Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 29671-29686 [06-4736]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 99 (Tuesday, May 23, 2006)]
[Notices]
[Pages 29671-29686]
From the Federal Register Online via the Government Printing 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, https://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for leave to intervene is filed within 60 days, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, any hearing held would take place before the 
issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-

[[Page 29673]]

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

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 st