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

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.

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 [Federal Register: February 21, 2001 (Volume 66, Number 35)]
[Notices]
[Page 11050-11070]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr21fe01-117]

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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 January 29, 2001, through February 9, 2001.
The last biweekly notice was published on February 7, 2001.

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 Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, 11555 Rockville Pike, Room O-1F15, Rockville, Maryland.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
    By March 23, 2001, 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, 11555 Rockville
Pike, Room O-1F15, Rockville, Maryland, 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

[[Page 11051]]

the proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves
no 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: Rules and
Adjudications Branch, or may be delivered to the Commission's Public
Document Room, 11555 Rockville Pike, Room O-1F15, Rockville, Maryland,
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).

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of amendment request: January 15, 2001.
    Description of amendment request: The proposed amendment revises
the Technical Specification (TS) Design Features Section 5.4.2(f),
``Spent Fuel Storage,'' to remove the existing TS fuel assembly U\235\
loading criterion for fuel assemblies stored in the spent fuel storage
pool.
    Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

    1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed change has no effect on the normal
operating, design basis accident, or transient analyses applicable
to the TMI [Three Mile Island] Unit 1 fuel storage requirements.
Other existing TMI Unit 1 Technical Specification provisions ensure
sub-criticality for normal and postulated accident conditions.
    Therefore, the proposed change 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. Fuel assembly U\235\ loading is not an initial condition
of a design basis accident or transient that either assumes the
failure of or presents a challenge to the integrity of a fission
product barrier. Discussion of fuel assembly U\235\ loading in the
TMI Unit 1 UFSAR [Updated Final Safety Analysis Report] ensures that
changes to fuel designs that increase fuel reactivity relative to
design assumptions for fuel storage are evaluated in accordance with
the requirements of 10 CFR 50.59.
    Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
    3. The proposed change does not involve a significant reduction
in a margin of safety. The proposed change does not affect existing
TMI Unit 1 Technical Specification requirements controlling maximum
fuel enrichment, allowable enrichment vs. burnup, soluble boron
requirements, storage rack spacing, allowable rack locations for
fuel assembly storage or sub-criticality requirements for normal and
accident conditions. These existing Technical Specification
requirements ensure that the current margin of safety is not
reduced. The fuel assembly U\235\ loading criterion does not
represent an input parameter or limiting design condition for any
supporting design basis analyses applicable to the TMI Unit 1 spent
fuel storage requirements.
    Therefore, the proposed change does not involve a significant
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: Edward J. Cullen, Jr., Esq., PECO Energy
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
    NRC Section Chief: Marsha Gamberoni.

[[Page 11052]]

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County,
North Carolina

    Date of amendment request: January 17, 2001.
    Description of amendment request: The proposed amendments would
relax Surveillance Requirement 3.6.1.3.7 by allowing a ``representative
sample'' of excess flow check valves to be tested every 24 months, such
that each excess flow check valve will be tested at least once every 10
years.
    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 license amendments do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
    The current surveillance requirement frequency requires each
reactor instrumentation line excess flow check valve to be tested
every 24 months. The excess flow check valves at BSEP are designed
to close automatically in the event of a line break downstream of
the valve. The proposed change allows a reduction in the number of
excess flow check valves to be tested every 24 months to
approximately 20 percent of the valves each operating cycle.
Industry operating experience demonstrates a high level of
reliability for these excess flow check valves. A failure of an
excess flow check valve to isolate cannot initiate previously
evaluated accidents. Therefore, there is no increase in the
probability of occurrence of an accident as a result of this
proposed change. The postulated failure of an excess flow check
valve to isolate is bounded by the limiting analysis in the Updated
Final Safety Analysis Report (UFSAR). For a postulated break of an
instrument line upstream of an excess flow check valve, leakage from
the line rupture would be minimized by the line size or the flow-
restricting orifice in the line. The rate and quantity of process
fluid loss from an instrument line rupture is well within the
capability of the reactor coolant make-up systems. The proposed
change does not alter the design of the plants' instrument lines in
any manner, and the integrity and functional performance of the
secondary containment and Standby Gas Treatment system are not
affected by this proposed change. The potential offsite radiological
exposure associated with a postulated instrument line rupture
upstream of an excess flow check valve is bounded by the main steam
line break analysis and is substantially below the guidelines of 10
CFR 100. Therefore, the proposed license amendments do not involve a
significant increase in the consequences of an accident previously
evaluated.
    2. The proposed license amendments will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
    The proposed change allows a reduced number of excess flow check
valves to be tested each operating cycle. No other change in
requirements are being proposed. Industry operating experience
demonstrates the high reliability of the excess flow check valves.
The potential failure of an excess flow check valve to isolate is
bounded by the main steam line break analysis. The proposed license
amendments do not physically alter the plants and will not alter the
operation of the structures, systems, and components described in
the UFSAR. Therefore, a new or different kind of accident will not
be created.
    3. The proposed license amendments do not involve a significant
reduction in a margin of safety.
    Industry experience with excess flow check valves indicates that
they have very low failure rates. The postulated failure of an
excess flow check valve to isolate as a result of reduced testing is
bounded by the limiting analysis in the UFSAR, which is the main
steam line break analysis. For a postulated break of an instrument
line upstream of an excess flow check valve, leakage from the line
rupture would be minimized by the line size or the flow-restricting
orifice in the line. The rate and quantity of process fluid loss
from an instrument line rupture is well within the capability of the
reactor coolant make-up systems. The proposed change does not alter
the design of the plants' instrument line design in any manner, and
the integrity and functional performance of the secondary
containment and standby gas treatment system are not affected by
this proposed change. Therefore, the proposed license amendments 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.
    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. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois

    Date of amendment request: October 24, 2000.
    Description of amendment request: The proposed amendment would
revise the technical specifications to change the Westinghouse
references for Best Estimate Large Break Loss of Coolant Accident
(LOCA) analysis methodology. Reanalysis of large break LOCA transients,
utilizing the NRC approved Westinghouse Best Estimate LOCA model
WCOBRA/TRAC, was performed to demonstrate that 10 CFR 50.46 acceptance
criteria are satisfied at uprated power conditions.
    Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
    No physical plant changes are being made as a result of using
the Westinghouse Best Estimate Large Break LOCA analysis
methodology. The proposed TS changes simply involve updating the
references in TS 5.6.5.b, ``Core Operating Limits Report (COLR),''
to reference the Westinghouse Best Estimate Large Break LOCA
analysis methodology (i.e., Westinghouse topical report, WCAP-12945-
P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1,
``Code Qualification Document for Best Estimate LOCA Analysis,''
March 1998). The plant conditions assumed in the analysis are
bounded by the design conditions for all equipment in the plant;
therefore, there will be no increase in the probability of a LOCA.
The consequences of a LOCA are not being increased, since the
analysis has shown that the Emergency Core Cooling System (ECCS) is
designed such that its calculated cooling performance conforms to
the criteria contained in 10 CFR 50.46, ``Acceptance criteria for
emergency core cooling systems for light-water nuclear power
reactors.'' Furthermore, the re-performance of the Large Break LOCA
analysis has no effect on the performance of the ECCS equipment. No
other accident consequence is potentially affected by this change.
    All systems will continue to be operated in accordance with
current design requirements under the new analysis, therefore no new
components or system interactions have been identified that could
lead to an increase in the probability of any accident previously
evaluated in the Updated Final Safety Analysis Report (UFSAR). No
changes were required to the Reactor Protection System (RPS) or
Engineered Safety Features (ESF) setpoints because of the new
analysis methodology.
    Based on the analysis, it is concluded that the proposed TS
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or
different kind accident from any accident previously evaluated?
    There are no physical changes being made to the plant as a
result of using the Westinghouse Best Estimate Large Break LOCA
analysis methodology. No new modes of plant operation are being
introduced. The

[[Page 11053]]

configuration, operation and accident response of the Byron Station
and the Braidwood Station systems, structures or components are
unchanged by utilization of the new analysis methodology. Analyses
of transient events have confirmed that no transient event results
in a new sequence of events that could lead to a new accident
scenario. The parameters assumed in the analysis are within the
design limits of existing plant equipment.
    In addition, employing the Westinghouse Best Estimate Large
Break LOCA analysis methodology does not create any new failure
modes that could lead to a different kind of accident. The design of
all systems remains unchanged and no new equipment or systems have
been installed which could potentially introduce new failure modes
or accident sequences. No changes have been made to any RPS or ESF
actuation setpoints.
    Based on this review, it is concluded that no new accident
scenarios, failure mechanisms or limiting single failures are
introduced as a result of the proposed changes. Therefore, the
proposed TS changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a
margin of Safety?
    It has been shown that the analytic technique used in the
Westinghouse Best Estimate Large Break LOCA analysis methodology
realistically describes the expected behavior of the Byron Station
and Braidwood Station reactor system during a postulated LOCA.
Uncertainties have been accounted for as required by 10 CFR 50.46. A
sufficient number of LOCAs with different break sizes, different
locations, and other variations in properties have been considered
to provide assurance that the most severe postulated LOCAs have been
evaluated. The analysis has demonstrated that there is a high
probability that all acceptance criteria contained in 10 CFR 50.46,
paragraph b, continues to be satisfied. Based on this review, the
proposed TS changes 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
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. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois

