Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

Note: EPA no longer updates this information, but it may be useful as a reference or resource.
 [Federal Register: August 9, 2000 (Volume 65, Number 154)]
[Notices]
[Page 48744-48767]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr09au00-129]

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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 amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or
proposed to be issued from July 15, 2000, through July 28, 2000. The
last biweekly notice was published on July 26, 2000.

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
Review and Directives Branch, Division of Freedom of Information and
Publications 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 NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
    By September 8, 2000, 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 Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC, and electronically from
the ADAMS Public Library component on the NRC Web site, http://
www.nrc.gov (the Electronic Reading Room). 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

[[Page 48745]]

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 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: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and 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 Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site, http://www.nrc.gov (the
Electronic Reading Room).

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

    Date of amendment request: June 5, 2000.
    Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.7.8 to change the Required
Actions and Completion Times for the Ultimate Heat Sink (UHS) in the
event the service water (SW) temperature exceeds the 97 deg.F
surveillance acceptance 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:

    Carolina Power & Light (CP&L) Company has evaluated the proposed
Technical Specification change and has concluded that it does not
involve a significant hazards consideration. The CP&L conclusion is
in accordance with the criteria set forth in 10 CFR 50.92. The bases
for the conclusion that the proposed change does not involve a
significant hazards consideration are discussed below.
    1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
    The proposed change does not involve any physical alteration of
plant systems, structures or components. The proposed change
provides Required Actions for the plant condition where SW
temperature exceeds the TS limit. The SW system temperature is not
assumed to be an initiating condition of any accident analysis
evaluated in the safety analysis report (SAR). Therefore, the
revised limitations for SW temperature to be in excess of the design
limit does not involve an increase in the probability of an accident
previously evaluated in the safety analysis report. The SW system
supports operability of safety-related systems used to mitigate the
consequences of an accident. Plant equipment has been analyzed and
determined able to perform its safety-related function at [an] SW
temperature of 99 deg.F. Performance of the containment has been
analyzed in support of Amendment No. 187 to Technical Specifications
assuming 100 deg.F service water temperature and the results were
acceptable. The magnitude of any increase in SW temperature in
excess of the TS limit is expected to be small based on historical
data and experience for the UHS. An evaluation would be performed to
assure required cooling capability. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated in the SAR.
    2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical alteration of
plant systems, structures or components. The temperature of the SW
when near or slightly above the design temperature does not
introduce new failure mechanisms for systems, structures or
components not already considered in the SAR. Therefore, the
possibility of a new or different kind of accident from any accident
previously evaluated is not created.
    3. Does this change involve a significant reduction in a margin
of safety?
    The proposed change will not allow continued operation with the
SW temperature above the design basis limit. The proposed change
will allow continued operation provided the required cooling
capacity is verified and periodic monitoring is invoked to verify
the SW temperature remains less than or equal to 99 deg.F. Design
margins are affected which are associated with systems, structures
and components which are cooled by the SW system, and system
temperature is an input assumption for mitigating the effects of a
DBA [design-basis accident]. However, allowing SW temperature to
exceed the surveillance acceptance limit, as long as required
cooling is verified, will not significantly reduce the margin of
safety associated with this proposed change.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602 .
    NRC Section Chief: Richard P. Correia.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: May 31, 2000.
    Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) to delete the requirement to
remove the Reactor Protection System (RPS) circuitry shorting links
from TS Section 3/4.3.1, ``Reactor Protection System Instrumentation,''
3/4.9.2, ``Refueling Operations Instrumentation,'' and 3/4.10.3,
``Shutdown Margin Demonstrations,'' and to increase the required
signal-to-noise ratio for the source range monitor in (SRM) TS Sections
3/4.3.7.6, ``Source Range Monitors,'' and 3/4.9.2.

[[Page 48746]]

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

    Does the change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
    The proposed changes to TS Section 3/4.3.1, 3/4.9.2, and 3/
4.10.3 will relocate the requirement that the shorting links be
removed from the RPS circuitry prior to and during specified plant
conditions. The removal or installation of the RPS circuitry
shorting links does not have an effect on the probability of any
accident previously evaluated. The proposed changes to TS Sections
3/4.3.7.6 and 3/4.9.2 will increase the minimum signal-to-noise
ratio from  2:1 to  20:1, when the SRM count
rate is greater than or equal to 0.7 counts per second (cps) and
less than 3 cps.
    The operation of the SRM does not have an effect on the
probability of any accident previously evaluated. Thus, the
probability of any accident previously evaluated is not increased.
    The proposed changes do not affect the integrity of the fuel
cladding, reactor coolant system or secondary containment, because
no credit is taken in the current accident analyses for removal of
the RPS circuitry shorting links. Thus, the radiological
consequences of any accident previously evaluated are not increased.
    Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
    Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
    The proposed changes do not affect the assumed accident
performance of any LaSalle County Station structure, system or
component previously evaluated because accidents previously
evaluated assumed that the RPS circuitry shorting links were
installed and did not credit SRM operation. The proposed changes do
not introduce any new modes of system operation or failure
mechanisms.
    Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
    Does the change involve a significant reduction in a margin of
safety?
    The proposed changes to TS Sections 3/4.3.1, 3/4.9.2, and 3/
4.10.3 will relocate the requirement that the shorting links be
removed from the RPS circuitry prior to and during specified plant
conditions. The removal of the RPS circuitry shorting links in
Operations Condition 5, ``Refueling,'' modifies the RPS by
reconfiguring the scram signal for the intermediate range monitors
(IRMs) and average power range monitors (APRMs) to non-coincidental
and enabling the SRM non-coincidental high flux scram signal.
However, the SRM non-coincidental high flux scram signal is not
credited in any Design Basis Accident (DBA) and the IRM and APRM
one-out-of-two taken twice full scram provides the credited
protection with respect to safety analysis.
    Refueling interlocks and shutdown margin requirements ensure
that the reactor is maintained in a subcritical condition in
Operational Condition 5. The refueling interlocks are required to be
operable by TS Section 3/4.9.1, ``Reactor Mode Switch.'' The SRM,
IRM, and APRM control rod withdrawal block interlocks are not
affected by the removal or installation of the RPS circuitry
shorting links. Although shutdown margin may not yet have been
demonstrated in Operational Condition 5, shutdown margin
calculations performed prior to altering the reactor core, along
with procedural compliance for any Core Alterations, provides
indication that shutdown margin is available.
    The proposed changes to relocate the description and function of
the RPS circuitry shorting links to the UFSAR and be controlled in
accordance with the requirements of 10 CFR 50.59, are consistent
with the requirements of 10 CFR 50.36, ``Technical Specifications.''
The existing TS requirements to remove the RPS circuitry shorting
links do not satisfy any of the four criteria of 10 CFR 50.36 for
inclusion of a requirement into the TS. In accordance with NRC
guidance, existing TS requirements that do not satisfy the criteria
of 10 CFR 50.36 can be removed from the TS and relocated to other
controlled documents, such as the UFSAR. Changes to the LaSalle
County Station UFSAR are controlled in accordance with the
requirements of 10 CFR 50.59.
    The proposed changes to TS Sections 3/4.3.7.6 and 3/4.9.2 will
increase the statistical neutron monitoring confidence that the
indicated signal is correct when the SRMs indicate in the range form
0.7 cps to 3 cps. A SRM signal-to-noise ratio of  2:1
provides a statistical neutron monitoring confidence of 95% that the
indicated signal is correct with a minimum count rate of 3 cps. A
study was performed which concluded that a SRM signal-to-noise ratio
of 20:1 is required to provide a statistical neutron
monitoring confidence of 95% that the indicated signal is correct at
0.7 cps.
    Thus, the proposed changes do not involve a significant
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: December 27, 1999.
    Description of amendment request: The proposed amendment would
revise the technical specifications to increase the allowable out-of-
service times and surveillance test intervals for selected actuation
instrumentation.
    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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
    The proposed TS [technical specification] changes increases the
Allowable Outage Times and Surveillance Test Intervals (AOT/STI) for
actuation instrumentation based on analyses developed and approved
by the Nuclear Regulatory Commission (NRC). TS requirements that
govern operability or routine testing of plant instruments are not
assumed to be initiators of any analyzed event because these
instruments are intended to prevent, detect, or mitigate accidents.
Therefore, these changes will not involve an increase in the
probability of occurrence of an accident previously evaluated.
Additionally, these changes will not increase the consequences of an
accident previously evaluated because the proposed changes do not
involve any physical changes to plant systems, structures or
components (SSCs), or the manner in which these SSCs are operated.
These changes will not alter the operation of equipment assumed to
be available for the mitigation of accidents or transients by the
plant safety analysis or licensing basis. As justified and approved
in the AOT/STI licensing topical reports, the proposed changes
establish or maintain adequate assurance that components are
operable when necessary for the prevention or mitigation of
accidents or transients and that plant variables are maintained
within limits necessary to satisfy the assumptions for initial
conditions in the safety analyses. The proposed changes establish or
modify time limits allowable for operation with inoperable
instrument channels based on analyses which have been approved by
the NRC. Furthermore, there will be no change in the types or
significant increase in the amounts of any effluents released
offsite. For these reasons, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
    2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
    The proposed changes do not involve any physical changes to
SSCs, or the manner in which these SSCs function. Therefore, these
changes will not create the possibility of a

[[Page 48747]]

new or different kind of accident from any accident previously
evaluated. The changes in methods governing normal plant operation
are consistent with the current safety analysis assumptions.
Therefore, these changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin
of safety?
    The proposed changes increase the STIs and AOTs for actuation
instrumentation based on generic analyses completed by the Boiling
Water Reactor Owners' Group (BWROG). The NRC has reviewed and
approved the generic studies and has concurred with the BWROG that
the proposed changes do not significantly affect the probability of
failure or availability of the affected instrumentation systems. The
analysis determined that there is no significant change in the
availability and/or reliability of instrumentation as a result of
the proposed changes in STIs and AOTs. Furthermore, the change to
increase the frequency of the reactor protection system scram
contactor testing has been shown to improve plant safety. ComEd has
determined these studies are applicable to Quad Cities Nuclear Power
Station, Units 1 and 2. The proposed changes to AOTs provide
realistic times to complete required testing and maintenance actions
without increasing the overall instrument failure frequency.
Likewise, the extended STIs do not result in significant changes in
the probability of instrument failure. Furthermore, the proposed
changes will reduce the probability of test-induced plant transients
and equipment failures. Therefore, it is concluded that the proposed
changes will not result in a reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: December 30, 1999.
    Description of amendment request: The proposed amendment would
revise the technical specifications to (1) remove the Main Steam Line
Radiation Monitor (MSLRM) scram and main steam line isolation
functions, and (2) add a new requirement for the MSLRM mechanical
vacuum pump trip function.
    Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

    Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
    This proposed change involves the removal of existing Main Steam
Line Radiation Monitor (MSLRM) scram and the MSLRM MSL [main steam
line] Valve closure signal. The purpose of the MSLRM reactor scram
and the MSL isolation signal is to mitigate the radiological effects
of a fuel element failure. These functions do not serve as
initiators for any of the accidents evaluated in Chapter 15 of the
Updated Final Safety Analysis Report (UFSAR). Removal of these
functions will not increase the probability of any of the accidents
previously evaluated.
    The radiological effects of a Control Rod Drop Accident (CRDA)
have been evaluated for the Boiling Water Reactor Owners' Group
(BWROG) by General Electric (GE) in Report NEDO-31400A, ``Safety
Evaluation For Eliminating the Boiling Water Reactor Main Steam
Isolation Valve Closure Function and Scram Function of the Main
Steam Line Radiation Monitor.'' The GE report was evaluated by the
NRC and found acceptable by letter dated May 15, 1991, ``Acceptance
for Referencing of Licensing Topical Report NEDO-31400.'' The NRC
Safety Evaluation Report accepting the GE report required licensees
to demonstrate that the assumptions of the GE report analysis were
bounding for their plants. ComEd has evaluated the GE analysis for
applicability to Quad Cities Nuclear Power Station, Units 1 and 2.
    The GE analysis demonstrates that operation with the proposed
change does not represent a significant increase in the consequences
of a CRDA. Therefore, operation of Quad Cities Nuclear Power
Station, Units 1 and 2, under the proposed change does not represent
a significant increase in the probability or consequences of an
accident previously evaluated. A site specific radiological
evaluation was completed to confirm the applicability of the generic
GE analysis to Quad Cities Nuclear Power Station.
    Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
    This proposed change involves the removal of the existing MSLRM
scram and the MSL Valve closure input from the MSL Tunnel High
Radiation signal. Removal of these functions does not represent a
change in operating parameters for Quad Cities Nuclear Power
Station, Units 1 and 2. Removal of these functions does not add any
additional hardware and does not represent any new failure modes.
Operation of Quad Cities Nuclear Power Station, Units 1 and 2, under
the proposed change does not create the possibility of a new or
different type of accident previously evaluated.
    Does the change involve a significant reduction in a margin of
safety?
    The proposed change involves the elimination of the MSLRM scram
and the MSL Valve closure input from the MSL Tunnel High Radiation
signal. Operation under the proposed change will not change any
plant operation parameters, nor any protective system setpoints
other than removal of these functions. The GE report has
demonstrated that the consequences of the CRDA without the MSLRM
High scram and MSL Valve closure signal from the MSL Tunnel
Radiation detector results in doses which are well within 10 CFR
part 100, ``Reactor Site Criteria,'' limits. Therefore, the proposed
change 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
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: June 5, 2000.
    Description of amendment request: The proposed amendment, which
changes the Perry Nuclear Power Plant as described in the Updated
Safety Analysis Report, modifies the circuitry to the Reactor Core
Isolation Cooling (RCIC) System initiation logic. The proposed circuit
modification will include a time delay to the main turbine and
feedwater pump turbine trip signal associated with a RCIC system
automatic initiation. The addition of this time delay will prevent
potential main turbine and feedwater pump turbine trips that result in
unnecessary reactor scrams from inadvertent RCIC initiations.
    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 Reactor Core Isolation Cooling (RCIC) initiation turbine
trip circuit performs an operational protection of the main turbine
for commercial and reliability purposes. The proposed modification
slightly alters the methodology by which the turbine protective
features are performed but they have no

