'' to allow a one time, 5-year extension to the 
current 10-year test interval for the performance-based leakage rate 
test program for 10 CFR Part 50, Appendix J, Type A tests.
    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 for extending Type A test frequency does not 
significantly increase the probability of an accident previously 
evaluated since the change is not a modification to plant systems, 
nor a change to plant operation that could initiate an accident.
    TVA performed an evaluation of the risk significance for the 
proposed increase to the WBN Unit 1 Type A test frequency. The 
results of the TVA risk evaluation indicates that the increase in 
Large Early Release Frequency (LERF) remains below the level of risk 
significance defined in the NRC Regulatory Guide 1.174, ``An 
Approach for Using Probabilistic Risk Assessment In Risk-Informed 
Decisions On Plant-Specific Changes to the Licensing Basis.'' TVA's 
evaluation indicates that the calculated increase in frequency for 
all releases (small, large, early and late) and the increase in 
radiation dose to the population are also non-risk significant.
    The proposed test interval extension does not involve a 
significant increase in the consequences of an accident. Research 
documented in NUREG-1493, ``Performance-Based Containment Leakage-
Test Program,'' determined that generically, very few potential 
containment leakage paths fail to be identified by Type A tests. An 
analysis of 144 Type A test results, including 23 failures, found 
that no failures were due to containment liner breach. The NUREG 
concluded that reducing the Type A test frequency to once per 20 
years would lead to an imperceptible increase in risk. Furthermore, 
the NUREG concluded that Type B and C testing provides assurance 
that containment leakage from penetration leak paths (i.e., valves, 
flanges, containment air-locks) identify any leakage that would 
otherwise be detected by the Type A tests.
    In addition to the NUREG conclusions, TVA's American Society of 
Mechanical Engineers (ASME) IWE program performs containment 
inspections in order to detect evidence of degradation that may 
either affect the containment structural integrity or leak 
tightness.
    Therefore, the proposed extension of the Type A test interval 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to extend the Type A test interval does not 
create the possibility of a new or different type of accident 
because there are no physical changes made to the plant or plant 
equipment governing normal plant operation. There are no changes to 
the operation of the plant that would introduce a new failure mode 
creating the possibility of a new or different kind of accident. 
Therefore, the proposed extension does not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to extend the Type A test interval will not 
significantly reduce the margin of safety. A generic study 
documented in NUREG-1493 indicates that extending the Type A leak 
test interval to 20 years would result in an imperceptible increase 
in risk to the public. The NUREG also found that, generically, the 
containment leakage rate contributes a very small amount to the 
individual risk and that the decrease in the Type A test frequency 
would have a minimal effect on risk because most potential leakage 
paths are detected by Type C testing.
    Previous Type A leakage tests conducted on WBN Unit 1 indicate 
that leakage from containment have been less than the 10 CFR 50, 
Appendix J leakage limit of 1.0 La. A review of the 
previous Type A test results indicate a stable trend with an 
increase of less than 15 percent of La, well below the 
1.0 La leakage limit.
    Therefore, these test results, in conjunction with the research 
findings from NUREG-1493, provide assurance that the proposed 
extension to the Type A test interval 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Michael L. Marshall, Jr.

[[Page 10079]]

