[Federal Register Volume 71, Number 99 (Tuesday, May 23, 2006)]
[Notices]
[Pages 29671-29686]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-4736]


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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 April 28, 2006 to May 11, 2006. The last 
biweekly notice was published on May 9, 2006 (71 FR 26995).

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

[[Page 29672]]

proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, [email protected]; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-

[[Page 29673]]

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

Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin

    Date of amendment request: April 6, 2006.
    Description of amendment request: The proposed amendment would 
allow the use of a different methodology for determining the design 
requirements necessary for protecting safety-related equipment from 
damage by tornado generated missiles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The probability of occurrence of an accident previously 
evaluated is not significantly increased by the proposed change to 
permit probabilistic evaluation of missiles generated by natural 
phenomena. The actual frequency of tornado occurrence at Kewaunee is 
unaffected by the proposed change in assessment methodology. 
Furthermore, the projected frequency of tornado occurrence, as 
specified in the USAR [Updated Safety Analysis Report], is not 
significantly affected by this change. The value for the probability 
of tornado occurrence in the updated study is in general agreement 
with the original value in the USAR (i.e. 3.97E-4 vs. 4.86E-4). 
Similarly, the probability of a tornado-generated missile is not 
significantly affected by this change.
    Likewise, the consequences of an accident previously evaluated 
are not significantly increased by the proposed change. The actual 
probability of a tornado missile onsite remains unchanged. The 
actual probability of a tornado missile strike remains unchanged. 
For the limited number of components affected by this proposed 
change (i.e. exhaust ducts and fuel vent), the missile strike 
probability is approximately 5.75 x 10-7 per year, which 
is significantly lower than the SRP [Standard Review Plan] 
acceptance criteria of 1 x 10-6 per year. Therefore, the 
proposed change is not considered to constitute a significant 
increase in the consequences of an accident due to the low 
probability of occurrence.
    In addition, use of a probabilistic versus a deterministic 
methodology to assess missile hazard acceptability has no impact on 
accident initiation or consequence. 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 changes involve use of an evaluation methodology to 
determine protection requirements for two specific support 
components for safety-related equipment, which may be adversely 
affected by missiles during a tornado. A tornado at Kewaunee is 
considered in the USAR as a separate event and not occurring 
coincident with any of the design basis accidents in the USAR. As 
such, no new or different kind of accident is created by the 
proposed change to permit probabilistic evaluation of missiles 
generated by natural phenomena.
    This change involves recognition of the acceptability of 
performing tornado missile strike probability calculations in 
accordance with established regulatory guidance in lieu of using 
deterministic methodology alone. Therefore, the change would not 
create the possibility of, or be the initiator for, any new or 
different kind of accident. The acceptance criterion of the SRP 
guidance establishes a threshold for tornado missile damage to 
system components that is consistent with this conclusion.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The request does not involve a significant reduction in a margin 
of safety. The existing design basis for Kewaunee, with respect to a 
tornado affecting safety related equipment, is to provide positive 
missile barriers for all safety-related systems and components. The 
proposed change recognizes that for probability of occurrences below 
the SRP established acceptance limit, the extremely low probability 
associated with an ``important'' system or component being struck by 
a tornado missile does not represent a significant reduction in the 
margin of safety provided by use of the deterministic methodology. 
The change from ``protecting all safety-related systems and 
components'' to ``an extremely low probability of occurrence of 
tornado generated missile strikes on portions of important systems 
and components'' is not considered to constitute a significant 
reduction in the margin of safety.
    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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: March 28, 2006.
    Description of amendment request: The proposed amendment would 
eliminate redundant surveillance requirements [SRs] pertaining to post-
maintenance/post-modification testing. The associated TS bases will be 
updated to address the proposed changes.
    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 changes do not modify any plant equipment and do 
not impact any failure modes that could lead to an accident. Testing 
in accordance with the requirements of SR 4.0.1 will continue to 
provide the necessary assurance that the associated systems will 
function consistent with the assumptions used in the accident 
analyses. On this basis, the proposed amendment does not increase 
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 changes do not involve any physical changes to 
systems, structures, or components, or involve a change to the

[[Page 29674]]

method of plant operation. The requirement to perform post 
maintenance/post modification testing will continue to be 
implemented consistent with SR 4.0.1, through existing plant 
programs and procedures. As such, the proposed amendment does not 
introduce any new failure modes, 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 TS changes do not involve a significant reduction in a 
margin of safety because the requirements described in SR 4.0.1, as 
implemented through existing plant programs and procedures, will 
continue to ensure that post maintenance/post modification testing 
will be performed when necessary. The proposed change does not 
affect any of the assumptions used in the accident analyses, nor 
does it affect 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.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of amendment request: April 17, 2006.
    Description of amendment request: The proposed amendment would 
change the method for calculating fuel pool decay heat load from the 
original licensing basis methodology of ORIGEN and the Auxiliary 
Systems Branch Technical Position (ASBTP) 9-2, ``Residual Decay Heat 
Energy for Light Water Reactors for Long-Term Cooling,'' to ORIGEN-ARP.
    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 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 adoption of ORIGEN-ARP does not affect the probability or 
consequences of an accident previously evaluated. The calculation of 
the fuel pool decay heat load is used to evaluate and demonstrate 
the ability of the fuel pool cooling system to maintain the fuel 
pool temperatures within the acceptance limits specified in the 
Columbia Final Safety Analysis Report [FSAR]. The proposed change to 
the methodology used to calculate the fuel pool [decay] heat load 
has no bearing on the probability or consequences of any previously 
evaluated accident. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The change involves the use of a different methodology for 
calculating fuel pool decay heat load. This change does not involve 
any new equipment, it does not change any previously approved 
acceptance limits, and it does not affect or alter the operation of 
any equipment. Therefore[,] this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety provided by the fuel pool cooling system is 
primarily defined by the difference between the maximum allowed fuel 
pool temperature and the boiling point of water. The margin of 
safety is supplemented by the ability to make up water to the spent 
fuel pool if boiling were to occur. The proposed change in 
methodology for calculating the fuel pool [decay] heat load does not 
alter the current temperature limits or acceptance criteria 
specified in the FSAR and has no effect on the ability to provide 
make-up water if boiling were to occur. This change will allow 
Energy Northwest to more accurately calculate the fuel pool [decay] 
heat load to provide added confidence in the ability of the fuel 
pool cooling system to accommodate the heat load added to the spent 
fuel pool during refueling activities. Therefore, this 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: William A. Horin, Esq., Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of amendment request: April 18, 2006.
    Description of amendment request: The proposed change would modify 
technical specification surveillance requirement 3.6.1.1.2 by changing 
the test frequency of the drywell-to-suppression chamber bypass leakage 
test from 24 to 120 months. This proposed amendment also includes 
testing the suppression chamber-to-drywell vacuum breakers on a 24-
month frequency.
    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 operation of Columbia Generating Station in 
accordance with the proposed amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed changes would modify Technical Specification (TS) 
Surveillance Requirement (SR) 3.6.1.1.2 and add two new SRs, SR 
3.6.1.1.3 and SR 3.6.1.1.4. The proposed changes will extend the 
frequency for the drywell-to-suppression chamber bypass leakage test 
while maintaining the current leakage testing frequency for the 
suppression chamber-to-drywell vacuum breakers, and establish 
leakage acceptance criteria for the suppression chamber-to-drywell 
vacuum breakers when the valves are tested individually.
    The performance of a drywell-to-suppression chamber bypass 
leakage test or suppression chamber-to-drywell vacuum breaker 
leakage test is not a precursor to any accident previously 
evaluated. Thus, the proposed changes to the performance of the 
leakage tests do not have any affect on the probability of an 
accident previously evaluated.
    The performance of a drywell-to-suppression chamber bypass 
leakage test or a suppression chamber-to-drywell vacuum breaker 
leakage test continues to provide assurance that the containment 
will perform as designed. Thus, the radiological consequences of any 
accident previously evaluated are not impacted.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the operation of Columbia Generating Station in 
accordance with the proposed amendment create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to TS SR 3.6.1.1.2, and the addition of SR 
3.6.1.1.3, and SR 3.6.1.1.4 do not affect the assumed performance of 
any Columbia Generating