    Date of amendment request: November 7, 2000.
    Description of amendment request: The proposed amendment would
revise the technical specifications to extend the TS Surveillance Test
Interval (STI) from a 92-day STI to an 18-month STI, for the Solid
State Protection System (SSPS) slave relay types that meet the
acceptance criteria for the reliability assessments performed in
accordance with the methodology described in the NRC approved
Westinghouse Electric Corporation Topical Reports.
    Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
    The proposed changes are consistent with the NRC approved
Westinghouse Electric Corporation Topical Reports, WCAP-13877,
``Reliability Assessment of Westinghouse Type AR Relays Used as SSPS
Slave Relays,'' WCAP-13878, ``Reliability Assessment of Potter &
Brumfield MDR Series Relays,'' and WCAP-13900, ``Extension of Slave
Relay Surveillance Testing Intervals,'' that analyze extending the
Solid State Protection System (SSPS) slave relay surveillance test
interval (STI) for the Westinghouse Type AR slave relays and for the
Potter & Brumfield MDR Series slave relays. The reliability
assessment of the slave relays was comprised of a failure modes and
effects analysis (FMEA) and an aging assessment of the slave relays.
WCAP-13877 and WCAP-13878 verified that the Westinghouse Type AR and
the Potter & Brumfield MDR Series slave relays are highly reliable
and that degradation of the slave relays is sufficiently slow (i.e.,
the time to failure due to degradation is sufficiently long) that an
18-month STI will adequately identify slave relay failures. A 92-day
STI is no more likely to detect significant changes in the SSPS
slave relays than an 18-month STI. The results demonstrate that
extending the SSPS slave relay STI from 92 days to 18 months does
not adversely affect the reliability of the SSPS slave relays
utilized in Engineered Safety Features Actuation System (ESFAS)
functions.
    The high reliability of these slave relays precludes the need
for more frequent periodic surveillance testing to verify
operability.
    As stated in WCAP-13877 and WCAP-13878, the overly conservative
92-day STI can be extended to an 18-month STI without impact or
consequence to slave relay reliability. In addition, the proposed
changes will not adversely affect the ability of the SSPS to perform
its safety function. The same ESFAS instrumentation is being used
and the ESFAS reliability is being maintained with the proposed
changes. Because the reliability of the slave relays used in the
ESFAS applications is so high, elimination of the routine
surveillance testing of the slave relays when the reactor is at
power will have a positive impact on ESFAS availability and plant
safety. The proposed changes will not modify any system interface
and will not increase the likelihood of any accident initiator
because such events are independent of the proposed changes.
Therefore, the probability of an accident previously evaluated is
not increased.
    The proposed changes will not modify, degrade, or prevent
actions or alter any assumptions previously made in evaluating the
radiological consequences of any accident described in the Updated
Final Safety Analysis Report (UFSAR). The ESFAS instrumentation
remains capable of performing its intended safety function of
mitigation of consequences of accidents or transients. Therefore,
the consequences of an accident previously evaluated are not
increased.
    Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
    2. Does the proposed change create the possibility of a new or
different kind accident from any accident previously evaluated?
    The proposed changes do not alter the performance of the ESFAS.
The proposed changes only extend the STI, and no changes to the
testing methodology or the way in which the slave relays are tested
are being proposed. No new equipment is being installed, and no
installed equipment is being operated in a new or different manner
with the proposed changes. Extending the STI will maintain the
reliability of the slave relays as demonstrated by the NRC approved
FMEA and aging assessment, and may improve the reliability of the
system by reducing potential test-induced degradation. As documented
in WCAP-13877 and WCAP-13878, an STI of 92 days is no more likely to
detect significant changes in the SSPS Type AR and MDR Series slave
relays than a STI of 18 months.
    Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
    3. Does the proposed change involve a significant reduction in a
margin of safety?
    The proposed changes do not affect the total ESFAS response
assumed in the safety analysis. The periodic slave relay functional
verification is relaxed because of the demonstrated high reliability
of the slave relays and their insensitivity to any short-term wear
or aging effects. The Westinghouse Owners Group (WOG) program to
extend the STI for the slave relays, as documented in the NRC
approved WCAP-13877 and WCAP-13878, has concluded that the slave
relays used in the SSPS are highly reliable and that the
surveillance testing at a frequency of 18

[[Page 11054]]

months, instead of the 92-day STI currently required, does not
significantly decrease any margin of safety assumed in the safety
analysis. Plant safety will be improved by limiting the amount of
on-line testing that will be performed, because on-line testing of
the slave relays results in the removal of a train of equipment from
service or manipulation of specific safety-related equipment which
is then no longer able to perform its safety function if called upon
until the surveillance test is completed. The proposed changes also
act to improve plant safety by reducing equipment degradation and
reducing unnecessary burden on the operating personnel. There are no
changes in testing methodology or performance criteria.
    Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.929c) 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 60609-0767.
    NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois

    Date of amendment request: November 13, 2000.
    Description of amendment request: The proposed amendment would
revise the technical specifications to delete the ``Power Range Neutron
Flux High Negative Rate,'' Trip Function from Reactor Trip System
Instrumentation. The proposed change allows elimination of this
unnecessary function and thereby reduces the potential for a transient.
The proposed changes are consistent with the Westinghouse Topical
report previously accepted by the NRC.
    Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of he issue of no significant hazards
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
    The removal of the Power Range Neutron Flux High Negative Rate
Trip (i.e., Negative Flux Rate Trip (NFRT)) Function does not
increase the probability or consequences of reactor core damage
accidents resulting from dropper Rod Cluster Control Assembly (RCCA)
events previously analyzed. The safety functions of other safety
related systems and components, which are related to mitigation of
these events, have not been altered. All other primary Reactor Trip
System (RTS) and Engineered Safety Features Actuation Systems
(ESFAS) protection functions are not impacted by the elimination of
the NFRT Function. The NFRT circuitry detects and responds to
negative reactivity insertion due to RCCA misoperation events should
they occur. Therefore, the NFRT Function is not assumed in the
initiation of such events. Because the NFRT Function is being
eliminated from the plant, it can no longer actuate and cause a
transient. The consequences of accidents previously evaluated in the
Updated Final Safety Analysis Report (UFSAR) are unaffected by the
proposed changes because no change to any equipment response or
accident mitigation scenario has resulted.
    Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
    2. Does the proposed change create the possibility of a new or
different kind accident from any accident previously evaluated?
    The deletion of the NFRT Function does not create the
possibility of a new or different kind of accident than any accident
previously evaluated in the UFSAR. No new accident scenarios,
failure mechanisms, or limiting single failures are introduced as a
result of the proposed changes. The proposed changes do not
challenge the performance or integrity of any safety related
systems. It has been demonstrated that the NFRT Function can be
eliminated by the NRC approved methodology described in Westinghouse
Topical Report WCAP-11394-P-A, ``Methodology for the Analysis of the
Dropped Rod Event,'' dated January 1990. The Braidwood Station and
the Byron Station cycle-specific analyses have confirmed that for a
dropped RCCA(s) event, no direct reactor trip or automatic power
reduction is required to meet the Departure From Nucleate Boiling
(DNB) limits for this Condition II, ``Faults of Moderate
Frequency,'' event. The NFRT Function is not credited either as a
primary or backup mitigation feature for any other UFSAR event.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
    3. Does the proposed change involve a significant reduction in a
margin of safety?
    The margin of safety associated with the licensing basis
acceptance criteria for any postulated accident is unchanged. It has
been demonstrated that the NFRT Function can be eliminated by the
NRC approved methodology described in WCAP 11394-P-A. The Braidwood
Station and the Byron Station cycle-specific analyses have confirmed
that for a dropped RCCA(s) event, DNB limits are not exceeded with
the proposed changes. Conformance to our licensing basis acceptance
criteria for Design Basis Accidents (DBAs) and transients with the
deletion of the NFRT Function is demonstrated, and DNB limits are
not exceeded. The proposed changes will have no adverse effect on
the availability, operability, or performance of the safety related
systems and components assumed to actuate in the event of a DBA or
transient. Therefore, the proposed changes do not involve a
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of amendment request: February 29, 2000.
    Description of amendment request: The proposed amendment would
reduce the number of safety valves required for overpressure protection
at Dresden, Unit 2, by removing from Technical Specifications (TS)
Section 3.6.E, the safety valve function of the Target Rock safety/
relief valve (SRV). The proposed amendment would move the safety valve
lift pressure setpoints from TS Section 3.6.E to TS Section 4.6.E,
remove the Target Rock SRV setpoints from TS, and change the number of
safety valves from nine to eight. The proposed amendment would also
remove footnote ``c'' of Unit 3, TS Section 4.6.E, since this footnote
was only applicable to Unit 3, Cycle 15 which has been completed.
    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?
    The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those

[[Page 11055]]

consequences. Limits have been established, consistent with Nuclear
Regulatory Commission (NRC) approved methods to ensure that fuel
performance during normal, transient, and accident conditions is
acceptable. The proposed change to reduce the number of required
safety valves from nine (9) to eight (8) does not affect the ability
of plant systems to adequately mitigate the consequences of an
accident previously evaluated.
    This conclusion was derived by evaluating all applicable
analyses including thermal limit, American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel (B&PV) pressurization
events, margin to unpiped safety valve, anticipated transient
analysis without scram events, Loss Of Coolant Accident (LOCA),
station blackout, and 10 CFR 50, Appendix R analyses. Therefore,
there is no increase in the probability or consequences of an
accident previously evaluated because the analyses supports
operation without crediting the Target Rock Safety Relief Valve
safety mode function.
    Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
    The requested change has been previously evaluated by evaluating
all applicable analyses including thermal limit, ASME B&PV
pressurization events, margin to unpiped safety valve, anticipated
transient analysis without scram events, station blackout, LOCA, and
10 CFR 50, Appendix R analyses. The proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated because the analyses support operation
without crediting the Target Rock safety relief valve safety
function. No new failure modes will be introduced upon
implementation of the proposed changes, therefore, the possibility
of a new and different accident has not been created.
    Does the change involve a significant reduction in a margin of
safety?
    Changing the required number of safety valves from nine (9) to
eight (8) will not involve any reduction in margin of safety. This
conclusion was derived by evaluating all existing analyses including
thermal limit, ASME B&PV pressurization events, margin to unpiped
safety valve, anticipated transient analysis without scram events,
station blackout, LOCA, and 10 CFR 50, Appendix R analyses. The
analyses previously evaluated remain valid, therefore, a significant
reduction in the margin of safety does not exist.
    Therefore, based upon the above evaluation, ComEd has concluded
that these changes do not constitute 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 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.

Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: February 14, 2000.
    Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to correct various editorial
errors and make other administrative changes. Specifically, the
proposed amendment would make administrative changes that revise: (a)
Tables 3.6-1 and 4.4-1 to correct listing and editorial errors, (b) TS
3.8.B.10 to reflect the wording in 10 CFR 50.54(m)(2)(iv), (c) Figures
3.10-2 through 3.10-6 to remove these figures, (d) Table 4.1-1 to
reflect change in level indication components, (e) TS 4.19.B and
6.14.1.1 to correct editorial errors, (f) TS 6.12.1 to reference the
current sections of 10 CFR Part 20, (g) TS 6.12.1 to reflect an
organizational title change, and (h) TS 6.13.2 to correct a
typographical error.
    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) Changes To Tables 3.6-1 And 4.4-1 To Correct Listing And
Editorial Errors
    (1) Does the proposed license amendment involve a significant
increase in the probability or in the consequences of an accident
previously evaluated?
    No. The proposed changes are administrative in nature. The
changes involve correcting errors in Table 3.6-1 and additions to
Tables 3.6-1 and 4.4-1 to reflect UFSAR [Updated Final Safety
Analysis Report] Table 5.2-1 and the IST [inservice testing]
Program. These changes do not affect possible initiating events for
accidents previously evaluated or alter the configuration or
operation of the facility. The Limiting Safety System Settings and
Safety Limits specified in the current Technical Specifications
remain unchanged. Therefore, the proposed changes would not involve
a significant increase in the probability or in the consequences of
an accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
    No. The proposed changes are administrative in nature. The
safety analysis of the facility remains complete and accurate. There
are no physical changes to the facility and the plant conditions for
which the design basis accidents have been evaluated are still
valid. The operating procedures and emergency procedures are
unaffected. Consequently no new failure modes are introduced as a
result of the proposed changes. Therefore, the proposed changes
would not create the possibility of a new or different kind of
accident from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction
in a margin of safety?
    No. The proposed changes are administrative in nature. Since
there are no changes to the operation of the facility or the
physical design, the Updated Final Safety Analysis Report (UFSAR)
design basis, accident assumptions, or Technical Specification Bases
are not affected. Therefore, the proposed changes do not involve a
significant reduction in a margin of safety.
    (b) Change To Section 3.8.B.10 To Reflect The Wording In 10 CFR
50.54(m)(2)(iv)
    (1) Does the proposed license amendment involve a significant
increase in the probability or in the consequences of an accident
previously evaluated?
    No. The proposed change is administrative in nature. The change
involves updating Section 3.8.B.10 to reflect 10 CFR
50.54(m)(2)(iv). This change does not affect possible initiating
events for accidents previously evaluated or alter the configuration
or operation of the facility. The Limiting Safety System Settings
and Safety Limits specified in the current Technical Specifications
remain unchanged. Therefore, the proposed change would not involve a
significant increase in the probability or in the consequences of an
accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
    No. The proposed change is administrative in nature. The safety
analysis of the facility remains complete and accurate. There are no
physical changes to the facility and the plant conditions for which
the design basis accidents have been evaluated are still valid. The
operating procedures and emergency procedures are unaffected.
Consequently no new failure modes are introduced as a result of the
proposed change. Therefore, the proposed change would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
    (3) Does the proposed amendment involve a significant reduction
in a margin of safety?
    No. The proposed change is administrative in nature. Since there
are no changes to the operation of the facility or the physical
design, the Updated Final Safety Analysis Report (UFSAR) design
basis, accident assumptions, or Technical Specification Bases are
not affected. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
    (c) Deletion Of Figures 3.10-2 Through 3.10-6
    (1) Does the proposed license amendment involve a significant
increase in the

[[Page 11056]]

probability or in the consequences of an accident previously
evaluated?
    No. The proposed change is administrative in nature. The change
involves the deletion of Figures 3.10-2, 3.10-3, 3.10-4, 3.10-5 and
3.10-6. This change does not affect possible initiating events for
accidents previously evaluated or alter the configuration or
operation of the facility. The Limiting Safety System Settings and
Safety Limits specified in the current Technical Specifications
remain unchanged. Therefore, the proposed change would not involve a
significant increase in the probability or in the consequences of an
accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
    No. The proposed change is administrative in nature. The safety
analysis of the facility remains complete and accurate. There are no
physical changes to the facility and the plant conditions for which
the design basis accidents have been evaluated are still valid. The
operating procedures and emergency procedures are unaffected.
Consequently no new failure modes are introduced as a result of the
proposed change. Therefore, the proposed change would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
    (3) Does the proposed amendment involve a significant reduction
in a margin of safety?
    No. The proposed change is administrative in nature. Since there
are no changes to the operation of the facility or the physical
design, the Updated Final Safety Analysis Report (UFSAR) design
basis, accident assumptions, or Technical Specification Bases are
not affected. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
    (d) Change To Table 4.1-1 To Reflect Change In Level Indication
Components
    (1) Does the proposed license amendment involve a significant
increase in the probability or in the consequences of an accident
previously evaluated?
    No. This change does not affect possible initiating events for
accidents previously evaluated or operation of the facility. While
the configuration of the facility has changed with installation of
the new level sensors, the safety-related function of theses sensors
remains unchanged (i.e., at a predetermined level of approximately
35% of instrument span, a low level alarm will annunciate in the CCR
[control room]). The Limiting Safety System Settings and Safety
Limits specified in the current Technical Specifications remain
unchanged. Therefore, the proposed change would not involve a
significant increase in the probability or in the consequences of an
accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
    No. The safety analysis of the facility remains complete and
accurate. The plant conditions for which the design basis accidents
have been evaluated are still valid. While the configuration of the
facility has changed with installation of the new level sensors, the
safety-related function of theses [sic] sensors remains unchanged
(i.e., at a predetermined level of approximately 35% of instrument
span, a low level alarm will annunciate in the CCR). Consequently no
new failure modes are introduced as a result of the proposed change.
Therefore, the proposed change would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
    (3) Does the proposed amendment involve a significant reduction
in a margin of safety?
    No. While the configuration of the facility has changed with
installation of the new level sensors, the safety-related function
of theses sensors remains unchanged (i.e., at a predetermined level
of approximately 35% of instrument span, a low level alarm will
annunciate in the CCR). Also, there are no changes to the operation
of the facility. Thus the Updated Final Safety Analysis Report
(UFSAR) design basis, accident assumptions, or Technical
Specification Bases are not affected. Therefore, the proposed change
does not involve a significant reduction in a margin of safety.
    (e) Change To Sections 4.19.B And 6.14.1.1 To Correct Editorial
Errors
    (1) Does the proposed license amendment involve a significant
increase in the probability or in the consequences of an accident
previously evaluated?
    No. The proposed changes are administrative in nature. The
change in Sections 4.19.B and 6.14.1.1 involve amending ``the
Semiannual Radioactive Effluent Release Report'' to ``the Annual
Radioactive Effluent Release Report.'' These changes do not affect
possible initiating events for accidents previously evaluated or
alter the configuration or operation of the facility. The Limiting
Safety System Settings and Safety Limits specified in the current
Technical Specifications remain unchanged. Therefore, the proposed
changes would not involve a significant increase in the probability
or in the consequences of an accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
    No. The proposed changes are administrative in nature. The
safety analysis of the facility remains complete and accurate. There
are no physical changes to the facility and the plant conditions for
which the design basis accidents have been evaluated are still
valid. The operating procedures and emergency procedures are
unaffected. Consequently no new failure modes are introduced as a
result of the proposed change. Therefore, the proposed changes would
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction
in a margin of safety?
    No. The proposed changes are administrative in nature. Since
there are no changes to the operation of the facility or the
physical design, the Updated Final Safety Analysis Report (UFSAR)
design basis, accident assumptions, or Technical Specification Bases
are not affected. Therefore, the proposed changes do not involve a
significant reduction in a margin of safety.
    (f) Change To Section 6.12.1 To Reference The Current Sections
Of 10 CFR [Part] 20
    (1) Does the proposed license amendment involve a significant
increase in the probability or in the consequences of an accident
previously evaluated?
    No. The proposed change is administrative in nature. The change
involves updating Section 6.12.1 to reference 10 CFR 20.1601(a) and
10 CFR 20.1601(b). This change does not affect possible initiating
events for accidents previously evaluated or alter the configuration
or operation of the facility. The Limiting Safety System Settings
and Safety Limits specified in the current Technical Specifications
remain unchanged. Therefore, the proposed change would not involve a
significant increase in the probability or in the consequences of an
accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
    No. The proposed change is administrative in nature. The safety
analysis of the facility remains complete and accurate. There are no
physical changes to the facility and the plant conditions for which
the design basis accidents have been evaluated are still valid. The
operating procedures and emergency procedures are unaffected.
Consequently no new failure modes are introduced as a result of the
proposed change. Therefore, the proposed change would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
    (3) Does the proposed amendment involve a significant reduction
in a margin of safety?
    No. The proposed change is administrative in nature. Since there
are no changes to the operation of the facility or the physical
design, the Updated Final Safety Analysis Report (UFSAR) design
basis, accident assumptions, or Technical Specification Bases are
not affected. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
    (g) Change To Section 6.12.1 To Reflect An Organizational Title
Change
    (1) Does the proposed license amendment involve a significant
increase in the probability or in the consequences of an accident
previously evaluated?
    No. The proposed change is administrative in nature. The change
involves updating Section 6.12.1 to use the title ``Shift Manager''
instead of ``Senior Watch Supervisor.'' This change does not affect
possible initiating events for accidents previously evaluated or
alter the configuration or operation of the facility. The Limiting
Safety System Settings and Safety Limits specified in the current
Technical Specifications remain unchanged. Therefore, the proposed
change would not involve a significant increase in the probability
or in the consequences of an accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new
or different kind of