[[Page 48748]]

influence on any of the accidents previously evaluated. The
associated circuits do not interfere with higher priority protection
systems.
    Installation of circuits associated with the proposed
modification cannot initiate an accident, nor are they used to
mitigate the consequences of any previously defined accident. Their
function is to provide turbine protection that is separate and
distinct from the turbine overspeed protection system. The circuits
modified by this modification will still result in actions taken
(auto or manual) that meet the bases for the present design. Also,
this modification does not alter or adversely affect the turbine
overspeed function in any manner.
    The proposed modification reduces the probability of occurrence
of spurious turbine trips due to spurious RCIC initiation.
Therefore, with the implementation of this modification, the
boundaries of the accident analysis will be less challenged and
result in fewer false scrams.
    The proposed modification provides assurance for compliance with
the current licensing basis regarding dose limits of General Design
Criteria (GDC) 19 of Appendix A to 10 CFR [Part] 50 and 10 CFR
[Part] 100. The proposed modification ensures originally stated
design criteria are met and therefore does not affect the precursors
for accidents or transients analyzed in Chapter 15 of the Perry
Nuclear Power Plant (PNPP) Updated Safety Analysis Report (USAR).
With the proposed modification, the radiological consequences are
the same as previously stated in the USAR. Therefore, the
implementation of the proposed modification does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
    2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
    The USAR addresses accident analysis of the reactor based on
events such as turbine trips, including spurious trips and turbine
missiles. The present RCIC initiation turbine trip circuit is a
potential contributor to spurious turbine trips. The addition of the
time delay relay reduces this potential. A time delay relay failure
that fails to trip the turbine would have the same effect on the
turbine as the failure of the present trip circuit that has no time
delay relay. The consequence of the failure of this circuit to
protect the turbine remains unchanged with the addition of a time
delay relay and is bounded by the existing accident analysis. The
accident analysis for missile protection of those systems,
structures, components required for the safe shutdown of the plant
remain unchanged.
    The probability of external missile generation has not changed
with implementation of the proposed modification. The Main Turbine
casing and surrounding structures will not be changed by the
proposed modification. The location of equipment important to safety
as it relates to the turbine missiles will not be changed. Therefore
the missile strike probability will not be increased by the 4\1/2\
minute time delay.
    The proposed modification provides assurance for compliance with
the current licensing basis regarding dose limits of GDC 19 of
Appendix A to 10 CFR [Part] 50 and 10 CFR [Part] 100. The proposed
modification does not change the assumptions used in any accident
analysis and no new or different kind of accident is created. The
proposed modification ensures originally stated design criteria are
met and therefore does not affect the precursors for accidents or
transients analyzed in Chapter 15 of the PNPP USAR. Therefore, the
implementation of the proposed modification 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 margin of safety by which this modification is evaluated
against is the design/criteria of the turbine overspeed protective
system relative to the PNPP USAR, SER, GDC4 and Reg[ulatory] Guide
1.115, [``Protection Against Low-Trajectory Turbine Missiles.''] The
change in response time of the main turbine RCIC initiation trip
circuit does not affect the margin of safety as reflected in these
documents. There is no safety margin criteria associated with this
circuit, as defined in the USAR or the bases for any Technical
Specifications.
    Although there is no margin of safety associated with the
turbine, the regulatory requirement for acceptance of the turbine
for use at PNPP is based upon a calculated value of probability of
external turbine missile interaction with safety related equipment.
    The barriers (Turbine casing and surrounding structures) and
barrier interaction as previously analyzed will not be changed by
this modification. The location of safety related equipment as it
relates to the turbine missiles will not be changed. The probability
of external missile generation has not changed with implementation
of the proposed modification. Therefore, there is no reduction in
the margin of safety by the proposed modification.

    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.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,
Unit No. 1, St. Lucie County, Florida

    Date of amendment request: July 19, 2000.
    Description of amendment request: To revise the license: (1) to
implement Siemens Power Corporation (SPC) high thermal performance
(HTP) fuel assembly design in Cycle 17, (2) relocate shutdown margin
(SDM) requirements in Modes 1 to 5 to the Core Operating Limits Report
(COLR), (3) update the COLR methodologies listed in the Technical
Specification (TS) Section 6.9.1.11, and (4) request relief from the
SPC fuel assembly reconstitution restrictions for peripheral low power
fuel assemblies. Applicable TS surveillance requirements are changed to
be consistent with the proposed license amendment. Additionally,
administrative changes are proposed to the boron concentration
specifications related to the boration requirements.
    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) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
    The proposed amendment would allow the implementation of HTP
fuel design for Cycle 17. The design of this fuel will be evaluated
to meet all the mechanical, neutronics and thermal-hydraulics
requirements, and acceptance criteria based on the approved
methodology. The relocation of shutdown margin to the COLR and other
proposed changes have no adverse impact on the operation of the
plant and have no relevance to the accident initiators. There are no
changes to the plant configuration, and thus the frequency of
occurrence of previously analyzed accidents is not affected by the
proposed changes. The changes proposed to the fuel reconstitution
methodology would not impact the design acceptance criteria for the
reconstituted fuel assemblies.
    The proposed change for the relocation of shutdown margin to the
COLR has no impact on current safety analyses and their
consequences. Changes to the COLR limits will be controlled per
Generic Letter 88-16 under the provisions of 10 CFR 50.59 and the
requirements of TS 6.9.1.11.c. The application of the added
methodology, which includes the approved HTP DNB [departure from
nucleate boiling] correlation, would remain consistent with the
design basis requirements and would not involve a significant
increase in the consequences of design basis accidents. Other
proposed TS and TS bases changes do not affect safety analysis
results. The changes proposed to the fuel reconstitution methodology
would not impact the safety analysis consequences as the changes are
related to the non-limiting rod locations.
    Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
    (2) Use of the modified specification would not create the
possibility of a new or different

[[Page 48749]]

kind of accident from any previously evaluated.
    The proposed amendment updates the list of approved methodology
in TS 6.9.1.11, relocates shutdown margin requirements to the COLR
and requests relief for fuel reconstitution requirements. None of
these changes would create the possibility of a new kind of accident
since the reload analysis with these changes would continue to meet
all applicable design limits. There is no change to plant
configuration, systems or components which would create new failure
modes. The modes of operation of the plant would remain unchanged.
    Therefore, operation of the facility in accordance with the
proposed amendment would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
    (3) Use of the modified specification would not involve a
significant reduction in a margin of safety.
    The proposed changes have no significant adverse impact on the
safety analysis. As such, these changes would continue to provide
margin to the acceptance criteria for specified acceptable fuel
design limits (SAFDL), 10 CFR 50.46(b) requirements, primary and
secondary overpressurization, peak containment pressure, potential
radioactive releases, and existing limiting conditions for
operation. The future use of updated approved methodologies will
follow all design basis requirements to ensure that a safety margin
to the acceptance criteria would continue to remain available for
full power operation of St. Lucie Unit 1.
    Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant reduction in a
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

Florida Power and Light Company (FPL), Docket No. 50-335, St. Lucie
Plant, Unit No. 1, St. Lucie County, Florida

    Date of amendment request: July 19, 2000.
    Description of amendment request: The amendment would revise the
St. Lucie Unit 1 Technical Specifications (TS) to require laboratory
testing of activated charcoal samples for applicable engineered safety
feature ventilation systems using the ASTM D3803-1989 protocol. In
addition the proposed changes revise the TS test criteria for methyl
iodide removal efficiency to be consistent with the guidance of NRC
Generic Letter (GL) 99-02. The affected Unit 1 TS are the shield
building ventilation system (SBVS), TS 4.6.6.1; control room emergency
ventilation system (CREVS), TS 4.7.7.1; emergency core cooling system
(ECCS) area ventilation system, TS 4.7.8.1; and fuel pool ventilation
system--fuel storage, TS 4.9.12.
    The July 19, 2000, application is a complete replacement of the
proposed Unit 1 TS amendment previously submitted by FPL letter L-99-
241 on November 17, 1999. The NRC staff had previously published a
Federal Register notice on January 12, 2000 (Vol. 65, page 1923),
regarding the proposed amendments for St. Lucie Units 1 and 2, but
subsequently, issued the licence amendment for St. Lucie, Unit 2 only,
on February 17, 2000. This revised amendment request increases the TS-
required removal efficiency of the Unit 1 SBVS, ECCS area ventilation
system, and CREVS charcoal adsorbers to 97.5% when tested in accordance
with ASTM D3803-1989 at 30 deg.C, 70% relative humidity. The revised
testing requirements align the TS acceptance criteria and methodology
with the Unit 1 accident analysis assumptions and GL 99-02
recommendations.
    Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
    The proposed amendment does not involve a significant increase
in the probability or consequences of any accident previously
evaluated. The new charcoal testing protocol is performed offsite on
samples extracted from the safety related ventilation systems.
Therefore, there is no impact on any accident initiator and results
in no changes in the probability. The proposed testing protocol is
more conservative than previous tests; therefore, the efficiency of
charcoal for the affected safety related systems would not be
overestimated. With the new testing protocol, more conservative
testing results are expected since the temperature at which testing
is performed is lower and the charcoal retention capability is more
consistent with actual accident conditions. The proposed change thus
ensures that the charcoal in service will comply with the
penetration requirements to meet the design basis accident
conditions.
    Therefore, operation of the facility in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
    The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed charcoal testing protocol only affects
surveillance testing requirements for safety related ventilation
systems. The functions of these systems remain unchanged and
unaffected. No new system interactions have been introduced by the
proposed amendment, which would create a new or different type of
accident than previously analyzed. No physical changes are being
made to any structure, system, or component. The operation of the
facility will not be altered by the proposed amendment. The systems
involved are not initiators of any accidents as previously
evaluated.
    The proposed amendment will not change the physical plant or the
modes of operation defined in the facility license. The changes do
not involve the addition of new equipment or the modification of
existing equipment, nor do they alter the design of St. Lucie Unit 1
systems. Therefore, operation of the facility in accordance with the
proposed amendment would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
    The proposed amendment does not involve a reduction in the
margin of safety. The margin of safety of the Technical
Specifications, its Bases, the Updated Final Safety Analysis Report,
the Safety Evaluation Report or in any other design document has
been increased by the use of a safety factor of two for the TS
affected by the proposed amendment. The change provided in this
proposed amendment is related to introducing an improved testing
protocol for the activated charcoal in safety related ventilation
systems. The change consists of testing the charcoal with a new
testing protocol, higher efficiencies, and with lower test
temperatures to more closely reflect accident conditions and to
eliminate potential overestimation of charcoal efficiency.
    Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant reduction in a
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

[[Page 48750]]