[[Page 29675]]

Station structure, system or component previously evaluated. The 
proposed changes do not introduce any new modes of system operation 
or any new failure mechanisms. This is an administrative change and 
does not involve the modification, addition or removal of any plant 
equipment.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the operation of Columbia Generating Station in 
accordance with the proposed amendment involve a significant 
reduction in the margin of safety?
    Response: No.
    The current frequency associated with a drywell-to-suppression 
chamber bypass leakage test in TS SR 3.6.1.1.2 is 24 months or 12 
months if two consecutive tests fail and continues at this frequency 
until two consecutive tests pass. The proposed change will modify 
this leakage test frequency to 120 months, or 48 months following 
one test failure or 24 months if two consecutive tests fail and 
continues at this frequency until two consecutive tests pass. The 
proposed change in SR 3.6.1.1.2 frequency is acceptable as the 
results from previous tests show that the measured drywell-to-
suppression chamber bypass leakage at the current TS frequency has 
been a small percentage of the allowable leakage. Acceptability is 
further demonstrated by the design requirements applied to the 
primary containment components and other periodically performed 
primary containment inspections.
    The proposed SR 3.6.1.1.3 will establish a leakage test 
frequency of 24 months for each suppression chamber-to-drywell 
vacuum breaker except when the leakage test of SR 3.6.1.1.2 has been 
performed within the past 24 months. SR 3.6.1.1.3 specifies a 
leakage limit for each suppression chamber-to-drywell vacuum breaker 
pathway of less than or equal to 12 percent of the bypass leakage 
limit of SR 3.6.1.1.2. The proposed SR 3.6.1.1.4 will establish a 
total leakage limit of less than or equal to 30 percent of the 
bypass leakage limit of SR 3.6.1.1.2 when the suppression chamber-
to-drywell vacuum breakers are tested in accordance with SR 
3.6.1.1.3.
    TS SR 3.6.1.1.2 drywell-to-suppression chamber bypass leakage 
test monitors the combined leakage of three types of pathways: (1) 
The drywell floor and downcomers, (2) piping externally connected to 
both the drywell and suppression chamber air space, and (3) the 
suppression chamber-to-drywell vacuum breakers. This amendment would 
extend the surveillance interval on the passive components of the 
test (the first two types of pathways), while retaining the current 
surveillance interval on the active components (suppression chamber-
to-drywell vacuum breakers). The proposed changes establish leakage 
limits for both individual suppression chamber-to-drywell vacuum 
breakers and the total leakage. Additional testing is required if 
acceptable results are not achieved.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: April 12, 2006.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification reactor pressure vessel Pressure and 
Temperature (P-T) curves.
    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 License Amendment (LA) does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. There are no physical changes to the 
plant being introduced by the proposed changes to the pressure-
temperature curves. The proposed change does not modify the reactor 
coolant pressure boundary, (i.e., there are no changes in operating 
pressure, materials, or seismic loading). The proposed change does 
not adversely affect the integrity of the reactor coolant pressure 
boundary such that its function in the control of radiological 
consequences is affected.
    The proposed pressure-temperature curves are generated in 
accordance with the fracture toughness requirements of 10 CFR 50 
Appendix G, and American Society of Mechanical Engineers (ASME) 
Boiler and Pressure Vessel (B&PV) Code, Section Xl, Appendix G and 
Regulatory Guide (R.G.) 1.99, Revision 2[,] ``Radiation 
Embrittlement of Reactor Vessel Materials.'' A best-estimate 
calculation of reactor vessel 34 effective full power years (EFPYs) 
neutron fluence and associated uncertainty has been completed for 
Pilgrim using the Radiation Analysis Modeling Application (RAMA) 
methodology. This methodology was previously approved by the NRC. 
The resulting reactor vessel neutron fluence value was then used in 
conjunction with R.G. 1.99, [Revision] 2 to determine the adjusted 
reference temperature (ART) and with ASME Section Xl Appendix G to 
develop revised P-T curves.
    This provides sufficient assurance that the Pilgrim reactor 
vessel will be operated in a manner that will protect it from 
brittle fracture under all operating conditions. This proposed 
license amendment provides compliance with the intent of 10 CFR 
[Part 50] Appendix G and provides margins of safety that assure 
reactor vessel integrity.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. [Does] the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed license amendment does not create the possibility 
of new or different kind of accident from any accident previously 
evaluated. The revised pressure-temperature curves are generated in 
accordance with the fracture toughness requirements of 10 CFR Part 
50 Appendix G and ASME Section Xl Appendix G. Compliance with the 
proposed pressure-temperature curves will ensure the avoidance of 
conditions in which brittle fracture of primary coolant pressure 
boundary materials is possible because such compliance with the 
pressure-temperature curves provides sufficient protection against a 
non-ductile-type fracture of the reactor pressure vessel. No new 
modes of operation are introduced by the proposed change. The 
proposed change will not create any failure mode not bounded by 
previously evaluated accidents. Further, the proposed change does 
not affect any activities or equipment and is not assumed in any 
safety analysis to initiate any accident sequence. This provides 
sufficient assurance that Pilgrim reactor vessel will be operated in 
a manner that will protect it from brittle fracture under all 
operating conditions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. [Does] the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The proposed license amendment requests the use of revised P-T 
curves that are based on established NRC and ASME methodologies. A 
best-estimate calculation of reactor vessel neutron fluence and 
associated uncertainty has been completed for Pilgrim through 34 
EFPY using the NRC approved RAMA methodology. The 34 EFPY reactor 
vessel neutron fluence value was used in conjunction with R.G. 1.99, 
[Revision 2] to compute reference temperature shift, and with ASME 
Section Xl Appendix G to develop revised P-T curves. This provides 
sufficient margin such that the Pilgrim reactor vessel will be 
operated in a manner that will protect it from brittle fracture 
under all operating conditions. Operation within the proposed limits 
ensures that the reactor vessel materials will continue to behave in 
a non-brittle manner, thereby preserving the original safety design 
bases. No plant safetylimits, set points, or design parameters are 
adversely affected by the proposed changes.