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

    Date of application request: August 26, 2005, as supplemented by 
letter dated December 16, 2005.
    Description of amendment request: The amendment would authorize 
changes to the Final Safety Analysis Report (FSAR) for the Callaway 
Plant, Unit 1, that would revise the methodology for the reactor 
coolant system (RCS) leak detection instrumentation. This revision 
would clarify the requirements of the containment atmosphere gaseous 
radioactivity monitor with regard to the RCS leak detection capability 
and would justify that the monitor can be considered operable in 
compliance with Limiting Condition for Operation 3.4.15, in Technical 
Specification (TS) 3.4.15, ``RCS Leakage Detection Instrumentation,'' 
during all applicable reactor modes. There are no proposed changes to 
the TS.
    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 proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change has been evaluated and determined to not 
increase the probability or consequences of an accident previously 
evaluated. The proposed change does not make hardware changes and 
does not alter the configuration of any plant system, structure, or 
component (SSC). The proposed change only clarifies the design and 
OPERABILITY requirements for the containment atmosphere gaseous 
radioactivity monitor[s] and identifies the capabilities of the 
containment atmosphere gaseous radioactivity monitors at low RCS 
[radio]activity levels. The containment radiation monitors are not 
initiators of any accident; therefore, the probability of occurrence 
of an accident is not increased. The FSAR and TS will continue to 
require diverse means of [RCS] leakage detection equipment, thus 
ensuring that leakage due to cracks [in the RCS] would continue to 
be identified prior to propagating to the point of a [RCS] pipe 
break. Therefore, the consequences of an accident [previously 
evaluated] are not increased.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve the use or installation of 
new equipment and the currently installed equipment will not be 
operated in a new or different manner. No new or different system 
interactions are created and no new processes are introduced. The 
proposed changes will not introduce any new failure mechanisms, 
malfunctions, or accident initiators not already considered in the 
design and licensing bas[i]s [for the Callaway Plant]. The proposed 
change does not affect any SSC associated with an accident 
initiator. Based on this evaluation, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not alter any RCS leakage detection 
components. The proposed change only clarifies the design and 
OPERABILITY requirements for the containment atmosphere gaseous 
radioactivity monitor[s] and identifies the capabilities of the 
containment atmosphere gaseous radioactivity monitors at low RCS 
[radio]activity levels. This change is required since the level of 
radioactivity in the Callaway Plant reactor coolant has become much 
lower than what was assumed in the FSAR [when the plant was 
licensed] and the gaseous channel [(monitor)] can no longer promptly 
detect a small RCS leak under all operating conditions. The proposed 
amendment continues to require diverse means of [RCS] leakage 
detection equipment with [the] capability to promptly detect RCS 
leakage. Although not required by TS, additional diverse means of 
leakage detection capability are available as described in the FSAR 
Section 5.2.5. Early detection of [RCS] leakage, as the potential 
indicator of a crack(s) in the RCS pressure boundary, will thus 
continue to be in place so that such a condition is known and 
appropriate actions taken well before any such crack would propagate 
to a more severe condition. Based on this evaluation, 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: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: David Terao.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: February 1, 2006.
    Description of amendment request: The amendment would revise the 
Inservice Testing Program in Section 5.5.8 of the Administrative 
Controls, Programs and Manuals, section of the Technical Specifications 
(TSs). The licensee is adopting NRC-approved Technical Specification 
Task Force (TSTF) 479, Revision 0, ``Changes to Reflect Revision of 10 
CFR 50.55a.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS 5.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 50.55a(f) 
regarding the inservice testing of pumps and valves. The proposed 
change incorporates revisions to the ASME [American Society of 
Mechanical Engineers] Code [for Operation and Maintenance of Nuclear 
Power Plants] that result in a net improvement in the measures for 
testing pumps and valves.
    The proposed change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. They do not involve the addition or removal of any 
equipment, or any design changes to the facility. Therefore, the 
proposed change does not represent a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises TS 5.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 50.55a(f) 
regarding the inservice testing of pumps and valves. The proposed 
change incorporates revisions to the ASME Code that result in a net 
improvement in the measures for testing pumps and valves.
    The proposed change does not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. Additionally, there is no 
change in the types or increases in the amounts of any effluent that 
may be released off-site and there is no increase in individual or 
cumulative occupational exposure. Therefore, this proposed change 
does not create the possibility of an accident of a different kind 
than previously evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The proposed change revises TS 5.5.8, ``Inservice Testing 
Program,'' for consistency with the requirements of 10 CFR 50.55a(f) 
regarding the inservice testing of pumps and valves. The proposed 
change incorporates revisions to the ASME Code that result in a net 
improvement in the measures for testing pumps and valves. The safety 
function of the affected pumps and valves will be

