[Federal Register Volume 71, Number 157 (Tuesday, August 15, 2006)]
[Notices]
[Pages 46929-46946]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-6921]


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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 July 21, 2006, to August 3, 2006. The last 
biweekly notice was published on August 1, 2006 (71 FR 43528).

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

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

[[Page 46930]]

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-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].

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

    Date of amendment request: June 30, 2006.
    Description of amendment request: The proposed change would revise 
the Note preceding Technical Specification (TS) Surveillance 
Requirement (SR) 3.4.6.1 to be consistent with the wording in NUREG-
1434, ``Standard Technical Specifications General Electric Plants, BWR/
6,'' Revision 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment revises the note associated with TS SR 
3.4.6.1, which requires verification that the leakage past the 
Reactor Coolant System (RCS) Pressure Isolation Valves (PIVs) is 
less than a specified limit. The proposed revision provides 
clarification that performance of this SR is allowed during plant 
shutdown (i.e., a Mode other than Modes 1 and 2).
    The proposed change does not require modification to the 
facility. The proposed change does not affect the operation of any 
facility equipment, the interface between facility systems, or the 
reliability of any equipment. In addition, the proposed change does 
not alter the requirement to perform the leakage testing of the RCS 
PIVs and does not revise the leakage limits associated with this SR. 
The function of the RCS PIVs is to separate the high pressure RCS 
from an attached low pressure system. Periodic testing of PIVs can 
substantially reduce intersystem Loss of Coolant Accident (LOCA) 
probability. Since the proposed change does not alter the method or 
limits associated with the leak rate testing of the RCS PIVs there 
is no significant increase in the probability of a LOCA. Therefore, 
the proposed amendment does not involve a significant increase in 
the

[[Page 46931]]

probability of an accident previously evaluated.
    The consequences of a previously analyzed event are dependent on 
the initial conditions assumed in the analysis, the availability and 
successful functioning of equipment assumed to operate in response 
to the analyzed event, and the setpoints at which these actions are 
initiated. The method for performing the leakage testing of the RCS 
PIVs and the specified leakage limit for this testing will not 
change as a result of the proposed revision and, therefore, there is 
no change in the consequences associated with the LOCA analysis. The 
radiological consequences remain within applicable regulatory 
limits. The proposed change does not alter any system's performance 
measures or the ability to perform its accident mitigation 
functions. The radiological consequences associated with any 
previously evaluated accident do not change as a result of the 
proposed revision. Therefore, the proposed change does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Based on the above, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change to the wording of the Note to TS SR 3.4.6.1 
clarifies the plant conditions for when the surveillance is required 
to be performed. The proposed change does not affect the design, 
functional performance or operation of the facility. No new 
equipment is being introduced and installed equipment is not being 
operated in a new or different manner. Similarly, the proposed 
change does not affect the design or operation of any structures, 
systems or components involved in the mitigation of any accidents, 
nor does it affect the design or operation of any component in the 
facility such that new equipment failure modes are created. There 
are no setpoints at which protective or mitigative actions are 
initiated that are affected by this proposed action. No change is 
being made to procedures relied upon to respond to an off-normal 
event.
    As such the proposed amendment will 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.
    Margins of safety are established in the design of components, 
the configuration of components to meet certain performance 
parameters, and in the establishment of setpoints to initiate alarms 
or actions. The proposed change revises a note associated with a 
surveillance requirement to clarify the plant conditions for when 
the surveillance needs to be performed. This change involves an 
administrative clarification to reflect the original intent of the 
TS. The equipment will continue to be tested in a manner and at a 
frequency necessary to provide confidence that the equipment can 
perform its intended safety function. There is no change in the 
design of the affected systems, no alteration of the setpoints at 
which alarms or actions are initiated, and no change in plant 
configuration from original design. There is no impact on the plant 
safety analyses.
    Therefore, operation of CPS in accordance with the proposed 
change will not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Daniel S. Collins.

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

    Date of amendment request: June 14, 2006.
    Description of amendment request: The proposed change will delete 
Waterford 3 Technical Specification (TS) Surveillance Requirement (SR) 
4.8.1.1.2.f. This SR requires that the emergency diesel generator be 
subjected to an inspection in accordance with procedures prepared in 
conjunction with its manufacturer's recommendations.
    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 ability of the emergency diesel generator to perform its 
safety function is not proven by the performance of the 
manufacturer's recommended inspections. The inspections are not 
considered an initiator or mitigating factor in any previously 
evaluated accidents.
    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 results in the deletion of the SR associated 
with the performance of manufacturer's inspections. No modifications 
to plant structures, systems, or components, or changes in the 
design of the plant structures, systems, or components are required 
to support the proposed TS change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The ability of the emergency diesel generator to perform its 
safety function is not proven by the performance of the 
manufacturer's recommended inspections. Inspection activities will 
continue to be performed.
    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: N.S. Reynolds, Esquire, Winston & Strawn, 
1700 K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: David Terao.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois

    Date of amendment request: June 2, 2006.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Surveillance Requirement (SR) 
3.4.3.1 to increase the allowable as-found main steam safety valve 
(MSSV) lift set point tolerance from +/-1 percent to +/-3 percent. The 
proposed change would also revise the SR 3.1.7.10 to increase the 
enrichment of sodium pentaborate used in the Standby Liquid Control 
(SLC) system from greater than or equal to 30 atom percent boron-10 to 
greater than or equal to 45 atom percent boron-10.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change increases the allowable as-found MSSV lift 
setpoint tolerance, determined by test after the valves have been 
removed from service, from +/-1 percent to +/-3 percent. The 
proposed change does not alter the TS requirements for

[[Page 46932]]

the number of MSSVs required to be operable, the nominal lift 
setpoints, the allowable as-left lift setpoint tolerance, the MSSV 
testing frequency, or the manner in which the valves are operated.
    Consistent with current TS requirements, the proposed change 
continues to require that the MSSVs be adjusted to within +/-1 
percent of their nominal lift setpoints following testing. Since the 
proposed change does not alter the manner in which the valves are 
operated, there is no significant impact on reactor operation.
    The proposed change does not involve a physical change to the 
valves, nor does it change the safety function of the valves. The 
proposed TS revision involves no significant changes to the 
operation of any systems or components in normal or accident 
operating conditions and no changes to existing structures, systems, 
or components, with the exception of the SLC system enrichment 
change. The proposed change to increase the enrichment of sodium 
pentaborate used in the SLC system by a design modification using a 
single SLC pump will ensure that the requirements of 10 CFR 50.62, 
``Requirements for reduction of risk from anticipated transients 
without scram (ATWS) events for light-water-cooled nuclear power 
plants,'' continue to be met. The SLC system is not an initiator to 
an accident; rather, the SLC system is used to mitigate a postulated 
anticipated transient without scram (ATWS) event. Therefore, these 
changes will not increase the probability of an accident previously 
evaluated.
    Generic considerations related to the change in setpoint 
tolerance were addressed in NEDC-31753P, ``BWROG In-Service Pressure 
Relief Technical Specification Revision Licensing Topical Report,'' 
and were reviewed and approved by the NRC in a safety evaluation 
dated March 8, 1993. The plant specific evaluations, required by the 
NRC's safety evaluation and performed to support this proposed 
change, show that there is no change to the design core thermal 
limits and adequate margin to the reactor vessel pressure limits 
using a +/-3 percent lift setpoint tolerance. These analyses also 
show that operation of Emergency Core Cooling Systems is not 
affected, and the containment response following a loss-of-coolant 
accident is acceptable. The plant systems associated with these 
proposed changes are capable of meeting applicable design basis 
requirements and retain the capability to mitigate the consequences 
of accidents described in the Updated Final Safety Analysis Report. 
Therefore, these changes do not involve an increase in the 
consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change increases the allowable as-found lift 
setpoint tolerance for the DNPS MSSVs, and increases the required 
enrichment of sodium pentaborate used in the SLC system. The 
proposed change to increase the enrichment of sodium pentaborate 
used in the SLC system will ensure that the requirements of 10 CFR 
50.62 continue to be met.
    The proposed change to increase the MSSV tolerance was developed 
in accordance with the provisions contained in the NRC safety 
evaluation for NEDC-31753P. MSSVs installed in the plant following 
testing or refurbishment will continue to meet the current tolerance 
as-left acceptance criteria of +/-1 percent of the nominal setpoint. 
The proposed change does not affect the manner in which the 
overpressure protection system is operated; therefore, there are no 
new failure mechanisms for the overpressure protection system.
    The proposed change to allow an increase in the MSSV setpoint 
tolerance does not alter the nominal MSSV lift setpoints or the 
number of MSSVs currently required to be operable by DNPS TS. The 
proposed change does not involve physical changes to the valves, nor 
does it change the safety function of the valves. There is no 
alteration to the parameters within which the plant is normally 
operated. As a result, no new failure modes are being introduced.
    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 margin of safety is established through the design of the 
plant structures, systems, and components, the parameters within 
which the plant is operated, and the establishment of the setpoints 
for the actuation of equipment relied upon to respond to an event. 
The proposed change does not modify the safety limits or setpoints 
at which protective actions are initiated, and does not change the 
requirements governing operation or availability of safety equipment 
assumed to operate to preserve the margin of safety.
    Establishment of the 3 percent MSSV setpoint 
tolerance limit does not adversely impact the operation of any 
safety-related component or equipment. Evaluations performed in 
accordance with the NRC safety evaluation for NEDC-31753P have 
concluded that all design limits will continue to be met.
    The proposed change to increase the enrichment of sodium 
pentaborate used in the SLC system will ensure that the requirements 
of 10 CFR 50.62 continue to be met.
    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 Exelon Way, Kennett 
Square, PA 19348.
    NRC Branch Chief: Daniel S. Collins.