[[Page 29676]]

    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Travis C. McCullough, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    Branch Chief: Richard Laufer.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: June 2, 2005.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) reactor coolant system leakage 
detection instrumentation requirements and actions.
    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 relocation is administrative in 
nature and does not involve the modification of any plant equipment 
or affect basic plant operation. The associated instrumentation and 
surveillances are not assumed to be an initiator of any analyzed 
event, nor are these functions assumed in the mitigation of 
consequences of accidents. Additionally, the associated required 
actions for inoperable components do not impact the initiation or 
mitigation of any accident. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The proposed change does not involve any physical 
alteration of plant equipment and does not change the method by 
which any safety-related system performs its function. As such, no 
new or different types of equipment will be installed, and the basic 
operation of installed equipment is unchanged. The methods governing 
plant operation and testing remain consistent with current safety 
analysis assumptions. 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 to relocate current TS 
requirements to the FSAR [Final Safety Analysis Report], consistent 
with regulatory guidance and previously approved changes for other 
stations, are administrative in nature. These changes do not negate 
any existing requirement, and do not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to 
operate in the safety analysis. As such, there are no changes being 
made to safety analysis assumptions, safety limits or safety system 
settings that would adversely affect plant safety as a result of the 
proposed change. Margins of safety are unaffected by requirements 
that are retained, but relocated from the Technical Specifications 
to the FSAR. Additionally, the changes being made to allow 
additional repair time for inoperable instrumentation will not 
affect the required leakage limits, which will continue to be 
monitored at the same required frequency. These compensatory 
measures, operational limitations, and administrative functions that 
will be modified are not credited in any design-basis event and do 
not reflect a margin of safety. 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Travis C. McCullough, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White 
Plains, NY 10601.
    Branch Chief: Richard Laufer.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 
1 and 2, Will County, Illinois.
    Date of amendment request: November 18, 2005.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to adopt NRC-approved Revision 
4 to Technical Specification Task Force (TSTF) Standard Technical 
Specification Change Traveler, TSTF-449, ``Steam Generator Tube 
Integrity.'' The proposed amendment would also include changes to the 
TS definition of Leakage, TS 3.4.13, ``RCS [Reactor Coolant System] 
Operational LEAKAGE,'' TS 5.5.9, ``Steam Generator (SG) Program,'' TS 
5.6.9, Steam Generator Tube Inspection Report,'' and would add TS 
3.4.19, ``Steam Generator (SG) Tube Integrity.'' The proposed changes 
are necessary in order to implement the guidance for the industry 
initiative on Nuclear Energy Institute (NEI) 97-06, ``Steam Generator 
Program Guidelines.''
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments 
adopting TSTF-449, 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 6, 2005 
(70 FR 24126). The licensee affirmed the applicability of the published 
NSHC determination in its application dated November 18, 2005.
    The licensee included a variation from TSTF-449 for Braidwood, Unit 
2 and Byron, Unit 2 in that the proposed amendment would also include 
an effective change to the definition of primary pressure boundary from 
the hot-leg tube end weld to 17 inches below the top of the hot-leg 
tube sheet. The proposed amendment would also delete the current TS 
allowance to use Westinghouse laser welded sleeves as a SG tube repair 
method. The licensee provided an analyses of the NSHC issue in its 
application for the plant-specific variations from TSTF-449.
    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:

    Exelon Generation Company, LLC, (EGC) has reviewed the proposed 
no significant hazards consideration determination published on 
March 2, 2005 (i.e., 70 FR 10298) as part of the consolidated line 
item improvement process (CLIIP) item. EGC has concluded that the 
proposed determination presented in the notice is applicable to 
Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, 
and the determination is hereby incorporated by reference to satisfy 
the requirements of 10 CFR 50.91 (a), except as discussed below.
    The proposed amendment also revises the Technical Specification 
Task Force (TSTF) Standard Technical Specification Change Traveler, 
TSTF-449, ``Steam Generator Tube Integrity,'' Revision 4, version of 
TS 5.5.9, Steam Generator Program, to exclude the portion of the 
tube below 17 inches from the top of the hot leg tubesheet in the 
Braidwood Station, Unit 2, and Byron Station, Unit 2, steam 
generators from TS 5.5.9.d, ``Provisions for SG tube inspections.'' 
This proposed

[[Page 29677]]

license amendment request, in effect, redefines the Braidwood 
Station, Unit 2, and Byron Station, Unit 2, primary pressure 
boundary from the hot leg tube end weld to 17 inches below the top 
of the hot leg tube sheet. This proposed license amendment also 
deletes the current TS 5.5.9.e.6 and TS 5.5.9.e.10 allowance to use 
Westinghouse laser welded sleeves as a SG tube repair method.
    EGC has evaluated whether or not a significant hazards 
consideration is involved with the proposed TS change by focusing on 
the three criteria set forth in 10 CFR 50.92 as discussed below:

Criterion 1.--Does the proposed change involve a significant increase 
in the probability or consequences of an accident previously evaluated?

    Response: No.
    The previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed changes 
that alter the SG inspection criteria and delete the allowance to 
repair SG tubes using Westinghouse laser welded sleeves do not have 
a detrimental impact on the integrity of any plant structure, 
system, or component that initiates an analyzed event. The proposed 
changes will not alter the operation of, or otherwise increase the 
failure probability of any plant equipment that initiates an 
analyzed accident. Therefore, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    Of the applicable accidents previously evaluated, the limiting 
transients with consideration to the proposed changes to the SG tube 
inspection criteria, are the SG tube rupture (SGTR) event and the 
steam line break (SLB) accident.
    During the SGTR event, the required structural integrity margins 
of the SG tubes will be maintained by the presence of the SG 
tubesheet. SG tubes are hydraulically expanded in the tubesheet 
area. Tube rupture in tubes with cracks in the tubesheet is 
precluded by the constraint provided by the tubesheet. This 
constraint results from the hydraulic expansion process, thermal 
expansion mismatch between the tube and tubesheet and from the 
differential pressure between the primary and secondary side. Based 
on this design, the structural margins against burst, discussed in 
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR 
[Pressurized Water Reactor] SG Tubes,'' are maintained for both 
normal and postulated accident conditions.
    The proposed changes do not affect other systems, structures, 
components or operational features. Therefore, the proposed changes 
result in no significant increase in the probability of the 
occurrence of a SGTR accident.
    At normal operating pressures, leakage from primary water stress 
corrosion cracking (PWSCC) below the proposed limited inspection 
depth is limited by both the tube-to-tubesheet crevice and the 
limited crack opening permitted by the tubesheet constraint. 
Consequently, negligible normal operating leakage is expected from 
cracks within the tubesheet region. The consequences of an SGTR 
event are affected by the primary-to-secondary leakage flow during 
the event. Primary-to-secondary leakage flow through a postulated 
broken tube is not affected by the proposed change since the 
tubesheet enhances the tube integrity in the region of the hydraulic 
expansion by precluding tube deformation beyond its initial 
hydraulically expanded outside diameter.
    The probability of a SLB is unaffected by the potential failure 
of a SG tube as this failure is not an initiator for a SLB.
    The consequences of a SLB are also not significantly affected by 
the proposed changes. During a SLB accident, the reduction in 
pressure above the tubesheet on the shell side of the SG creates an 
axially uniformly distributed load on the tubesheet due to the 
reactor coolant system pressure on the underside of the tubesheet. 
The resulting bending action constrains the tubes in the tubesheet 
thereby restricting primary-to-secondary leakage below the midplane.
    Primary-to-secondary leakage from tube degradation in the 
tubesheet area during the limiting accident (i.e., SLB) is limited 
by flow restrictions resulting from the crack and tube-to-tubesheet 
contact pressures that provide a restricted leakage path above the 
indications and also limit the degree of potential crack face 
opening as compared to free span indications. The primary-to-
secondary leak rate during postulated SLB accident conditions would 
be expected to be less than that during normal operation for 
indications near the bottom of the tubesheet (i.e., including 
indications in the tube end welds). This conclusion is based on the 
observation that while the driving pressure causing leakage 
increases by approximately a factor of two, the flow resistance 
associated with an increase in the tube-to-tubesheet contact 
pressure, during a SLB, increases by up to approximately a factor of 
three. While such a leakage decrease is logically expected, the 
postulated accident leak rate could be conservatively bounded by 
twice the normal operating leak rate if the increase in contact 
pressure is ignored. Since normal operating leakage is limited to 
less than 0.104 gpm [gallons per minute] (150 gpd [gallons per day]) 
per TS 3.4.13, ``RCS Operational Leakage,'' the associated accident 
condition leak rate, assuming all leakage to be from lower tubesheet 
indications, would be bounded by approximately 0.2 gpm. This value 
is well within the assumed accident leakage rate of 0.5 gpm 
discussed in Updated Final Safety Analysis Table 15.1-3, 
``Parameters Used in Steam Line Break Analyses.'' Hence it is 
reasonable to omit any consideration of inspection of the tube, tube 
end weld, bulges/overexpansions or other anomalies below 17 inches 
from the top of the hot leg tubesheet. Therefore, the consequences 
of a SLB accident remain unaffected.
    Based on the above discussion, the proposed changes do not 
involve an increase in the consequences of an accident previously 
evaluated.