[[Page 11057]]

accident from any accident previously evaluated?
    No. The proposed change is administrative in nature. The safety
analysis of the facility remains complete and accurate. There are no
physical changes to the facility and the plant conditions for which
the design basis accidents have been evaluated are still valid. The
operating procedures and emergency procedures are unaffected.
Consequently no new failure modes are introduced as a result of the
proposed change. Therefore, the proposed change would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
    (3) Does the proposed amendment involve a significant reduction
in a margin of safety?
    No. The proposed change is administrative in nature. Since there
are no changes to the operation of the facility or the physical
design, the Updated Final Safety Analysis Report (UFSAR) design
basis, accident assumptions, or Technical Specification Bases are
not affected. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
    (h) Change To Section 6.13.2 To Correct A Typographical Error
    (1) Does the proposed license amendment involve a significant
increase in the probability or in the consequences of an accident
previously evaluated?
    No. The proposed change is administrative in nature. The change
involves updating Section 6.13.2 from ``DOR [Division of Operating
Reactors] Guidelines of NUREG-0588'' to ``DOR Guidelines or NUREG-
0588.'' This change does not affect possible initiating events for
accidents previously evaluated or alter the configuration or
operation of the facility. The Limiting Safety System Settings and
Safety Limits specified in the current Technical Specifications
remain unchanged. Therefore, the proposed change would not involve a
significant increase in the probability or in the consequences of an
accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
    No. The proposed change is administrative in nature. The safety
analysis of the facility remains complete and accurate. There are no
physical changes to the facility and the plant conditions for which
the design basis accidents have been evaluated are still valid. The
operating procedures and emergency procedures are unaffected.
Consequently no new failure modes are introduced as a result of the
proposed change. Therefore, the proposed change would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
    (3) Does the proposed amendment involve a significant reduction
in a margin of safety?
    No. The proposed change is administrative in nature. Since there
are no changes to the operation of the facility or the physical
design, the Updated Final Safety Analysis Report (UFSAR) design
basis, accident assumptions, or Technical Specification Bases are
not affected. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
    NRC Section Chief: Marsha Gamberoni.

Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: December 11, 2000.
    Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to provide editorial
revisions, clarifications, and corrections. Specifically, the proposed
amendment would: (1) Provide updated information and corrections to the
TS cover page, table of contents, and list of figures, (2) revise TS
4.5.E, ``Control Room Air Filtration System,'' to remove an incorrect
system test description and provide consistent test values for system
flow rate and filter efficiency, (3) revise TS 6.2.1.a, ``Facility
Management and Technical Support,'' to reference the Quality Assurance
Program Description as the location of the documentation rather than
the Updated Final Safety Analysis Report, (4) revise TS 6.9.1.7,
``Monthly Operating Report,'' to change the recipient of the Monthly
Operating Report, and (5) correct the periodicity of the Radioactive
Effluent Release Report from annual to semiannual in TS 6.15, ``Offsite
Dose Calculation Manual'' and TS 6.16, ``Major Changes to Radioactive
Liquid, Gaseous and Solid Waste Systems.'' In addition, the proposed
change revises TS Figure 5.1-1B concerning the indicated vent location
associated with Indian Point Unit 3 (IP3). The labels for the plant
vent and the machine shop are reversed.
    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?
    The proposed changes consist of editorial changes,
administrative changes, clarifications, and corrections to existing
TSs. These changes do not involve a change to the design or
operation of any plant system nor are any of the safety analyses
affected as a result of these changes. Accordingly, the initiators
of any accident as well as any system relied upon for the mitigation
of an accident are not affected by the proposed changes. Therefore,
there is no increase in the probability or in the consequences of an
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
    The proposed changes do not involve a change to the design or
operation of any plant system. These changes include editorial
changes, administrative changes, clarifications, and corrections of
existing TSs and, therefore, do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
    3. Does the proposed amendment involve a significant reduction
in a margin of safety?
    The proposed changes consist of editorial changes,
administrative changes, and clarifications to existing TSs and do
not involve changes to any 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: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
    NRC Section Chief: Marsha Gamberoni.

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

    Date of amendment request: January 8, 2001.
    Description of amendment request: The proposed change revises the
lower limit of the allowable containment internal pressure in Technical
Specification (TS) 3.6.1.4, ``Containment Systems--Internal Pressure,''
from 14.375 pounds per square inch, absolute (psia) to 14.275 psia.
    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. Will the operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
    Response: The proposed change revises the lower limit of the
allowable containment internal pressure in TS 3.6.1.4 from 14.375 to
14.275 psia. This change will allow

[[Page 11058]]

additional operating margin for the containment atmosphere purge
(CAP) system during conditions of low atmospheric pressure. The
containment minimum pressure parameter is not an accident initiator
and does not affect the probability of any initiating event
scenario. Although the TSs will allow a lower initial containment
internal pressure, the current analyses for the associated design
events are not affected since the lower pressure has already been
conservatively included. The proposed change in initial containment
internal pressure is bounded in the current design. Therefore, this
proposed change does not involve an increase in the probability or
consequences of an accident previously evaluated.
    2. Will the operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
    Response: The proposed change affects the TS allowed lower limit
on containment internal pressure and consequently the atmospheric
range in which the CAP system can be operated. The change in the
lower limit on containment internal pressure is encompassed by
current design analyses and does not result in a change of analyzed
conditions or analyzed operating ranges.
    Based on the proposed TS change, CAP system operation will be
allowed at a lower atmospheric pressure. This change does not change
the function of the system or its method of operation. Although the
initial atmospheric pressure at which the CAP system can be
initiated is being lowered, this is within the current design of the
CAP system and does not change the differential pressures at which
it will be operated.
    Therefore, the proposed change[d] does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
    3. Will the operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
    Response: The proposed change makes use of the initial
containment pressure assumption values used in the current analyses
to provide additional operating margin for the CAP system. The
margin of safety that was inherent in the results of these safety
analyses has been preserved. The associated analyses ensure the
negative pressure differential associated with an inadvertent
actuation of the containment spray system is acceptable, and ensure
that the emergency core cooling system can satisfy its design safety
function under worst case conditions. The calculated maximum
differential pressure is 0.49 psid [pounds per square inch
differential] which is within the design limit of 0.65 psid. The
peak clad temperature for the worst case large break loss of coolant
accident is 2177 deg.F which is within the acceptance criteria given
in 10CFR50.46. Since the proposed change does not affect the initial
containment pressure utilized in these analyses, the results of the
analyses are unchanged. Therefore, there is no affect on any margin
of safety associated with this parameter.
    Based on the above No Significant Hazards Consideration
Determination, it is concluded that: (1) The proposed change does
not constitute a significant hazards consideration as defined by
10CFR50.92; (2) there is a reasonable assurance that the health and
safety of the public will not be endangered by the proposed change;
and (3) this action will not result in a condition which
significantly alters the impact of the station on the environment as
described in the NRC final environmental statement.

    The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
    Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: September 1, 2000.
    Description of amendment request: The proposed amendments would
revise the technical specifications to add a new requirement for the
Main Steam Line Radiation Monitor 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 of occurrence or consequences of an accident previously
evaluated?
    The addition of the MSLRM [Main Steam Line Radiation Monitor]
automatic trip signal to the MVP [mechanical vacuum pump] has no
adverse effect on safety. The addition of Surveillance Requirements
(SRs) and the Limiting Condition for Operation (LCO) to our TS
enhances current safety features of the plant by establishing
controls for a required, and currently functional, safety feature.
The automatic trip function of the MVP does not serve as an
initiator for any accidents evaluated in Chapter 15, ``Accident and
Transient Analysis,'' of the Updated Final Safety Analysis Report.
Therefore, this change will not result in an increase of either the
probability or consequences of an accident.
    Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
    These proposed changes involve the addition of the MVP trip
input from the Main Steam Line Tunnel High Radiation signal. The
addition of this function does not represent a change in operating
parameters or equipment configuration for DNPS [Dresden Nuclear
Power Station], Units 2 and 3. Operation of DNPS, Units 2 and 3,
under the proposed changes 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?
    These proposed changes create a TS LCO and identify SRs for the
MVP trip input from the MSLRM signal. Operation under the proposed
change will not change any plant operation parameters, nor any
protective system setpoints. The calculations of off site dose
demonstrate that with the MVP trip instrumentation operating
properly, the doses that result from a CRDA [control rod drop
accident] with the MVP operating 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.