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

    Date of amendment request: March 16, 2006.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 3.3.6.1, ``Primary Containment 
Isolation Instrumentation,'' Table 3.3.6.1-1 to revise the allowable 
values (AVs) for the reactor core isolation cooling (RCIC) temperature-
based leak detection. The proposed change is a result of revising the 
setpoint calculation for the subject temperature instruments based on 
the current reactor coolant leak detection analytical limit. The 
temperature limits correspond to a 25-gallon per minute (gpm) leak as 
determined by LSCS calculations. The proposed changes would revise TS 
Table 3.3.6.1-1 AVs for the following four RCIC system isolation 
functions:

Item 3.e. RCIC Equipment Room Temperature--High
Item 3.f. RCIC Equipment Room Differential Temperature--High
Item 3.g. RCIC Steam Line Tunnel Temperature--High
Item 3.h. RCIC Steam Line Tunnel Differential Temperature--High

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change is a result of revising the setpoint 
calculation for the subject temperature instruments based on the 
current reactor coolant leak detection calculation analytical limit. 
The proposed changes will revise TS Table 3.3.6.1-1 Allowable Values 
for the following four RCIC system isolation functions as noted 
below.
 Increase the Allowable Value for Function 3.e., ``RCIC 
Equipment Room Temperature--High,'' from <= 291.0 [deg]F to <= 297.0 
[deg]F
 Decrease the Allowable Value for Function 3.f., ``RCIC 
Equipment Room Differential Temperature--High,'' from <= 189.0 
[deg]F to <= 188.0 [deg]F
 Decrease the Allowable Value for Function 3.g., ``RCIC 
Steam Line Tunnel Temperature--High,'' from <= 277.0 [deg]F to <= 
267.0 [deg]F
 Increase the Allowable Value for Function 3.h., ``RCIC 
Steam Line Tunnel Differential Temperature--High,'' from <= 155.0 
[deg]F to <= 163.0 [deg]F


[[Page 46933]]


    The function of the instrumentation listed on TS Table 3.3.6.1-
1, in combination with other accident mitigation features, is to 
limit fission product release during and following postulated Design 
Basis Accidents to within allowable limits. The Allowable Values 
specified in TS Table 3.3.6.1-1 provide assurance that the 
instrumentation will perform as designed.
    The Allowable Values for RCIC system isolation are not a 
precursor to any accident previously evaluated. Accidents are 
assumed to be initiated by equipment failure. The proposed change 
does not alter the initiation conditions or operational parameters 
for the system. There is no increase in the failure probability of 
the system. As such, the probability of occurrence for a previously 
evaluated accident is not increased.
    The Allowable Values specified in Table 3.3.6.1-1 provide 
assurance that the RCIC system will perform as designed. The 
proposed revision to the Allowable Values does not change any of the 
RCIC system leak detection isolation actuation setpoints. Thus, the 
radiological consequences of any accident previously evaluated are 
not increased.
    Based on the above information, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not affect the control parameters 
governing unit operation or the response of plant equipment to 
transient conditions. The proposed change does not change or 
introduce any new equipment, modes of system operation or failure 
mechanisms.
    The proposed change is based on revised reactor coolant leak 
detection calculation analytical limits determined by the most 
current revision to the heat rise calculation. Setpoint calculations 
have been performed to determine the nominal trip setpoints and 
Allowable Values for the instrumentation associated with the leak 
detection function based on the revised analytical limits determined 
by the heat rise calculations. The proposed revision to the 
Allowable Values does not change any of the RCIC system leak 
detection isolation actuation setpoints.
    Based on the above information, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change will revise TS Table 3.3.6.1-1 Allowable 
Values for the instrument functions associated with RCIC Isolation.
    The current Allowable Values for these functions are:

<= 291.0 [deg]F for RCIC Equipment Room Temperature--High
<= 189.0 [deg]F for RCIC Equipment Room Differential Temperature--
High
<= 277.0 [deg]F for the RCIC Steam Line Tunnel Temperature--High
<= 155.0 [deg]F for the RCIC Steam Line Tunnel Differential 
Temperature--High

    The proposed change revises the Allowable Values to the 
following:

<= 297.0 [deg]F for RCIC Equipment Room Temperature--High
<= 188.0 [deg]F for RCIC Equipment Room Differential Temperature--
High
<= 267.0 [deg]F for the RCIC Steam Line Tunnel Temperature--High
<= 163.0 [deg]F for the RCIC Steam Line Tunnel Differential 
Temperature--High

    The proposed change is a result of revising the setpoint 
calculation for the subject temperature instruments based on the 
current analytical limit. The proposed changes will revise TS Table 
3.3.6.1-1 Allowable Values for the subject four RCIC system 
isolation functions and will provide assurance that the RCIC system 
will perform as designed. The proposed revision to the Allowable 
Values does not change any of the RCIC system leak detection 
isolation actuation setpoints.
    Margin of safety is established by the design and qualification 
of plant equipment, the operation of the plant within analyzed 
limits, and the point at which protective or mitigative actions are 
being initiated. The proposed change does not alter these 
considerations. The proposed allowable values will still ensure that 
the results of the accident analysis remain valid.
    Based on this information, 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley 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. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: April 4, 2006.
    Description of amendment request: The proposed amendment request 
will add one NRC approved topical report reference to the list of 
analytical methods in Technical Specification (TS) 5.6.5, ``Core 
Operating Limits Report (COLR),'' that can be used to determine core 
operating limits, and will delete seven obsolete references from the 
same TS Section.
    The proposed changes are:

    1. Add an NRC previously approved Topical Report ANF-1358(P)(A), 
Revision 3, ``The Loss of Feedwater Heating Transient in Boiling 
Water Reactors,'' (LOFWH), which will list FRA-ANP method for 
evaluating the LOFWH transient.
    2. Delete seven references describing previously approved Global 
Nuclear Fuel (GNF) and FRA-ANP methodologies for the analyses of 
ATRIUM-9B and GE9 fuel. Both of these fuel types have been or will 
be completely discharged from both Lasalle County Station (LSCS) 
reactors after the loading of ATRIUM-10 fuel during the LSCS Unit 2 
refuel outage currently scheduled to begin in February 2007 (i.e., 
L2R11).

    The proposed changes support the continued irradiation of ATRIUM-10 
fuel in the LSCS reactors and the use of the NRC-approved analytical 
methodology for evaluation of LOFWH transients.
    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.
    Technical Specification (TS) 5.6.5 lists NRC-approved analytical 
methods used at LaSalle County Station (LSCS) to determine core 
operating limits. The proposed changes will add an NRC-approved 
topical report reference to the list of administratively controlled 
analytical methods in TS 5.6.5, ``Core Operating Limits Report 
(COLR),'' that can be used to determine core operating limits, and 
delete seven obsolete references.
    The addition of a Framatome ANP (FRA-ANP) methodology to 
determine overall core operating limits for future LSCS core 
configurations was approved by the NRC in Reference 2. LSCS Unit 2 
will continue to load Framatome ANP ATRIUM-10 fuel during the Unit 2 
Refueling Outage 11 currently scheduled for February 2007. The 
proposed change to TS 5.6.5 will add a FRA-ANP methodology as a 
reference to determine core operating limits for loss of feedwater 
heater (LOFWH) conditions. Thus, the proposed change will allow LSCS 
to use the most recent FRA-ANP methodology for analysis of LOFWH 
conditions.
    The addition and deletion of approved analytical methods in TS 
Section 5.6.5 has no effect on any accident initiator or precursor 
previously evaluated and does not change the manner in which the 
core is operated. The NRC-approved methods ensure that the output 
accurately models predicted core behavior, have no effect on the 
type or amount of radiation released, and have no effect on 
predicted offsite doses in the event of an accident. Additionally, 
the NRC-approved methods do not change any key core parameters that 
influence any accident consequences. Thus, the proposed changes do 
not have any effect on the probability of an accident previously 
evaluated.
    The methodology conservatively establishes acceptable core 
operating limits such that the consequences of previously analyzed 
events are not significantly increased.

