[Federal Register Volume 68, Number 189 (Tuesday, September 30, 2003)]
[Notices]
[Pages 56340-56350]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-24477]


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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 Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any

[[Page 56341]]

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 September 5, 2003, through September 18, 
2003. The last biweekly notice was published on September 18, 2003 (68 
FR 54747).

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.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing 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 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By October 30, 2003, 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 2.714, which is 
available at the Commission's PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, 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 factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide 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 to relief. A 
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, 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

[[Page 56342]]

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 with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland, by the above date. 
Because of continuing disruptions in delivery of mail to United States 
Government offices, it is requested that petitions for leave to 
intervene and requests for hearing be transmitted to the Secretary of 
the Commission either by means of facsimile transmission to 301-415-
1101 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 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and because of continuing disruptions in 
delivery of mail to United States Government offices, 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 filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    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 Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to 
[email protected].

    Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-
455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois.
    Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 
1 and 2, Will County, Illinois.
    Date of amendment request: June 11, 2003.
    Description of amendment request: The proposed amendment would add 
a license condition to increase the completion time (CT) from 72 hours 
to 144 hours required to restore a unit specific essential service 
water (SX) train to operable status. The proposed change would be a one 
time change applicable to Braidwood Station, Unit 1, and both units at 
Byron Station.
    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 changes have been evaluated using the risk informed 
processes described in RG 1.174, ``An Approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' dated July 1998 and RG 
1.177, ``An Approach for Plant-Specific, Risk-Informed 
Decisionmaking: Technical Specifications, ``dated August 1998. The 
risk associated with the proposed change was found to be acceptable.
    The previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The SX system is not 
considered an initiator for any of these previously analyzed events. 
The proposed change does not have a detrimental impact on the 
integrity of any plant structure, system, or component that 
initiated an analyzed event. No active or passive failure mechanisms 
that could lead to an accident are affected. The proposed change 
will not alter the operation of, or otherwise increase the failure 
probability of any plant equipment that initiates an analyzed 
accident. Therefore, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The unit-specific SX system consists of two separate, 
electrically independent, 100% capacity, safety related, cooling 
water trains. Each train consists of a 100% capacity pump, piping, 
valving, and instrumentation. The pumps and valves are remote and 
manually aligned, except in the unlikely event of a loss of coolant 
accident (LOCA). The pumps are automatically started upon receipt of 
a safety injection signal or an undervoltage on the engineered 
safety features (ESF) bus, and all essential valves are aligned to 
their post accident positions. The SX system is also the backup 
water supply to the auxiliary feedwater system and fire protection 
system.
    The design basis of the SX system is for one SX train, in 
conjunction with the component cooling water (CC) system and a 100% 
capacity containment cooling system, to remove core decay heat 
following a design basis LOCA as discussed in the UFSAR, Section 
6.2, ``Containment Systems.'' This prevents the containment sump 
fluid from increasing in temperature during the recirculation phase 
following a LOCA and provides for a gradual reduction in the 
temperature of this fluid as it is supplied to the reactor coolant 
system by the emergency core cooling system pumps. The SX system is 
designed to perform its function with a single failure or any active 
component, assuming the loss of offsite power. The proposed one-time 
increase in the CT of the operating unit's SX pump is consistent 
with the philosophy of the current Technical Specification LCO which 
allows one train of SX to be inoperable for 72 hours. This change 
only extends the 72 hour perspective; therefore, the proposed change 
does not involve a significant increase in the consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve the use or installation of 
new equipment and the currently installed equipment will not be 
operated in a new or different manner. No new or different system 
interactions are created and no new processes are introduced. The 
proposed changes will not introduce any new failure mechanisms, 
malfunctions, or accident initiators not already considered in the 
design and licensing bases. Based on this evaluation, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not alter any existing setpoints at 
which protective actions are initiated and no new setpoints or 
protective actions are introduced. The design and operations of the 
SX system remains unchanged. The risk associated with the proposed 
increase in the time an SX pump is allowed to be inoperable was 
evaluated using the risk informed processes described in RG 1.174, 
``An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing 
Basis,'' dated July 1998 and RG 1.177, ``An Approach for Plant-
Specific, Risk-Informed Decisionmaking: Technical Specifications,'' 
dated August 1998. The risk was shown to be acceptable. Based on 
this evaluation, the proposed change does not involve a significant 
reduction in a margin of safety.