Criterion 2.--Does the proposed change create the possibility of a new 
or different kind of accident from any accident previously evaluated?

    Response: No.
    The proposed changes do not involve the 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 bases.
    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.

Criterion 3.--Does the proposed change involve a significant reduction 
in a margin of safety?

    Response: No.
    The proposed changes maintain the required structural margins of 
the SG tubes for both normal and accident conditions. Nuclear Energy 
Institute (NEI) 97-06, ``Steam Generator Program Guidelines,'' 
Revision 1 and Regulatory Guide (RG) 1.121, ``Bases for Plugging 
Degraded PWR Steam Generator Tubes,'' are used as the bases in the 
development of the limited hot leg tubesheet inspection depth 
methodology for determining that SG tube integrity considerations 
are maintained within acceptable limits. RG 1.121 describes a method 
acceptable to the NRC for meeting General Design Criteria (GDC) 14, 
``Reactor coolant pressure boundary,'' GDC 15, ``Reactor coolant 
system design,'' GDC 31, ``Fracture prevention of reactor coolant 
pressure boundary,'' and GDC 32, ``Inspection of reactor coolant 
pressure boundary,'' by reducing the probability and consequences of 
a SGTR. RG 1.121 concludes that by determining the limiting safe 
conditions for tube wall degradation the probability and 
consequences of a SGTR are reduced. This RG uses safety factors on 
loads for tube burst that are consistent with the requirements of 
Section III of the American Society of Mechanical Engineers (ASME) 
Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, Westinghouse letter LTR-CDME-
05-32, ``Limited Inspection of the Steam Generator Tube Portion 
Within the Tubesheet at Byron Unit 2 and Braidwood Unit 2,'' 
Revision 2, dated August 2005, defines a length of degradation free 
expanded tubing that provides the necessary resistance to tube 
pullout due to the pressure induced forces, with applicable safety 
factors applied. Application of the limited hot leg tubesheet 
inspection depth criteria will preclude unacceptable primary-to-
secondary leakage during all plant conditions. The methodology for 
determining leakage provides for large margins between calculated 
and actual leakage values in the proposed limited hot leg tubesheet 
inspection depth criteria.
    Therefore, the proposed changes do not involve a significant 
hazards consideration under the criteria set forth in 10 CFR 
50.92(c).

    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 29678]]

review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Brad J. Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Daniel S. Collins.

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 amendment request: January 25, 2006.
    Description of amendment request: The proposed amendment would 
revise the Updated Final Safety Analysis Report (UFSAR) to allow the 
use of automatic load tap changers (LTCs) to operate in automatic mode 
on the reserve auxiliary transformers (RATs) to compensate for 
potential offsite power voltage fluctuations, in order to ensure that 
acceptable voltage is maintained for safety related equipment.
    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 requested change allows the automatic operation mode of the 
LTC. The only accident previously evaluated for which the 
probability is potentially affected by the change is the loss of 
offsite power (LOOP). A failure of the LTC while in automatic 
operation mode that results in decreased voltage to the ESS 
[essential service system] buses could cause a LOOP. This could 
occur in two ways. A failure of the LTC controller that results in 
rapidly decreasing the voltage to the emergency buses is the most 
severe failure mode. However, a backup controller is provided with 
the LTC that makes this failure unlikely. A failure of the LTC 
controller to respond to decreasing grid voltage is less severe, 
since grid voltage changes occur slowly. In both of the above 
potential failure modes, operators will take manual control of the 
LTC to mitigate the effects of the failure. Thus, the probability of 
a LOOP is not significantly increased.
    The proposed change has no effect on the consequences of a LOOP, 
since the emergency diesel generators provide power to safety 
related equipment following a LOOP. The emergency diesel generators 
are not affected by the proposed change.
    The probability of other accidents previously evaluated is not 
affected, since the proposed change does not affect the way plant 
equipment is operated and thus does not contribute to the initiation 
of any of the previously evaluated accidents.
    The LTC is equipped with a backup controller, which controls the 
LTC in the event of primary controller failure. Additionally, 
operator action is available to prevent a sustained high voltage 
condition from occurring. Damage due to over-voltage is time-
dependent. Therefore, damage of safety related equipment is 
extremely unlikely, and the consequences of these accidents are not 
significantly increased. The only way in which the consequences of 
other previously evaluated accidents could be affected is if a 
failure of the LTC, while in automatic operation mode, led to a 
sustained high voltage condition, which resulted in damage to safety 
related equipment that is used to mitigate an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change involves functions that provide offsite 
power to safety related equipment for accident mitigation. Thus, the 
proposed change potentially affects the consequences of previously 
evaluated accidents (as addressed in Question 1), but does not 
result in any new mechanisms that could initiate damage to the 
reactor and its principal safety barriers (i.e., fuel cladding, 
reactor coolant system, or primary containment).
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not affect the inputs or assumptions of 
any of the analyses that demonstrate the integrity of the fuel 
cladding, reactor coolant system, or containment during accident 
conditions. The allowable values for the degraded voltage protection 
function are unchanged and will continue to ensure that the degraded 
voltage protection function actuates when required, but does not 
actuate prematurely to cause a LOOP. Automatic operation of the LTC 
increases margin by reducing the potential for transferring to the 
EDGs [emergency diesel generators] during an event.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelong Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Daniel S. Collins.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry 
Nuclear Power Plant, Unit 1, Lake County, Ohio

    Date of amendment request: February 10, 2006.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 3.3.5.1, ``Emergency Core Cooling 
System (ECCS) Instrumentation,'' to correct a Perry Nuclear Power Plant 
(PNPP)-specific issue and establish consistency with the improved 
standard technical specifications (ISTS). Specifically, Sub-actions 
B.1.2.1 and B.1.2.2, which were added into PNPP TS 3.3.5.1 during the 
ISTS conversion process, will be deleted. PNPP Required Action B.1 will 
then match the ISTS Required Action B.1. As a result, actions with a 1-
hour completion time will only be required for the annulus exhaust gas 
treatment (AEGT) system if a loss of initiation capability in both 
divisions actually exists for an AEGT initiation function, as 
originally intended.
    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.
    There are no physical modifications being made to any plant 
system or component. The only change is to a Required Action within 
the Technical Specifications. The revised Technical Specification 
requirements do not impact initiators of previously evaluated 
accidents or transients.
    The specification being revised is associated with a system used 
to mitigate the consequences of accidents. The change does not 
affect how the AEGT system is controlled, operated, or tested. The 
intent of Required Action B.1 for the ECCS Instrumentation, 
specifically, a loss of initiation capability check, is maintained 
by the changes being proposed. The wording of Required Action B.1 
ensures appropriate actions are taken when a loss of initiation 
capability exists, by declaring the supported systems inoperable. 
This action is consistent with the current requirements.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no physical modifications being made to any plant 
system or component, and

[[Page 29679]]

the proposed change introduces no new method of operation for the 
plant, or its systems or components. Therefore, 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 change to the ECCS Instrumentation Required Action continues 
to ensure that a check is performed to determine if one or more of 
the ECCS Instrumentation Functions has lost its capability to 
actuate the Division 1 and 2 low-pressure ECCS, the AEGT subsystems, 
and the associated diesel generators. It continues to direct 
appropriate actions if such a loss of initiation capability is 
found. Therefore, the necessary function of the Technical 
Specification requirements is maintained, and the proposed changes 
do not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Daniel S. Collins.

