[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