Florida Power and Light Company, et al. (FPL), Docket Nos. 50-335 and
50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: June 21, 2000.
    Description of amendment request: The proposed amendments would
relocate Technical Specification Surveillance Requirement (SR)
4.8.1.1.2.e.1 to a licensee controlled maintenance program that will be
incorporated by reference into the next revision of each unit's Updated
Final Safety Analysis Report (UFSAR). SR 4.8.1.1.2.e.1 requires that
the emergency diesel generator (EDG) be inspected in accordance with
procedures prepared in conjunction with its manufacturer's
recommendations for this class of standby service, at least once every
18 months during shutdown. Upon relocation to the licensee controlled
maintenance program the requirement to perform the EDG inspections
every 18 months during shutdown will be eliminated. These amendments,
in combination with the previously submitted EDG risk informed allowed
outage time extension to 14 days, allows the EDG maintenance to be
performed in Modes 1 and 2. The licensee stated that approval of these
amendments is expected to reduce the complexity of activities performed
during refueling outages and, consequently, reduce human errors.
    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) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
    The proposed amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
There are no changes to the emergency diesel generator (EDG)
maintenance program. The actual EDG maintenance program is
unaffected.
    The only substantive change allows the periodic EDG inspection
to be performed in any operational mode instead of only during
shutdown. By FPL Letter L-99-228, dated November 17, 1999, FPL has
previously submitted a request for a risk informed EDG allowed
outage time (AOT) extension from 3 days to 14 days. An evaluation of
the impact on plant risk as expressed by the change in core damage
frequency (CDF), the incremental conditional core damage probability
(ICCDP), the change in large early release frequency (LERF), and the
incremental conditional large early release probability (ICLERP) was
provided as part of the EDG AOT extension submittal (L-99-228). The
EDG downtime (hours/train/year) assumed in the EDG AOT extension
risk assessment includes the out-of-service time that would be
incurred due to performing the proposed EDG inspections and
overhauls in Modes 1 and 2 instead of during shutdown. The risk
assessment for the proposed EDG AOT extension bounds the risk for
this change.
    NRC Regulatory Guide (RG) 1.177, An Approach for Plant-Specific
Risk-Informed Decision making: Technical Specifications, states that
an ICCDP of 5.0E-07 and an ICLERP of 5.0E-08 is considered small for
a single AOT change. Both the ICCDP and ICLERP for the proposed EDG
AOT extension and these proposed changes are below the RG 1.177
specified values and are thus considered small.
    NRC RG 1.174, An Approach for Using Probabilistic Risk
Assessment in Decisions on Plant Specific Changes to the Licensing
Basis, discusses acceptance criteria for changes in CDF and LERF. A
change in CDF of 1E-06 with a total CDF of 1E-04/year and a change
in LERF of 1E-07 with a total LERF of 1E-05 are considered very
small. The changes in CDF and LERF for the EDG AOT extension and
these proposed changes are below the RG 1.174 criteria and are thus
considered very small.
    The removal of the Mode restrictions from the maintenance
program are bounded by the risk assessment for the EDG AOT extension
and therefore do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
    (2) Use of the modified specification would not create the
possibility of a new or different kind of accident from any
previously evaluated.
    The use of the modified specifications cannot create the
possibility of a new or different kind of accident from any
previously evaluated since the proposed amendments will not change
the physical plant or the modes of plant operation defined in the
facility operating license. No new failure mode is introduced due to
implementation of this administrative change since the proposed
changes do not involve the addition or modification of equipment,
nor do they alter the design or operation of affected plant systems,
structures, or components.
    (3) Use of the modified specification would not involve a
significant reduction in a margin of safety.
    The operating limits and functional capabilities of the affected
systems, structures, and components remain unchanged by the proposed
amendments. Therefore, these changes do not involve a significant
reduction in the margin of safety. When the full scope of plant risk
is considered, the risks incurred by performing either corrective or
preventive EDG maintenance during power operation will be
substantially offset by plant benefits associated with avoiding
unnecessary plant transitions and/or reducing risks during shutdown
operations.
    Based on the above, we have determined that the proposed
amendments do not (1) involve a significant increase in the
probability or consequences of an accident previously evaluated, (2)
create the probability of a new or different kind of accident from
any previously evaluated, or (3) involve a significant reduction in
a margin of safety; and therefore does not involve a significant
hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

Florida Power and Light Company (FPL), Docket Nos. 50-250 and 50-251,
Turkey Point Plant, Units 3 and 4, Dade County, Florida

    Date of amendment request: May 22, 2000.
    Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) to incorporate the
requirements specified in the American Society of Mechanical Engineers
(ASME), Section XI, Subsection IWL, as modified and supplemented by the
requirements in Section 50.55a(b)(2)(viii), Examination of concrete
containments. In this regard, TS Section 3.6.1.6, ``Limiting Condition
for Operation,'' will be revised to conform to IWL tendon lift-off
force requirements, and TS Sections 4.6.1.6.1, 4.6.1.6.2, and 4.6.1.6.3
will be revised to conform to containment tendon and containment
surface inspection requirements specified in ASME Section XI,
Subsection IWL, 1992 Edition with the 1992 Addenda, and 10 CFR
50.55a(b)(2)(viii).
    The NRC Final Rule (61 FR 41303), dated August 8, 1996, requires
implementation of the revised requirements for containment examination
by September 9, 2001. FPL is planning to perform the containment tendon
surveillance for Turkey Point Units 3 and 4 in March 2001.
    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) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.

[[Page 48751]]

    Approval and implementation of this amendment will have no
effect on the probability or consequences of accident previously
evaluated. The containment is not an accident initiating system or
structure; therefore, there will be no impact on any accident
probabilities by the approval of this amendment. The containment
examination requirements in the proposed amendments are identical,
equivalent, or more rigorous than previous requirements. The
containment serves an important function to mitigate consequences of
postulated accidents evaluated and the examinations proposed in this
amendment will not result in a reduction in the capability of the
containment to meet its intended design function. Additionally, the
proposed changes to the Technical Specifications reflect the
adoption of ASME Section XI Subsection IWL containment inservice
inspections required by 10 CFR 55a(b)(2).
    Based on the above, it is concluded that the proposed amendments
do not involve a significant increase in the probability or
consequences of any accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any previously evaluated.
    The proposed changes do not alter the design, physical
configuration, or modes of operation of the plant. No changes are
being made to the plant that would introduce any new accident causal
mechanisms. The proposed Technical Specification changes do not
impact any plant systems that are accident initiators, since the
containment functions primarily as an accident mitigator and the
functional requirements of the containment structure are not
changed. No new accident causal mechanisms are created as a result
of NRC approval of the proposed amendments request. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously evaluated.
    (3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
    Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident situation, including the
performance of the containment. The containment is capable of
performing as intended, and its function is verified by visual
examination, post-tensioning system examinations, and leakage rate
testing. The containment examination requirements in the proposed
amendments are identical, equivalent, or more rigorous than previous
requirements. As such, the ability of the containment to perform its
design function will not be impaired by the implementation of the
proposed amendments request. Therefore, operation of the facility in
accordance with the proposed amendments would not involve a
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Dade County, Florida

    Date of amendment request: July 7, 2000.
    Description of amendment request: The proposed amendments would
revise the pressure-temperature (P/T) limits specified in Technical
Specification (TS) 3.4.9.1 and Figures 3.4-2, 3.4-3 and 3.4-4 to extend
their service period to a maximum of 32 effective full power years.
Also, the proposed amendments will revise TS 3.4.9.3, Cold Overpressure
Mitigation System (COMS) setpoints and its associated Surveillance
Requirements 4.4.9.3.1a and 4.4.9.3.1d. COMS is the Westinghouse
version of Low Temperature Overpressure Protection. Additionally, the
licensee's submittal requested two exemptions from the requirements of
10 CFR 50.60 based on the American Society of Mechanical Engineers
(ASME) Section XI, Code Cases N-588, ``Alternative to Reference Flaw
Orientation of Appendix G for Circumferential Welds in Reactor Vessels,
Section XI, Division 1'' and N-641, ``Alternative Pressure Temperature
Relationship and Low Temperature Overpressure Protection (LTOP) System
Requirements, Section XI, Division 1.'' The exemption requests will be
evaluated separately from the proposed license amendments.
    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) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
    The probability of occurrence of an accident previously
evaluated for Turkey Point is not altered by the proposed amendment
to the Technical Specifications. Each accident in the Turkey Point
UFSAR [Updated Final Safety Analysis Report] was examined with
respect to the changes to the proposed Pressure-Temperature (P/T)
limit curves and associated Cold Overpressure Mitigation System
(COMS) setpoint limitations.
    The proposed changes do not impact the integrity of the reactor
coolant system pressure boundary (i.e., no change in operating
pressure, materials, seismic loading, etc.) and therefore does not
increase the potential for the occurrence of a loss of coolant
accident (LOCA). The changes do not modify the reactor coolant
system pressure boundary, nor make any physical changes to the
facility design, material, or construction standards. The
probability of any design basis accident (DBA) is not affected by
this change, nor are the consequences of any DBA affected by this
change. The proposed P/T limit curves and COMS setpoint limit are
not considered to be an initiator or contributor to any accident
currently evaluated in the Turkey Point UFSAR.
    The curves and setpoint limit were generated in accordance with
approved NRC and ASME methodology. Code Cases N-588 and N-641 have
ASME Code Committee approval.
    Delaying performance of two of the COMS surveillances (PORV
[power operated relief valve] Channel Operational Test and the
backup nitrogen supply verification) until 12 hours after decreasing
the RCS cold leg temperature to 275 deg.F during cooldown
was also evaluated with respect to the plant accident analyses. The
change was determined to not represent a significant increase in the
probability or consequences of an accident because a) the likelihood
of a low temperature overpressure event occurring concurrently with
a loss of the redundant instrument air system is sufficiently small,
and b) the existing procedural controls will effectively prevent
challenges to the COMS.
    Additionally, delaying these surveillances for 12 hours will
allow the operators to focus their attention on transitioning the
plant to RHR [residual heat removal] cooling. Given the timing
sequence of the RHR system entry point to the COMS enable
temperature, the time extension is considered to be a prudent and
safety focused change to the method of performing a plant cooldown.
The proposed time extension is also consistent with the operational
flexibility currently provided in NUREG-1431, Standard Technical
Specifications for Westinghouse Plants.
    Based on the above, it is concluded that the proposed amendment
does not involve a significant increase in the probability or
consequences of any accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any previously evaluated.
    The proposed changes do not create a new accident scenario. The
requirements for the P/T limit curves and low temperature
overpressure protection have been in place for some time. The
fundamental approach follows approved ASME and Westinghouse topical
report methodology. The proposed curves reflect the change in
material properties acknowledged and managed by regulation and an
upgrade in technology, which has been approved by ASME.

[[Page 48752]]

    Delaying performance of two of the COMS surveillances (PORV
Channel Operational Test and the backup nitrogen supply
verification) until 12 hours after decreasing the RCS cold leg
temperature to 275 deg.F during cooldown was also
evaluated with respect to the plant accident analyses. The change
was determined to not represent a significant increase in the
probability or consequences of an accident because a) the likelihood
of a low temperature overpressure event occurring concurrently with
a loss of the redundant instrument air system is sufficiently small,
and b) the existing procedural controls will effectively prevent
challenges to the COMS.
    Additionally, delaying these surveillances for 12 hours is
consistent with the operational flexibility currently provided in
NUREG-1431, Standard Technical Specifications for Westinghouse
Plants.
    Since no new failure modes are associated with the proposed
changes, the activity does not create the possibility of a new or
different kind of accident from any previously evaluated.
    (3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
    The Technical Specifications for P/T limit curves and COMS
setpoints are expiring and must be updated. The COMS setpoint is
revised to incorporate additional margin in the instrument
uncertainty. Conservative ASME code methods including safety factors
have been used. The material properties used are from a much larger
database than in past submittals. This results in many more
datapoints available for the limiting weld metal than in past
submittals. A new master curve of irradiated and unirradiated
materials data has been developed for Turkey Point which shows that
these curves and associated setpoints are conservative and represent
an increase to the margin of safety. The new setpoint limit should
reduce the possibility of an inadvertent PORV actuation. They should
also reduce the potential for reactor coolant pump impeller
cavitation or seal damage when the pumps are operated during low
temperature conditions in the RCS. Changing the COMS surveillances
to allow completion up to 12 hours after decreasing RCS temperature
to 275 deg.F during cooldown does not result in a
reduction in the margin of safety. Acceptability is based on:
consistency with NUREG-1431, Standard Technical Specifications
Westinghouse Plants, COT [Channel Operational Test] Surveillance
Requirements; the inherent reliability and redundancy of the Turkey
Point Instrument Air System; and the existing procedural controls
established to prevent challenges to the LTOP System. The proposed
amendments will 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

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 requests: May 30, 2000.
    Description of amendment requests: The proposed amendments would
make changes to several Technical Specifications (TSs) to reflect
implementation of the revised 10 CFR Part 20, ``Standards for
Protection Against Radiation.'' In addition, the licensee proposed to
revise TS 6.8.4.a.7 to maintain existing instantaneous dose rate
limitations in the Offsite Dose Calculation Manual. Also, the licensee
proposed a revision to the requirements governing the annual tabulation
of radiation exposures.
    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 change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
    The proposed changes do not physically alter any plant
structures, systems, or components (SSCs), and do not affect or
create new accident initiators or precursors for any accident
evaluated in the Updated Final Safety Analysis Report. Therefore,
the probability of an accident previously evaluated is unchanged.
    The proposed changes do not affect the types or amounts of
radionuclides released following an accident, or the initiation and
duration of their release. The changes are administrative in nature.
Therefore, the consequences of an accident previously evaluated are
not increased.
    Therefore, the probability of occurrence or the consequences of
accidents previously evaluated are not significantly increased.
    2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
    The proposed changes do not physically alter any SSC and do not
affect or create new accident initiators or precursors. The accident
analysis assumptions and results are unchanged. No new failures or
interactions have been created.
    Therefore, the change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
    (3) Does the change involve a significant reduction in a margin
of safety?
    10 CFR 20.1301, Appendix I to 10 CFR 50, and 40 CFR 190
establish the controls and limitations on total effective dose
equivalent to individual members of the public from effluents
discharged to unrestricted areas. The proposed changes maintain
established limits for radioactive liquid effluents established in
10 CFR Part 20 and limits for radioactive gaseous effluents
established in the ODCM. I&M continues to comply with limits
specified in 10 CFR 20.1301, Appendix I to 10 CFR 50, and 40 CFR
190. Since compliance with these regulatory requirements has not
been compromised, the proposed changes do not involve a significant
reduction in the margin of safety.
    In summary, based upon the above evaluation, I&M has concluded
that the proposed amendment involves no significant hazards
consideration.
    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: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: June 20, 2000.
    Description of amendment request: The following technical
specification (TS) changes are being proposed to provide flexibility of
operation. These changes include: (1) the ability to have a standby
Safety Injection (SI) pump available during Reactor Coolant System
(RCS) reduced inventory conditions with the RCS pressure boundary
intact; (2) realigning a footnote to clarify the allowance of an
inoperable SI pump to be energized for testing or filling accumulators;
(3) allowance for an additional charging pump to be made capable of
injection during pump-swap operations; (4) recognition that a
substantial vent area exists for cold overpressure protection when the
reactor vessel head is on, and the studs are fully detensioned; (5)
limit maneuvering the plant beyond Hot Shutdown when one charging pump
is operable; and (6) establishes a new value for the open permissive
interlock associated with the Residual Heat Removal System suction
isolation valves.
    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. The NRC staff has reviewed the licensee's analysis
against