[[Page 46934]]

    The proposed changes in the list of analytical methods do not 
affect the ability of LSCS to successfully respond to previously 
evaluated accidents and does not affect radiological assumptions 
used in the evaluations. Thus, the radiological consequences of any 
accident previously evaluated are not increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to TS Section 5.6.5 do not affect the 
performance of any LSCS structure, system, or component credited 
with mitigating any accident previously evaluated. The NRC-approved 
analytical methodology for evaluating LOFWH transients will not 
affect the control parameters governing unit operation or the 
response of plant equipment to transient conditions. The proposed 
changes do not introduce any new modes of system operation or 
failure mechanism.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed changes will add a reference to the list of 
analytical methods in TS 5.6.5 that can be used to determine core 
operating limits and delete seven obsolete references. The proposed 
changes do not modify the safety limits or setpoints at which 
protective actions are initiated and do not change the requirements 
governing operation or availability of safety equipment assumed to 
operate to preserve the margin of safety. Therefore, the proposed 
changes provide an equivalent level of protection as that currently 
provided.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based on the above information, EGC concludes that the proposed 
amendment presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

    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. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Daniel S. Collins.

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 8, 2006.
    Description of amendment request: The proposed changes modify 
Technical Specifications (TSs) 3.1.3, ``Control Rod OPERABILITY''; 
3.1.6, ``Rod Pattern Control''; 3.3.2.1, ``Control Rod Block 
Instrumentation''; 3.10.7, ``Control Rod Testing--Operating''; and 
3.10.8, ``SHUTDOWN MARGIN (SDM) Test--Refueling'' to replace the 
current references to banked position withdrawal sequence (BPWS) with 
references to ``the analyzed rod position sequence.''
    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 change modifies Technical Specifications (TS) 
3.1.3, ``Control Rod OPERABILITY''; TS 3.1.6, ``Rod Pattern 
Control''; TS 3.3.2.1, ``Control Rod Block Instrumentation''; TS 
3.10.7, ``Control Rod Testing--Operating'', and; TS 3.10.8, SHUTDOWN 
MARGIN (SDM) Test--Refueling''. The proposed change would replace 
the current references to ``Banked Position Withdrawal Sequence 
(BPWS)'' with references to ``the analyzed rod position sequence''. 
The use of the ``the analyzed rod position sequence'' will continue 
to minimize the consequences of an accident previously evaluated 
including the Control Rod Drop Accident (CRDA). Additionally, the 
use of the words ``the analyzed rod position sequence'' will provide 
an equivalent level of protection during plant startups and 
shutdowns and therefore will not increase the consequences of an 
accident previously evaluated.
    Control rod patterns during startup and shutdown conditions will 
continue to be controlled by the operator and the Rod Worth 
Minimizer (RWM) (LCO [limiting condition of operation] 3.3.2.1, 
``Control Rod Block Instrumentation''), so that only specified 
control rod sequences and relative positions are allowed over the 
operating range of all control rods inserted to 10% of Rated Thermal 
Power. As a result of this change, these sequences will continue to 
limit the potential amount of reactivity addition that could occur 
in the event of a Control Rod Drop Accident (CRDA).
    Accidents are initiated by the malfunction of plant equipment, 
or the failure of plant structures, systems, or components. The 
proposed change will ensure that analyzed rod position sequences are 
developed to minimize incremental control rod reactivity worth in 
accordance with the ``General Electric Standard Application for 
Reactor Fuel,'' NEDE-24011-P-A-15 (GESTAR-II), and U.S. Supplement, 
NEDE-24011-P-A-15-US, September, 2005, NRC approved methodology, and 
reviewed and approved in accordance with the 10 CFR 50.59 process. 
These analyzed rod position sequences will limit the potential 
reactivity increase for a postulated CRDA during reactor startups 
and shutdowns below the Low Power Setpoint of 10% of Rated Thermal 
Power.
    The proposed change will continue to ensure that systems, 
structures and components are capable of performing their intended 
safety functions.
    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 affect the assumed accident 
performance of the control rods, nor any plant structure, system, or 
component previously evaluated.
    The proposed change does not involve the installation of new 
equipment, and installed equipment is not being operated in a new or 
different manner. The change ensures that control rods remain 
capable of performing their safety functions. No set points are 
being changed which would alter the dynamic response of plant 
equipment. Accordingly, no new failure modes are introduced.
    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 will ensure that analyzed rod position 
sequences are developed to minimize incremental control rod 
reactivity worth in accordance with the ``General Electric Standard 
Application for Reactor Fuel,'' NEDE-24011-P-A-15 (GESTAR-II), and 
U.S. Supplement, NEDE-24011-P-A-15-US, September, 2005, NRC approved 
methodology, and reviewed and approved in accordance with the 10 CFR 
50.59 process. The proposed change will not adversely impact the 
plant's response to an accident or transient. All current safety 
margins will be maintained. There are no changes proposed which 
alter the set points at which protective actions are initiated, and 
there is no change to the operability requirements for equipment 
assumed to operate for accident mitigation.
    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

[[Page 46935]]

amendment request involves no significant hazards consideration.
    Attorney for Licensee: Mr. Brad Fewell, Assistant General Counsel, 
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA 
19348.
    NRC Branch Chief (Acting): Brooke D. Poole.

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

    Date of amendment request: June 14, 2006.
    Description of amendment request: The amendments would incorporate 
the results of a new spent fuel pool criticality analysis documented in 
WCAP-16518-P/WCAP-16518-NP, ``Beaver Valley Unit 2 Spent Fuel Pool 
Criticality Analysis,'' Revision 1, May 2006 for the BVPS-2 spent fuel 
storage pool. The revised criticality analysis will permit utilization 
of vacant storage locations dictated by the existing Technical 
Specification (TS) storage configurations in the BVPS-2 spent fuel 
storage pool.
    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 relevant accidents previously evaluated are 
limited to the fuel handling and criticality accidents.
    Administrative controls during fuel fabrication ensure that the 
fuel is fabricated to ensure proper loading of fuel in the fuel 
assemblies. Administrative and operational controls used to load 
fuel assemblies into the spent fuel pool ensure the fuel assemblies 
are stored in compliance with the allowed storage configurations. 
Fuel handling is performed under administrative controls and 
physical limitations. These controls will remain in effect and 
continue to protect against criticality and fuel handling accidents 
involving new storage configurations dictated by the new analysis. 
There is therefore no impact on the probability of fuel handling or 
criticality accidents.
    The new criticality analysis defines new spent fuel storage 
configurations with new enrichment and burnup limits. Integral Fuel 
Burnable Absorber (IFBA) limits are used to comply with the 1-out-
of-4 configuration for fresh fuel. The boron dilution evaluation 
that supported Amendment [No.] 128 [February 11, 2002, Agencywide 
Documents Access and Management System Accession No. ML020020373], 
permitting credit for soluble boron at BVPS Unit No. 2 continues to 
remain valid. The new analysis demonstrates that keff 
remains below unity for the various storage configurations 
considered with zero soluble boron, and that keff remains 
less than or equal to 0.95 for the entire pool with credit for 
soluble boron under non-accident and accident conditions with a 95% 
probability at a 95% confidence level (95/95). Potential 
consequences of accidents previously analyzed remain unchanged.
    The editorial changes made to the table numbers and the LCO 
[Limiting Condition for Operation] and Surveillance Requirement 
wording do not impact probability or consequences of an accident 
previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No. The relevant types of accidents previously 
evaluated are limited to criticality and fuel handling accidents. 
Although the new analysis will allow utilization of additional 
storage capacity, implementation of fuel loading configurations and 
fuel handling activities will continue to be performed under 
administrative and operational controls. No new or different 
activities are introduced as a result of the proposed changes. The 
utilization of additional storage capacity within the allowances of 
the revised analysis will introduce no new or other kind of 
accident.
    The editorial changes made to the table numbers and the LCO and 
Surveillance Requirement wording do not impact any previously 
evaluated accident.
    Therefore, the proposed changes do 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 margin to safety with respect to analyzed 
accidents involves maintaining keff through fuel storage 
configurations and boron concentration controls in the spent fuel 
pool. The boron dilution evaluation that supported that supported 
Amendment [No.] 128 permitting credit for soluble boron at BVPS Unit 
No. 2 remains valid. The Amendment [No.] 128 evaluation concluded 
that a boron dilution event is not credible for BVPS Unit No. 2. The 
new analysis calculates the non-accident soluble boron concentration 
to be less than was determined in the Amendment [No.] 128 
evaluation. Thus, there is no significant reduction in a margin of 
safety because of the new analysis and the conclusions of the 
Amendment [No.] 128 dilution evaluation remain valid.
    Under accident conditions, the soluble boron needed to maintain 
keff below 0.95 with the new storage configurations is 
less than what is assumed in current analysis. The proposed change 
does not involve a significant reduction in a margin of safety for 
accident conditions.
    The editorial changes made to the table numbers and the LCO and 
Surveillance Requirement wording do not impact a margin of safety.
    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: Mary O'Reilly, FirstEnergy Nuclear Operating 
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 
44308.
    NRC Branch Chief: Richard J. Laufer.

