Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 46929-46946 [06-6921]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 157 (Tuesday, August 15, 2006)]
[Notices]
[Pages 46929-46946]
From the Federal Register Online via the Government Printing 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, https://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, HearingDocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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 analy