[[Page 56343]]


    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. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

    Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-
455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois.
    Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 
1 and 2, Will County, Illinois.

    Date of amendment request: June 27, 2003.
    Description of amendment request: The proposed amendment would 
revise TS 3.4.10, ``Pressurizer Safety Valves,'' by changing the 
existing pressurizer safety valve (PSV) lift settings from `` \3\ 2460 
psig and [pound] 2510 psig,'' to `` \3\ 2411 psig and [pound] 2509 
psig.'' The existing TS represents a +/-1% tolerance band around a lift 
setting of 2485 psig. The proposed lift setting range of `` \3\ 2411 
psig and [pound] 2509 psig'' represents a +/-2% tolerance band around a 
lift setting of 2460 psig.
    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 TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Reanalysis/evaluations were performed to assess all transients 
that could be potentially impacted by the proposed PSV lift setting 
and tolerance band change. The proposed change in the PSV tolerance 
from +/-1% to +/-2% with a reduction in the lift setting from 2485 
psig to 2460 psig allows a decrease in the valve minimum opening 
pressure and therefore, provides earlier pressurizer relief and a 
reduced RCS pressure. The proposed change does not affect the 
maximum opening pressure assumed in the non-LOCA analyses since the 
proposed change in maximum PSV opening pressure is insignificant and 
in the conservative direction. Therefore, only those transients for 
which it is conservative to minimize the RCS pressure (i.e., DNB and 
pressurizer overfill concerns) are potentially impacted by the 
proposed change. The reanalyses/evaluations of all the affected 
transients demonstrated acceptable results with no significant 
increase in the probability or consequences.
    Further, any evaluations performed on an overpressure transient 
conservatively assume the upper limit of the PSV tolerance. The 
proposed change to the lower tolerance limit of the PSV lift setting 
means that an overpressure transient may be terminated at a pressure 
that is lower than assumed in the analysis. It has also been 
determined that the transient analyses are not adversely affected 
because the limiting transients are not sensitive to the pressure 
tolerance decrease. Therefore, the primary system pressure boundary 
is not challenged by the PSV lower tolerance limit change. The 
assumed maximum PSV lift setting was not changed, and therefore, 
does not impact analyses performed for overpressure transients. It 
has been determined that the design relieving capacity of the PSVs 
can still be met with the reduction in PSV setpoint. Except for the 
PSV lower lift setting and increased tolerance, the design and 
operation of the PSVs remains unchanged.
    Based on this analysis, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated in the Byron/Braidwood Stations 
Updated Final Safety Analysis Report.
    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 in the PSV tolerance from +/-1% to +/-2% 
with a reduction in the lift setting from 2485 psig to 2460 psig 
allows a decrease in the valve minimum opening pressure and 
therefore provides earlier pressurizer relief and a reduced RCS 
pressure. The proposed change does not affect the maximum opening 
pressure assumed in the accident analyses since the proposed change 
in maximum PSV opening pressure is insignificant and in the 
conservative direction. The pressurizer PORVs serve to minimize 
challenges to the PSVs. An assessment of the impact of reducing the 
PSV lift setpoint and increasing the tolerance has determined that 
the resulting margin is sufficient to ensure that the PORVs will 
actuate prior to the PSVs. Except for the PSV lower lift setting and 
increased tolerance, the design and operation of the PSVs remain 
unchanged.
    The proposed change does not involve the use or installation of 
new equipment and all currently installed equipment will not be 
operated in a new or different manner. No new or different system 
interactions are created and no new processes are introduced. The 
proposed change will not introduce any new failure mechanisms, 
malfunctions, or accident initiators not already considered in the 
design and licensing bases.
    Based on this evaluation, the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The PSVs provide, in conjunction with the reactor protection 
system, overpressure protection for the RCS. The PSVs are designed 
to prevent the system pressure from exceeding the system safety 
limit, 2735 psig, which is 110% of the design pressure. The change 
in the upper limit of the PSV tolerance from +1% to +2% with a 
reduction in the nominal setpoint from 2485 psig to 2460 psig does 
not challenge the upper limit of the overpressure protection. The 
change in PSV maximum opening lift setting is insignificant and in 
the conservative direction with respect to overpressure protection, 
therefore, the proposed change does not impact analyses performed 
for overpressure transients. For all non-LOCA events, the analyses/
evaluations support the change in PSV lift setting and tolerance 
from 2485 psig +/-1% to 2460 psig +/-2%. The LOCA analyses are not 
impacted because the transient results in a decrease in RCS pressure 
and therefore, will not challenge the PSV opening pressure lift 
setting. The change in the PSV lift setting and tolerance also has 
no effect on the reactor protection or engineered safety features 
systems trip set points. Thus, the proposed change does not involve 
a significant reduction in any margin of safety.
    Based on the above discussions, it has been determined that the 
requested TS change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated; or 
create the possibility of a new or different kind of accident from 
any accident previously evaluated; or 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. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

    FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-
Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio.
    Date of amendment request: August 25, 2003.
    Description of amendment request: The proposed amendment would 
revise the Steam and Feedwater Rupture Control System (SFRCS) 
instrumentation Technical Specifications (TSs) to clearly identify the 
appropriate actions to be taken if an SFRCS instrumentation channel's 
output logic becomes inoperable; relocate the SFRCS instrumentation 
trip setpoints from the TSs to the Updated Safety Analysis Report; and 
decrease the SFRCS instrument channel functional test frequency from 
monthly to quarterly and make associated changes to the trip setpoint 
allowable values.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 56344]]

As required by 10 CFR 50.91(a), the licensees have provided their 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not change any accident initiator, 
initiating condition, or assumption, and do not involve a 
significant change to plant design or operation. In addition, the 
proposed changes do not increase the likelihood of a malfunction of 
any plant structures, systems, or components, do not invalidate 
assumptions used in evaluating the radiological consequences of an 
accident, do not alter the source term or containment isolation, and 
do not provide a new radiation release path or alter radiological 
consequences. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not introduce a new or different 
accident initiator or introduce a new or different equipment failure 
mode or mechanism. 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 SFRCS instrumentation setpoint analyses will continue to 
adequately preserve the margin of safety. In addition, there are no 
new or significant changes to the initial conditions contributing to 
accident severity or consequences. 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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

    Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska.
    Date of amendment request: August 25, 2003.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) Surveillance Requirement (SR) 
3.3.2.1.4 and TS Table 3.3.2.1-1 for mathematical symbols and use of 
Allowable Values in the place of Analytical Limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change to the Cooper Nuclear Station (CNS) 
Technical Specifications (TS) corrects the mathematical symbols for 
the RBM [Rod Block Monitor] LPSP [Low Power Setpoint], IPSP 
[Intermediate Power Setpoint], and the HPSP [High Power Setpoint] to 
clarify the power ranges at which the RBM upscale trips are in 
affect. In addition, the change incorporates the use of Allowable 
Values in the place of Analytical Limits.
    Calculation NEC 98-024 Rev. 3, which documents the Analytical 
Limits and calculates the Allowable Values for the [RBM LPSP, IPSP, 
and HPSP] have not been altered. The calculation results implemented 
in procedures 6.1/2RBM.302 remain unchanged. The proposed TS change 
does not change or invalidate the Analytical Limits.
    Based on the above, NPPD [Nebraska Public Power District] 
concludes that the proposed TS change to modify the mathematical 
symbols in TS SR 3.3.2.1.4 and TS Table 3.3.2.1-1 footnotes (a), 
(b), (c), and (e) does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change to the [CNS TS] corrects the mathematical 
symbols for the RBM LPSP, IPSP, and the HPSP to clarify the power 
ranges at which the RBM upscale trips are in affect. In addition, 
the change incorporates the use of Allowable Values in the place of 
Analytical Limits. The values for the RBM trip setpoints, Analytical 
Limits, and Allowable Values are not being altered in any way.
    Based on the above, NPPD concludes that the proposed TS change 
to modify the mathematical symbols in TS SR 3.3.2.1.4 and TS Table 
3.3.2.1-1 footnotes (a), (b), (c), and (e) does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    The proposed change to the [CNS TS] corrects the mathematical 
symbols for the RBM LPSP, IPSP, and the HPSP to clarify the power 
ranges at which the RBM upscale trips are in affect. In addition, 
the change incorporates the use of Allowable Values in the place of 
Analytical Limits. This TS change does not change any Analytical 
Limits or Allowable Value calculations. The methodology by which the 
RBM Trip Setpoints, Analytical Limits, and Allowable Values are 
derived has not changed.
    Based on the above, NPPD concludes that the proposed TS change 
to modify the mathematical symbols in TS SR 3.3.2.1.4 and TS Table 
3.3.2.1-1 footnotes (a), (b), (c), and (e) does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