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

    Date of amendment request: June 1, 2006.
    Description of amendment request: The proposed amendment would 
modify Technical Specification 3.4.10, ``Residual Heat Removal (RHR) 
Shutdown Cooling System--Cold Shutdown'' by adding a default Condition 
to address situations when an RHR shutdown cooling subsystem becomes 
inoperable in MODE 4 and, within the completion time of 1 hour, an 
alternate method of decay heat removal can not be verified to be 
available.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No. The proposed amendment does not change the design 
of any structures, systems or components (SSCs), and does not affect 
the manner in which plant systems are operated. It is a change to 
the Technical Specifications only, to provide guidance to plant 
operators on appropriate actions to take, where no Technical 
Specification guidance currently exists. Since the design of plant 
SSCs is not changed and plant systems and components are not 
operated in a different manner, there is no change to previously 
identified accident initiators, and the proposed amendment would not 
impact the probability of any of the previously evaluated accidents 
in the Updated Safety Analysis Report (USAR).
    The USAR event that evaluates the consequences of a loss of RHR 
Shutdown Cooling is included in Section 15.2.9 entitled ``Failure of 
RHR Shutdown Cooling''. This event examines the consequences of a 
loss of not only an RHR shutdown cooling

[[Page 46936]]

subsystem, but also the loss of the suction source from the 
recirculation system leading to both RHR Shutdown Cooling 
subsystems, and a loss of offsite power. Even with these multiple 
failures, this event is not one of the limiting transients. As noted 
in Section 15.2.9.5, ``Radiological Consequences,'' there are no 
fuel failures, and the consequences of the event are much less than 
those for the ``Main Steam Isolation Valve Closure'' transient, 
which is evaluated with acceptable results in USAR Section 15.2.4.5. 
Since the proposed amendment only involves the addition of a 
Required Action where no guidance currently exists, and the design 
of plant SSCs is not changed and plant systems and components are 
not operated in a different manner, the proposed amendment does not 
affect the consequences of the Section 15.2.9 analysis, nor does it 
affect the ability of the installed RHR subsystems to perform their 
shutdown cooling function. The change adds a default Condition to 
provide guidance to the operators in those situations when a 
subsystem becomes inoperable with the plant in MODE 4 and an 
alternate cannot be verified to be available within an hour, which 
does not impact the consequences of the previously evaluated 
accidents in the USAR.
    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. This change to the required Technical 
Specification actions does not involve a change in the design 
function or operation of plant SSCs. It does not introduce credible 
new failure mechanisms, malfunctions, or accident initiators not 
considered in the existing plant design and licensing basis.
    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. This proposed amendment only involves a change to 
the required Technical Specification actions. It does not involve a 
change in the evaluation and analysis methods used to demonstrate 
compliance with regulatory and licensing requirements, and does not 
exceed or alter a design basis or safety limit. The safety margin 
before the change remains unchanged after the proposed amendment.
    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: David W. Jenkins, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    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: June 1, 2006.
    Description of amendment request: The proposed amendment would 
modify Technical Specification 3.4.9, ``Residual Heat Removal (RHR) 
Shutdown Cooling System--Hot Shutdown,'' to revise the Required Actions 
when both RHR shutdown cooling subsystems are inoperable in MODE 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No. The proposed amendment does not change the design 
of any structures, systems or components (SSCs), and does not affect 
the manner in which plant systems are operated. It is a change to 
the Technical Specifications only, to provide guidance to plant 
operators on appropriate actions to take, when both RHR shutdown 
cooling subsystems are inoperable. Since the design of plant SSCs is 
not changed and plant systems and components are not operated in a 
different manner, there is no change to previously identified 
accident initiators, and the proposed amendment would not impact the 
probability of any of the previously evaluated accidents in the 
Updated Safety Analysis Report (USAR).
    The USAR event that evaluates the consequences of a loss of RHR 
Shutdown Cooling is included in Section 15.2.9 entitled ``Failure of 
RHR Shutdown Cooling.'' This event examines the consequences of a 
loss of not only an RHR shutdown cooling subsystem, but also the 
loss of the suction source from the recirculation system leading to 
both RHR Shutdown Cooling subsystems, and a loss of offsite power. 
Even with these multiple failures, this event is not one of the 
limiting transients. As noted in Section 15.2.9.5, ``Radiological 
Consequences,'' there are no fuel failures, and the consequences of 
the event are much less than those for the ``Main Steam Isolation 
Valve Closure'' transient, which is evaluated with acceptable 
results in USAR Section 15.2.4.5. Since the proposed amendment only 
involves the addition of a Required Action where no guidance 
currently exists, and the design of plant SSCs is not changed and 
plant systems and components are not operated in a different manner, 
the proposed amendment does not affect the consequences of the 
Section 15.2.9 analysis, nor does it affect the ability of the 
installed RHR subsystems to perform their shutdown cooling function.
    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. This change to the required Technical 
Specification actions does not involve a change in the design 
function or operation of plant SSCs. It does not introduce credible 
new failure mechanisms, malfunctions, or accident initiators not 
considered in the existing plant design and licensing basis.
    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. This proposed amendment only involves a change to 
the required Technical Specification actions. It does not involve a 
change in the evaluation and analysis methods used to demonstrate 
compliance with regulatory and licensing requirements, and does not 
exceed or alter a design basis or safety limit. The safety margin 
before the change remains unchanged after the proposed amendment.
    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: David W. Jenkins, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Daniel S. Collins.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County, 
Pennsylvania

    Date of amendment request: April 28, 2006.
    Description of amendment request: The proposed amendment would 
change the SSES 1 and 2 Technical Specifications (TSs) to modify the 
standby liquid control system for single loop pump operation and use of 
enriched sodium pentaborate solution.
    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:

    Does the proposed change involve a significant increase in the 
probability or

[[Page 46937]]

consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise Technical Specification 3.1.7 for 
the Standby Liquid Control (SLC) system to reflect new boron weight-
percent and enrichment requirements. In addition, the change to 
single pump operation reduces the required SLC pump flow and 
discharge pressure required to satisfy 10 CFR 50.62, thus increasing 
the reliability of the system. The changes do not otherwise alter 
the design or operation of the SLC system, and the existing design 
of the system is sufficient to support operation with the enriched 
sodium pentaborate solution. The SLC system is not considered to be 
the initiator of any event currently analyzed in the FSAR [Final 
Safety Analysis Report]. Therefore, the proposed changes do not 
increase the probability of a previously evaluated accident.
    The SSES ATWS [anticipated transient without scram] analysis was 
performed using standard accepted assumptions, inputs, and codes. 
That analysis, which demonstrated that the acceptance criteria for 
peak vessel pressure, peak cladding temperature, peak local cladding 
oxidation, peak suppression pool temperature, and peak containment 
pressure, established the requirements for the proposed boron 
weight-percent and concentration, and pump flow rate. The analysis 
assumed the use of only a single pump, versus two pumps. The results 
of the analysis are that no fission product barriers are adversely 
challenged, and the radiological consequences of previously 
evaluated accidents (i.e., ATWS) are not increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes revise Technical Specification 3.1.7 for 
the SLC system to reflect new boron weight-percent and enrichment 
requirements. In addition, the change to single pump operation 
reduces the required SLC pump flow and discharge pressure required 
to satisfy 10 CFR 50.62, thus increasing the reliability of the 
system. A new Surveillance Requirement (SR 3.1.7.10) is also added 
to verify the correct solution enrichment prior to addition of 
inventory to the SLC tank. The changes do not otherwise alter the 
design or operation of the SLC system, and the existing design of 
the system is sufficient to process the enriched sodium pentaborate 
solution. With the exception of these changes, no other physical 
changes to plant structures or systems are proposed. Thus, the 
proposed changes do not create a new initiating event for the 
spectrum of events currently postulated in the FSAR.
    Therefore, the proposed changes do 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 revise Technical Specification 3.1.7 for 
the SLC system to reflect new boron weight-percent and enrichment 
requirements. In addition, the change to single pump operation 
reduces the required SLC pump flow and discharge pressure required 
to satisfy 10 CFR 50.62, thus increasing the reliability of the 
system. The changes do not otherwise alter the design or operation 
of the SLC system, and the existing design of the system is 
sufficient to process the enriched sodium pentaborate solution.
    The analysis was performed using standard accepted assumptions, 
inputs, and codes. That analysis, which demonstrated that ATWS 
acceptance criteria are satisfied, established the requirements for 
the proposed boron weight-percent and concentration, and pump flow 
rate. Further, the analysis assumed only a single pump is in 
operation verses two pumps. The evaluation demonstrated that the SLC 
system meets this post-LOCA [loss-of-coolant accident] suppression 
pool pH control design function.
    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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Richard J. Laufer.