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

    Date of amendment request: February 16, 2006.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) requirements related to steam 
generator (SG) tube integrity. The change is consistent with NRC-
approved Revision 4 to Technical Specification Task Force (TSTF) 
Standard Technical Specification Change Traveler, TSTF-449, ``Steam 
Generator Tube Integrity.'' The availability of this TS improvement was 
announced in the Federal Register on May 6, 2005 (70 FR 24126) as part 
of the consolidated line item improvement process (CLIIP).
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on March 2, 
2005 (70 FR 10298) as part of the CLIIP. The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
February 16, 2006.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1.--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change requires a SG Program that includes 
performance criteria that will provide reasonable assurance that the 
SG tubing will retain integrity over the full range of operating 
conditions (including startup, operation in the power range, hot 
standby, cooldown and all anticipated transients included in the 
design specification). The SG performance criteria are based on tube 
structural integrity, accident induced leakage, and operational 
LEAKAGE.
    A SGTR [steam generator tube rupture] event is one of the design 
basis accidents that are analyzed as part of a plant's licensing 
basis. In the analysis of a SGTR event, a bounding primary to 
secondary LEAKAGE rate equal to the operational LEAKAGE rate limits 
in the licensing basis plus the LEAKAGE rate associated with a 
double-ended rupture of a single tube is assumed.
    For other design basis accidents such as MSLB [main steamline 
break], rod ejection, and reactor coolant pump locked rotor the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). These analyses typically assume that 
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or 
increases to 1 gallon per minute as a result of accident induced 
stresses. The accident induced leakage criterion introduced by the 
proposed changes accounts for tubes that may leak during design 
basis accidents. The accident induced leakage criterion limits this 
leakage to no more than the value assumed in the accident analysis.
    The SG performance criteria proposed change to the TS identify 
the standards against which tube integrity is to be measured. 
Meeting the performance criteria provides reasonable assurance that 
the SG tubing will remain capable of fulfilling its specific safety 
function of maintaining reactor coolant pressure boundary integrity 
throughout each operating cycle and in the unlikely event of a 
design basis accident. The performance criteria are only a part of 
the SG Program required by the proposed change to the TS. The 
program, defined by NEI [Nuclear Energy Institute] 97-06, Steam 
Generator Program Guidelines, includes a framework that incorporates 
a balance of prevention, inspection, evaluation, repair, and leakage 
monitoring. The proposed changes do not, therefore, significantly 
increase the probability of an accident previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT I-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT I-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The typical analysis of the limiting design basis 
accident assumes that primary to secondary leak rate after the 
accident is 1 gallon per minute with no more than [500 gallons per 
day or 720 gallons per day] in any one SG, and that the reactor 
coolant activity levels of DOSE EQUIVALENT I-131 are at the TS 
values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design basis 
accident and is an improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of a SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.

Criterion 2.--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed performance based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed SG Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the SG Program will be an 
enhancement of SG tube performance. Primary to secondary LEAKAGE 
that may be experienced during all plant conditions will be 
monitored to ensure it remains within current accident analysis 
assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    Criterion 3.--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change

[[Page 29680]]

does not affect tube design or operating environment. The proposed 
change is expected to result in an improvement in the tube integrity 
by implementing the SG Program to manage SG tube inspection, 
assessment, repair, and plugging. The requirements established by 
the SG Program are consistent with those in the applicable design 
codes and standards and are an improvement over the requirements in 
the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: July 9, 2004 (TS-436).
    Description of amendment request: The proposed amendment would 
revise Technical Specification Surveillance Requirement 3.6.1.3.10 to 
increase the allowed main steam isolation valve (MSIV) leak rate from 
11.5 standard cubic feet per hour (scfh) per valve, to 100 scfh for 
individual MSIVs with a 150 scfh combined leakage for all four main 
steam lines.
    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.
    TVA proposes to utilize the main steam drain lines to 
preferentially direct MSIV leakage to the main condenser. This drain 
path takes advantage of the large volume of the steam lines and 
condenser to provide holdup and plate-out of fission products that 
may leak through the closed MSIVs. In this approach, the main steam 
lines, steam drain piping, and the main condenser are used to 
mitigate the consequences of an accident to limit potential doses 
below the limits prescribed in 10 CFR 50.67(b)(2)(i) for the 
exclusion area, 10 CFR 50.67(b)(2)(ii) for the low population zone, 
and in 10 CFR 50.67(b)(2)(iii) for control room personnel.
    Seismic verification walkdowns and evaluations of bounding 
piping/supports were performed to demonstrate the main steam line 
piping and components that comprise the Alternate Leakage Treatment 
(ALT) path were rugged and able to perform the safety function of 
MSIV leakage control following a Design Basis Earthquake (DBE). 
Thus, it has been concluded the components in the MSIV alternate 
leakage treatment flow path can be relied upon to maintain 
structural integrity.
    Therefore, the proposed amendment does not involve changes to 
structures, components, or systems which would affect the 
probability of an accident previously evaluated in the Browns Ferry 
Updated Final Safety Analysis Report (UFSAR).
    A plant-specific radiological analysis has been performed to 
assess the effects of the proposed increase in MSIV leakage 
acceptance criteria in terms of off-site doses and main control room 
dose. The analysis shows the dose contribution from the proposed 
increase in leakage acceptance criteria is acceptable compared to 
doses limits prescribed in 10 CFR 50.67(b)(2)(i) for the exclusion 
area, 10 CFR 50.67(b)(2)(ii) for the low population zone, and in 10 
CFR 50.67(b)(2)(iii) for control room personnel.
    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 require the use of the main steam piping 
and the condenser to process MSIV leakage. This additional function 
does not compromise the reliability of these systems. They will 
continue to function as intended and not be subject to a failure of 
a different kind than previously considered. In addition, MSIV 
functionality will not be adversely impacted by the increased 
leakage limit. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to Surveillance Requirement 3.6.1.3.10, to 
increase the allowable MSIV leakage, does not involve a significant 
reduction in the margin of safety. The allowable leak rate specified 
for the MSIVs is used to quantify a maximum amount of leakage 
assumed to bypass containment. The results of the re-analysis 
supporting these changes were evaluated against the dose limits 
contained in 10 CFR 50.67(b)(2)(i) for the exclusion area, 10 CFR 
50.67(b)(2)(ii) for the low population zone, and in 10 CFR 
50.67(b)(2)(iii) for control room personnel. Sufficient margin 
relative to the regulatory limits is maintained even when 
conservative assumptions and methods are utilized. 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Michael L. Marshall, Jr.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of amendment request: August 16, 2004 (TS--447).
    Description of amendment request: The proposed amendment would 
extend the channel calibration frequency requirements for 
instrumentation in the high pressure coolant injection, reactor core 
isolation cooling, and reactor water core isolation cooling systems.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes extend the channel calibration surveillance 
frequency of instrumentation used for the high area temperature 
isolation of the high pressure coolant injection (HPCI), reactor 
core isolation cooling (RCIC), and the reactor water clean-up (RWCU) 
systems. The allowable trip point value for three sets of RCIC 
instruments on each unit and for two sets of RWCU instruments on 
Unit 1 are also revised. The calibration surveillance frequency is 
extended to 24 months from 92 days (for the HPCI and RCIC high area 
temperature instrumentation) and from 122 days (for the RWCU high 
area temperature instrumentation). Under certain circumstances, 
Technical Specifications (TS) SR [Surveillance Requirement] 3.0.2 
would allow a maximum surveillance interval of 30 months for an SR 
having a nominal 24-month performance frequency. Instrumentation 
scaling and setpoint calculations performed in accordance with the 
guidelines of Generic Letter 91-04 have shown that the reliability 
of the affected protection instrumentation will be preserved for the 
maximum allowable calibration surveillance interval. The Unit 1 
instrumentation will be physically modified to be essentially 
identical to that installed on Unit 2 and Unit 3 prior to restart of 
Unit 1. Therefore, the proposed change does not involve a 
significant increase in the