[[Page 10080]]

maintained. Therefore, this proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: David Terao.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: February 7, 2006.
    Description of amendment request: The amendment would add 
Surveillance Requirement (SR) 3.3.1.16, to verify the reactor trip 
system response time, to Function 3.a, power range neutron flux--high 
positive rate trip function, in Table 3.3.1-1, ``Reactor Trip System 
Instrumentation,'' of the Technical Specifications (TSs).
    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.
    Overall protection system performance will remain within the 
bounds of the accident analysis since there are no hardware changes. 
The design of the Reactor Trip System (RTS) instrumentation, 
specifically the positive [neutron] flux rate trip (PFRT) function, 
will be unaffected. The reactor protection system will continue to 
function in a manner consistent with the plant design basis. All 
design, material, and construction standards that were applicable 
prior to the request [(i.e., this amendment application)] are 
maintained.
    The proposed change imposes additional surveillance requirements 
to assure safety related structures, systems, and components are 
verified to be consistent with the [plant] safety analysis and 
licensing basis. In this specific case, a response time verification 
requirement will be added to the PFRT Function [in TS Table 3.3.1-
1].
    The proposed [change] will not modify any system interface. The 
proposed [change] will not affect the probability of any event 
initiators. There will be no degradation in the performance of or an 
increase in the number of challenges imposed on safety-related 
equipment assumed to function during an accident situation. There 
will be no change to normal plant operating parameters or accident 
mitigation performance. The proposed [change] will not alter any 
assumptions or change any mitigation actions in the radiological 
consequence evaluations in the Updated Safety Analysis Report (USAR) 
[for Wolf Creek Generating Station].
    The proposed [change does] not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, or configuration of the facility or the manner in which 
the plant is operated or maintained. The proposed [change does] not 
alter or prevent the ability of structures, systems, and components 
(SSCs) from performing their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed [change does] not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of an accident previously 
evaluated. The proposed [change is] consistent with the safety 
analysis assumptions and resultant consequences.
    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.
    There are no hardware changes nor are there any changes in the 
method by which any safety related plant system performs its safety 
function. This change will not affect the normal method of plant 
operation or change any operating parameters. No performance 
requirements will be affected; however, the proposed change does 
impose additional surveillance requirements. The additional 
requirements are consistent with assumptions made in the safety 
analysis and licensing basis.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of [the change]. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of [the change].
    Therefore, the proposed change does not create the possibility 
of a new or different [kind of] accident from any accident 
previously evaluated.
    (3) Does the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The proposed [change does] not affect the acceptance criteria 
for any analyzed event nor is there a change to any Safety Analysis 
Limit (SAL). There will be no effect on the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions. There will be no impact on the overpower limit, DNBR 
[departure from nucleate boiling ratio] limit, FQ [heat 
flux hot channel factor], F[utri]H [nuclear enthalpy rise hot 
channel factor], LOCA PCT [loss-of-coolant accident peak cladding 
temperature], peak local power density, or any other margin of 
safety. The radiological dose consequence acceptance criteria listed 
in the [NRC] Standard Review Plan [NUREG-0800] will continue to be 
met.
    The safety analysis limits assumed in the transient and accident 
analyses are unchanged. None of the acceptance criteria for any 
accident analysis is changed. The imposition of additional 
surveillance requirements increases the margin of safety by assuring 
that the affected safety analysis assumptions on equipment response 
time are verified on a periodic frequency. 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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Branch Chief: David Terao.

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.