    Nine Mile Point Nuclear Station, LLC (NMPNS), Docket No. 50-410, 
Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York.
    Date of amendment request: August 22, 2003.
    Description of amendment request: The licensee proposed to revise 
Section 3.7.1, ``Service Water (SW) System and Ultimate Heat Sink 
(UHS),'' of the Technical Specifications (TS) to allow continued 
operation with short-term elevated UHS temperatures. The proposed 
revision is based on an NRC-approved Technical Specification Task Force 
(TSTF) Standard Technical Specification change, identified as TSTF-330, 
``Allowed Outage Time--Ultimate Heat Sink,'' Revision 3, dated October 
16, 2000. Adoption of TSTF-330 would allow continued plant operation 
with UHS temperatures that temporarily exceed the 82 [deg]F limit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change allows plant operation to continue if the 
temperature of the UHS exceeds the TS limit of 82 [deg]F provided 
that (1) the water temperature, averaged over the previous 24 hour 
period, is at or below 82 [deg]F, and (2) the UHS temperature is 
less than or equal to 84 [deg]F. This increase in UHS temperature 
will not affect the normal operation of the plant to the extent that 
it would make any accident more likely to occur. The UHS is not an 
accident initiator. In addition, the proposed change assures 
adequate margin in the safety systems

[[Page 56345]]

and safety-related heat exchangers to meet the design safety 
functions at the higher temperature. Thus, the proposed change will 
have no adverse effect on plant operation, or the availability or 
operation of any accident mitigation equipment. Furthermore, the 
proposed change cannot cause an accident, nor will the change 
significantly affect the plant response to any accidents. Therefore, 
there will be no 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 will not alter the current plant 
configuration (no new or different type of equipment will be 
installed) or require any new or unusual operator actions. The 
proposed change will not alter the way any structure, system, or 
component functions and will not cause an adverse effect on plant 
operation or accident mitigation equipment. The response of the 
plant and the operators following a design-basis accident is 
unaffected by the change. The proposed change does not introduce any 
new failure modes and the design basis heat removal capability of 
the affected safety-related components is maintained at the 
increased UHS temperature limit. Therefore, the proposed change will 
not create the possibility of a new or different kind of accident 
from any previously analyzed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    NMPNS has performed an evaluation of the safety systems to 
ensure their safety functions can be met with a UHS water 
temperature of 84 [deg]F. The higher UHS temperature represents a 
slight reduction in the margins of safety in terms of these systems' 
abilities to remove accident heat loads. As part of the evaluation, 
however, it was verified that these safety systems will still be 
capable of performing their design-basis functions. The proposed 
change will have no adverse effect on plant operation or equipment 
important to safety. The plant responses to accidents will not be 
significantly affected and the accident mitigation equipment will 
continue to function as assumed in the accident analysis. Therefore, 
there will be no 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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

    Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine 
Mile Point Nuclear Station, Unit 2, Oswego County, New York.
    Date of amendment request: August 28, 2003.
    Description of amendment request: The licensee proposed to 
change Section 3.1.7, ``Standby Liquid Control (SLC) System,'' of 
the Technical Specifications (TS) to raise the required average 
boron concentration in the reactor, resulting from injection of 
sodium pentaborate solution by the SLC system, to support a 
transition to the General Electric (GE) 14 fuel design. This design 
change includes the use of sodium pentaborate solution enriched with 
the boron-10 isotope. The proposed amendment would add a new 
surveillance requirement to verify the required boron-10 enrichment 
of the sodium pentaborate solution prior to addition to the SLC 
tank. It would also revise the figure that depicts acceptable values 
of SLC storage tank volume and sodium pentaborate solution 
concentration by adding a notation regarding the required boron-10 
enrichment, and by making a minor adjustment to one of the 
coordinates that define the Acceptable Operation region on the 
figure.
    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 SLC system is designed to provide sufficient negative 
reactivity to bring the reactor from full power to a subcritical 
condition at any time in a fuel cycle, without taking credit for 
control rod movement. The proposed changes to the SLC sodium 
pentaborate solution requirements maintain the capability of the SLC 
system to perform this reactivity control function, and assure 
continued compliance with the requirements of 10 CFR 50.62 for 
anticipated transients without scram (ATWS). The SLC system is 
provided to mitigate ATWS events and, as such, is not considered to 
be an initiator of the ATWS event or any other analyzed accident. 
The use of sodium pentaborate solution enriched with the Boron-10 
isotope, which is chemically and physically similar to the current 
solution, does not alter the design or operation of the SLC system 
or increase the likelihood of a system malfunction that could 
increase the consequences of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Injection of sodium pentaborate solution into the reactor vessel 
has been considered in the plant design. The proposed changes revise 
the SLC boron solution requirements such that the capability of the 
SLC system to bring the reactor to a subcritical condition without 
taking credit for control rod movement is maintained, considering 
operation with an equilibrium core of GE14 fuel. The use of sodium 
pentaborate solution enriched with the Boron-10 isotope, which is 
chemically and physically similar to the current solution, does not 
alter the design, function, or operation of the SLC system. The 
correct Boron-10 enrichment is assured by the proposed revisions to 
the TS surveillance requirements. The impact on the solubility limit 
of enriching the sodium pentaborate solution with the Boron-10 
isotope is insignificant; thus, the existing minimum solution and 
piping temperature specified in the TS will ensure that the boron 
remains in solution and does not precipitate out in the SLC storage 
tank or in the SLC pump suction piping. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes revise the SLC boron solution requirements 
to maintain the capability of the SLC system to bring the reactor to 
a subcritical condition without taking credit for control rod 
movement. These changes support operation with an equilibrium core 
of GE14 fuel and assure continued compliance with the requirements 
of 10 CFR 50.62. The minimum required average boron concentration in 
the reactor core, resulting from the injection of sodium pentaborate 
solution by the SLC system, has been determined using approved 
analytical methods. The analysis demonstrates that sufficient 
shutdown margin is maintained in the reactor such that the 
reactivity control function of the SLC system is assured. The 
additional quantity of boron included to allow for imperfect mixing 
and leakage is being increased from 20 percent to 25 percent. Thus, 
additional safety margin is provided to bring the reactor 
subcritical in the event of an ATWS. 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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.
    Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York.
    Date of amendment request: September 3, 2003.
    Brief description of amendments: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
Technical Specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee

[[Page 56346]]

performs a risk assessment and manages risk consistent with the program 
in place for complying with the requirements of 10 CFR 50.65(a)(4). 
Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual 
TS would be eliminated, and Surveillance Requirement (SR) 3.0.4 revised 
to reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
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 April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated September 3, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

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

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in [a] Margin of Safety

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS LCO. The 
risk associated with this allowance is managed by the imposition of 
required actions that must be performed within the prescribed 
completion times. The net effect of being in a TS condition on the 
margin of safety is not considered significant. The proposed change 
does not alter the required actions or completion times of the TS. 
The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing risk. The net change to the margin of safety 
is insignificant. Therefore, this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

    South Carolina Electric & Gas Company (SCE&G), South Carolina 
Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear 
Station, Unit No. 1, Fairfield County, South Carolina.
    Date of amendment request: July 23, 2003.
    Description of amendment request: The proposed change will revise 
the near-end of life (EOL) Moderator Temperature Coefficient (MTC) 
Surveillance Requirement 4.1.1.3.b by placing a set of conditions on 
core operation, which if met, would allow exemption from the required 
MTC measurement. The conditional exemption will be determined on a 
cycle-specific basis by considering the margin predicted to the 
surveillance requirement MTC limit and the performance of other core 
parameters, such as beginning of life MTC measurements and the critical 
boron concentration as a function of cycle length. The conditional 
exemption will improve plant availability and minimize disruptions to 
normal plant operations. Plant safety criteria will not be compromised 
by the conditional exemption of this one measurement.
    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?
    The probability or consequences of accidents previously 
evaluated in the VCSNS FSAR [Final Safety Analysis Report] are 
unaffected by this proposed change because there is no change to any 
equipment response or accident mitigation scenario. There are no 
additional challenges to fission product barrier integrity. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The proposed change does not challenge the performance or integrity 
of any safety-related system. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Does this change involve a significant reduction in margin of 
safety?
    The margin of safety associated with the acceptance criteria of 
any accident is unchanged. The proposed change will have no effect 
on the availability, operability, or performance of the safety-
related systems and components. A change to the surveillance 
requirement is proposed, but the limiting conditions for operation 
required by TS [technical specifications] are not changed.
    The TS Bases are founded in part on the ability of the 
regulatory criteria to be satisfied assuming the limiting conditions 
for operation are met for the various systems. Conformance to 
regulatory criteria for operation with the conditional exemption 
from the near-EOL MTC measurement is demonstrated and the regulatory 
limits are not exceeded. Therefore, the margin of safety as defined 
in the TS is not reduced and the proposed change does not involve a 
significant reduction in a margin of safety.
    Pursuant to 10 CFR 50.91, the preceding analyses provide a 
determination that the