[[Page 48753]]

the standards of 10 CFR 50.92(c). The NRC staff's review is presented
below:
    1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
    The proposed changes do not affect plant systems such that their
function in the control of radiological consequences is adversely
affected. The proposed changes do not adversely affect accident
initiators or precursors, nor alter the design assumptions,
conditions, or manner in which structures, systems, and components
(SSCs) perform their intended safety function to mitigate the
consequences of an initiating event within the acceptance limits
assumed in the Updated Final Safety Analysis Report (UFSAR). The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
Since there are no changes to previous accident analyses, the
radiological consequences associated with these analyses remain
unchanged; therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident
from any accidents previously evaluated.
    The proposed changes do not result in a change to the design
basis of any plant SSC. All equipment important to safety will
operate as designed. The proposed TS changes in conjunction with
administrative controls will provide adequate control measures to
ensure component integrity is not challenged. The proposed changes
do not cause the initiation of any accident nor create any new
failure mechanisms. The changes do not result in any event
previously deemed incredible being made credible. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not adversely affect equipment design or
operation and there are no changes being made to the TS-required
safety limits or safety system settings that would adversely affect
plant safety. The proposed TS changes in conjunction with
administrative controls will provide adequate control measures to
ensure component integrity is not challenged. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
    Based on this review, it appears that the three standards of
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: June 20, 2000.
    Description of amendment request: The licensee proposes revising
the Technical Specifications (TS) by removing the prescriptive
requirement for determining the reactor coolant system flow rate by
precision heat balance in Surveillance Requirement 4.2.5.3 and
incorporating a time limit for completion of the surveillance
requirement. The change would also revise TS Table 2.2-1 to reflect the
allowed calibration tolerance of the protection racks and note that the
Trip Setpoint for Functional Unit 12, Reactor Coolant Flow-Low reactor
trip is based on an indicated value rather than a measured value.
    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 changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
    The proposed changes do not adversely affect accident initiators
or precursors nor alter the design, conditions, and configuration of
the facility or the manner in which the plant is operated. The
proposed changes do not alter or prevent the ability of structures,
systems, and components (SSCs) to perform their intended function to
mitigate the consequences of an initiating event within the
acceptance limits assumed in the Updated Final Safety Analysis
Report (UFSAR).
    Determination of RCS [Reactor Coolant System] total flow rate by
elbow tap P measurement will not subject the reactor core
to conditions adverse to nuclear safety. The proposed change does
not affect the source term; containment isolation or radiological
release assumptions used in evaluating the radiological consequences
of an accident previously evaluated in the Seabrook Station UFSAR.
The initial conditions for all accident scenarios modeled are the
same. Therefore, the consequences of an accident occurring remain
unchanged.
    The evaluation for use of elbow tap P measurement
determined that sufficient margin exists to account for all
reasonable instrument uncertainties, therefore no changes to
installed equipment or hardware in the plant are required. Though
the calibration process of the elbow tap P transmitters has
changed, i.e., normalization to previously performed precision RCS
flow calorimetrics for Cycles 1 and 2 instead of normalization to a
precision RCS flow calorimetric each cycle, this has been accounted
for by the addition of instrument uncertainties usually considered
to be zeroed out by normalization performed each cycle. Accounting
for the additional instrument uncertainties yields a flow
uncertainty that is slightly less (2.3 percent) than the current NRC
[Nuclear Regulatory Commission] licensed value (2.4 percent), thus
no change is required to the nominal reactor trip setpoint for RCS
flow. The proposed change has no adverse affect on component or
system interactions. Therefore, the proposed changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
    4. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
    The proposed changes do not alter the design, conditions and
configuration of the facility or the manner in which the plant is
operated and maintained in a state of readiness. Existing system and
component redundancy is not being changed by the proposed changes.
Though the calibration process of the elbow tap P
transmitters has changed, i.e., normalization to previously
performed precision RCS flow calorimetrics for Cycles 1 and 2
instead of normalization to a precision RCS flow calorimetric each
cycle, this has been accounted for by the addition of instrument
uncertainties usually considered to be zeroed out by normalization
performed each cycle. Accounting for the additional instrument
uncertainties yields a flow uncertainty that is slightly less than
the current NRC licensed value, thus no change is required to the
nominal reactor trip setpoint for RCS flow. The proposed change has
no adverse affect on component or system interactions. The time of
reactor trip remains the same. Therefore, since there are no changes
to the design, conditions, configuration of the facility, or the
manner in which the plant is operated and maintained in a state of
readiness, the proposed changes do not create the possibility of a
new or different kind of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not adversely affect equipment design or
operation and there are no changes being made to the Technical
Specification required safety limits or safety system settings that
would adversely affect plant safety. The additional instrument
uncertainties resulting from use of elbow tap P
transmitters without the requirement to normalize to a precision RCS
flow calorimetric each cycle have been accounted for and no change
in the nominal Trip Setpoint is required. The calculated instrument
uncertainty is 2.3 percent flow. This uncertainty is slightly less
than the current licensed value of 2.4 percent flow. The time of
reactor trip, as modeled in the various safety analyses, is
maintained. Therefore, there is 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request

[[Page 48754]]

involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

Northeast Nuclear Energy Company, et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London
County, Connecticut

    Date of amendment request: February 22, 2000
    Description of amendment request: The proposed changes to the
Technical Specifications (TSs) are associated with radiological
effluent. The proposed changes will relocate selected radiological
effluent TSs and the associated Bases to the Millstone Radiological
Effluent Monitoring and Offsite Dose Calculation Manual in accordance
with the Nuclear Regulatory Commission's (NRC) Generic Letter 89-01.
    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:

    In accordance with 10 CFR 50.92, NNECO [Northeast Nuclear Energy
Company] has reviewed the proposed changes and has concluded that
they do not involve a Significant Hazards Consideration (SHC). The
basis for this conclusion is that the three criteria of 10 CFR
50.92(c) are not compromised. The proposed changes do not involve an
SHC because the changes would not:
    1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
    The purpose of the Radiological Liquid and Gaseous Effluent
Monitoring Instrumentation is to monitor routine radioactive
releases. [This] instrumentation provide[s] a surveillance of
potential release points and initiates automatic alarm and trip
functions which will terminate the release prior to exceeding the
limits of 10 CFR Part 20 (1993 version). Relocation of Technical
Specification 3.3.3.9, ``Radioactive Liquid Effluent Monitoring
Instrumentation,'' and Technical Specification 3.3.3.10,
``Radioactive Gaseous Effluent Monitoring Instrumentation,'' to the
Radiological Effluent Monitoring and Offsite Dose Calculation Manual
(REMODCM) does not imply any reduction in its importance in
monitoring routine radioactive releases. These instruments are
neither used for, nor capable of, detecting a significant abnormal
degradation of the reactor coolant pressure boundary before a design
basis accident, nor do they function as a primary success path to
mitigate events which assume a failure of or a challenge to the
integrity of fission product barriers. These monitors are not an
active design feature needed to preclude analyzed accidents or
transients. Therefore, this change will not significantly increase
the probability or consequences of an accident previously evaluated.
    Technical Specification 3.11.1.1 ensure[s] the concentration of
radioactive materials released in liquid waste effluents from the
site will be less than the concentration levels specified in 10 CFR
Part 20 (1993 version), Appendix B, Table II. Technical
Specification 3.11.1.2 ensures the dose or dose commitment from
radioactive materials released in liquid waste effluents will not
exceed the requirements of Sections II.A, III.A and IV.A of Appendix
I, 10 CFR Part 50. Technical Specification 3.11.2.1 ensures the dose
rate from gaseous effluents released from all units on site will be
less than dose limits specified in 10 CFR Part 20 (1993 version),
Appendix B, Table II. Technical Specification 3.11.2.2 ensures the
dose from noble gases released in gaseous effluents will not exceed
the requirements of Sections II.B, III.A and IV.A of Appendix I, 10
CFR Part 50. Technical Specification 3.11.2.3 implements the
requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR
Part 50. Technical Specification 3.11.3 ensures the reporting
requirements of 40 CFR 190 are met. Relocation of these Technical
Specifications to REMODCM does not imply any reduction in its
importance in ensuring that the regulatory limits are met. The
instrumentation covered by these Technical Specifications [is]
neither used for, nor capable of, detecting a significant abnormal
degradation of the reactor coolant pressure boundary before a design
basis accident, nor [does it] function as a primary success path to
mitigate events which assume a failure of or a challenge to the
integrity of fission product barriers. [This] instrumentation [is]
not an active design feature needed to preclude analyzed accidents
or transients. Therefore, this change will not significantly
increase the probability or consequences of an accident previously
evaluated.
    As a result of the relocation of the Radiological Effluent
Technical Specifications (RETS) to the REMODCM, there are no
Technical Specifications remaining that use definitions 1.31 and
1.26, ``Radiological Effluent Monitoring and Offsite Dose
Calculation Manual (REMODCM),'' of Unit Nos. 2 and 3 respectively.
The guidelines and procedures addressing the use of radioactive
waste treatment systems are covered by Specifications 6.15 and 6.13
of unit Nos. 2 and 3 respectively, which describes the REMODCM.
Therefore, definitions 1.33 and 1.25, ``Radioactive Waste Treatment
Systems,'' of Unit Nos. 2 and 3 respectively are no longer needed.
In addition, there are no Specifications that use this phrase in the
context of a defined term. These changes do not impact the
assumptions used in any accident analysis, affect plant equipment,
plant configuration, or the way the plant is operated. Therefore,
this change will not significantly increase the probability or
consequences of an accident previously evaluated.
    Replacing Technical Specification 6.9.1.6 of Millstone Unit No.
2 with Technical Specifications 6.9.1.6a and 6.9.1.6b and revising
Technical Specifications 6.9.1.3 and 6.9.1.4 of Millstone Unit No. 3
will provide descriptions which satisfy the requirements of parts 10
CFR 50.36a and 10 CFR 50, Appendix I, Sections IV.B.1, IV.B.2,
IV.B.3, and IV.C. These changes are consistent with NUREG-1432 and
NUREG-1431. These changes do not impact the assumptions used in any
accident analysis, affect plant equipment, plant configuration, or
the way the plant is operated. Therefore, this change will not
significantly increase the probability or consequences of an
accident previously evaluated.
    The description of the REMODCM contained in Technical
Specifications 6.15 and 6.13 of Millstone Unit Nos. 2 and 3
respectively will be modified to be consistent with the guidance of
GL 89-01, and with NUREG-1432 and NUREG-1431. Additional minor
changes have been made to be consistent with the proposed changes to
Technical Specification 6.9.1.6 of Millstone Unit No. 2 and
Technical Specifications 6.9.1.3 and 6.9.1.4 of Millstone Unit No.
3. These changes do not impact the assumptions used in any accident
analysis, affect plant equipment, plant configuration, or the way
the plant is operated. Therefore, this change will not significantly
increase the probability consequences of an accident previously
evaluated.
    Adding Technical Specifications 6.20 and 6.15, Radiological
Effluent Controls Program, to Millstone Unit Nos. 2 and 3
respectively, and 6.21 and 6.16, Radiological Environmental
Monitoring Program, to Millstone Unit Nos. 2 and 3 respectively is
consistent with the guidance contained in Generic Letter 89-01 for
the relocation of the Radiological Effluents Technical
Specifications and with NUREG-1432 and NUREG-1431. Additional minor
changes have been made to be consistent with the version of 10 CFR
20, Appendix B, Table II, Column 1 which is being used by Millstone
Unit Nos. 2 and 3, namely the 1993 version. These changes do not
impact the assumptions used in any accident analysis, affect plant
equipment, plant configuration, or the way the plant is operated.
Therefore, this change will not significantly increase the
probability or consequences of an accident previously evaluated.
    The following proposed changes are administrative in nature.
Therefore, these changes will not significantly increase the
probability or consequences of an accident previously evaluated.
     Revise Index Pages of Unit Nos. 2 and 3 Technical
Specifications to reflect the proposed changes to relocate the RETS
to the REMODCM.
     Address additional changes to the Millstone Unit No. 2
Technical Specifications to resolve issues not related to
transferring the RETS to the REMODCM.
     Relocate to the associated Bases sections.
    The proposed changes do not alter how any structure, system, or
component functions. There will be no effect on equipment important
to safety. The proposed changes have no effect on any of the design
basis accidents previously evaluated. Therefore, this License
Amendment Request