Exelon Generating Company, LLC (Exelon), Docket No. 50-353, Limerick
Generating Station, Unit 2, Montgomery County, Pennsylvania

    Date of amendment request: November 20, 2000.
    Description of amendment request: PECO Energy Company (PECO)
proposed changes to the Technical Specifications (TSs) that would
revise the heatup, cooldown, and inservice test Pressure-Temperature
(P-T) limitations (TS Figure 3.4.6.1-1) of the Limerick Generating
Station (LGS), Unit 2, Reactor Pressure Vessel (RPV) to a maximum of 32
Effective Full Power Years (EFPY). In addition, the licensee proposed
text changes to both Limiting Condition for Operation 3.4.6.1 and
Surveillance Requirement 4.4.6.1.1 to delete the reference to the A'
curve on TS Figure 3.4.6.1-1 since this curve will not be included in
the proposed Figure 3.4.6.1-1. The licensee also proposed adding an
intermediate hydrotest curve (Curve A22) to TS Figure
3.4.6.1-1, which is valid to 22 EFPY. By letter dated January 30, 2001,
Exelon stated that it has assumed responsibility, as of the date of the
transfer, for the active items on the Limerick Units 1 and 2

[[Page 11059]]

dockets previously submitted by PECO, including the subject amendment
request.
    Moreover, Exelon is revising its TS Bases Section B 3/4.4.6 to
update several RPV material chemistry parameters. The licensee
identified the need for these revisions during a Certified Material
Test Report data search performed by General Electric Company during
development of the proposed P-T curves.
    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 analysis is
presented below:

    1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
    There are no physical changes to the plant being introduced by
the proposed changes to the P-T curves. The proposed changes do not
modify the reactor coolant pressure boundary, i.e., there are no
changes in operating pressure, materials or seismic loading. The
proposed changes do not adversely affect the integrity of the
reactor coolant pressure boundary such that its function in the
control of radiological consequences is affected. The proposed P-T
curves were generated in accordance with the fracture toughness
requirements of 10 CFR 50, Appendix G, and American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,
Section XI, Appendix G, in conjunction with ASME Code Case N-640.
The proposed P-T curves were established in compliance with the
methodology used to calculate the predicted irradiation effects on
vessel beltline materials. Usage of these procedures provides
compliance with the intent of 10 CFR 50, Appendix G, and provides
margins of safety that ensure that failure of the reactor vessel
will not occur. The proposed P-T curves prohibit operational
conditions in which brittle fracture of reactor vessel materials is
possible. Consequently, the primary coolant pressure boundary
integrity will be maintained. Therefore, the proposed changes do not
involve a significant increase in the probability or 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 to the P-T curves were generated in
accordance with the fracture toughness requirements of 10 CFR 50,
Appendix G, and ASME B&PV Code, Section XI, Appendix G, in
conjunction with ASME Code Case N-640. Compliance with the proposed
P-T curves will ensure that conditions in which brittle fracture of
primary coolant pressure boundary materials are possible will be
avoided. No new modes of operation are introduced by the proposed
changes. The proposed changes will not create any new failure mode
from previously evaluated accidents. Further, the proposed changes
to the P-T curves do not affect any activities or equipment, and are
not assumed in any safety analysis to initiate nor mitigate any
accident sequence. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
    3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
    The proposed changes reflect an update of the P-T curves to
extend the reactor pressure vessel operating limit to 32 Effective
Full Power Years (EFPY). The revised curves are based on the latest
ASME guidance. The revised P-T limits, which provide more
operational flexibility than the current limits, were established in
accordance with current regulations and the latest ASME Code
information. No plant safety limits, set points, or design
parameters are adversely affected by the proposed TS changes. These
proposed changes maintain the relative margin of safety commensurate
with that which existed at the time that the ASME B&PV Code, Section
XI, Appendix G, was approved in 1974.
    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 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, Exelon Generating Company, 2301 Market Street,
Philadelphia, PA 19101
    NRC Section Chief: James W. Clifford.

Exelon Generation Company, LLC (Exelon), Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania

    Date of amendment request: January 18, 2001.
    Description of amendment request: Exelon requested a Technical
Specification (TS) change which will revise Surveillance Requirement
(SR) 4.9.2.d.1 to clarify that ``shorting links'' do not need to be
removed, if adequate shutdown margin has been demonstrated, when moving
a control rod during Operational Condition 5.
    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 change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
    This TS Change Request revises SR 4.9.2.d.1 to clarify that
``shorting links'' do not need to be removed if adequate shutdown
margin has been demonstrated when a control rod is withdrawn during
Operational Condition 5. This revision ensures that the words and
intent of the SR 4.9.2.d.1 match the words and intent of Limiting
Condition for Operation (LCO) 3.9.2.d, and will improve the
readability of the SR for plant operators. This change to SR
4.9.2.d.1 will clarify that ``shorting links'' can remain installed
if adequate shutdown margin has been demonstrated any time a control
rod is withdrawn in Operational Condition 5. This revision does not
impact any accident or transient events. There are no new initiators
created by this revision. Additionally, this revision will not
impact any existing analyses or requirements contained in the
Updated Final Safety Analysis Report. No changes in the operation of
the plant during power operation or refueling will occur as a result
of this revision. Therefore, the proposed TS revision does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
    2. The proposed TS revision does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
    The proposed TS revision will not impact any physical changes to
plant structures, systems, or components. The design, function, and
reliability of the Reactor Protection System will not be impacted by
this revision. This revision does not adversely impact any equipment
which is required for the prevention or mitigation of accidents or
transients. This revision ensures that the words and intent of the
SR 4.9.2.d.1 match the words and intent of LCO 3.9.2.d, and will
improve the readability of the SR for plant operators. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
    3. The proposed TS change does not involve a significant
reduction in a margin of safety.
    This proposed revision to SR 4.9.2.d.1 does not affect any
safety limits or analytical limits. There are also no changes to
accident or transient core thermal hydraulic conditions, minimum
combustible concentration limits, or fuel or reactor coolant
boundary design limits, as a result of this proposed change. This
revision ensures that the words and intent of the SR 4.9.2.d.1 match
the words and intent of LCO 3.9.2.d. Therefore, the proposed change
does not involve a significant reduction in a 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

[[Page 11060]]

Counsel, Exelon Generating Company, 2301 Market Street, Philadelphia,
PA 19101.
    NRC Section Chief: James W. Clifford.

Exelon Generation Company, LLC (Exelon), Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
    Date of amendment request: January 18, 2001.
    Description of amendment request: Exelon requested a Technical
Specification (TS) change which will revise the Units 1 and 2 TS Table
1.2, ``Operational Conditions,'' to allow placing the reactor mode
switch to the REFUEL position during Operational Conditions 3 and 4 to
accommodate post maintenance and surveillance testing on control rod
drives.
    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 change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
    The proposed revision allows a single control rod to be
withdrawn under control of the reactor mode switch REFUEL position
and the one-rod-out interlock in Operational Conditions 3 and 4.
This change does not affect any existing accident initiators. There
is no change to the coupling integrity of the control rod during
this accident. Although this change would allow an increase in the
frequency of single control rod withdrawals in Operational
Conditions 3 and 4, the probability of the previously analyzed
accidents is not affected.
    The onsite and offsite radiological consequences of previously
analyzed accidents in Operational Conditions 3 and 4 are not
affected by this proposed change. This change does not affect any
existing accident mitigators. The shutdown margin combined with the
refueling interlocks prevent a rod withdrawal error while in
refueling thereby preventing inadvertent criticality. There is no
impact on the ability of the Reactor Protection System (RPS)
circuitry to mitigate a Control Rod Drop Accident as described in
the Safety Analysis Report, nor is there an increase in the number
of fuel failures from this accident. As a result, the probability
and consequences of previously analyzed accidents are not
significantly increased.
    2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
    There are no new accident initiators created by the proposed
revision to Table 1.2. Single control rods can be withdrawn in
Operational Conditions 3 and 4 under the existing Technical
Specifications to permit control rod recoupling. The proposed
revision would expand this provision to other control rod
maintenance and testing activities performed in Operational
Conditions 3 and 4. The withdrawal of individual control rods in
Operational Conditions 3 and 4 is a mode of operation permitted
under limited circumstances by the existing TSs. The additional
control rod maintenance and testing activities which could be
performed in Operational Conditions 3 and 4, are already permitted
by the existing TSs in Operational Conditions 1, 2, 4, and 5.
    Based on the above, this change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
    3. The proposed TS change does not involve a significant
reduction in a margin of safety.
    The TSs currently permit single control rod withdrawal for the
purpose of control rod recoupling when in Operational Conditions 3
or 4 if the one-rod-out interlock is Operable. This change allows
additional activities for which a single control rod may be
withdrawn when in Operational Conditions 3 or 4, with the same
restriction that the one-rod-out interlock be Operable.
    The operability requirements for the one-rod-out interlock and
the shutdown margin requirements of TS 3.1.1 ensure the reactor will
be maintained subcritical during single control rod withdrawals.
Therefore, this change will not involve a significant reduction in a
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, Exelon Generation Company, 2301 Market Street,
Philadelphia, PA 19101.
    NRC Section Chief: James W. Clifford.

Exelon Generating Company, LLC (Exelon), Docket No. 50-353, Limerick
Generating Station, Unit 2, Montgomery County, Pennsylvania.