[[Page 29681]]

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 extend the channel calibration surveillance 
frequency of instrumentation used for the high area temperature 
isolation of the high pressure coolant injection (HPCI), reactor 
core isolation cooling (RCIC), and the reactor water clean-up (RWCU) 
systems. The allowable trip point value for three sets of RCIC 
instruments on each unit and for two sets of RWCU instruments on 
Unit 1 are also revised. The instrumentation will function in the 
same way following the amendment as it functions currently. Hence, 
the changes do not create the possibility of any new failure 
mechanisms. Note that the Unit 1 instrumentation will be modified to 
be essentially identical to that installed on Unit 2 and Unit 3 
prior to restart of Unit 1. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes extend the channel calibration surveillance 
frequency of instrumentation used for the high area temperature 
isolation of the high pressure coolant injection (HPCI), reactor 
core isolation cooling (RCIC), and the reactor water clean-up (RWCU) 
systems. The allowable trip point value for three sets of RCIC 
instruments on each unit and for two sets of RWCU instruments on 
Unit 1 are also revised. Instrumentation scaling and setpoint 
calculations performed in accordance with the guidelines of Generic 
Letter 91-04 have shown safety margins are preserved with the 
extended surveillance frequency and the revised TS allowable values. 
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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: May 8, 2006 (TS-06-09).
    Description of amendment request: The proposed amendment would 
revise the limiting condition for operation for Technical Specification 
(TS) Section 3.7.9, ``Ultimate Heat Sink.'' The maximum essential raw 
cooling water (ERCW) temperature limit associated with Surveillance 
Requirement 3.7.9.1 would increase from 85 degrees Fahrenheit ([deg]F) 
to 88 [deg]F. This proposed change is based on evaluations of the ERCW 
system and the ultimate heat sink (UHS) functions and maximum 
temperatures that will satisfy the associated safety functions.
    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 increase the UHS maximum temperature will 
not adversely alter the function, design, or operating practices for 
plant systems or components. The UHS is utilized to remove heat 
loads from plant systems during normal and accident conditions. This 
function is not expected or postulated to result in the generation 
of any accident and continues to adequately satisfy the associated 
safety functions with the proposed changes. Therefore, the 
probability of an accident presently evaluated in the safety 
analyses will not be increased. The heat loads, that the UHS is 
designed to accommodate, have been evaluated with the higher 
temperature limit. The result of these evaluations is that there is 
existing margin associated with the systems that utilize the UHS for 
normal and accident conditions. These margins are sufficient to 
accommodate the postulated normal and accident heat loads with the 
proposed changes to the UHS. Since the safety functions of the UHS 
are maintained, the systems that ensure acceptable offsite dose 
consequences will continue to operate as designed. The change in the 
maximum calculated containment pressure associated with the design 
basis loss-of-coolant-accident (LOCA) remains below the American 
Society of Mechanical Engineers (ASME) Code design internal 
pressure. Therefore, the consequence of any accident will be the 
same as those previously analyzed.
    Since the UHS safety function will continue to meet accident 
mitigation requirements and limit dose consequences to acceptable 
levels, TVA has concluded that the proposed TS change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The UHS function provides accident mitigation capabilities and 
serves as a heat sink for normal and upset plant conditions; the UHS 
is not an initiator of any accident. By allowing the proposed change 
in the UHS temperature requirements, only the parameters for UHS 
operation are changed while the safety functions of the UHS and 
systems that transfer the heat sink capability continue to be 
maintained. The proposed change does not impact the response of the 
systems and components assumed in the safety analysis. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any accident evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change has been evaluated for systems that are 
needed to support accident mitigation functions as well as normal 
operational evolutions. Operational margins were found to exist in 
the systems that utilize the UHS capabilities such that these 
proposed changes will not result in the loss of any safety function 
necessary for normal or accident conditions. The ERCW system has 
excess flow capacity that will accommodate the increased flows 
necessary for the proposed temperature increase. While operating 
margins have been reduced by the proposed changes, safety margins 
have been maintained as assumed in the accident analyses for 
postulated events. The proposed change results in an increase in the 
maximum calculated containment peak pressure. However, the change in 
the maximum calculated containment peak pressure associated with the 
design basis LOCA is a small percentage of the margin between the 
current maximum calculated containment peak pressure and the ASME 
Code design internal pressure. This aspect of the proposed change 
does not involve a significant reduction in a margin of safety. 
Additionally, the proposed changes do not require any further 
modification of component setpoints or operating provisions that are 
necessary to maintain margins of safety established by the WBN 
design (the shutdown board room chillers were physically modified to 
operate properly at the 88 degree F UHS temperature). 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Michael L. Marshall, Jr.

[[Page 29682]]

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

    Date of amendment request: April 14, 2005, as supplemented by 
letter dated December 21, 2005.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) by (1) adding a new TS 3.1.9, ``RCS 
[Reactor Coolant System] Boron Limitations <500 [deg]F,'' and (2) 
revising TS 3.3.1, ``Reactor Trip System (RTS) Instrumentation.''
    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.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since there are 
no design changes. The design of the reactor trip system (RTS) 
instrumentation and engineered safety feature actuation system 
(ESFAS) instrumentation will be unaffected and these protection 
systems will continue to function in a manner consistent with the 
plant design basis. All design, material, and construction standards 
that were applicable prior to this amendment request will be 
maintained.
    The proposed changes will not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained other than extending the 
OPERABILITY requirements for RTS trip Function 2.b (Power Range 
Neutron Flux--Low) to the upper portion of MODE 3. The proposed 
changes will not alter or prevent the ability of structures, 
systems, and components (SSCs) from performing their intended 
functions to mitigate the consequences of an initiating event within 
the assumed acceptance limits.
    As discussed previously [in the application,] the proposed 
change[s] will add more restrictive requirements in the form of a 
new LCO [limiting condition for operation] 3.1.9 and an expanded LCO 
Applicability for RTS trip Function 2.b, Power Range Neutron Flux--
Low, to provide mitigative capability in the event of an 
uncontrolled RCCA [rod cluster control assembly] bank withdrawal 
event postulated to occur during low power or subcritical (startup) 
conditions.
    There will be no change[s] to normal plant operating parameters 
or accident mitigation performance. None of the proposed changes 
will initiate any accidents; therefore, the probability of an 
accident will not be increased. There will be no degradation in the 
performance of, nor an increase in the number of challenges imposed 
on, safety-related equipment assumed to function during an accident 
situation.
    All accident analysis acceptance criteria will continue to be 
met with the proposed changes. The proposed changes will not affect 
the source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of an 
accident previously evaluated. The proposed changes will not alter 
any assumptions or change any mitigation actions in the radiological 
consequence evaluations in the FSAR [Final Safety Analysis Report 
for Callaway]. The applicable radiological dose acceptance criteria 
will continue to be met.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    There are no proposed design changes nor are there any changes 
in the method by which any safety-related plant SSC performs its 
safety function. [These changes] will not affect the normal method 
of plant operation or change any operating parameters. No equipment 
performance requirements will be affected other than the more 
restrictive Applicability requirements being imposed on RTS trip 
Function 2.b, Power Range Neutron Flux--Low, in the upper portion of 
MODE 3. The proposed changes will not alter any assumptions made in 
the safety analyses.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of this amendment. There will be no adverse effect or 
challenges imposed on any safety-related system as a result of this 
amendment.
    The proposed amendment will not alter the design or performance 
of the 7300 Process Protection System, Nuclear Instrumentation 
System, or Solid State Protection System used in the plant 
protection systems.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings 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, departure from nucleate boiling ratio (DNBR) limits, heat 
flux hot channel factor (FQ), nuclear enthalpy rise hot 
channel factor (F[Delta]H), loss of coolant accident peak cladding 
temperature (LOCA PCT), peak local power density, or any other 
margin of safety. The applicable radiological dose consequence 
acceptance criteria will continue to be met.
    The proposed changes do not eliminate any RTS or ESFAS 
surveillances or alter the Frequency of surveillances required by 
the Technical Specifications. More restrictive changes are proposed 
by virtue of a new LCO 3.1.9 on [RCS] boron requirements when the 
RCS temperature is below 500 [deg]F and by virtue of extending the 
Applicability of RTS trip Function 2.b, Power Range Neutron Flux--
Low, to the upper portion of MODE 3. The nominal RTS and ESFAS trip 
setpoints will remain unchanged. None of the acceptance criteria for 
any accident analysis will be changed.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