[[Page 48755]]

does not impact the probability of an accident previously evaluated,
nor does it involve a significant increase in the consequences of an
accident previously evaluated.
    (2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
    The proposed changes do not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. They do not alter the way any
structure, system, or component functions and do not alter the
manner in which the plant is operated. The proposed changes do not
introduce any new failure modes. Therefore, the proposed changes
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Relocation of Technical Specifications 3.3.3.9, 3.3.3.10,
3.11.1.1, 3.11.1.2, 3.11.2.1, 3.11.2.2, 3.11.2.3, and 3.11.3 to
REMODCM does not imply any reduction in its importance in monitoring
and ensuring that the regulatory limits are met. As a result of the
relocation of the RETS to the REMODCM, there are no Technical
Specifications remaining that use definitions 1.31 and 1.26.
Additionally, the guidelines and procedures addressing the use of
radioactive waste treatment systems which are covered by
Specifications 6.15 and 6.13 remove the need for definitions 1.33
and 1.25 of Unit Nos. 2 and 3 respectively. Replacing Technical
Specification 6.9.1.6 of Millstone Unit No. 2 with Technical
Specifications 6.9.1.6a and 6.9.1.6b and revising Technical
Specifications 6.9.1.3 and 6.9.1.4 of Millstone Unit No. 3 will
provide descriptions which satisfy the requirements of parts 10 CFR
50.36a and 10 CFR 50, Appendix I, Sections IV.B.1, IV.B.2, IV.B.3,
and IV.C. Modifying the description of the REMODCM contained in
Technical Specifications 6.15 and 6.13 of Millstone Unit Nos. 2 and
3 respectively and adding Technical Specifications 6.20, 6.21 and
6.15, 6.16 to Millstone Unit Nos. 2 and 3 respectively is consistent
with the guidance contained in Generic Letter 89-01 for the
relocation of the Radiological Effluents Technical Specifications
and with NUREG-1432 and NUREG-1431.
    The proposed changes do not affect any of the assumptions used
in the accident analysis, nor do they affect any operability
requirements for equipment important to plant safety. Therefore, the
proposed changes will not result in a significant reduction in the
margin of safety as defined in the Bases for Technical
Specifications covered in this License Amendment Request.
    As described above, this License Amendment Request does not
involve a significant increase in the probability of an accident
previously evaluated, does not involve a significant increase in the
consequences of an accident previously evaluated, does not create
the possibility of a new or different kind of accident from any
accident previously evaluated, and does not result in a significant
reduction in a margin of safety. Therefore, NNECO has concluded that
the proposed changes do not involve an SHC.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
    NRC Section Chief: James W. Clifford.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: June 26, 2000.
    Description of amendment request: The proposed changes to
Millstone, Unit 3, Technical Specifications (TS) revise TS Section
1.13, Definitions, ``Engineered Safety Features Response Time'', TS
Section 1.28, ``Reactor Trip System Response Time,'' TS Section 3.3.1,
``Instrumentation--Reactor Trip System Instrumentation,'' and TS
Section 3.3.2, ``Instrumentation--Engineered Safety Features Actuation
System Instrumentation'' to provide for verification of response time
for selected components provided that the components and the
methodology for verification have been previously reviewed and approved
by the Nuclear Regulatory Commission.
    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. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
    This change to the Technical Specifications does not result in a
condition where the design, material, and construction standards
that were applicable prior to the change are altered. The same RTS
[Reactor Trip System] and ESFAS [Emergency Safety Features Actuation
System] instrumentation is being used; the time response
allocations/modeling assumptions in the Chapter 15 analyses are
still the same; only the method of verifying time response is
changed. The proposed change will not modify any system interface
and could not increase the likelihood of an accident since these
events are independent of this change. The proposed activity will
not change, degrade or prevent actions or alter any assumptions
previously made in evaluating the radiological consequences of an
accident described in the SAR [Safety Evaluation Report]. Therefore,
there will be no significant increase in the probability or
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
    This change does not alter the performance of the pressure and
differential pressure transmitters, Process Protection racks,
Nuclear Instrumentation, and Logic Systems used in the plant
protection systems. These sensors and systems will still have
response time verified by test before being placed in operational
service. Changing the method of periodically verifying instrument
response for these sensors and systems (assuring equipment
operability) from time response testing or calibration and channel
checks will not create any new accident initiators or scenarios.
Periodic surveillance of these sensors and systems will continue and
may be used to (a) detect significant degradation in the sensor
responses characteristic, and (b) other degradation that could cause
the response time characteristic to exceed the total allowance. The
total time response allowance for each function bounds all
degradation that cannot be detected by periodic surveillance.
Therefore, the proposed changes will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
    3. Involve a significant reduction in a margin of safety.
    This change does not affect the total system response time
assumed in the safety analysis. The periodic system response time
verification method for selected pressure and differential pressure
sensors, the Process Protection racks, Nuclear Instrumentation, and
Logic Systems is modified to allow use of actual test data or
engineering data. The method of verification still provides
assurance that the total system response is within that defined in
the safety analysis, since calibration tests will continue to be
performed and may be used to detect any degradation which (a) might
significantly affect sensor response time, or (b) might cause the
response time to exceed the total allowance. The total system time
response allowance for each function bounds all degradation that
cannot be detected by periodic surveillance. Based on the above, it
is concluded that the proposed license amendment request does not
result in a significant reduction in margin with respect to plant
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: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
    NRC Section Chief: James W. Clifford.

[[Page 48756]]

PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Dockets
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2
and 3, York County, Pennsylvania

    Date of application for amendments: May 31, 2000.
    Description of amendment request: The proposed amendments would
revise the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3,
Technical Specifications (TSs) Surveillance Requirement (SR) 3.6.1.3.11
to allow a representative sample of reactor instrumentation line excess
flow check valves (EFCVs) to be tested every 24 months, instead of
testing each EFCV every 24 months.
    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. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:

    1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
    The current SR frequency requires each reactor instrumentation
line EFCV to be tested every 24 months. The EFCVs at PBAPS, Units 2
and 3 are designed to not close accidentally during normal
operation, but will close automatically in the event of a line break
downstream of the valve. The proposed changes would allow a reduced
number of EFCVs to be tested each operating cycle. Since the EFCVs
are an accident mitigation feature, their postulated failure to
isolate cannot initiate previously evaluated accidents. In addition,
since the proposed changes will only change the surveillance
frequency, there can be no increase in the probability of occurrence
of an accident as a result of this proposed change.
    The postulated break of an instrument line attached to the
reactor coolant pressure boundary is discussed and evaluated in the
Updated Final Safety Analysis Report (UFSAR), Section 5.2.3.5. The
proposed change will continue to verify the operability of the EFCVs
to perform their mitigating functions. Industry operating experience
as documented in the Boiling Water Reactors Owners Group (BWROG)
Report B21-00658-01 provides supporting evidence that the reduced
testing frequency will not affect the high reliability of these
valves. The radiation dose consequences of such a break are not
impacted by this proposed change. Therefore, the proposed TS changes
do not involve a significant increase in the consequences of an
accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
    The proposed changes would allow a reduced number of EFCVs to be
tested each operating cycle. No other changes in requirements are
being proposed. The changes are not a physical alteration of the
plant and will not alter the operation of the structures, systems
and components as described in the UFSAR. Therefore, a new or
different kind of accident will not be created.
    3. The proposed TS changes do not involve a significant
reduction in a margin of safety. The consequences of an unisolable
rupture of an instrument line has been previously evaluated in the
PBAPS, Units 2 and 3 UFSAR, Section 5.2.3.5. That evaluation assumed
a continuous discharge of reactor water for the duration of the
detection and cooldown sequence. The integrity and functional
performance of the secondary containment and standby gas treatment
system are not impaired by this event, and the calculated potential
offsite exposures are substantially below the guidelines of 10 CFR
Part 100. Therefore, a failure of an EFCV, though not expected as a
result of this TS change, is bounded by the previous evaluation of
an instrument line break. Since the proposed changes are only
affecting the surveillance frequency, the accident analyses are
unaffected and this change does not involve a significant reduction
in the margin of safety.

    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: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
    NRC Section Chief: James W. Clifford.

Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: June 7, 2000.
    Description of amendment request: The proposed amendment to the
Indian Point Nuclear Generating Unit No. 3 (IP3) Technical
Specifications (TSs) would require either the Operations Manager or the
Assistant Operations Manager to hold a Senior Reactor Operator (SRO)
license. The proposed amendment would also remove the title of ``Shift
Manager'' from the TS.
    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 license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
    No. This change allows either the Operations Manager or
Assistant Operations Manager to be SRO licensed. This is an
administrative change. The Operations department will still have an
SRO licensed individual overseeing the operating crews. Therefore,
there will be no increase in the probability or consequences of an
evaluated accident. This is consistent with the qualifications
required to be a manager in TS 6.3.1.
    The change also deletes the title of Shift Manager. At IP3,
``Shift Manager'' is the NYPA [New York Power Authority] specific
title for the person meeting the requirements of 10 CFR
50.54(m)(2)(ii) as the SRO assigned responsibility for overall plant
operation. This requirement is redundant to 10 CFR 50.54(m)(2)(ii)
and TS section 6.2.2 requirements for an SRO and therefore removal
is an administrative change with no increase in the probability or
consequences of an accident.
    2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
    No. The change allows either the Operations Manager or Assistant
Operations Manager to hold the SRO license. The proposed change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated since they do not affect
plant configuration or plant design. The Operations Manager and the
Assistant Operations Manager are still required to maintain a
knowledge of IP3 plant design and operations through job position
requirements.
    The change also deletes the title of Shift Manager. At IP3,
``Shift Manager'' is the NYPA specific title for the person meeting
the requirements of 10 CFR 50.54(m)(2)(ii) as the SRO assigned
responsibility for overall plant operation. This requirement is
redundant to 10 CFR 50.54(m)(2)(ii) and TS section 6.2.2
requirements and it is therefore an administrative change that
cannot create the possibility of a new or different accident.
    3. Does the proposed amendment involve a significant reduction
in a margin of safety?
    No. The change allows either the Operations Manager or Assistant
Operations Manager to hold the SRO License. The proposed amendment
does not involve a significant reduction in a margin of safety
because the Operations Manager and/or the Assistant Operations
Manager is still required to maintain a current SRO license.
Administrative Controls ensure that shift activities are directed by
an individual holding an SRO license. Technical Specification 6.3.1
ensure that the Operations Manager will be a knowledgeable and
qualified individual.
    The change also deletes the title of Shift Manager. At IP3,
``Shift Manager'' is the NYPA specific title for the person meeting
the requirements of 10 CFR 50.54(m)(2)(ii) as the SRO assigned
responsibility for overall plant operation. This requirement is
redundant to 10 CFR 50.54(m)(2)(ii) therefore the change has no
effect on requirements and cannot offset the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on
this

[[Page 48757]]

review, it appears that the three standards of 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New
York, New York 10019.
    NRC Section Chief: Marsha Gamberoni.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: July 21, 2000.
    Description of amendment request: The proposed amendment would
delete the requirement to have the Control Room Emergency Air Treatment
System (CREATS) Actuation Instrumentation and CREATS operable in Modes
5 and 6 except during core alterations and fuel movement.
    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:

Evaluation of More Restrictive Changes

    The more restrictive changes (which is a conservative
characterization, as these changes are implied by the current
specifications) associated with amending the Applicability section
for LCO [limiting condition for operation] 3.3.6 and LCO 3.7.9, and
Condition C of LCO 3.3.6 and Condition D and F of LCO 3.7.9, to
include ``during CORE ALTERATIONS'', do not involve a significant
hazards consideration as discussed below:
    (1) Operation of Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability
or consequences of an accident previously evaluated. The changes add
a conservative Mode of Applicability for the Control Room Emergency
Air Treatment System (CREATS) and CREATS actuation instrumentation.
This does not increase the probability of an accident previously
evaluated since the CREATS and CREATS actuation instrumentation
themselves are not accident initiators. The proposed changes are
consistent with the guidance of NUREG-1431 and provide assurance
that the CREATS is in the conservative mode of operation for a
response to an accident. Therefore, the probability or consequences
of an accident previously evaluated are not significantly increased.
    (2) Operation of Ginna Station in accordance with the proposed
changes does not create the possibility of a new or different kind
of accident from any accident previously evaluated. The proposed
change for a new mode of applicability does not of itself involve a
physical alteration of the plant or change in the methods governing
normal plant operation. The change only involves a conservative
increase in the requirement of when the CREATS and CREATS actuation
instrumentation are operable. Therefore, the possibility for a new
or different kind of accident from any accident previously evaluated
are not created.
    (3) Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in a margin of
safety. The proposed change requires the CREATS and CREATS actuation
instrumentation to be in the conservative mode of operation for a
response to an accident. The change adds conservatism as determined
by the guidance of NUREG-1431. Therefore, this change does not
involve a significant reduction in a margin of safety.
    Based upon the preceding information, it has been determined
that the proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated,
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. Therefore, it is concluded that the
proposed changes meet the requirements of 10 CFR 50.92(c) and do not
involve a significant hazards consideration.