[[Page 10081]]

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

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of application for amendments: August 11, 2005.
    Brief Description of amendments: The amendments revise Technical 
Specification (TS) 5.5.12, ``Primary Containment Leakage Rate Testing 
Program,'' by removing an exception that allows for compensation of 
flow meter instrument inaccuracies in accordance with ANSI/ANS-56.8-
1987 rather than ANSI/ANS-56.8-1994.
    Date of issuance: February 8, 2006.
    Effective date: Date of issuance to be implemented within 60 days.
    Amendment Nos.: 238 and 266.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the TS.
    Date of initial notice in Federal Register: September 13, 2005 (70 
FR 54087).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 8, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 8, 2005.
    Brief description of amendment: The proposed changes would add 
Limiting Condition for Operation 3.0.8 to address conditions where one 
or more snubbers are unable to perform their associated support 
function.
    Date of issuance: February 13, 2006.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 245 and 229.
    Facility Operating License Nos. DPR-26 and DPR-64: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 16, 2005 (70 FR 
48203).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 13, 2006.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: March 7, 2005, as supplemented 
by letter dated December 5, 2005.
    Brief description of amendments: The amendments will add two 
Nuclear Regulatory Commission (NRC) approved topical report references 
to the list of analytical methods in Technical Specification 5.6.5, 
``Core Operating Limits Report,'' that can be used to determine core 
operating limits.
    Date of issuance: February 1, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 174 and 160.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 16, 2005 (70 FR 
48205).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 1, 2006.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of application for amendments: December 17, 2004.
    Brief description of amendments: The amendments revised the 
Appendix B, Environmental Protection Plan (non-radiological), of the 
Quad Cities Station Renewed Facility Operating Licenses.
    Date of issuance: February 2, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 229 and 224.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Environmental Protection Plan.
    Date of initial notice in Federal Register: April 12, 2005 (70 FR 
19115).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 2, 2006.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County, 
Pennsylvania

    Date of application for amendment: April 13, 2005, as supplemented 
by letters dated August 26, October 28 and 31, November 18, and 
December 6 and 16, 2005.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to allow replacement of the BVPS-1 steam 
generators (SGs). These changes include revising the fuel assembly-
specific departure from nucleate boiling ratios and correlations, 
modifying the Overtemperature [Delta]T and Overpower [Delta]T 
equations, revising the SG water level low-low and high-high setpoints, 
revising the SG secondary side level in Modes 4 and 5, revising the SG 
TSs to reflect the replacement SGs and remove TS requirements that are 
no longer applicable to the new SGs, revising the required charging 
pump discharge pressure for reactor coolant pump seal injection flow, 
raising the accumulator pressure, and adding WCAP-14565-P-A (VIPRE) and 
WCAP-15025-P-A (WRB-2M) Topical Reports to the list of NRC-approved 
methodologies listed in TS 6.9.5. The amendment also approves an 
expanded selective alternate source term methodology implementation in 
accordance with Regulatory Guide 1.183, ``Alternate Radiological Source 
Terms for Evaluating Design Basis Accidents at Nuclear Power 
Reactors,'' and approves use of the 1979 ANS Decay Heat + 2[sigma] 
model for mass and energy releases for a main steam line break outside 
containment.
    Date of issuance: February 9, 2005.
    Effective date: As of its date of issuance and shall be implemented 
prior to entry into Mode 4 upon startup from refueling outage 1R17 
which begins on or about February 10, 2006.
    Amendment No: 273.

[[Page 10082]]

    Facility Operating License No. DPR-66: The Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 21, 2005 (70 FR 
35737). The supplements dated August 26, October 28 and 31, November 
18, and December 6 and 16, 2005, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 9, 2006.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania

    Date of application for amendments: October 4, 2004, as 
supplemented July 8, and November 14, 2005.
    Brief description of amendments: These amendments approved 
application of the Westinghouse best-estimate loss-of-coolant accident 
(LOCA) analysis methodology to BVPS-1 and 2 for large-break LOCA 
analysis.
    Date of issuance: February 6, 2006.
    Effective date: These license amendments are effective as of the 
date of issuance and shall be implemented for BVPS-1, prior to Mode 4 
entry during startup from refueling outage 1R17 which begins on or 
about February 10, 2006, and for BVPS-2, prior to Mode 4 entry during 
startup from refueling outage 2R12 which begins October 2006.
    Amendment Nos.: 272 and 154.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 7, 2004 (69 FR 
70718). The supplements dated July 8, and November 14, 2005, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the Nuclear Regulatory Commission staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 6, 2006.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendment: August 10, 2005.
    Brief description of amendment: The amendments deleted the power 
range neutron flux high negative rate trip function from Table 3.3.1-1, 
``Reactor Trip System Instrumentation.''
    Date of issuance: February 10, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 293, 275.
    Facility Operating License No. DPR-58: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 6, 2005 (70 FR 
72674). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 10, 2006.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of application for amendment: June 7, 2004, as supplemented by 
letters dated February 18, May 20, June 16, July 8, August 3, September 
23, and November 16, 2005, and February 6, 2006.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to reflect an expanded operating domain resulting 
from the implementation of the Average Power Range Monitor, Rod Block 
Monitor TSs/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA).
    Date of issuance: February 8, 2006.
    Effective date: As of the date of issuance, to be implemented 
within 120 days.
    Amendment No.: 163.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: September 14, 2004 (69 
FR 55471). The supplements dated February 18, May 20, June 16, July 8, 
August 3, September 23, and November 16, 2005, and February 6, 2006, 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination or expand 
the application beyond the scope of the original Federal Register 
notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 8, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 1, 2005, supplemented 
by letters dated February 22, September 16, December 2, 2005, and 
January 5, 2006.
    Brief description of amendments: The amendments revise the spent 
fuel pool (SFP) criticality analysis methodology and technical 
specifications governing the storage of irradiated fuel in the SFP. The 
licensee's amendment request stated that subcritical conditions would 
be maintained in the SFP under the revised technical specification 
storage requirements.
    Date of issuance: February 5, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 172, 162.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 15, 2005, (70 FR 
12748). The supplemental letters contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice. The Commission's related evaluation of the amendments 
is contained in a Safety Evaluation dated February 5, 2006.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendments request: January 19, 2005, as supplemented on 
June 9 (two letters) and November 18, 2005.
    Brief Description of amendments: The amendment authorizes revision 
of the Updated Final Safety Analysis Report (UFSAR) to reflect the 
utilization of fire-rated electrical Mineral Insulated cables in lieu 
of Appendix R, Section III.G.2 1-hour rated fire barriers.
    Date of issuance: February 13, 2006.
    Effective date: As of the date of issuance, to be incorporated into 
the UFSAR at the time of its next update.
    Amendment No.: 162.
    Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendment 
authorizes revision to the UFSAR.
    Date of initial notice in Federal Register: April 26, 2005 (70 FR 
21464). The supplemental letters provided clarifying information that 
was within

[[Page 10083]]

the scope of the initial notice and did not change the initial proposed 
no significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated February 13, 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 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/

[[Page 10084]]

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.
---------------------------------------------------------------------------

    \1\ To the extent that the applications contain attachments and 
supporting documents that are not publicly available 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).

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: February 5, 2006, as supplemented 
February 5, 2006.
    Description of amendment request: The amendment revised Technical 
Specification 3.8.1, ``AC Sources--Operating,'' to extend the allowed 
outage time for Emergency Diesel Generator 12 from seven days to 14 
days for one specific incident.
    Date of issuance: February 6, 2006.
    Effective date: As of the date of issuance and shall be implemented 
immediately.
    Amendment No.: 171.
    Facility Operating License No. 50-341: Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No. The Commission's related evaluation of the 
amendment, finding of emergency circumstances, state consultation, and 
final NSHC determination are contained in a safety evaluation dated 
February 6, 2006.
    Attorney for licensee: David G. Pettinari, Legal Department, 688 
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
    NRC Branch Chief: Timothy J. Kobetz, Acting.

    Dated at Rockville, Maryland, this 16th day of February, 2006.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 06-1737 Filed 2-27-06; 8:45 am]
BILLING CODE 7590-01-P