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

    Date of amendment request: October 12, 2004.
    Description of amendment request: As part of Nuclear Regulatory 
Commission's (NRC) approval of the Improved Technical Specifications 
for Browns Ferry Nuclear Plant, Unit 1, by Amendment No. 234, NRC 
imposed License Condition 2.C(4) to ensure that the required analyses 
and modifications needed to support the Technical Specification (TS) 
changes made by License Amendment No. 234 and any subsequent TS 
changes, were completed by licensee prior to entering the mode for 
which the TS applies. The proposed amendment would remove this license 
condition from the license.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not affect any precursors for 
accidents described in Chapter 14 of the Browns Ferry Updated Final 
Safety Analysis Report (UFSAR). The proposed amendment does not 
change the conditions, operating configurations, or minimum amount 
of operating equipment assumed in the safety analysis for accident 
mitigation. No changes are proposed in plant protection or which 
create new modes of plant operation. Therefore, the proposed 
amendment 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 amendment does not introduce new equipment, which 
could create a new or different kind of accident. No new external 
threats, release pathways, or equipment failure modes are created. 
Therefore, the implementation of the proposed amendment will not 
create a possibility for an accident of a new or different type than 
those previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment does not impact the redundancy or 
availability of equipment credited in the response to accidents 
described in Chapter 14 of the UFSAR. For these reasons, the 
proposed amendment does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: L. Raghavan.

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

    Date of amendment request: May 1, 2006 (TS-455).
    Description of amendment request: The proposed amendment would 
revise the numeric values of the safety limit minimum critical power 
ratio (SLMCPR) in the Technical Specification (TS) Section 2.1.1.2 for 
single and two reactor recirculation loop operation to incorporate the 
results of the Browns Ferry Nuclear Plant, Unit 1 Cycle 7 SLMCPR 
analysis.

[[Page 46938]]

    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 Technical Specification change involve a 
significant increase in the probability or consequences of an 
accident previously evaluated?
    Response: No.
    The proposed amendment establishes a revised SLMCPR value for 
single and two recirculation loop operation. The probability of an 
evaluated accident is derived from the probabilities of the 
individual precursors to that accident. The proposed SLMCPR values 
preserve the existing margin to transition boiling and the 
probability of fuel damage is not increased. Since the change does 
not require any physical plant modifications or physically affect 
any plant components, no individual precursors of an accident are 
affected and the probability of an evaluated accident is not 
increased by revising the SLMCPR values.
    The consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. The revised SLMCPR values have been determined using 
NRC-approved methods and procedures. The basis of the MCPR Safety 
Limit is to ensure no mechanistic fuel damage is calculated to occur 
if the limit is not violated. These calculations do not change the 
method of operating the plant and have no effect on the consequences 
of an evaluated accident. Therefore, the proposed TS change does not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed Technical Specification change create the 
possibility of a new or different kind of accident from any accident 
previously evaluated?
    Response: No.
    The proposed license amendment involves a revision of the SLMCPR 
value for single and two recirculation loop operation based on the 
results of an analysis of the Unit 1 Cycle 7 core. Creation of the 
possibility of a new or different kind of accident would require the 
creation of one or more new precursors of that accident. New 
accident precursors may be created by modifications of the plant 
configuration, including changes in the allowable methods of 
operating the facility. This proposed license amendment does not 
involve any modifications of the plant configuration or changes in 
the allowable methods of operation. Therefore, the proposed TS 
change does not create the possibility of a new or different kind of 
accident previously evaluated.
    3. Does the proposed Technical Specification change involve a 
significant reduction in a margin of safety?
    Response: No.
    The margin of safety as defined in the TS bases will remain the 
same. The new SLMCPR values were calculated using referenced fuel 
vendor methods and procedures, which are in accordance with the fuel 
design and licensing criteria. The SLMCPR remains high enough to 
ensure that greater than 99.9 percent of all fuel rods in the core 
are expected to avoid transition boiling if the limit is not 
violated, thereby preserving the fuel cladding integrity. Therefore, 
the proposed TS change does not involve a reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC (Acting) Branch Chief: L. Raghavan.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: July 6, 2006 (TS-06-04).
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) for the Sequoyah Nuclear 
Plant, Units 1 and 2. Action a.1 of TS 3.1.3.2, ``Position Indication 
Systems--Operating,'' requires the verification of rod position by use 
of the moveable incore detectors. Tennessee Valley Authority (the 
licensee, TVA) is proposing a revision to TS 3.1.3.2 to allow the 
position of the control and shutdown rods to be monitored by a means 
other than the moveable incore detectors. The amendment will provide a 
less burdensome monitoring method should problems with the analog rod 
position indication (ARPI) system be experienced. When a recurring 
problem in the system requires the monitoring of a rod's position by 
the alternate means, TVA plans to continue unit operation and to use 
the alternate means until the unit enters Mode 5 and repairs to the 
system can safely be implemented.
    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 provides an alternative method for the 
monitoring of the position of a rod once the position of the rod is 
verified using the moveable incore detector system. The proposed 
monitoring of rod control system parameters provides a reasonably 
similar approach to rod position monitoring as that provided by the 
movable incore detector system. In particular, the ability to 
immediately detect a rod drop or misalignment is not directly 
provided by the movable incore detector system or by the monitoring 
of rod control system parameters. Additionally, neither the movable 
incore detector system, nor the monitoring of rod control system 
parameters, provides the capability to verify rod position following 
a reactor trip or shutdown. Therefore, the monitoring of rod control 
system parameters, in lieu of the use of the movable incore detector 
system, provides an equivalent and acceptable method of monitoring 
rod position while a position indicator is inoperable.
    The proposed change does not alter plant equipment that is 
considered to have the potential to alter the probability of an 
accident. The affected components are for monitoring only and do not 
actively affect equipment that interacts with the control of the 
reactor. Likewise, the affected components are for monitoring and 
provide an equivalent level of indication of rod position as the 
current action. This maintains an acceptable level of rod position 
indication for normal plant operations, as well as post accident 
mitigation actions. 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.
    As described above, the proposed change provides only an 
alternative method of monitoring the position of a rod. No new 
accident initiators are introduced by the proposed alternative 
manner of performing rod position monitoring. The proposed change 
does not affect the reactor protection system or the reactor control 
system. Hence, no new failure modes are created that would cause a 
new or different kind of accident from any accident previously 
evaluated.
    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 rod position indicators are required to determine control 
rod positions and thereby ensure compliance with the control rod 
alignment and insertion limits. The proposed change does not alter 
the requirement to determine rod position but provides an 
alternative method for monitoring the position of the affected rod 
after the position of the rod is verified using the moveable incore 
detector system. As a result, the initial conditions of the accident 
analysis are preserved. The components affected by the alternate rod 
monitoring will not affect plant setpoints utilized for automatic 
mitigation of accident conditions or other equipment necessary for 
accident mitigation.

[[Page 46939]]

    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. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: July 12, 2006 (TS-06-03).
    Description of amendment request: The proposed amendment would 
revise the limiting condition for operation for the Sequoyah Nuclear 
Plant, Units 1 and 2, Technical Specification (TS) Section 3.7.5, 
``Ultimate Heat Sink.'' This revision would change the minimum ultimate 
heat sink (UHS) water elevation in TS 3.7.5.a from 670 feet to 674 
feet. The essential raw cooling water (ERCW) temperature requirement in 
TS 3.7.5.b would be increased from 83 degrees Fahrenheit ([deg]F) to 87 
[deg]F. The conditional requirements of TS 3.7.5.c would no longer be 
required and would be deleted by the proposed change. This change would 
also delete a footnote that established a temporary UHS temperature 
limit of 87 [deg]F through September 30, 1995. These proposed changes 
are supported by a combination of design basis re-analysis, bounding 
analysis, and sensitivity analysis of the ERCW system, the UHS, and 
supported 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 change to increase the UHS maximum temperature and 
the minimum water level does not alter the function, design, or 
operating practices for plant systems or components. One exception 
is the elimination of non-safety-related station air compressor 
loads located in the turbine building. 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 because the UHS function does not 
have the potential to be the source of an accident. The heat loads 
that the UHS is designed to accommodate have been evaluated for 
functionality with the higher temperature and elevation 
requirements. The result of these evaluations is that there is 
existing margins 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. 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 UHS function is not an initiator of any accident and only 
serves as a heat sink for normal and upset plant conditions. By 
allowing the proposed change in the UHS temperature and elevation 
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 UHS function 
provides accident mitigation capabilities and does not reflect the 
potential for accident generation. Therefore, the possibility for 
creating a new or different kind of accident is not created because 
the UHS is only utilized for heat removal functions that are not a 
potential source for accident generation. 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 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 margins 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.
    Additionally, the proposed changes do not require the 
modification of component setpoints utilized for automatic 
mitigation of accident conditions or other equipment necessary for 
accident mitigation. Therefore, a significant reduction in the 
margin to safety is not created by this proposed change. 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: L. Raghavan.