    Date of amendment request: February 1, 2001.
    Description of amendment request: Exelon proposed changes that
would revise Technical Specification (TS) 2.1 to incorporate revised
Safety Limit Minimum Critical Power Ratios due to the cycle-specific
analysis performed by Global Nuclear Fuel for Limerick Generating
Station, Unit 2, Cycle 7, which will include the use of the GE-14 fuel
product line.
    Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

    1. The proposed TS change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
    The derivation of the cycle specific Safety Limit Minimum
Critical Power Ratios (SLMCPRs) for incorporation into the Technical
Specifications (TS), and its use to determine cycle specific thermal
limits, has been performed using the methodology discussed in
``General Electric Standard Application for Reactor Fuel,'' NEDE-
24011-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-24011-P-A-14-US,
June, 2000, which incorporates Amendment 25. Amendment 25 was
approved by the NRC [Nuclear Regulatory Commission] in a March 11,
1999 safety evaluation report.
    The basis of the SLMCPR calculation is to ensure that greater
than 99.9% of all fuel rods in the core avoid transition boiling if
the limit is not violated. The new SLMCPRs preserve the existing
margin to transition boiling. The GE-14 fuel is in compliance with
Amendment 22 to ``General Electric Standard Application for Reactor
Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and U.S. Supplement, NEDE-
24011-P-A-14-US, June, 2000, which provides the fuel licensing
acceptance criteria. The probability of fuel damage will not be
increased as a result of this change. Therefore, the proposed TS
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
    The SLMCPR is a TS numerical value, calculated to ensure that
transition boiling does not occur in 99.9% of all fuel rods in the
core if the limit is not violated. The new SLMCPRs are calculated
using NRC approved methodology discussed in ``General Electric
Standard Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-
II), and U.S. Supplement, NEDE-24011-P-A-14-US, June, 2000, which
incorporates Amendment 25. Additionally, the GE-14 fuel is in
compliance with Amendment 22 to ``General Electric Standard
Application for Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and
U.S. Supplement, NEDE-24011-P-A-14-US, June 2000, which provides the
fuel licensing acceptance criteria. The SLMCPR is not an accident
initiator, and its revision will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
    3. The proposed TS change does not involve a significant
reduction in a margin of safety.
    There is no significant reduction in the margin of safety
previously approved by the NRC as a result of the proposed change to
the SLMCPRs, which includes the use of GE-14 fuel. The new SLMCPRs
are calculated using

[[Page 11061]]

methodology discussed in ``General Electric Standard Application for
Reactor Fuel,'' NEDE-24011-P-A-14 (GESTAR-II), and U.S. Supplement,
NEDE-24011-P-A-14-US, June, 2000, which incorporates Amendment 25.
The SLMCPRs ensure that greater than 99.9% of all fuel rods in the
core will avoid transition boiling if the limit is not violated when
all uncertainties are considered, thereby preserving the fuel
cladding integrity. Therefore, the proposed TS change will not
involve a significant reduction in the margin of safety previously
approved by the NRC.

    Based on the staff's review of the licensee's evaluation, 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, Exelon Generating Company, 2301 Market Street,
Philadelphia, PA 19101.
    NRC Section Chief: James W. Clifford.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: January 13, 2000.
    Description of amendment request: The proposed amendment would
change the Kewaunee Nuclear Power Plant Technical Specification 3.6,
``Containment.'' The proposed amendment would add limiting condition
for operation and allowed outage times for containment penetrations and
associated isolation devices to provide clear guidance. Also, the
proposed amendment would provide additional information, clarification,
and uniformity to the basis of the associated technical specification.
    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 Technical Specification [TS] change provides definition for
the [Allowable Outage Time] AOT for a containment isolation valve
and containment air lock. The original design and design basis of
the plant is still maintained and the probability and consequences
of previously evaluated accidents is unchanged. In our current
Technical Specifications the allowed outage time for a safeguards
480-volt bus is 24 hours. The basis for this outage time states:
    ``The intent of this TS is to provide assurance that at least
one external source and one standby source of electrical power is
always available to accomplish safe shutdown and containment
isolation and to operate required engineered safety features
equipment following an accident.''
    With one 480-volt safeguards bus out of service an associated
motor operated containment isolation valve is also out of service.
Since the 24-hour AOT is part of Kewaunee's original design basis,
allowing the containment isolation valves to be out of service for
24 hours does not increase the probability or consequences of an
accident previously evaluated.
    A risk assessment of the probability of a -loss-of-coolant-
accident with a train of containment isolation failing during a 4-
hour verse a 24-hour time span was conducted. The probability of
[loss-of-coolant accident] LOCA coincident with the failure of
containment isolation occurring during a 4-hour period versus a 24-
hour period is shown on Figure 1[ in licensee's submittal]. This
change in probability is considered insignificant.
    The proposed TS changes do not involve any physical or
operational changes to structures, systems or components. The
current safety analysis and design basis for the accident mitigation
functions of the containment, the airlocks, and the containment
isolation valves are maintained. On-site and off-site dose
consequences remain unaffected.
    2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
    The function of the containment vessel is to contain the
radiologically hazardous material following a LOCA. By maintaining
at least one containment isolation barrier intact the vessel can
perform its function. This amendment still ensures that at least one
barrier is intact or the leakage is evaluated not to exceed that
which is already evaluated and allowed by current technical
specification.
    The accidents considered are found in the Safety Analysis,
Section 14 of the [Updated Safety Analysis Report] USAR. The
proposed change does not involve a change to the plant design
(structures, systems or components) or operation. No new failure
mechanisms beyond those already considered in the current plant
Safety Analysis are introduced. No new accident is introduced and no
safety-related equipment or safety functions are altered. The
proposed change does not affect any of the parameters or conditions
that contribute to initiation of any accidents.
    3. Involve a significant reduction in a margin of safety.
    With one containment barrier intact during plant operation the
isolation of containment is still ensured. The plant's original
design basis addressed the inability of one of the two containment
isolation valves to operate for a 24-hour period. As this AOT has
been previously considered, there therefore is no 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: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: January 18, 2001.
    Description of amendment request: The proposed amendment would
change the Kewaunee Nuclear Power Plant Technical Specification 3.10.m
for reactor coolant minimum flow from the current value of 85,500
gallons per minute (gpm) to 93,000 gpm due to the replacement of steam
generators scheduled for the fall 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) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
    The change in Reactor Coolant Minimum Flow value for TS 3.10.m
proposed in this amendment request is needed to reflect operating
characteristics of the new [Replacement Steam Generators] RSGs.
Accident analyses affected by the RSGs have each been evaluated to
establish that there is no significant change in the documented
results (Attachment 3). These evaluations have shown that the
proposed value for Reactor Coolant Minimum Flow is bounded by the
Thermal Design Flow value used in the analyses and provides greater
margin to safety analysis acceptance criteria (e.g., [Departure from
Nucleate Boiling] DNB). All safety analysis acceptance criteria are
satisfied. Since Reactor Coolant flow values for the RSG conform to
the design bases and are bounded by the existing safety analyses,
changing the technical specification within limits of the bounding
accident analyses will not cause an increase in the probability or
consequences of an accident previously evaluated.
    (2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
    The proposed change is fully consistent with current plant
design bases and does not adversely affect any fission product
barrier, nor does it alter the safety function of safety related
systems, structures, and components depended upon for accident
prevention or mitigation. Thus, it does not create the possibility
of a new or different kind of accident.
    (3) Involve a significant reduction in the margin of safety.
    The proposed change does not alter the manner in which Safety
Limits, Limiting

[[Page 11062]]

Safety System Setpoints, or Limiting Conditions for Operation are
determined. It returns TS 3.10.m for Reactor Coolant Minimum Flow to
a value slightly higher, thus more conservative, than the value
specified for the [Original Steam Generators] OSG when new. It
conforms to plant design bases, is consistent with current safety
analyses, and limits actual plant operation. Analysis of the effect
of the proposed Reactor Coolant Minimum Flow limitation on [Loss-of-
Coolant-Accident] LOCA and non-LOCA transients determined that all
safety analysis acceptance criteria are satisfied at a [Thermal
Design Flow] TDF that bounds the revised Reactor Coolant Minimum
Flow and all [Kewaunee Nuclear Power Plant] KNPP safety requirements
continue to be met. 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
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: February 5, 2001.
    Description of amendment request: The proposed amendment would
change the Kewaunee Nuclear Power Plant Technical Specification 3.1.d.2
to reduce the maximum allowable leakage of primary system reactor
coolant to the secondary system from 500 gallons per day (gpd) through
any one steam generator to 150 gpd through any one steam generator. In
addition, the proposed amendment would remove reference to voltage
based repair criteria.
    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.
    The change in Leakage of Reactor Coolant value proposed by this
request for [Technical Specification] TS 3.1.d.2 complies with
[Nuclear Energy Institute] NEI 97-06, ``Steam Generator Program
Guidelines.'' [Nuclear Management Company, LLC] NMC evaluated
accident analyses affected by [steam generator] SG tube leakage and
determined that this change continues to be bounded by the existing
licensing and design basis. Design basis accidents and transients,
including steam generator tube rupture (SGTR), were analyzed using
Westinghouse Model 54F steam generator assumptions as part of steam
generator replacement. These evaluations show that the proposed 150
gpd [gallons per day] value for Leakage of Reactor Coolant is
bounded by the larger value used in applicable existing design basis
accident and transient analyses. The 150 gpd leak rate provides
increased margin to acceptance criteria found in these analyses. All
acceptance criteria are satisfied and SG primary to secondary
leakage values for the [replacement steam generator] RSG conform to
the existing design bases and are bounded by the existing safety
analyses. Changing the technical specification within limits of the
bounding accident analyses cannot change the probability or
consequence of an accident previously evaluated. Removal of an
allowance for voltage-based alternate repair criteria defaults to a
more conservative repair criteria. Thus, nothing in this proposal
will cause an increase in the probability or consequence of an
accident previously evaluated.
    (2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
    The 150 gpd value proposed for maximum allowable Leakage of
Reactor Coolant is consistent with current plant design bases and
does not adversely affect any fission product barrier, nor does it
alter the safety function of safety significant systems, structures
and components or their roles in accident prevention or mitigation.
The proposed value for maximum allowable leakage through any one
steam generator is bounded by currently licensed design basis
accident and transient analyses of record. Removal of a reference in
the TS to voltage-based repair criteria leaves in its place a more
conservative, more restrictive criteria for repair or plugging of
steam generator tubes. Thus, this proposal does not create the
possibility of a new or different kind of accident.
    (3) Involve a significant reduction in the margin of safety.
    The proposed change does not alter the manner in which Safety
Limits, Limiting Safety System Setpoints, or Limiting Conditions for
Operation are determined. It sets TS 3.1.d.2 for Leakage of Reactor
Coolant to a lower, thus more conservative, value than that
previously specified and approved for use by the NRC [Nuclear
Regulatory Commission]. It conforms to plant design bases, is
consistent with current safety analyses, and limits actual plant
operation within analyzed and licensed boundaries. Analyses of
applicable transients were performed using a primary to secondary
leakage rate greater than the rate proposed by this request. All
safety analysis acceptance criteria are satisfied at this value and
all [Kewaunee Nuclear Power Plant] KNPP safety requirements continue
to be met. The 150 gpd leak rate proposed by this amendment request
is bounded by these analyses. Removal of reference to use of
voltage-based repair criteria from TS 3.1.d.2 and its basis leaves
an existing and more conservative repair criteria in place. Thus,
changes proposed by this 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.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: Claudia M. Craig.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California