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

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

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

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

    Date of amendment request: April 20, 2006.
    Brief description of amendment request: The proposed amendments 
would reinstate the previous reactor coolant system pressure and 
temperature limits, low temperature overpressure protection system 
(LTOPS) setpoint, and (LTOPS) enable temperature basis that were 
approved by the NRC staff on December 28, 1995, as License Amendments 
Nos. 207 and 207 for Surry 1 and 2.
    Date of publication of individual notice in Federal Register: April 
28, 2006 (71 FR 25249)
    Expiration date of individual notice: 30 day expiration date, May 
30, 2006,

[[Page 29683]]

and 60 day expiration date, June 27, 2006.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendments: June 20, 2005.
    Brief Description of amendments: The amendments revise the 
Technical Specification (TS) Surveillance Requirement 3.6.1.6.2 of 
3.6.1.6, ``Suppression Chamber-to-Drywell Vacuum Breakers'' for the 
frequency of functionally testing the suppression chamber-to-drywell 
vacuum breakers.
    Date of issuance: May 5, 2006.
    Effective date: May 5, 2006.
    Amendment Nos.: 240 and 268.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the TS.
    Date of initial notice in Federal Register: August 16, 2005 (70 FR 
48202).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 5, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 18, 2005, as supplemented 
by letter dated February 15, 2006.
    Brief description of amendment: This amendment authorizes the use 
of fire-resistive electrical cables in lieu of the alternatives 
specified in Section C5.b.2 of Branch Technical Position Chemical 
Engineering Branch 9.5-1 (NUREG-0800), `` Guidelines for Fire 
Protection for Nuclear Power Plants,'' dated July 1981, for Fire Areas 
12-A-CR, 1-A-CSRA, 1-A-CSRB, 1-A-SWGRA, 1-A-SWGRB, and 1-A-BAL-B.
    Date of issuance: May 1, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No. 123.
    Facility Operating License No. NPF-63: Amendment revises the 
License.
    Date of initial notice in Federal Register: November 8, 2005.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 1, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 14, 2005, as supplemented 
January 11, 2006.
    Brief description of amendment: The proposed change modifies the 
Millstone Power Station, Unit No. 2 reactor coolant system heatup and 
cooldown limits Technical Specification (TS) 3.4.9.1, ``Reactor Coolant 
System''. The associated TS bases will be updated to address the 
proposed change.
    Date of issuance: May 3, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 292.
    Facility Operating License No. DPR-65: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: August 30, 2005 (70 FR 
51379). The supplement dated January 11, 2006, provided clarifying 
information that did not change the scope of the proposed amendment as 
described in the original notice, and did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 3, 2006.
    No significant hazards consideration comments received: No.

Duke Power Company, LLC Docket No. 50-287, Oconee Nuclear Station, Unit 
3, Oconee County, South Carolina

    Date of application of amendment: August 18, 2005, supplemented 
September 15, 2005, and January 5 and April 6, 2006.
    Brief description of amendment: The amendment revised Technical 
Specifications 3.5.2.6 and 3.5.3.6 to accommodate the replacement of 
the reactor building emergency sump suction inlet trash racks and 
screens with strainers. Similar amendments were issued for Units 1 and 
2 on November 1, 2005; however, the amendment for Unit 3 was not issued 
at that time since the licensee had not completed its evaluation of the 
impact of pipe whip, jet impingement and internally generated missiles 
for Unit 3.
    Date of Issuance: May 4, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 350.
    Renewed Facility Operating License No. DPR-55: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: August 31, 2005 (70 FR 
51852).
    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 initial Federal

[[Page 29684]]

Register notice. The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated May 4, 2006.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania

    Date of application for amendments: June 24, 2004.
    Brief description of amendments: These amendments implement 25 
generic Technical Specification (TS) changes previously approved by the 
NRC staff as part of the Technical Specifications Task Force (TSTF). 
The TSTF change travelers and proposed changes are:
    1. TSTF-5, an administrative change to TS 2.2 to remove reporting 
requirements that are already in the regulations 10 CFR, Sections 50.36 
and 50.73;
    2. TSTF-208, an extension of the time allowed to reach MODE 2 once 
a TS 3.0.3 condition is identified, from the current 7 hours to 10 
hours;
    3. TSTFs-222 and 229, changes to TS 3.1.4 to allow scram time 
testing on only affected rods when an outage is short and only a 
limited number of fuel assemblies are moved and to require the Minimum 
Critical Power Ratio to be determined after scram time testing;
    4. TSTFs-297 and 227, changes to TSs 3.3.2.2, 3.3.4.1, and 3.3.4.2 
to allow reactor feedwater pumps and main turbine valves to be removed 
from service if their trip function is compromised;
    5. TSTF-295, a clarification in Table 3.3.3.1-1 that penetration 
flow paths, not just valve positions, are to be considered;
    6. TSTF-275, a clarification Table 3.3.5.1-1 that certain emergency 
core cooling system (ECCS) instrumentation needs to be operable when 
ECCS and ECCS support systems are required to be operable;
    7. TSTF-306, changes to TS 3.3.6.1 to allow penetration flow paths 
to be opened intermittently under administrative controls and to set 
apart the Traversing In-core Probe system isolation as a separate 
function;
    8. TSTF-416, changes to TSs 3.5.1 and 3.5.2 to allow the low 
pressure coolant injection subsystems to be considered operable during 
alignment and operation in the decay heat removal mode;
    9. TSTF-17, a change to TS 3.6.1.2 to extend the containment air 
lock interlock mechanism testing frequency from 6 months to 2 years to 
coincide with refueling outage frequency;
    10. TSTFs-30, 323, 45, 46, and 269, changes to TSs 3.6.1.3 and 
3.6.4.2 related to primary and secondary containment isolation valve 
completion times, isolation times, and status verification;
    11. TSTF-322, a clarification in TS 3.6.4.1 of the intent of 
secondary containment drawdown tests;
    12. TSTF-276, Revision 2, a change to TS 3.8.1 to allow certain 
emergency diesel generator (EDG) testing to continue even if the stated 
power factor cannot be attained;
    13. TSTF-404, a change to TS 3.1.8 to revise required actions when 
one valve is inoperable in one or more scram discharge volume vent and 
drain lines, as part of the consolidated line item improvement process;
    14. TSTF-65 Revision 1, a change to allow the use of generic 
organizational titles in the TSs, as opposed to plant-specific titles;
    15. TSTF-299, a clarification in TS 5.2.2 of the intent of 
refueling cycle intervals with respect to system leak test 
requirements;
    16. TSTF-279, a deletion in TS 5.5.6 of the reference to 
``applicable supports'' as part of the description of the Inservice 
Testing Program;
    17. TSTF-118, a change to TS 5.5.9 to apply the provisions of 
Surveillance Requirement (SR) 3.0.2 (25% extension interval) and SR 
3.0.3 (missed surveillance actions) to EDG fuel oil testing 
surveillances;
    18. TSTF-106, Revision 1, a clarification in TS 5.5.9 that the 
American Society for Testing and Materials standard for EDG fuel oil 
applies only to new fuel being received; and
    19. TSTF-152, a change to the Radioactive Effluent Release Report 
to ensure that a common report for both units combines sections common 
to both units.
    Date of issuance: May 10, 2006.
    Effective date: As of the date of issuance, to be implemented 
within 90 days.
    Amendments Nos.: 259 and 262.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the TSs.
    Date of initial notice in Federal Register: September 28, 2004 (69 
FR 57985) and October 26, 2004 (69 FR 62476).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 10, 2006.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 19, 2006.
    Description of amendment request: The amendment deletes Technical 
Specification (TS) 6.8.1.2a, ``Occupational Radiation Exposure Report 
[ORER],'' TS 6.8.1.2.c, regarding challenges to pressurizer relief and 
safety valves and TS 6.8.1.5, ``Monthly Operating Report [MOR],'' as 
described in the Notice of Availability published in the Federal 
Register on June 23, 2004 (69 FR 35067).
    Date of issuance: May 5, 2006.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 109.
    Facility Operating License No. NPF-86: The amendment revised the 
TSs.
    Date of initial notice in Federal Register: February 14, 2006 (71 
FR 7808).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 5, 2006.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 12, 2005.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Section 3.4.9, ``RCS [reactor coolant system] 
Pressure and Temperature (P/T) Limits,'' curves 3.4.9-1, ``Pressure/
Temperature Limits for Non-Nuclear Heatup or Cooldown Following Nuclear 
Shutdown,'' 3.4.9-2, ``Pressure/Temperature Limits for Inservice 
Hydrostatic and Inservice Leakage Tests, and 3.4.9-3, ``Pressure/
Temperature Limits for Criticality,'' to remove the cycle operating 
restriction and replace it with a limitation of 30 effective full-power 
years (EFPY).
    Date of issuance: April 27, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 219.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 3, 2006 (71 FR 
150).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 27, 2006.
    No significant hazards consideration comments received: No.

[[Page 29685]]

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

    Date of amendment request: December 30, 2005.
    Brief description of amendment: The amendment established 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 leakage rate limit.
    Date of issuance: May 2, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 220.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 2006 (71 
FR 10073)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 2, 2006.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: January 30, 2006.
    Brief description of amendment: The amendment 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.
    Date of issuance: May 2, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 221.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 2006 (71 
FR 10074).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 2, 2006.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear 
Generating Plant (MNGP), Wright County, Minnesota

    Date of application for amendment: April 29, 2004, as supplemented 
on November 23, 2004; January 20, February 28, April 12, 2005; and 
March 10, 2006.
    Brief description of amendment: The amendment revised the MNGP 
licensing basis by selectively implementing the alternative source term 
for the postulated fuel handling accident, leading to revision of 
portions of the Technical Specifications to reflect this change in 
licensing basis.
    Date of issuance: April 24, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 145.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2891)
    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 amendment is contained 
in a Safety Evaluation dated April 24, 2006.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: August 11, 2005.
    Brief description of amendment: The 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.
    Date of issuance: March 1, 2006.
    Effective date: As of its date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 238.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised 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 March 1, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 19, 2005.
    Brief description of amendments: The 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.
    Date of issuance: March 7, 2006.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: Unit 1--185; Unit 2--187
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 2005 (70 
FR 75495).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 7, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 9, 2004, as 
supplemented by letters dated November 18 and December 5, 2005.
    Brief description of amendment: The amendment authorizes 
modification to

[[Page 29686]]

the Updated Final Safety Analysis Report (UFSAR) to include a revision 
to the methodology for splicing reinforcing steel bars during 
restoration of the Unit 1 concrete shield building dome as part of the 
steam generator replacement project.
    Date of issuance: April 27, 2006.
    Effective date: As of the date of issuance and shall be implemented 
as part of the next UFSAR update made in accordance with 10 CFR 
50.71(e).
    Amendment No. 60.
    Facility Operating License No. NPF-90: Amendment authorizes 
revision of the Updated Final Safety Analysis Report.
    Date of initial notice in the Federal Register: January 4, 2005 (70 
FR 405). The supplemental letters provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 27, 2006.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 30, 2003, as supplemented by 
letters dated August 31 and November 18, 2005, and March 6, 2006.
    Brief description of amendment: The amendment increases the 
completion times (CTs) for Technical Specification (TS) 3.8.1, ``AC 
Sources--Operating,'' and adds requirements on the diesel generators at 
the Sharpe Station when a diesel generator at Wolf Creek Generating 
Station is in an extended CT greater than 72 hours. The proposed 
changes to TS 3.8.9, ``Distribution Systems--Operating,'' are 
withdrawn. The amendment also revises a page in the license and adds 
conditions to Appendix D, ``Additional Conditions,'' of the license.
    Date of issuance: April 26, 2006.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days of the date of issuance.
    Amendment No.: 163.
    Facility Operating License No. NPF-42. The amendment revised the 
license including Appendix D, ``Additional Conditions,'' and Appendix 
A, ``Technical Specifications.''
    Date of initial notice in Federal Register: January 6, 2004 (69 FR 
700).
    The supplemental letters dated August 31 and November 18, 2005, and 
March 2, 2006, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the NRC staff's original proposed no 
significant hazards consideration determination published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 26, 2006.
    No significant hazards consideration comments received: No

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

    Date of amendment request: November 3, 2005, and supplemental 
letters dated February 21 and March 28, 2006.
    Brief description of amendment: The amendment revised the Technical 
Specifications associated with steam generator tube integrity 
consistent with Revision 4 to Technical Specification Task Force (TSTF) 
Standard Technical Specification Change Traveler, TSTF-449, ``Steam 
Generator Tube Integrity.'' A notice of availability for this TS 
improvement using the consolidated line item improvement process was 
published in the Federal Register on May 6, 2005 (70 FR 24126).
    Date of issuance: May 8, 2006.
    Effective date: The license amendment is effective as of its date 
of issuance and shall be implemented prior to the entry into Mode 5 in 
the restart from Refueling Outage 15, which is scheduled to begin in 
October 2006.
    Amendment No.: 164.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 6, 2005 (70 FR 
72676) The supplemental letters dated February 21 and March 28, 2006, 
provided additional clarifying information, 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 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 8, 2006.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 15th day of May 2006.

    For the Nuclear Regulatory Commission
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 06-4736 Filed 5-22-06; 8:45 am]
BILLING CODE 7590-01-P