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

    Date of amendment request: June 16, 2006 (WBN-TS-06-04).
    Description of amendment request: The proposed amendment change 
would revise Technical Specification (TS) 5.7.2.11, ``Inservice Testing 
Program,'' to remove ``applicable supports'' from the Inservice Testing 
(IST) Program and revise the IST Program for pumps and valves to meet 
the requirements of the latest Edition and Addenda of the American 
Society of Mechanical Engineers (ASME) Code approved by the NRC for use 
on the date 12-months prior to the start of the 10-year IST Interval. 
For the Watts Bar Nuclear Plant (WBN), Unit 1, the second 10-year IST 
Interval will begin on December 27, 2006. The ASME Code that was 
approved in 10 CFR 50.55a(f)(4) for use on December 27, 2005, was ASME 
Operations and Maintenance (OM) Code, 2001 Edition, with Addenda 
through 2003. The proposed change provides consistency with the 
requirements in 10 CFR 50.55a(f)(4) by replacing the reference to ASME 
Boiler and Pressure Vessel Code, Section XI, with ASME OM Code. This 
proposed change is based on Technical Specification Task Force (TSTF) 
Traveler 479, Revision 0, ``Changes to Reflect Revision of 10 CFR 
50.55a.'' TSTF 279-A, Revision 0, ``Remove `applicable supports' from 
Inservice Testing Program,'' was approved by NRC and incorporated into 
Revision 2 of NUREG-1431, ``Standard Technical Specification 
Westinghouse Plants.'' In addition, the proposed amendment would add 
provisions to TS 5.7.2.11, Item b, to only apply Surveillance 
Requirement 3.0.2 to those IST frequencies of 2 years or less.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 46940]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises Technical Specification Section 
5.7.2.11 for WBN Unit 1 to conform to the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves 
which are classified as ASME Code Class 1, 2, and 3.
    ASME has in the last several years, transitioned the 
requirements for inservice testing of pumps and valves out of ASME 
Section XI and into a separate, stand alone code entitled the ``Code 
for Operation and Maintenance of Nuclear Power Plants,'' (ASME OM 
Code). The ASME OM Code has been endorsed by the NRC in 10 CFR 
50.55a and is the Code that will be required for inservice testing 
of pumps and valves during the WBN Second Inservice Interval. The 
proposed change incorporates revisions to the ASME Code that result 
in a net improvement in the measures for testing pumps and valves. 
The proposed change also deletes the reference to supports from the 
Inservice Testing Program as supports are already inspected under 
the Inservice Inspection Program.
    The proposed changes do not involve any hardware changes, nor do 
the changes affect the probability of any event initiators. There 
will be no change to normal plant operating parameters, accident 
mitigation capabilities, or accident analysis assumptions or inputs. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the Technical Specifications to 
delete the reference to ``applicable supports'' from the Inservice 
Testing Program and to incorporate the latest Code requirements in 
10 CFR 50.55a(f)(4) for Code Class 1, 2, and 3 pumps and valves for 
WBN's next ten year interval. The testing requirements are similar 
and reflect the same type testing. Valves are still stroke timed; 
remote position indicators are still verified to be accurate; seat 
leakage measurements of critical valves are still performed; relief 
valves still have their setpoints and seat leakages verified; pumps 
are still tested for hydraulic performance and mechanical condition; 
check valves are verified to open and close properly; and supports 
are still inspected under the appropriate inspection program.
    The proposed changes do not involve a modification to the 
physical configuration of the plant or change methods governing 
normal plant operation. No test methods are added or deleted. 
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 revises the TS for consistency with the 
Standard Technical Specification and with the requirements in 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves 
which are classified as ASME Code Class 1, 2, and 3. This change 
incorporates revisions to the ASME Code that result in a net 
improvement in the measures of testing. Incorporation of the ASME OM 
Code does not alter the limiting values and acceptance criteria used 
to judge the continued acceptability of components tested by the 
Inservice Testing Program. Deletion of the reference to supports in 
the Inservice Testing Program does not alter the support inspection 
program as the program is currently under the Inservice Inspection 
Program. Since these limits are not altered, the margin of safety is 
not altered. 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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: May 30, 2006.
    Description of amendment request: The amendment would revise 
Surveillance Requirements (SRs) 3.5.2.8 and 3.6.7.1 in the Technical 
Specifications (TSs), and delete the footnote to the frequency for SR 
3.5.2.5. SR 3.5.2.8 would be revised by replacing the phrase ``trash 
racks and screens'' with the word ``strainers.'' This reflects (1) the 
replacement of the existing containment recirculation sump suction 
inlet trash racks and screens with strainers with significantly greater 
effective surface area, and (2) the resulting relocation of the 
recirculation fluid pH control system in Refueling Outage 15 schedule 
for the spring of 2007. The footnote to SR 3.5.2.5 would be deleted 
because it is no longer applicable to the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do[es] the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    None of the changes impact the initiation or probability of 
occurrence of any accident [previously evaluated].
    The consequences of accidents evaluated in the FSAR [Final 
Safety Analysis Report for the Callaway Plant] that could be 
affected by this proposed change are those involving the 
pressurization of the containment and associated flooding of the 
containment and recirculation of this fluid within the Emergency 
Core Cooling System (ECCS) or the Containment Spray System (CSS) 
(e.g., LOCAs [Loss-of-Coolant Accidents]). [The containment sump 
trash racks and screens, and the sump strainers that are replacing 
the trash racks and screens are not initiators of accidents.]
    Although the configurations of the existing sump screen and the 
replacement strainer assemblies are different, they serve the same 
fundamental purpose of passively removing debris from the suction of 
the supported system pumps. Removal of trash racks does not impact 
the adequacy of the pump NPSH [net positive suction head] assumed in 
the safety analyses. Likewise the change does not reduce the 
reliability of any supported systems or introduce any new system 
interactions. The greatly increased surface area of the new strainer 
is designed to reduce head loss [at the containment sump] and reduce 
the approach velocity at the strainer face significantly, decreasing 
the risk of impact from large debris entrained in the sump flow 
stream.
    The recirculation fluid pH control system storage baskets serve 
a passive function to provide a buffering agent to neutralize the 
sump solution. The redesign and relocation of the storage baskets 
are considered a like kind replacement. The baskets will be located 
within the flood plain and will continue to ensure that the 
buffering agent is dissolved in the sump fluid to ensure an 
equilibrium pH >= 7.1. Failure of a basket would not initiate an 
accident. The ECCS and CSS will continue to function in a manner 
consistent with the plant design basis.
    As such, the proposed change to the Technical Specifications 
Surveillance Requirements does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The installed quantity of trisodium phosphate Crystalline will 
provide a minimum equilibrium sump pH of 7.1 following dissolution 
and mixing. [Deleting the footnote to SR 3.5.2.5 is an 
administrative change to remove a one-time required verification 
that has already been performed and is no longer a requirement in 
the current TSs.] Therefore, there is not a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Do[es] the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The containment recirculation sump strainers and recirculation 
fluid pH control

[[Page 46941]]

system are passive systems used for accident mitigation. As such, 
they cannot be accident initiators. Therefore, there is no 
possibility that this change could create any accident of any kind. 
[The containment recirculation sump suction inlet trash racks and 
screens are being replaced with a complex strainer design with 
significantly larger effective surface area to reduce head loss and 
reduce the approach velocity at the strainer face significantly, 
decreasing the risk of impact from large debris entrained in the 
sump flow stream. This will result in the recirculation fluid pH 
control system being relocated.]
    No new accident scenarios, transient precursors, or limiting 
single failures are introduced as a result of these changes. There 
will be no adverse effect[s] or challenges imposed on any safety-
related system as a result of these changes. The quantity of 
trisodium phosphate crystalline will provide a minimum equilibrium 
sump pH of >= 7.1 following dissolution and mixing. Therefore, the 
possibility of a new or different type of accident is not created.
    There are no changes which would cause the malfunction of 
safety-related equipment, assumed to be operable in the accident 
analyses, as a result of the proposed Technical Specification 
changes. No new equipment performance burdens are imposed. The 
possibility of a malfunction of safety-related equipment with a 
different result is not created. [Deleting the footnote to SR 
3.5.2.5 is an administrative change to remove a one-time required 
verification that has already been performed and is no longer a 
requirement in the current TSs.] Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Do[es] the proposed change involve a significant reduction in 
a margin of safety?
    Response: No.
    The proposed changes do not adversely affect any plant safety 
limits, setpoints, or design parameters. The changes also do not 
adversely affect the fuel, fuel cladding, Reactor Coolant System 
(RCS), or containment integrity. [The radiological dose consequence 
acceptance criteria in the Standard Review Plan for accidents will 
continue to be met. Deleting the footnote to SR 3.5.2.5 is an 
administrative change to remove a one-time required verification 
that has already been performed and is no longer a requirement in 
the current TSs.] Therefore, the proposed TS change does not involve 
a significant reduction in a margin of safety.