    Date of amendment requests: January 11, 2001.
    Description of amendment requests: The proposed amendment deletes
requirements from the Technical Specifications (and, as applicable,
other elements of the licensing bases) to maintain a Post Accident
Sampling System (PASS). Licensees were generally required to implement
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI, Unit 2. Requirements related to
PASS were imposed by Order for many facilities and were added to or
included in the technical specifications (TS) for nuclear power
reactors currently licensed to operate. Lessons learned and
improvements implemented over the last 20 years have shown that the
information obtained from PASS can be readily obtained through other
means or is of little use in the assessment and mitigation of accident
conditions.
    The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271) on possible
amendments to eliminate PASS, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 31, 2000 (65 FR

[[Page 11063]]

65018). The licensee affirmed the applicability of the following NSHC
determination in its application dated January 11, 2001.
    Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
    The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
    The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radionuclides
within the containment building.
    Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
    The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
    Therefore, this change does not involve a significant reduction
in the margin of safety.
    Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
    NRC Section Chief: Stephen Dembek.

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

    Date of application for amendment: January 18, 2001.
    Brief description of amendment request: The amendment request
identifies an unreviewed safety question related to the planned
replacement of the engineered safety features (ESF) transformers with
new transformers having active automatic load tap changers (LTCs).
Markups to the Final Safety Analysis Report (FSAR) were included in the
application.
    Basis for proposed no significant hazards determination: As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:

    1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
    Based on the review of the modification details there is an
insignificant increase in the probability of a malfunction of
equipment important to safety, however there is no increase in the
probability of an accident previously evaluated. The modification
has no effect on the radiological consequences of accidents
previously evaluated. Installation of the LTCs does not impact
accident initiators though a failure mode has been identified that
can increase the probability of malfunction, a risk study shows this
risk is insignificant.
    2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
    The overall effect of the malfunction of the LTC controllers
would lead to a loss of the associated ESF bus which is not a new
failure mode that can lead to a new or different kind of accident
than previously evaluated. The LTC failure effects are limited to
the associated ESF train, therefore this type of failure meets the
definition of a single failure as defined in 10 CFR 50 Appendix A
for operation under normal (Non T/S [technical specification]
action) conditions. Additionally, during the 10 CFR 50.59 evaluation
for the LTCs criteria (a)(2)(ii) with respect to accidents of a
different type was not met.
    3. Does the proposed change involve a significant reduction in
margin of safety?
    The installation of the replacement transformers with load tap
changers will help assure the required minimum NB bus voltage
established by Reference 7.10 [design calculations] under a wider
variation of grid voltage.
    Current Technical Specification Bases for the offsite power
distribution system are covered in sections B3.8.1-AC Sources-
Operating, B3.8.9-Distribution Systems-Operating, B3.8.2-AC Sources-
Shutdown, and B3.8.10-Distribution Systems-Shutdown. These bases
ensure that sufficient power will be available to supply the safety-
related equipment required for: (1) The safe shutdown of the
facility; and (2) The mitigation and control of accident conditions
within the facility. The minimum specified independent and redundant
AC power and distribution systems satisfy the requirements of
General Design Criterion 17 of Appendix A to 10 CFR Part 50. The
ACTIONS sections of the applicable Technical Specifications provide
requirements specified for various levels of degradation of the
power sources and provide restrictions upon continued facility
operation commensurate with the

[[Page 11064]]

level of degradation. The Operability of the power sources are
consistent with the initial condition assumptions of the safety
analyses and are based upon maintaining at least one redundant set
of onsite AC power sources and associated distribution systems
operable during accident conditions coincident with an assumed loss
of offsite power and single failure of the other onsite AC source.
    The installation of the transformers with automatic load tap
changers reduces the possibility of the loss of the offsite power
system due to the increased grid voltage variations as documented in
the description of the change in section 4.1.4. Therefore, the
installation of the transformers with load tap changers will not
reduce 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: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.

Virginia Electric and Power Company, Docket No. 50-338, North Anna
Power Station, Unit No. 1, Louisa County, Virginia

    Date of amendment request: January 9, 2001.
    Description of amendment request: The proposed administrative
changes will remove obsolete license conditions from the Facility
Operating License (FOL) and implement associated changes to the
Technical Specifications (TS). These changes involve editorial
revisions, relocation of license conditions, removal of redundant
license conditions covered throughout the license, removal of expired
license conditions, and removal of license conditions and TS associated
with completed modifications.
    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:

    Criterion 1--The proposed license amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
    The proposed change to the North Anna Unit 1 Facility Operating
License, NPF-4, is administrative (and in part editorial) in nature.
The removal of license conditions regarding completed, no longer
needed, and expired requirements has no impact on plant operations
since these requirements no longer have meaningful applications. The
renumbering and/or relocation within the FOL of various license
conditions in this proposed administrative change does not alter the
technical basis, requirements or the implementation of the affected
items. The proposed change is within the current design and
licensing bases of the facility. Since this change is administrative
only and neither station operations nor design are affected by the
change, it does not involve any significant increase in the
probability or the consequences of any accident or malfunction of
equipment important to safety previously evaluated.
    Criterion 2--The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
    The proposed change is administrative (and in part editorial) in
nature. The license conditions that are being removed or relocated
by this proposed change do not impact station operations or station
equipment in any manner. The proposed change does not involve a
physical alteration of the plant, nor a change in the methods used
to respond to plant transients that has not been previously
analyzed. No new or different equipment is being installed and no
installed equipment is being removed or operated in a different
manner. Consequently, no new failure modes are introduced and the
proposed administrative change to the North Anna Unit 1 Facility
Operating License does not create the possibility of a new or
different kind of accident or malfunction of equipment important to
safety from any previously evaluated.
    Criterion 3--The proposed license amendment does not involve a
significant reduction in a margin of safety.
    The proposed change is administrative (and in part editorial) in
nature and neither station operations nor design are affected by the
change. Since station operations are not affected by the proposed
administrative change and no physical change is being made to the
station, the change does not impact the condition, design, or
performance of any station structure, system or component.
Therefore, the proposed administrative change to the North Anna Unit
1 Facility Operating License does not involve a significant
reduction in any margin of safety described in the bases of the
Technical Specifications.

    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: Maitri Banerjee, Acting.

Virginia Electric and Power Company, Docket No. 50-339, North Anna
Power Station, Unit No. 2, Louisa County, Virginia

    Date of amendment request: January 9, 2001.
    Description of amendment request: The proposed administrative
changes will remove obsolete license conditions from the Facility
Operating License (FOL). These changes involve editorial revisions,
relocation of license conditions, removal of expired license
conditions, and removal of license conditions associated with completed
modifications.
    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:

    Criterion 1--The proposed license amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
    The proposed change to the North Anna Unit 2 Facility Operating
License, NPF-7, is administrative (and in part editorial) in nature.
The removal of license conditions regarding completed, no longer
needed, and expired requirements has no impact on plant operations
since these requirements no longer have meaningful applications. The
renumbering and/or relocation within the FOL of various license
conditions in this proposed administrative change does not alter the
technical basis, requirements or the implementation of the affected
items. The proposed change is within the current design and
licensing bases of the facility. Since this change is administrative
only and neither station operations nor design are affected by the
change, it does not involve any significant increase in the
probability or the consequences of any accident or malfunction of
equipment important to safety previously evaluated.
    Criterion 2--The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
    The proposed change is administrative (and in part editorial) in
nature. The license conditions that are being removed or relocated
by this proposed change do not impact station operations or station
equipment in any manner. The proposed change does not involve a
physical alteration of the plant, nor a change in the methods used
to respond to plant transients that has not been previously
analyzed. No new or different equipment is being installed and no
installed equipment is being removed or operated in a different
manner. Consequently, no new failure modes are introduced and the
proposed administrative change to the North Anna Unit 2 Facility
Operating License does not create the possibility of a new or
different kind of accident or malfunction of equipment

[[Page 11065]]

important to safety from any previously evaluated.
    Criterion 3--The proposed license amendment does not involve a
significant reduction in a margin of safety.
    The proposed change is administrative (and in part editorial) in
nature and neither station operations nor design are affected by the
change. Since station operations are not affected by the proposed
administrative change and no physical change is being made to the
station, the change does not impact the condition, design, or
performance of any station structure, system or component.
Therefore, the proposed administrative change to the North Anna Unit
2 Facility Operating License does not involve a significant
reduction in any margin of safety described in the bases of the
Technical Specifications.