Evaluation of Less Restrictive Changes

    The less restrictive changes associated with amending the
applicability sections for LCO 3.3.6 and LCO 3.7.9, and Condition C
of LCO 3.3.6 and Condition D and F of LCO 3.7.9, to delete Modes 5
and 6 from these sections do not involve a significant hazards
consideration as discussed below:
    1. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability
or consequences of an accident previously evaluated. The changes are
the result of an analysis performed of the control room dose
consequences which could occur as the result of a potential waste
gas decay tank failure. This does not increase the probability of an
accident previously evaluated since the Control Room Emergency Air
Treatment System (CREATS) and CREATS actuation instrumentation
themselves are not accident initiators. The results of the analysis
show that if no credit is taken for the CREATS, the control room
doses remain well within the limits specified in 10 CFR 50, Appendix
A, GDC [General Design Criteria] 19 and the guidance provided by the
NRC in NUREG-0737 Section ll.B.2, Dose Rate Criteria, and NUREG-0800
Section 6.4, Control Room Habitability Program. The proposed Mode of
Applicability change is consistent with the guidance of NUREG-1431
which allows plant-specific changes with respect to Modes 5 and 6.
Therefore, the probability or consequences of an accident previously
evaluated are not significantly increased.
    (2) Operation of Ginna Station in accordance with the proposed
changes does not create the possibility of a new or different kind
of accident from any accident previously evaluated. The proposed
changes associated with the modes of applicability for the CREATS
and CREATS actuation instrumentation are not of themselves nor do
they affect potential accident initiators. Therefore, the
possibility for a new or different kind of accident from any
accident previously evaluated are not created.
    (3) Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in a margin of
safety. The proposed changes remove the requirements for the control
room ventilation system, which has been shown by analysis to not be
required to meet regulatory limits. The changes are consistent with
the guidance of NUREG-1431. Therefore, these changes do not involve
a significant reduction in a margin of safety.
    Based upon the preceding information, it has been determined
that the proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated,
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. Therefore, it is concluded that the
proposed changes meet the requirements of 10 CFR 50.92(c) and do not
involve a significant hazards consideration.
    The less restrictive change associated with amending the
Required Action and Completion Time of Condition C of LCO 3.3.6 and
Condition F of LCO 3.7.9 to remove a required action, do not involve
a significant hazards consideration as discussed below:
    (1) Operation of Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability
or consequences of an accident previously evaluated. The proposed
changes to remove a required action of restoring equipment to
operable status do not affect the probability of an accident as the
Control Room Emergency Air Treatment System (CREATS) and CREATS
actuation instrumentation, in and of themselves, have no failure
modes or effects which are precursors to accidents. The proposed
changes do not introduce any new failure modes or effects to any
other system or component which is a precursor to an accident. The
remaining Required Actions within the referenced Conditions place
the plant outside of the Mode of Applicability for these systems.
Therefore, the probability or consequences of an accident previously
evaluated are not significantly increased.
    (2) Operation of Ginna Station in accordance with the proposed
changes does not create the possibility of a new or different kind
of accident from any accident previously evaluated. The changes do
not of themselves involve a physical alteration of the plant or
change in the methods governing normal plant operation. The proposed
changes create no new functional interactions with existing plant
equipment nor do they introduce any new failure modes or mechanisms
which could lead to reactor core damage or fission product release.
Therefore, because the changes do not affect any system that can act
as an accident precursor, the possibility for a new or different
kind of accident from any accident previously evaluated are not
created.

[[Page 48758]]

    (3) Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in a margin of
safety. The proposed changes remove requirements for restoring
systems which are no longer required. The changes are consistent
with the guidance of NUREG-1431. Therefore, these changes do not
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400
L Street, NW., Washington, DC 20005.
    NRC Section Chief: Marsha K. Gamberoni.

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama

    Date of amendment request: June 29, 2000.
    Description of amendment request: The proposed amendment would
change the Farley Nuclear Plant, Units 1 and 2, design bases described
in the Final Safety Analysis Report. The change adds a description of
the methodology Southern Nuclear Operating Company uses to determine
what systems and components need to be protected from tornado missiles.
    Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

    1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
    Proposed for NRC review and approval are changes to the Farley
Nuclear Plant (FNP) Final Safety Analysis Report (FSAR) which in
essence constitute a license amendment to incorporate use of an
NRC[-]approved methodology to assess the need for additional
positive (physical) tornado missile protection of specific features
at FNP. The FSAR changes will reflect use of the Electric Power
Research Institute (EPRI) Topical Report ``Tornado Missile Risk
Evaluation Methodology'' (EPRI NP-2005), Volumes I and II. As noted
in the NRC Safety Evaluation Report on this topic dated October 26,
1983, the current licensing criteria governing tornado missile
protection are contained in Standard Review Plan (SRP) Sections
3.5.1.4 and 3.5.2. These criteria generally specify that safety-
related systems be provided positive tornado missile protection
(barriers) from the maximum credible tornado threat. However, SRP
Section 3.5.1.4 includes acceptance criteria permitting relaxation
of the above deterministic guidance, if it can be demonstrated that
the probability of damage to unprotected essential safety-related
features is sufficiently small.
    As permitted in NRC Standard Review Plan (NUREG-0800) sections,
the combined probability will be maintained below an allowable
level, i.e., an acceptance criterion threshold, which reflects an
extremely low probability of occurrence. The FNP approach assumes
that if the probability calculation result for the total plant
identifies that the probability of a combination of tornado missiles
striking and damaging a portion of an important system or component
is greater than or equal to 10^-\6\ then installation of
unique missile barriers would be needed to lower the total combined
probability below the acceptance criterion of 10^-\6\.
    With respect to the probability of occurrence or the
consequences of an accident previously evaluated in the FSAR, the
possibility of a tornado reaching the FNP site and causing damage to
plant structures, systems and components is a design basis event
considered in the [FSAR]. The changes being proposed do not affect
the probability that the natural phenomenon (a tornado) will reach
the plant, but from a licensing basis perspective they do affect the
probability that missiles generated by the winds of the tornado
might strike and damage certain plant systems or components. There
are a limited number of safety-related components that could
theoretically be struck and consequently damaged by tornado-
generated missiles. The probability of tornado-generated missile
strikes on ``important'' systems and components (as discussed in
Regulatory Guide 1.117) is what is to be analyzed using the
probability methods discussed above. The combined probability of
damage will be maintained below an extremely low acceptance
criterion to ensure overall plant safety. The proposed change is not
considered to constitute a significant increase in the probability
of occurrence or the consequences of an accident, due to the
extremely low probability of damage due to tornado-generated
missiles and thus an extremely low probability of a radiological
release. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of
previously evaluated accidents.
    2. The proposed change will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
    The possibility of a tornado reaching the FNP site is a design
basis event considered in the [FSAR]. This change involves
recognition of the acceptability of performing tornado missile
probability calculations in accordance with established regulatory
guidance. The change therefore deals with an established design
basis event (the tornado). Therefore, the proposed change would not
contribute to the possibility of a new or different kind of accident
from those previously analyzed. The probability and consequences of
such a design basis event are addressed in Question 1 above. Based
on the above discussions, the proposed change will not create the
possibility of a new or different kind of accident than those
previously evaluated.
    3. The proposed change will not involve a significant reduction
in a margin of safety.
    The existing licensing basis for FNP with respect to the design
basis event of a tornado reaching the plant, generating missiles and
directing them toward safety-related systems and components is to
provide positive missile barriers for all safety-related systems and
components. With the change, it will be recognized that there is an
extremely low probability, below an established acceptance limit,
that a limited subset of the ``important'' systems and components
could be struck and consequently damaged. The change from protecting
all safety-related systems and components to ensuring an extremely
low probability of occurrence of tornado-generated missile strikes
and consequential damage on portions of important systems and
components is not considered to constitute a significant decrease in
the margin of safety due to that extremely low probability.
Therefore, the changes associated with this license amendment
request do 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.
    Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama.
    NRC Section Chief: L. Raghavan, Acting.

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

    Date of amendment request: May 16, 2000.
    Description of amendments request: Amend Technical Specification
(TS) 4.8.1.1.2 to revise the emergency diesel generator fuel oil
surveillance requirements to adopt more current industry standards.
    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:

    A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.

[[Page 48759]]

    The probability of occurrence or the consequences for an
accident is not increased by this request. The proposal to establish
a Diesel Fuel Oil Program and specifying the ASTM [American Society
for Testing and Materials] standards in the TS Bases does not modify
the manner in which the plant is operated. Deletion of the portion
of the surveillance requirement (SR) that specifies the use of
sodium hypochlorite solution in cleaning of the fuel oil storage
tanks, and the deletion of the SR to perform a pressure test of
those portions of the diesel fuel oil system designed as Section
III, subsection ND of the ASME [American Society of Mechanical
Engineers] Code do not alter the way any structure, system, or
component functions and does not modify the manner in which the
plant is operated.
    This request will ensure that the fuel oil continues to be
properly evaluated to ensure that the fuel oil will not degrade the
ability of the D/G [diesel generator] to perform its intended
function. The fuel oil storage tanks will be cleaned at the required
frequency. The deletion of the SR to perform a pressure test of
those portions of the diesel fuel oil system designed to Section
III, subsection ND of the ASME Code, removes potential confusion
about testing of the fuel oil system since no portion of the system
is designed to Section III, subsection ND of the ASME Code.
Therefore, these changes will not change or impact previously
evaluated accidents and the D/Gs ability to perform their intended
function.
    B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
    The proposed changes are procedural in nature concerning fuel
oil testing, cleaning chemical to be used on the fuel oil storage
tanks, and deletion of the pressure test of those portions of the
diesel fuel oil system designed as Section III, subsection ND of the
ASME Code. The possibility for an accident or malfunction of a
different type than any evaluated previously in SQN's [Sequoyah's]
Final Safety Analysis Report are not created. The proposal does not
alter the way any structure, system, or component functions and does
not modify the manner in which the plant is operated. The fuel oil
quality will not be reduced and will not result in a decrease in D/G
operability. The fuel oil storage tanks will be cleaned at the
required frequency. Therefore, the possibility of a new or different
kind of accident previously evaluated is not created.
    C. The proposed amendment does not involve a significant
reduction in a margin of safety.
    The proposed changes are procedural in nature concerning fuel
oil testing, cleaning chemical to be used on the fuel oil storage
tanks, and deletion of the pressure test of the diesel fuel oil
system. The margin of safety has not been reduced since the change
in test methodologies are NRC approved and will continue to ensure
the quality of the fuel oil. Also, deletion of the portion of the SR
that specifies the use of sodium hypochlorite does not change the
requirement to clean the fuel oil storage tanks. ASME Code
requirements will continue to be met. Therefore, the proposed
changes do not involve a reduction in the margin of safety.

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

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

    Date of amendment request: July 10, 2000 (TS 00-08).
    Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) as follows:

Part A--Channel Operational Test (COT) 12 Hour Limit

    Channel operational tests (COTs) are performed for the Power
Range and Intermediate Range neutron monitors in accordance with
Reactor Trip System (RTS) Surveillance Requirements (SRs) 3.3.1.7
and 3.3.1.8. While the unit is in Modes 1 or 2, SR 3.3.1.7 is
performed for the Power Range monitors every 92 days. SR 3.3.1.8 is
performed for the Intermediate Range monitors prior to startup of
the reactor and at various points during power escalation or
reduction. In addition, SR 3.1.10.1 currently requires that a COT be
performed on the Power Range and Intermediate Range neutron monitors
within 12 hours prior to initiation of a physics test, even though
SR 3.3.1.7 and SR 3.3.1.8 have been performed on the required
frequency.
    TVA proposes to eliminate the 12 hour requirement for the
testing required by SR 3.1.10.1 so that the testing performed for SR
3.3.1.7 and SR 3.3.1.8 can be used to satisfy SR 3.1.10.1. This
issue was addressed by Technical Specification Task Force (TSTF)
Traveler 108. The proposed amendment revises SR 3.1.10.1 to
implement the portion of the approved TSTF 108 applicable to Watts
Bar.

Part B--Trip System Logic for Physics Testing TSTF Traveler 315

    During the performance of physics testing one power range
channel is used to provide input to the reactivity computer. In
preparation for the test, the fuses to the electronics drawer for
the channel are removed and the channel is placed in a tripped
condition and results in the NIS trip logic being in a one-out-of-
three logic status. Therefore, any spurious signals received on one
channel will result in a reactor trip. The changes proposed by TSTF-
315 allows the fuses to remain in the NIS channel that is connected
to the reactivity computer and avoid tripping the bistables
associated with the NIS channel. This configuration results in the
channel being in a bypassed state and places the overall logic in a
two-out-of-three logic status. The advantage of this configuration
is that a single spurious signal would not result in a reactor trip.
The proposed amendment does not deviate from the version of TSTF-315
that was approved by NRC on June 29, 1999.

    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:
    A. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.

Part A--Channel Operational Test (COT) 12 Hour Limit

    The proposed amendment removes the requirement to perform an
additional Channel Operational Test (COT) on the Intermediate and
Power Range functions within 12 hours of performing a physics test.
The Intermediate and Power Range instrumentation is determined to be
OPERABLE by periodic surveillance requirements which must be
confirmed to be within frequency prior to making the reactor
critical. A COT for the Intermediate or Power Range instrumentation
is not a precursor to, or assumed to be an initiator of any analyzed
accident. Therefore, this change does not involve a significant
increase in the probability of an accident previously evaluated.
    Regarding a significant increase in the consequences of an
accident, several factors must be considered. First the physics
tests are performed in accordance with the Technical Specifications
in Mode 2. Therefore, the power level of the reactor is limited to 5
percent or less. Along with this, the reactor trip function of the
intermediate range detectors will be unaffected by the proposed
amendment and therefore, will be available to mitigate a reactivity
transient at low power. Further, the trip setpoint for the power
range monitors are decreased during startup of the reactor from the
normal 109% setpoint to a value less than or equal to 85%. This
setpoint reduction provides an additional measure to limit a
reactivity excursion. Considering these factors, the proposed change
will not involve a significant increase in the consequences of an
accident previously evaluated.