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

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

    Date of amendment request: May 30, 2006, as supplemented by letter 
dated June 30, 2006.
    Description of amendment request: The proposed amendments would 
relocate the American Society for Testing and Materials (ASTM) standard 
being used to test the total particulate concentration of the stored 
fuel oil to the TS Bases. This proposed change is described in TS Task 
force (TSTF) Standard TS Change Traveler TSTF-374-A, Rev. 0, ``Revision 
to TS 5.5.13 and Associated TS Bases for Diesel Fuel Oil.'' In 
addition, the licensee has proposed to use a ``water and sediment 
test'' instead of the ``clear and bright'' test provided in TSTF-374.
    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 changes involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    The proposed change relocates the specific ASTM reference from 
the Administrative Controls Section of Technical Specifications (TS) 
to a licensee-controlled document. Relocating the specific ASTM 
Standard reference from the TS to a licensee-controlled document 
will not affect nor degrade the ability of the EDGs [emergency 
diesel generators] to perform their specified safety function. Fuel 
oil quality will continue to meet the current ASTM requirements for 
particulate concentration.
    The proposed change is administrative in nature and does 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. The 
proposed change does not alter or prevent the ability of structures, 
systems or components from performing their intended function to 
mitigate the consequences on an initiating event with the assumed 
acceptance limits. The proposed change does not affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed change does not increase 
the types and amounts of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposure.
    Therefore, the change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Do changes create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    The proposed change relocates the specific ASTM reference from 
the Administrative Controls Section of Technical Specifications to a 
licensee-controlled document.
    The change does not involve a physical alteration of the plant 
or a change in the methods governing normal plant conditions. In 
addition, the change does not impose any new or different 
requirements or eliminate any existing requirements. The change does 
not alter assumptions made in the safety analysis and licensing 
basis. Therefore, the change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Do changes involve a significant reduction in the margin of 
safety?
    The proposed change relocates the specific ASTM reference from 
the Administrative Controls Section of TS to a licensee-controlled 
document. The detail associated with the specific ASTM standard 
reference is not required to be in the TS to provide adequate 
protection of the public health and safety, since the TS still 
retain the requirement for compliance with the applicable ASTM 
standard.
    The level of safety of facility operation is unaffected by the 
proposed change since there is no change in the intent of the TS 
requirements of assuring fuel oil is of the appropriate quality for 
EDG use. The proposed change provides the flexibility needed to 
maintain state-of-the-art technology in fuel oil sampling and 
analysis methodology.
    The proposed change does not reduce a margin of safety since it 
has no impact on any transient or safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: May 26, 2006.
    Description of amendment request: Item 1: The proposed amendments 
would revise the Technical Specification (TS) requirements related to 
Reactor Coolant System (RCS) leakage definitions and requirements and 
steam generator tube integrity. The licensee requested this change to 
implement TS Task Force (TSTF) Standard TS Change Traveler, TSTF-449, 
``Steam Generator

[[Page 46942]]

Tube Integrity,'' (TSTF-449, Rev. 4). Item 2: In addition, in its 
submittal dated May 26, 2006, the licensee proposed minor deviations 
from the TS changes described in TSTF-449, Rev. 4, to provide 
consistency with Surry's custom TSs.
    Basis for proposed no significant hazards consideration 
determination: Item 1: As required by 10 CFR 50.91(a), an analysis of 
the issue of no significant hazards consideration is presented below:

    1. The Proposed Change Does Not Involve a Significant Increase 
in the Probability or Consequences of an Accident Previously 
Evaluated.
    The proposed change 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 SG tube rupture (TR) 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 main steam line break 
(MSLB), rod ejection, and reactor coolant pump locked rotor the 
tubes are assumed to retain their structural integrity (i.e., they 
are assumed not to rupture). These analyses typically assume that 
primary to secondary leakage for all SGs is 1 gallon per minute or 
increases to 1 gallon per minute as a result of accident induced 
stresses. The accident induced leakage criterion introduced by the 
proposed changes accounts for tubes that may leak during design 
basis accidents. The accident induced leakage criterion limits this 
leakage to no more than the value assumed in the accident analysis.
    The SG performance criteria proposed change to the TS identify 
the standards against which tube integrity is to be measured. 
Meeting the performance criteria provides reasonable assurance that 
the SG tubing will remain capable of fulfilling its specific safety 
function of maintaining reactor coolant pressure boundary integrity 
throughout each operating cycle and in the unlikely event of a 
design basis accident. The performance criteria are only a part of 
the SG Program required by the proposed change to the TS. The 
program, defined by NEI 97-06, Steam Generator Program Guidelines, 
includes a framework that incorporates a balance of prevention, 
inspection, evaluation, repair, and leakage monitoring. The proposed 
changes do not, therefore, significantly increase the probability of 
an accident previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT 1-131 in the primary coolant and 
the primary to secondary leakage rates resulting from an accident. 
Therefore, limits are included in the plant TS for operational 
leakage and for DOSE EQUIVALENT 1-131 in primary coolant to ensure 
the plant is operated within its analyzed condition. The typical 
analysis of the limiting design basis accident assumes that primary 
to secondary leak rate after the accident is 1 gallon per minute 
with no more than 500 gallons per day in any one SG, and that the 
reactor coolant activity levels of DOSE EQUIVALENT 1-131 are at the 
TS values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design basis 
accident and is an improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of a SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of an MSLB, rod ejection, or a reactor coolant pump 
locked rotor event, or other previously evaluated accident.
    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 [TS]. 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.
    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.
    [SG] tube integrity is a function of the design, environment, 
and the physical condition of the tube. The proposed change does not 
affect tube design or operating environment. The proposed change is 
expected to result in an improvement in the tube integrity by 
implementing the SG Program to manage SG tube inspection, 
assessment, repair, and plugging. The requirements established by 
the SG Program are consistent with those in the applicable design 
codes and standards and are an improvement over the requirements in 
the current TSs.
    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.

    The NRC staff has reviewed the licensee's incorporation of the 
above analysis by reference 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.
    Item 2: 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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes involve adding a new definition for RCS 
[reactor coolant system] leakage and rewording certain [TSs] for 
consistency with NUREG-1431, Revision 3. These changes do not 
involve any physical plant modifications or changes in plant 
operation; consequently, no technical changes are being made to the 
existing TS. As such, these changes are administrative in nature and 
do not affect initiators of analyzed events or assumed mitigation of 
accident or transient events. Therefore, these changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes involve adding a new definition for RCS 
leakage and rewording certain [TSs] for consistency with NUREG-1431, 
Revision 3. These administrative changes do not involve physical 
alteration of the plant (no new or different type of equipment will 
be installed) or changes in methods governing normal plant 
operation. The changes will not impose any new or different 
requirements or eliminate any existing requirements. Therefore, 
these changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

[[Page 46943]]

    3. Involve a significant reduction in a margin of safety.
    The proposed changes involve adding a new definition for RCS 
leakage and rewording certain [TS] for consistency with NUREG-1431, 
Revision 3. The changes are administrative in nature and will not 
involve any technical changes. The changes will not reduce a margin 
of safety because they have no impact on any safety analysis 
assumptions. Also, since these changes are administrative in nature, 
no question of safety is involved. Therefore, the 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Branch Chief: Evangelos C. Marinos.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendments: May 26, 2005, as supplemented 
by letters dated May 23 and June 20, 2006.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 1.1, ``Definitions,'' TS 3.4.14, ``RCS [reactor 
coolant system] Operational Leakage,'' TS 5.5.9, ``Steam Generator (SG) 
Program,'' and TS 5.6.8, ``Steam Generator Tube Inspection Report,'' 
and added a new specification, TS 3.4.18, ``Steam Generator (SG) Tube 
Integrity.'' The changes are consistent with TS Task Force (TSTF) 
Change TSTF-449, Revision 4, ``Steam Generator Tube Integrity.''
    Date of issuance: July 27, 2006.
    Effective date: As of the date of issuance to be implemented within 
150 days from the date of issuance.
    Amendment Nos.: Unit 1-161, Unit 2-161, Unit 3-161.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Operating Licenses and the Technical 
Specifications for all three units.
    Date of initial notice in Federal Register: July 5, 2005 (70 FR 
38714). The May 23 and June 20, 2006, supplemental letters provided 
additional clarifying information, 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.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 27, 2006.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2 (HBRSEP2), Darlington County, South Carolina

    Date of application for amendment: January 21, 2005, as 
supplemented by letters dated May 26, 2005, September 19, 2005, and 
March 31, 2006.
    Brief description of amendment: The amendment approves the 
implementation of the alternative source term methodology for a loss-
of-coolant accident at HBRSEP2.
    Date of issuance: July 11, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No. 207.
    Renewed Facility Operating License No. DPR-23. Amendment does not 
revise the Technical Specifications.
    Date of initial notice in Federal Register: May 24, 2005 (70 FR 
29786). The supplemental letters dated May 26, 2005, September 19, 
2005, and March 31, 2006, provided clarifying information that 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 July 11, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 12, 2006, as 
supplemented by letter dated June 2, 2006.
    Brief description of amendment: The amendment revises the existing 
steam generator (SG) tube surveillance program to be consistent with TS 
Task Force (TSTF) Change TSTF-449, Revision 4, ``Steam Generator Tube 
Integrity,'' and the model safety evaluation prepared by the Nuclear 
Regulatory Commission (NRC) and published in the Federal Register on 
March 2, 2005 (70 FR 10298) under the consolidated line item 
improvement process (CLIIP).
    Date of issuance: July 18, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 188.
    Facility Operating License No. DPR-43: Amendment revised the 
Facility Operating License and Technical Specifications.