[[Page 56347]]

proposed Technical Specifications change poses no significant hazard 
as delineated by 10 CFR 50.92.

    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: Thomas G. Eppink, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Section Chief: John A. Nakoski.

    Southern California Edison Company, et al., Docket No. 50-361, San 
Onofre Nuclear Generating Station, Unit 2, San Diego County, 
California.
    Date of amendment requests: August 26, 2003.
    Description of amendment requests: The proposed change would revise 
Technical Specifications (TS) 1.1 ``Definitions,'' 3.4 ``Reactor 
Coolant System [RCS],'' and 5.7 ``Reporting Requirements. Specifically, 
the licensee requests to relocate the RCS pressure-temperature curves 
and limits from the TSs to a licensee-controlled document identified as 
the PTLR [Pressure and Temperature Limits Report].
    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.
    Updating the Reactor Coolant System (RCS) pressure and 
temperature curves and limits in accordance with 10 CFR [Part] 50 
Appendices G and H ensures the reactor coolant system's pressure 
boundary integrity will be protected until End Of Life (EOL) and 
does not contribute to the probability of or the initiation of 
accidents. There is no change to the safety analysis.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    These changes are required to maintain the RCS pressure boundary 
integrity until EOL. Changes to the RCS pressure and temperature 
curve and limits will not create a new or different kind of 
accident. There is no change to the safety analysis.
    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.
    Pressure and temperature curves and limits are provided as 
limits to plant operation for ensuring RCS pressure boundary 
integrity is maintained until EOL. No margin of safety is impacted 
by changes to the RCS pressure and temperature curves and limits. 
There is no change to the safety analysis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SCE 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 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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

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

    Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan.
    Date of amendment request: September 3, 2003.
    Brief description of amendment request: The proposed amendment 
would revise Technical Specification (TS) Limiting Condition for 
Operation (LCO) 3.6.5.1.d to replace the phrase ``Each ice basket'' 
with the phrase ``Ice baskets.'' This change would make the LCO 
consistent with associated TS Surveillance Requirement (SR) 4.6.5.1.b.2 
and would allow the SR to define the detailed requirements for ice 
basket weight.
    Date of publication of individual notice in Federal Register: 
September 10, 2003 (68 FR 53402).
    Expiration date of individual notice: October 10, 2003.

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, 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

[[Page 56348]]

have access to ADAMS or if there are problems in accessing the 
documents located in ADAMS, contact the NRC Public Document Room (PDR) 
Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to 
[email protected].

    Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina.
    Date of application of amendments: February 19, 2003.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) 5.5.10, ``Steam Generator (SG) Tube 
Surveillance Program.'' Specifically, the proposed changes would revise 
the SG surveillance requirements in the Oconee Units 1, 2, and 3 TSs. 
Since steam generator replacement outages are respectively scheduled 
for Fall 2003, Spring 2004, and Fall 2004, the licensee proposes to 
relocate the program requirements applicable to the original SGs, 
existing TS 5.5.10 requirements, to TS 5.5.21 and to provide program 
requirements applicable to the replacement SGs, in TS 5.5.10.
    Date of Issuance: September 4, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 334, 334, & 335.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 18, 2003 (68 FR 
12949).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 4, 2003.
    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, 2002, as 
supplemented by letter dated July 18, 2003.
    Brief description of amendment: The proposed amendment extends the 
allowed outage time (AOT) for a single inoperable low pressure safety 
injection (LPSI) train from 72 hours to 7 days. In addition, an AOT of 
72 hours is included for other conditions where the equivalent of a 
single emergency core cooling system (ECCS) subsystem flow is still 
available to both the LPSI and high pressure safety injection (HPSI) 
trains. Also, an action statement is added to restore at least one of 
each HPSI and LPSI train to operable status within one hour if 100% of 
ECCS flow is unavailable due to two inoperable HPSI or LPSI trains.
    Date of issuance: September 11, 2003.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 251.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2002.
    The July 18, 2003, supplemental letter provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 11, 2003.
    No significant hazards consideration comments received: No.

    Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois.
    Date of application for amendments: December 20, 2002, as 
supplemented May 30, 2003.
    Brief description of amendments: The amendments revise the 
licensing basis as described in the Updated Final Safety Analysis 
Report to implement the Boiling-Water Reactor Vessel and Internals 
Project reactor pressure vessel integrated surveillance program as the 
basis for demonstrating compliance with the requirements of Appendix H 
to 10 CFR part 50.
    Date of issuance: August 28, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 217/211.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the licensing basis.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5669).
    The supplement dated May 30, 2003, 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. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated August 28, 2003.
    No significant hazards consideration comments received: No.

    FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 
and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 
2), Beaver County, Pennsylvania.
    Date of application for amendments: June 5, 2002, as supplemented 
August 19 and December 2, 2002, and January 30, February 14, March 19 
and 31, June 6 and 24, and September 5, 2003.
    Brief description of amendments: The amendments approved selective 
implementation of an alternative source term methodology for the loss-
of-coolant accident (LOCA) and the control rod ejection accident 
(CREA), incorporation of ARCON96 methodology for release points 
associated with the LOCA and CREA, elimination of the control room 
emergency bottled air pressurization system, changes to the control 
room emergency ventilation system (CREVS), and a change to the BVPS-1 
CREVS filter bypass leakage acceptance test criteria.
    Date of issuance: September 10, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 257 and 139.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75876). The supplements dated August 19 and December 2, 2002, and 
January 30, February 14, March 19 and 31, June 6 and 24, and September 
5, 2003, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed except as noted below, and did not change the staff's original 
proposed no significant hazards consideration determination. The 
February 14, 2003, submittal requested the scope of the review be 
expanded by including in the scope of the review related Updated Final 
Safety Analysis Report (UFSAR) page changes, but this request was 
withdrawn in the March 31, 2003, submittal. Additionally, a portion of 
the requested review was withdrawn in the March 19, 2003, submittal, as 
these changes were no longer necessary. The portion of the proposed 
application related to conversion of the BVPS-1 and 2 containments from 
subatmospheric to atmospheric operating conditions was withdrawn by 
letter dated September 5, 2003.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 10, 2003.
    No significant hazards consideration comments received: No.

    FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-
Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio.

[[Page 56349]]

    Date of application for amendment: May 14, 2003, as supplemented by 
letters dated June 16, August 2, August 7, and August 20, 2003.
    Brief description of amendment: This amendment revised the 
Technical Specifications to allow a one time exception, only during the 
Restart Test Plan, to allow entry into Mode 3 of operation without the 
high-pressure injection pumps being able of taking suction from the 
low-pressure injection trains when aligned for containment sump 
recirculation. The exception cannot be used for entry into Mode 2 or 
Mode 1.
    Date of issuance: September 5, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 257.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34668).
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 5, 2003.
    No significant hazards consideration comments received: No.

    Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida.
    Date of application for amendment: February 17, 2003.
    Brief description of amendment: The amendment revises Technical 
Specification (ITS) 3.6.3 ``Containment Isolation Valves,'' to allow 
verification by administrative means of isolation devices in high 
radiation areas, and isolation devices that are locked, sealed or 
otherwise secured.
    Date of issuance: September 8, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 209.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18277).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 8, 2003.
    No significant hazards consideration comments received: No.

    FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
No. 1, Rockingham County, New Hampshire.
    Date of amendment request: October 11, 2002, as supplemented by 
letter dated May 29, 2003.
    Description of amendment request: The amendment revises Technical 
Specification (TS) 3.4.9.1, ``Reactor Coolant System [RCS]--Pressure/
Temperature Limits,'' and TS 3.4.9.3, ``Reactor Coolant System--
Overpressure Protection Systems'' and their associated Bases sections. 
Specifically, the changes replace TS Figures 3.4-2 ``Reactor Coolant 
System Heatup Limitations,'' 3.4-3 ``Reactor Coolant System Cooldown 
Limitations,'' and 3.4-4 ``RCS Cold Overpressure Protection'' to allow 
operation to 20 Effective Full Power Years.
    Date of issuance: September 11, 2003.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 89.
    Facility Operating License No. NPF-86: Amendment revises the TS.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75879). The May 29, 2003, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination nor expand the amendment beyond the scope 
of the initial notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 11, 2003.
    No significant hazards consideration comments received: No.

    FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
No. 1, Rockingham County, New Hampshire.
    Date of amendment request: April 24, 2002.
    Description of amendment request: The amendment revises 
surveillance requirements (SRs) in Technical Specification (TS) 
4.6.2.1, ``Containment Spray System,'' and TS 4.7.1.2.1b, ``Auxiliary 
Feedwater System,'' and associated Bases Section 3/4.7.1.2. 
Specifically, the proposed changes would move SR acceptance criteria 
for containment spray and auxiliary feedwater pumps from the TSs to the 
Seabrook Station Technical Requirements Manual.
    Date of issuance: September 12, 2003.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 90.
    Facility Operating License No. NPF-86: The amendment revises the 
TSs.
    Date of initial notice in Federal Register: June 11, 2002 (67 FR 
40024).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 12, 2003.
    No significant hazards consideration comments received: No.

    Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin
    Date of application for amendments: September 12, 2002, as 
supplemented March 27 and May 30, 2003.
    Brief description of amendments: The amendments add surveillance 
requirements for Technical Specification (TS) 3.5.2, ``ECCS--
Operating,'' and TS 3.5.3, ``ECCS--Shutdown,'' to verify, every 31 
days, that the emergency core cooling system piping is full of water.
    Date of issuance: September 5, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 209 and 214.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5679).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 5, 2003.
    No significant hazards consideration comments received: No.

    Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama.
    Date of application for amendments: February 13, 2003, as 
supplemented April 14, 2003.
    Description of amendment request: The amendments revised Technical 
Specification (TS) 4.2.1 ``Fuel assemblies,'' to modify the fuel design 
description to encompass Framatome Advanced Nuclear Power fuel 
assemblies, and also to modify TS 4.3 ``Fuel Storage,'' to remove 
nomenclature specific to Global Nuclear Fuels analysis methods.
    Date of issuance: September 5, 2003.
    Effective date: September 5, 2003.
    Amendment Nos.: 247, 284, 242.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68. 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 1, 2003 (68 FR 
15763). The April 14, 2003, letter provided clarifying information that 
did not

[[Page 56350]]

change the scope of the original request or the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated September 5, 2003.
    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: May 1, 2003, as supplemented on 
July 8, 2003.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.8.7, ``Inverters--Operating.'' The revised TS 
requires only one inverter for each of the four 120V AC Vital 
Instrument channels. This amendment is the initial phase of a project 
that will update the 120V AC Vital Instrument Power System.
    Date of issuance: September 8, 2003.
    Effective date: As of the date of issuance and shall be implemented 
prior to Mode 4 entry following the next refueling outage in the fall 
of 2003.
    Amendment No.: 45.
    Facility Operating License No. NPF-90: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28859). The supplemental letter provided clarifying information that 
did not expand the scope of the original request 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 September 8, 2003.
    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: June 2, 2003.
    Brief description of amendment: The amendment revises the technical 
specifications (TSs) to increase the specified minimum fuel oil 
inventories maintained in the fuel oil storage tanks for the diesel 
generators.
    Date of issuance: September 9, 2003.
    Effective date: September 9, 2003, and shall be implemented within 
60 days from the date of issuance.
    Amendment No.: 156.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 22, 2003 (68 FR 
43393).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 9, 2003.
    No significant hazards consideration comments received: No.

    Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 
50-281, Surry Power Station, Units 1 and 2, Surry County, Virginia.
    Date of application for amendments: September 5, 2002, as 
supplemented on April 16, June 9, and July 7, 2003.
    Brief Description of amendments: These amendments revise the 
Technical Specifications to add provisions to permit inspection and 
related repair of a buried fuel oil storage tank during plant operation 
by extending the allowed outage time for a buried fuel oil storage tank 
to 7 days from 24 hours for this purpose.
    Date of issuance: September 10, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment Nos.: 236 and 235.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: 
Amendments change the Technical Specifications.
    Date of initial notice in Federal Register: August 5, 2003 (68 FR 
46247).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 10, 2003.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 22nd day of September 2003.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 03-24477 Filed 9-29-03; 8:45 am]
BILLING CODE 7590-01-P