Part B--Trip System Logic for Physics Testing

    During the performance of physics testing one power range
channel is used to provide input to the reactivity computer. In
preparation for the test, the fuses to the electronics drawer for
the channel are removed and the channel is placed in a tripped
condition and results in the NIS trip logic being in a one-out-of-
three logic status. Therefore, any spurious signals received on one
channel will result in a reactor trip. The changes proposed by TSTF-
315 allows the

[[Page 48760]]

fuses to remain in the NIS channel that is connected to the
reactivity computer. This configuration results in the channel being
in a bypassed state and places the overall logic in a two-out-of-
three logic status. The advantage of this configuration is that a
single spurious signal will not result in a reactor trip. In
addition, the physics tests required by LCO 3.1.10 are performed
while the reactor is in Mode 2. Therefore, the thermal power of the
reactor is restricted to 5 percent or less. Neutron flux, which is
monitored by the NIS, is only one of several RTS variables which may
initiate a reactor trip in Mode 2. The other variables include
reactor coolant temperature, pressurizer pressure and steam
generator water level. These variables are unaffected by the
proposed amendment. Considering this, the low thermal power level of
the reactor, and a potential reduction in unnecessary plant
transients due to the one-out-of-three logic, the proposed amendment
will not significantly impact the safe operation of the plant.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
    B. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.

Part A--Channel Operational Test (COT) 12 Hour Limit

    The proposed amendment is not based on a change in the design or
configuration of the plant. Also, the proposed amendment does not
change the manner in which the plant is operated. The amendment
deletes the requirement for the performance of a COT for the
Intermediate and Power Range instrumentation within 12 hours of
starting a physics test. Therefore, the proposed change will not
create the possibility of a new or different kind of accident than
any previously evaluated.

Part B--Trip System Logic for Physics Testing

    The NIS provides indication, alarm, control, and trip signals
along with the capability to monitor neutron flux over the complete
range from reactor shutdown to 120 percent full power. The system
also generates permissive and level trip signals, which are then
coupled to the logic matrices of the RTS. This interface either
allows power changes based upon proper functioning of the next range
of measurement instrumentation or shuts down the reactor as unsafe
operating limits are approached. The changes in the operation of the
NIS proposed by this amendment for TSTF-315, do not inhibit the
capabilities of the system to initiate a reactor trip, if required.
Therefore, the proposed amendment will not create the possibility of
a new or different kind of accident.
    C. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in the margin of
safety.

Part A--Channel Operational Test (COT) 12 Hour Limit

    As stated previously, the proposed change deletes the
requirement to perform an additional COT for the Intermediate and
Power Range functions within 12 hours of the start of physics test.
The Intermediate and Power Range instrumentation channels are
determined to be operable by meeting the requirements of the
periodic surveillances. These surveillance requirements are not
affected by the proposed amendment. Since the equipment will be
determined to be operable by periodic surveillances, the performance
of the a surveillance prior to the initiation of a physics test does
not provide any additional assurance that the functions are more
reliable. Considering this, the proposed amendment does not
significantly reduce the margin of safety.

Part B--Trip System Logic for Physics Testing

    During the low power physics testing, implementation of the
proposed amendment will result in one power range channel being in a
bypassed state. In this configuration, there will be three available
channels with a two-out-of-three logic required to actuate the
neutron flux trip function. As required by LCO 3.1.10, the testing
will be performed while the reactor is in Mode 2 and therefore,
restricted by the Technical Specifications to a power level of less
than or equal to 5 percent.
    There are two power range control functions, rod control and
steam generator level control. At the 5 percent or less power level,
rod control is in manual and is not affected by the testing
configuration. Steam generator level control is not affected since
its input from the NIS channel connected to the Reactivity Computer
is placed in bypass when establishing the test configuration.
Therefore, an assumed failure affecting these control functions does
not have to be considered for the testing configuration. Also while
in this configuration, an assumed single failure will not prevent
the power range monitors from actuating as designed.
    The reactor trip function of the intermediate range detectors
will be unaffected by the proposed amendment and therefore, will be
available to mitigate a reactivity transient at low power. Further,
the trip setpoint for the power range monitors are decreased during
startup of the reactor from the normal 109% setpoint to a value less
than or equal to 85%. This setpoint reduction provides an additional
measure to limit a reactivity excursion.
    Based on the preceding, TVA concludes that there is no significant
reduction in the margin of safety due to the implementation of the
proposed amendment.
    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: June 22, 2000.
    Description of amendment request: The proposed amendments to the
Technical Specification Figures 3.4-2, 3.4-3, and associated Bases
would extend the cumulative core burnup applicability limits for the
reactor coolant system pressure-temperature (P/T) operating limits, Low
Temperature Overpressure Protection System (LTOPS) setpoints, and LTOPS
enable temperature (T enable). Implementation of American Society of
Mechanical Engineers (ASME) Section XI Code Cases N-640 and N-514 will
require exemptions from the requirements of 10 CFR 50, Appendix G.
    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 change involve a significant increase in the
probability or consequences of an accident previously evaluated[?]
    The proposed changes extend the cumulative core burnup
applicability of the existing North Anna Units 1 and 2 P/T limits,
LTOPS setpoints, and T enable values. No changes to plant systems,
structures, or components are proposed, and no new allowable
operating modes are established, The P/T limits, LTOPS setpoints,
and T enable values do not contribute to the probability of
occurrence or consequences of accidents previously analyzed. The
revised licensing basis analyses utilize acceptable analytical
methods, and continue to demonstrate that established accident
analysis acceptance criteria are met. Therefore, there is no
increase in the probability or consequences of any accident
previously evaluated.
    2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated[?]
    The proposed changes extend the cumulative core burnup
applicability of the existing North Anna Units 1 and 2 P/T limits,
LTOPS setpoints, and T enable values. No changes to plant systems,
structures, or components are proposed, and no new allowable
operating modes are established. Therefore, the proposed changes do
not create the possibility of any accident or malfunction of a
different type previously evaluated.
    3. Does the change involve a significant reduction in the margin
of safety[?]
    The proposed revised analysis bases use the ASME Section XI code
Case N-640 K1c stress intensity formulation and a plant

[[Page 48761]]

specific application of the analysis methodology which supports ASME
Section XI Code Case N-514. These analysis features are less
restrictive than those associated with the existing analyses, but
are conservative with respect to [those] established by ASME Section
XI margins. The proposed revised analyses support continued use of
the existing North Anna Units 1 and 2 Technical specification P/T
limit curves, LTOPS setpoints, LTOPS enable temperatures for North
Anna Units 1 and 2 cumulative core burnups up to 32.3 effective full
power years (EFPY) and 34.3 EFPY, respectively. The analyses
demonstrate that established analysis acceptance criteria continue
to be met. Specifically, the existing P/T limit curves, LTOPS
setpoints, and LTOPS T enable values provide acceptable margin to
vessel fracture under both normal operation and LTOPS design basis
(mass addition and heat addition) accident conditions. Therefore,
the proposed changes do not result in a significant reduction in a
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
    NRC Section Chief: L. Raghavan, Acting.

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

    Date of amendment request: June 22, 2000.
    Description of amendment request: The proposed changes would modify
Facility Operating Licenses NPF4 and NPF-7, along with the associated
Bases, to permit the elimination of the assumed increase in the rod
control cluster assembly (RCCA) drop time resulting from a concurrent
trip and seismic event, when determining if the measured rod drop times
meet the Technical Specifications limit of 2.7 seconds.
    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 change involve a significant increase in the
probability or consequences of an accident previously evaluated[?]
    Elimination of the assumed increase in the RCCA drop time
resulting from a concurrent trip and seismic event when determining
if the measured rod drop times, including measurement uncertainties,
meet the accident analysis limit[,] does not contribute to the
probability of previously analyzed accidents. The proposed change
will not alter the limiting results of the safety analyses presented
in Chapter 15 of the UFSAR [Updated Final Safety Analysis Report].
Although the proposed change eliminates an accident consideration
that is currently addressed in the UFSAR accident analyses (i.e. any
Chapter 15 accident with the effects of a concurrent seismic
occurrence reflected in the RCCA drop time), there is no significant
increae in the probability or consequences of any accident
previously evaluated.
    2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated[?]
    There are no modifications to the plant as a result of the
changes. No new accident or event initiators are created by
eliminating the assumed increase in the RCCA drop time resulting
from a concurrent trip and seismic event. The proposed change will
not alter the ability of the reactor protection and control system
to perform their design functions or to meet the applicable criteria
set forth in the IEEE [Institute of Electrical and Electronics
Engineers] and ANSI [American National Standards Institute]
standards and in 10 CFR 50 Appendix A. Therefore, the proposed
changes do not create the possibility of any accident or malfunction
of a different type previously evaluated.
    3. Does the change involve a significant reduction in the margin
of safety[?]
    The proposed change will not alter the limiting results of the
safety analyses presented in Chapter 15 of the UFSAR. Elimination of
the assumed increase in the RCCA drop time resulting from a
concurrent trip and seismic event when determining if the measured
rod drop times, including measurement uncertainties, [meet] the
accident analysis limit maintains adequate safety margin in the
safety analysis. Therefore, the proposed change does not
significantly reduce a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
    NRC Section Chief: L. Raghavan, Acting.

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

    Date of amendment request: March 29, 2000.
    Description of amendment request: The proposed change would revise
Technical Specification (TS) 3.19 and TS 4.1. The change would reflect
two redundant trains of bottled air for the main control room (MCR),
include remedial action statements for one train and two trains
inoperable, eliminate the extension of 8 hours to 24 hours currently
permitted by TS 3.19.B, add requirements for an inoperable control room
pressure boundary, and include additional surveillance testing
requirements. The TS 3.19 Basis and TS 4.1 Basis would be revised for
consistency with the respective TS.
    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:

    The proposed TS change includes train specific requirements,
adds requirements for an inoperable control room pressure boundary,
imposes additional surveillance testing requirements for the MCR
bottled air system, and is consistent with the existing accident
analyses. We have reviewed the proposed TS change relative to the
requirements of 10 CFR 50.92 and determined that a significant
hazards consideration is not involved. Specifically, operation of
Surry Power Station with the proposed change will not:
    1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
    The proposed change does not involve a physical modification and
does not modify the design or operation of the MCR bottled air
system or the plant. Since the MCR bottled air system functions to
respond to--not prevent--an accident, the probability of occurrence
of an accident is not affected. The elimination of the currently
allowed extension of the remedial action time, the addition of train
specific requirements and inoperable boundary requirements, and the
imposition of additional surveillance testing requirements serve to
ensure no increase in the consequences of an accident. Therefore,
the proposed change does not significantly increase the probability
of occurrence or the consequences of any previously analyzed
accident.
    2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
    The proposed change does not involve a physical modification and
does not affect the design or operation of the MCR bottled air
system or the plant. Consequently, no new or unique operational
modes or accident precursors are introduced. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    The proposed change does not involve a physical modification and
does not modify the design or operation of the MCR bottled air
system or the plant. The elimination of

[[Page 48762]]

the currently allowed extension of the remedial action time, the
addition of train specific requirements and inoperable boundary
requirements, and the imposition of additional surveillance testing
requirements serve to ensure the bottled air system's ability to
pressurize the main control room for one hour following a design
basis accident, which is consistent with the existing accident
analyses. Therefore, the proposed change does not result in a
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams,
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia
23219.
    NRC Acting Section Chief: L. Raghavan.

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.

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

    Date of application for amendments: July 13, 2000.
    Description of amendments request: Amend Technical Specification
3.7.5.c to allow an increase in the average essential raw cooling water
supply header temperature from 84.5 deg.F to 87 deg.F until September
30, 2000.
    Date of publication of individual notice in the Federal Register:
July 20, 2000 (65 FR 45113).
    Expiration date of individual notice: August 3, 2000.

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, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site, http://www.nrc.gov (the
Electronic Reading Room).

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

    Date of application for amendment: August 23, 1999, as supplemented
January 8, 2000.
    Brief description of amendment: The amendment deletes certain
license conditions that are obsolete and no longer apply.
    Date of issuance: July 24, 2000.
    Effective date: As of the date of issuance and shall be implemented
within 30 days.
    Amendment No.: 130.
    Facility Operating License No. NPF-62: The amendment revised the
License.
    Date of initial notice in Federal Register: September 22, 1999 (64
FR 51346). The January 8, 2000, submittal identified an additional
license condition that was no longer applicable and thus did not change
the scope of the action noticed or alter the initial no significant
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 24, 2000.
    No significant hazards consideration comments received: No.
    Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
    Date of application for amendment: August 2, 1999, as supplemented
April 7 and July 5, 2000.
    Brief description of amendment: This amendment revises Technical
Specification 6.2.2.e, ``Administrative Controls--Unit Staff.'' The
license requirements for operations management have been modified.
    Date of issuance: July 19, 2000.
    Effective date: July 19, 2000.
    Amendment No.: 99.
    Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR
46426). The supplemental letters dated April 7 and July 5, 2000,
contained clarifying information only, and did not change the initial
no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 19, 2000.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina

    Date of application for amendment: April 12, 2000, as supplemented
June 2, 2000.
    Brief description of amendment: This amendment revises Technical
Specification (TS) 3/4.4.9.2, ``Pressure/Temperature (P-T) Limits--
Reactor Coolant System,'' and TS 3/4.4.9.4, ``Overpressure Protection
System,'' and the associated Bases. Specifically, the amendment
incorporates results of the Reactor Vessel Surveillance Program capsule
analysis and an exemption from 10 CFR 50.60(a), based on American
Society of Mechanical Engineers Code Case N-640.