[[Page 46944]]

    Date of initial notice in Federal Register: February 14, 2006 (71 
FR 7806). The supplement letter contained clarifying information and 
did not change the initial no significant hazards consideration 
determination, and did not expand the scope of the orginal Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 18, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 27, 2004.
    Brief description of amendments: The amendments revised the 
facility operating licenses by removal of license condition 2.F, 
``Reporting Requirements'', with regard to maximum power level, Updated 
Final Safety Analysis Report, antitrust conditions, fire protection, 
and additional conditions.
    Date of issuance: July 31, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 230, 226.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the licenses.
    Date of initial notice in Federal Register: July 5, 2005 (70 FR 
38717).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 31, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 17, 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 April 17, 2006.
    Date of issuance: July 11, 2006.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 198.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 9, 2006 (71 FR 
26998).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 11, 2006.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: September 26, 2005, as 
supplemented by letter dated April 11, 2006.
    Brief description of amendment: The amendment revises the analysis 
method used for the large-break loss-of-coolant accident.
    Date of issuance: July 24, 2006.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 248.
    Facility Operating License No. DPR-26: The amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: November 8, 2005 (70 FR 
67747). The April 11, 2006, supplement 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 as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 24, 2006.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant (JAFNPP), Oswego County, New York

    Date of application for amendment: January 26, 2006, as 
supplemented by letter dated April 12, 2006.
    Brief description of amendment: The amendment approves the 
implementation of the Boiling Water Reactor Vessel and Internals 
Project reactor pressure vessel integrated surveillance program as the 
basis for demonstrating the compliance of JAFNPP with the requirements 
of Appendix H to Title 10 of the Code of Federal Regulations part 50.
    Date of issuance: July 26, 2006.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 285.
    Facility Operating License No. DPR-59: The amendment revised the 
Updated Final Safety Analysis Report and the License.
    Date of initial notice in Federal Register: March 14, 2006 (71 FR 
13174). The April 12, 2006, supplement 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 as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 26, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 19, 2005.
    Brief description of amendment: The amendment modified ANO-2 
Surveillance Requirement TS 3.1.1.4, ``Moderator Temperature 
Coefficient,'' and allowed the use of WCAP-16011-P-A, ``Startup Test 
Activity Reduction Program.''
    Date of issuance: August 2, 2006.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 265.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications/license.
    Date of initial notice in Federal Register: December 6, 2005 (70 FR 
72671).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 2, 2006.

[[Page 46945]]

    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 19, 2005, as 
supplemented by letters dated May 11 and June 19, 2006.
    Brief description of amendment: The amendment revised the existing 
steam generator tube surveillance program to be consistent with the 
U.S. Nuclear Regulatory Commission's approved Technical Specification 
Task Force Standard Technical Specification Change Traveler, TSTF-449, 
``Steam Generator Tube Integrity,'' Revision 4. TSTF-449 is part of the 
consolidated line item improvement process.
    Date of issuance: August 2, 2006.
    Effective date: As of the date of issuance to be implemented within 
90 days from the date of issuance.
    Amendment No.: 266.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications and Renewed Facility Operating License.
    Date of initial notice in Federal Register: January 3, 2006 (71 FR 
147). The supplements dated May 11 and June 19, 2006, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 2, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 25, 2006, as 
supplemented by letter dated May 17, 2006.
    Brief description of amendments: The amendment revised the Quad 
Cities licensing basis, as described in the Updated Final Safety 
Analysis Report, to allow the use of automatic load tap changers to 
operate in automatic mode on the reserve auxiliary transformers to 
compensate for potential offsite power voltage fluctuations, in order 
to ensure that acceptable voltage is maintained for safety-related 
equipment.
    Date of issuance: July 24, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 232 and 228.
    Renewed Facility Operating License Nos. DPR-29 and DPR-30: The 
amendments revised the License.
    Date of initial notice in Federal Register: May 23, 2006 (71 FR 
29678). The May 17, 2006, supplement contained clarifying information 
and did not change the NRC staff's initial proposed finding of no 
significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 24, 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: August 23, 2005, as 
supplemented on April 6, 2006.
    Brief description of amendments: The amendments extended the 
licensed lives of the Diablo Canyon Power Plant, Unit Nos. 1 and 2 
reactors by the amount of time the licensee had expended to perform 
low-power testing of the reactors prior to initial startup.
    Date of issuance: July 17, 2006.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: Unit 1-188; Unit 2-190.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: October 11, 2005 (70 FR 
59087). The April 6, 2006, supplemental letter provided additional 
information that clarified the application, and did not expand the 
scope of the application as originally noticed.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 17, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 4, 2005, as supplemented 
by letters dated February 9, July 18, and August 1, 2006.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.7.1.3, ``Ultimate Heat Sink,'' to permit continued 
plant operation if the temperature of the ultimate heat sink (UHS) 
exceeds 89 [deg]F, provided the UHS temperature averaged over the 
previous 24-hour period is verified at least once per hour to be less 
than or equal to 89 [deg]F, and the UHS temperature does not exceed a 
maximum value of 91.4 [deg]F.
    Date of issuance: August 1, 2006.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 168.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: August 30, 2005 (70 FR 
51382).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 1, 2006.
    No significant hazards consideration comments received: No.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: November 7, 2005, as 
supplemented on May 5, 2006.
    Brief description of amendment: The amendment revises Technical 
Specification 3.9.3, ``Containment Penetrations,'' to allow an 
emergency egress door, access door, or roll up door, as associated with 
the equipment hatch penetration, to be open, but capable of being 
closed, during core alterations or movement of irradiated fuel within 
containment.
    Date of issuance: July 26, 2006.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 98.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: January 3, 2006 (71 FR 
154). The May 5, 2006, letter provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 26, 2006.
    No significant hazards consideration comments received: No.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: November 18, 2005.
    Brief description of amendment: The amendment revises the frequency 
in Technical Specification Surveillance Requirement 3.6.6.15, which 
verifies

[[Page 46946]]

that each containment spray nozzle is unobstructed. The frequency is 
changed from ``10 years'' to ``following maintenance which could result 
in nozzle blockage.''
    Date of issuance: July 31, 2006.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 99.
    Renewed Facility Operating License No. DPR-18: Amendment revised 
the Technical Specifications and the License.
    Date of initial notice in Federal Register: January 3, 2006 (71 FR 
154).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 31, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 6, 2004 (TS 428) as 
supplemented by letter dated June 16, 2005.
    Brief description of amendment: The amendment revised the reactor 
vessel Pressure-Temperature curves depicted in the Technical 
Specification (TS) Figure 3.4.9-1 and adds a new TS Figure 3.4.9-2.
    Date of issuance: July 26, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 256.
    Facility Operating License No. DPR-33: Amendment revised the TS.
    Date of initial notice in Federal Register: January 18, 2005 (70 FR 
2899). The supplement dated June 16, 2005, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 26, 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 15, 2005 (TS-05-09), as 
supplemented by letter dated June 7, 2006.
    Brief description of amendment: The amendment revises the Watts Bar 
Nuclear Plant (WBN) Technical Specification Surveillance Requirements 
to increase the minimum required average ice basket weight, thus, 
increasing the corresponding total weight of the stored ice in the WBN 
ice condenser. The changes to the ice basket and total ice weights are 
due to the additional energy associated with the Replacement Steam 
Generators.
    Date of issuance: July 25, 2006.
    Effective date: As of the date of issuance and shall be implemented 
prior to Mode 4 at startup to begin Cycle 8 fuel cycle.
    Amendment No. 62.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 14, 2006 (71 
FR 7814). The supplemental letter 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 July 25, 2006.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 28, 2006.
    Brief description of amendment: The amendment revised Technical 
Specification 5.0, ``Administrative Controls,'' by changing a position 
title and department name.
    Date of issuance: July 11, 2006.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days of the date of issuance.
    Amendment No.: 173.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 9, 2006 (71 FR 
27005).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 11, 2006.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

    Date of application for amendment: July 5, 2005, as supplemented by 
letters dated March 30, April 13, and May 11, 2006.
    Brief description of amendment: The amendments revised the 
Technical Specifications (TSs) to add a reference in TS 5.65.b, ``Core 
Operating Limits Report (COLR),'' to permit the use of an alternate 
methodology to perform a thermal-hydraulic analysis to predict the 
critical heat flux and departure from nucleate boiling ratio for the 
AREVA Advanced Mark-BW fuel in the North Anna 1 and 2 cores.
    Date of issuance: July 21, 2006.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 247, 227.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments 
changed the Licenses and the TSs.
    Date of initial notice in Federal Register: August 16, 2005 (70 FR 
48208). The supplements dated March 30, April 13, and May 11, 2006, 
contained clarifying information only and did not change the initial no 
significant hazards consideration determination or expand the scope of 
the initial application.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 21, 2006.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 8th day of August, 2006.

    For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 06-6921 Filed 8-14-06; 8:45 am]
BILLING CODE 7590-01-P