[[Page 48763]]

    Date of issuance: July 28, 2000.
    Effective date: July 28, 2000.
    Amendment No. 100.
    Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
    Date of initial notice in Federal Register: May 3, 2000 (65 FR
25762). The supplemental letter dated June 2, 2000, contained
clarifying information only, and did not change the initial no
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 28, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 29, 1999, as supplemented by
letters dated August 24, 1999, January 27, 2000, May 22, 2000, and May
31, 2000.
    On June 14, 2000, the Commission published in the Federal Register
(FR) Notice of consideration of issuance of amendment to facility
operating license, proposed no significant hazards consideration
determination, and opportunity for a hearing (65 FR 37425). In this
finding, incorrect reference is made to supplements dated August 8,
1999, and March 29, 2000. No supplements from the licensee with these
dates are related to this amendment.
    Brief description of amendment: The amendment modifies Technical
Specification 3.8.1.1 and associated Bases by extending the Emergency
Diesel Generator (EDG) allowed outage time from 72 hours to ten days.
In the supplemental letter dated May 22, 2000, an alternate source for
the onsite power system during the EDG maintenance outage, by way of a
temporary EDG (TEDG), was added. The application dated July 29, 1999,
did not include the TEDG.
    Date of issuance: July 21, 2000.
    Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
    Amendment No.: 166.
    Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR
37425). This notice is based on the supplement dated May 22, 2000, and
supercedes the notice dated February 9, 2000 (65 FR 6406), which is
based on the licensee's letter dated July 29, 1999. The May 31, 2000,
supplement did not expand the scope of the application as noticed or
change the proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 21, 2000.
    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: March 6, 2000.
    Brief description of amendment: Revised the Improved Technical
Specification Action Condition and Surveillance Requirement related to
the diesel-driven emergency feedwater pump (EFW-3) required lube oil
volume.
    Date of issuance: July 17, 2000.
    Effective date: July 17, 2000.
    Amendment No.: 192.
    Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: April 19, 2000 (65 FR
21036).
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 17, 2000.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Dade County, Florida

    Date of application for amendments: November 30, 1999, as
supplemented March 8, May 15, and July 5, 2000.
    Brief description of amendments: The proposed amendments would
revise the Technical Specifications to allow the use of credit for
soluble boron in the spent fuel pool criticality analyses. In addition,
a revised criticality analysis for the fresh fuel storage racks will be
used to update the licensing bases.
    Date of issuance: July 19, 2000.
    Effective date: July 19, 2000.
    Amendment Nos.: 206 and 200.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 3, 2000 (65 FR
25765). The May 15, and July 5, 2000, submittals provided clarifying
information that did not change the scope of the original request or
change the initial proposed no significant hazards consideration
determination.
    The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 19, 2000.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Dade County, Florida

    Date of application for amendments: April 27, 2000.
    Brief description of amendments: Incorporate references to the NRC
safety evaluations supporting exemptions granted for the Thermo-Lag
Upgrade project. In addition, the amendments modify Technical
Specification Section 6.0, Administrative Controls, Section 4.7.6.g, to
include page 3/4 7-21 which was inadvertently excluded from the
previous submittal and amendment.
    Date of issuance: July 20, 2000.
    Effective date: July 20, 2000.
    Amendment Nos.: 207 and 201.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications and the Operating Licenses.
    Date of initial notice in Federal Register: May 31, 2000 (65 FR
34746).
    The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 20, 2000.
    No significant hazards consideration comments received: No.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey

    Date of application for amendment: April 15, 1999, as supplemented
December 22, 1999, and February 24, 2000.
    Brief description of amendment: The amendment editorially revised
the Technical Specifications to enhance clarity.
    Date of Issuance: July 17, 2000.
    Effective date: July 17, 2000 and shall be implemented within 30
days of issuance.
    Amendment No.: 211.
    Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR
12293).
    The February 24, 2000, supplemental letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
    The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated July 17, 2000.
    No significant hazards consideration comments received: No.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey

    Date of application for amendment: June 3, 1999, as supplemented on
December 22, 1999.

[[Page 48764]]

    Brief description of amendment: The proposed amendment revised the
Technical Specifications to permit continued plant operation with a
maximum of two inoperable recirculation loops, provided certain
conditions are met. Oyster Creek's Technical Specifications (TSs),
Section 3.3.F.2 currently permit operation with 4 of the 5
recirculation loops with certain constraints. If only 3 loops are
operable, however, the TSs require plant shutdown within 12 hours.
Analysis indicates that the plant may be safely operated at 90 percent
power with three operable recirculation loops.
    Two definitions are added to Section 1 of the TSs to specify the
difference between an idle recirculation loop and an isolated
recirculation loop. These definitions have been incorporated into the
specification to provide an explicit description of acceptable valve
configurations. In addition, several paragraphs have been added to the
Bases of Section 3.3 and one paragraph in the Bases of Section 3.10 has
been modified. In each case the Bases section has been segmented from
the specification, which affects the pagination of the Bases.
    Date of Issuance: July 27, 2000.
    Effective date: July 27, 2000 and shall be implemented within 30
days of issuance.
    Amendment No.: 212.
    Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: May 3, 2000 (65 FR
25766). The December 22, 1999, supplemental letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
    The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated July 27, 2000.
    No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska

    Date of amendment request: December 11, 1997, as supplemented by
letter dated May 8, 2000.
    Brief description of amendment: The amendment revised the technical
specifications by adding a new limiting condition for operation (LCO)
for an inoperable engineered safety features logic subsystem. In
addition, administrative changes were made to either support the new
LCO or clarify existing text.
    Date of issuance: July 25, 2000.
    Effective date: July 25, 2000, and shall be implemented within 60
days from the date of issuance.
    Amendment No.: 194.
    Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998 (63
FR 6987). The May 8, 2000, supplemental letter provided additional
clarifying information, 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 amendment is contained in a Safety Evaluation
dated July 25, 2000.
    No significant hazards consideration comments received: No.

Public Service Electric & Gas Company, Docket No. 50-311, Salem Nuclear
Generating Station, Unit No. 2, Salem County, New Jersey

    Date of application for amendment: April 10, 2000.
    Brief description of amendment: This amendment modifies the
requirements contained in the Salem Unit No. 2 Technical Specifications
regarding the operation of the movable incore detector system and
allows continued operation of Salem Unit No. 2 through the remainder of
Cycle 11. The revision represents a one-time change to allow use of the
movable incore detector system for measurement of core peaking factors
with less than 75% and greater than or equal to 50% of the detector
thimbles available. Public Service Electric and Gas Company submitted
this request in response to degradation of the movable incore detector
system.
    Date of issuance: July 25, 2000.
    Effective date: As of the date of issuance, and shall be
implemented within 60 days of issuance.
    Amendment No.: 212.
    Facility Operating License No. DPR-75: This amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: May 23, 2000 (65 FR
33378).
    The Commission received comments which were addressed in the NRC
staff's Safety Evaluation dated July 25, 2000.
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 25, 2000.
    No significant hazards consideration comments received: Yes.

Southern California Edison Company, et al., Docket Nos. 50-206, San
Onofre Nuclear Generating Station (SONGS), Unit 1, San Diego County,
California

    Date of application for amendment: December 2, 1999, as
supplemented on May 16, 2000.
    Brief description of amendment: The amendment revised the SONGS
Unit 1 Technical Specifications by revising the administrative controls
to be consistent with the SONGS Unit 2 and 3 Technical Specification
administrative controls including changes to the administrative control
of working hours and working hour deviation approvals, position titles
and responsibilities and organizational description reference,
qualifications for a multi-discipline supervisor, quality assurance
program control of review and audit and record retention procedures,
high radiation area controls, description of the plant configuration
for environmental protection, and environmental protection related
document reporting. The amendment also incorporated changes related to
certified fuel handlers and 10 CFR 50.54(x).
    Date of issuance: July 19, 2000.
    Effective date: July 19, 2000, to be implemented within 30 days of
issuance.
    Amendment No.: 159.
    Facility Operating License No. DPR-13: The amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64
FR 73096).
    The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 19, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: December 17, 1999, as
supplemented on June 30, 2000.
    Brief description of amendments: Revises License Condition to allow
storage at the Sequoyah Nuclear Plant site of low-level radioactive
waste generated at Watts Bar Nuclear Plant, Unit 1.
    Date of issuance: July 18, 2000.
    Effective date: July 18, 2000.
    Amendment Nos.: 257 and 248.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the Operating Licenses.
    Date of initial notice in Federal Register: February 23, 2000 (65
FR 9012). The supplemental letter of June 30,

[[Page 48765]]

2000, did not change the initial No Significant Hazards Consideration
determination.
    The Commission's related evaluation of the amendment is contained
in an Environmental Assessment dated June 29, 2000 (65 FR 41739) and in
a Safety Evaluation dated July 18, 2000.
    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: November 20, 1998, as
supplemented July 19, 1999, and January 21, 2000.
    Brief description of amendment: The amendment revises the Technical
Specifications (TS) to change the surveillance requirements for an
inspection of the ice condenser flow channels that previously used a
0.38 inch ice/frost buildup criterion to a criterion that limits flow
blockage to the 15 percent value that was used in the accident
analysis. Changes to the Bases were also made. Tennessee Valley
Authority also indicated that its proposal is consistent with TS
Traveler Form No. 336.
    Date of issuance: July 17, 2000.
    Effective date: July 17, 2000.
    Amendment No.: 25.
    Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64
FR 70093). The January 21, 2000, letter contained clarifying
information that did not change the initial No Significant Hazards
Consideration Determination.
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 17, 2000.
    No significant hazards consideration comments received: No.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont

    Date of application for amendment: May 23, 2000.
    Brief description of amendment: The amendment relocates the
specifications for reactor coolant conductivity and chloride
concentration from the Technical Specifications to the Technical
Requirements Manual.
    Date of Issuance: July 18, 2000.
    Effective date: As of its date of issuance, and shall be
implemented within 60 days of issuance.
    Amendment No.: 190.
    Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR
37430).
    The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated July 18, 2000.
    No significant hazards consideration comments received: No.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont

    Date of application for amendment: May 23, 2000.
    Brief description of amendment: The amendment revises the Technical
Specifications to increase the interval between Local Power Range
Monitor calibrations from 1,000 equivalent full power hours to 2,000
megawatt-days/ton.
    Date of Issuance: July 18, 2000.
    Effective date: As of its date of issuance, and shall be
implemented within 60 days of issuance.
    Amendment No.: 191.
    Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR
37431).
    The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated July 18, 2000.
    No significant hazards consideration comments received: No.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont

    Date of application for amendment: May 22, 2000.
    Brief description of amendment: The amendment removes the Technical
Specifications surveillance requirement for visual inspection of
suppression chamber coating integrity once each refueling outage.
    Date of Issuance: July 19, 2000.
    Effective date: As of its date of issuance, and shall be
implemented within 60 days.
    Amendment No.: 192.
    Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR
37430).
    The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated July 19, 2000.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas

    Date of amendment request: March 31, 2000, as supplemented by
letter of July 7, 2000.
    Brief description of amendment: The amendment modifies the actions
for Limiting Condition for Operation (LCO) 3.7.9, ``Ultimate Heat Sink
(UHS),'' of the TSs. The new Action A for the LCO allows the plant to
operate with the plant inlet water temperature of the UHS above
90 deg.F, if the required lake water level is verified within 1 hour
and once per 12 hours thereafter, but would require that the plant be
shut down if the water temperature exceeded 94 deg.F. The amendment
replaces the requirement to shut down the plant if the UHS water
temperature exceeds 90 deg.F.
    Date of issuance: July 14, 2000.
    Effective date: July 14, 2000, to be implemented within 30 days
from the date of issuance.
    Amendment No.: 134.
    Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: April 19, 2000 (65 FR
21040). The supplemental letter of July 7, 2000, had minor
clarifications that are within the scope of the initial notice and does
not alter the no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 14, 2000.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)

    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 for the
amendment 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.
    Because of exigent or emergency circumstances associated with the
date

[[Page 48766]]

the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
    In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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
application for amendment, (2) the amendment to Facility Operating
License, 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, the Gelman Building, 2120 L Street, NW., Washington, DC, and
electronically from the ADAMS Public Library component on the NRC Web
site, http://www.nrc.gov (the Electronic Reading Room).
    The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By September 8, 2000, 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 Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and electronically from the ADAMS Public Library
component on the NRC Web site, http://www.nrc.gov (the Electronic
Reading Room). 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. Since the Commission has

[[Page 48767]]

made a final determination that the amendment involves no significant
hazards consideration, if a hearing is requested, it will not stay the
effectiveness of the amendment. Any hearing held would take place while
the amendment is in effect.
    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: Rulemakings and
Adjudications Staff or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of application for amendment: July 13, 2000, as supplemented
by letters dated July 14 and 21, 2000.
    Brief description of amendment: The amendment permitted a one-time
change to Technical Specification 4.4.5.0 and allowed alternate
inspection scope and expansion criteria for steam generator tube
inspections to be implemented during the mid-cycle outage scheduled for
summer 2000.
    Date of issuance: July 26, 2000.
    Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
    Amendment No.: 217.
    Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
    Public comments requested as to proposed no significant hazards
consideration: Yes.
    The NRC published a public notice of the proposed amendment, issued
a proposed finding of no significant hazards consideration, and
requested that any comments on the proposed no significant hazards
consideration be provided to the staff by the close of business on July
24, 2000. The notice was published in The Courier (in Russellville) and
the Arkansas Democrat-Gazette (in Little Rock) from July 20 through 22,
2000. No public comments were received.
    The Commission's related evaluation of the amendment, finding of
exigent circumstances, consultation with the State of Arkansas, and
final no significant hazards consideration determination are contained
in a Safety Evaluation dated July 26, 2000.

    Dated at Rockville, Maryland, this 3rd day of August 2000.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 00-20014 Filed 8-8-00; 8:45 am]
BILLING CODE 7590-01-P
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Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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