    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: Maitri Banerjee, Acting.

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.

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

    Date of amendment request: January 8, 2001.
    Brief description of amendment request: The proposed amendment
would revise the Technical Specifications (TS) to change the acceptance
values for Core Spray subsystem flow contained in TS 4.5.1.b.1 from the
current value of 6350 gallons per minute (gpm) to 6150 gpm.
    Date of publication of individual notice in Federal Register:
January 22, 2001 (66 FR 6701).
    Expiration date of individual notice: February 21, 2001.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
    For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Publicly available records will be accessible and electronically from
the ADAMS Public Library component on the NRC Web site, http://
www.nrc.gov (the Electronic Reading Room).

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

    Date of application for amendments: June 16, 2000.
    Brief description of amendments: The amendments revise TS Table
3.3.10-1, ``Post Accident Monitoring Instrumentation,'' to add the high
pressure safety injection (HPSI) cold leg flow and HPSI hot leg flow
instrumentation to the table.
    Date of issuance: February 8, 2001.
    Effective date: February 8, 2001, and shall be implemented within
30 days of the date of issuance.
    Amendment Nos.: Unit 1--131 , Unit 2--131, Unit 3--131.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 4, 2000 (65 FR
59220)
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 08, 2001.
    No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland

    Date of application for amendments: September 14, 2000.
    Brief description of amendments: The amendments incorporate changes
described below into the Technical Specifications for Culvert Cliffs
Units 1 and 2. On September 9, 1996, a final rule amending 10 CFR
50.55a was issued requiring owners to implement, by September 9, 2001,
the requirements of the 1992 Edition through the 1992 Addenda of the
American Society of Mechanical Engineers Boiler and Pressure Vessel
Code Section XI, Subsections IWE and IWL, as modified and supplemented
by 10 CFR 50.55a. Calvert Cliffs Nuclear Power Plant, Inc. has
developed a program to effect the implementation of Subsections IWE and
IWL.
    Date of issuance: January 30, 2001.
    Effective date: As of the date of issuance to be implemented within
30 days.
    Amendment Nos.: 240 and 214.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR
62384).

[[Page 11066]]

    The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated January 30, 2001.
    No significant hazards consideration comments received: No.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan

    Date of application for amendment: November 19, 1999, as
supplemented May 31, August 2, October 19, and November 21, 2000.
    Brief description of amendment: The amendment revises the Technical
Specifications (TSs) by changing (1) the design features description of
the fuel storage equipment and configuration to allow an increase in
the spent fuel pool (SFP) storage capacity and (2) the description of
the high-density spent fuel racks program to clarify that the
surveillance program is applicable only to racks containing Boraflex as
a neutron absorber. Specifically, the amendment revises the TSs for
Fermi 2 to increase the capacity of the SFP from 2,414 to 4,608 fuel
assemblies.
    Date of issuance: January 25, 2001.
    Effective date: As of the date of issuance and shall be implemented
within 90 days.
    Amendment No.: 141.
    Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications
    Date of initial in Federal Register March 13, 2000 (65 FR 13336)
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated a January 25, 2001.
    No significant hazards consideration comments received: No.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan

    Date of application for amendment: September 20, 2000.
    Brief description of amendment: The amendment changes Technical
Specification (TS) 5.5.7.d to decrease the maximum allowed pressure
drops across control room emergency filtration (CREF) make-up and
recirculation train filters and charcoal absorbers. The words ``(CREF
only)'' are also removed from the TS.
    Date of issuance: February 8, 2001.
    Effective date: As of the date of issuance and shall be implemented
within 60 days
    Amendment No.: 142.
    Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
    Date of initial notice in Federal Register: November 1, 2000 (65 FR

65340).
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 8, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 30, 2000.
    Brief description of amendment: The amendment relocated the
boration systems requirements from the Technical Specifications to the
Technical Requirements Manual.
    Date of issuance: January 31, 2001.
    Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
    Amendment No.: 229.
    Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65
FR 81916).
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 31, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 19, 1999, as
supplemented October 12, 2000.
    Brief description of amendment: The amendment changes the Technical
Specification surveillance testing requirements of the charcoal
adsorbers in the Standby Gas Treatment System and the Control Room
Emergency Ventilation Air Supply System to meet the requested actions
of Generic Letter 99-02.
    Date of issuance: February 5, 2001.
    Effective date: As of the date of issuance to be implemented within
30 days.
    Amendment No.: 269.
    Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: February 9, 2000 (65 FR
6410).
    The October 12, 2000, supplemental letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 5, 2001.
    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: May 22, 2000, as supplemented
October 4, 2000.
    Brief description of amendments: Changed the Technical
Specifications to incorporate that portion of the August 8, 1996, Final
Amended Rule (61 FR 41303) related to revised requirement of inservice
inspection of the containment post-tensioning system.
    Date of issuance: January 31, 2001.
    Effective date: January 31, 2001.
    Amendment Nos. 210 and 204.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR
48750). The October 4, 2000 letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
    The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 31, 2001.
    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: December 6, 2000.
    Brief description of amendments: The amendments delete Technical
Specifications (TS) Section 6.8.4.d, ``Post Accident Sampling,'' for
Turkey Point Units 3 and 4 and thereby eliminate the requirements to
have and maintain the post-accident sampling system (PASS) for those
units.
    Date of issuance: January 31, 2001.
    Effective date: January 31, 2001.
    Amendment Nos. 211 and 205.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments
revised the TSs.
    Date of initial notice in Federal Register: December 27, 2000 (65
FR 81923).
    The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 31, 2001.
    No significant hazards consideration comments received: No.

[[Page 11067]]

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit 2, Berrien County, Michigan

    Date of application for amendment: September 30, 2000, as
supplemented November 22, and December 20, 2000.
    Brief description of amendment: The amendment would allow an
extension of the steam generator tube inspection surveillance
requirements of Technical Specification Surveillance Requirement
4.4.5.3. Specifically, the licensee requested to extend the required
inspection interval from 40 to 56 calendar months.
    Date of issuance: January 30, 2001.
    Effective date: As of the date of issuance and shall be implemented
within 30 days.
    Amendment No.: 232.
    Facility Operating License No. DPR-74: Amendment revises the
Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 ( 65
FR
62387).
    The supplemental information contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 30, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 1, 2000.
    Brief description of amendment: The amendment approves changes to
Technical Specifications (TSs) 3.3.3.2, ``Instrumentation, Movable
Incore Detectors''; 3.3.3.3, ``Instrumentation, Seismic
Instrumentation''; 3.3.3.4, ``Instrumentation, Meteorological
Instrumentation''; 3.3.3.8, ``Loose-Part Detection System''; 3.3.4,
``Turbine Overspeed Protection''; and Index Pages vi and vii. The
changes relocate the requirements for the incore detectors, seismic
instrumentation, meteorological instrumentation, loose-part detection
system, and turbine overspeed protection system from the TSs to the
Technical Requirements Manual. The Bases for these TSs have been
modified to reflect the TS changes.
    Date of issuance: January 29, 2001.
    Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
    Amendment No.: 193.
    Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: November 29, 2000 (65
FR 71136).
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 29, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 31, 2000 as supplemented
January 5, 2001.
    Brief description of amendment: The amendment changes Technical
Specifications (TSs) 3.8.1.1, ``Electrical Power Systems--A.C.
Sources--Operating,'' and 3.8.1.2, ``Electrical Power Systems--A.C.
Sources--Shutdown.'' The changes allow certain EDG surveillance
requirements to be performed when the plant is operating instead of
shut down as currently required. The index and Bases for these TSs are
modified to reflect the changes.
    Date of issuance: February 2, 2001.
    Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
    Amendment No.: 194.
    Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: September 6, 2000 (65
FR
54087).
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 2, 2001.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: November 18, 1999, as
supplemented August 7, 2000.
    Brief description of amendment: The amendment to the Kewaunee
Nuclear Power Plant Technical Specifications approves an increase in
the allowable number of spent fuel assemblies in the spent fuel pools.
The addition of the 215 storage locations in the new north canal pool
will extend the full-core reserve capability until after the 2009
outage, and increase the total capacity to 1,205 spent fuel assemblies.
    Date of issuance: January 23, 2001.
    Effective date: As of the date of issuance and shall be implemented
within 30 days.
    Amendment No.: 150.
    Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: November 1 and December
21, 2000 (65 FR 65347 and 65 FR 80471 respectively)
    The supplemental information contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated January 23, 2001.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California

    Date of application for amendments: May 12, 2000, as supplemented
by letter dated January 25, 2001.
    Brief description of amendments: These amendments authorize (1) a
design upgrade of the refueling water purification (RWP) system to
permit reclassification of this system from Design Class II/non-Seismic
Category 1 to Design Class I/Seismic Category 1 to allow filtering of
the refueling water storage tank (RWST) water while the RWST is
required to be operable, and (2) the use of a temporary reverse osmosis
skid mounted system to reduce RWST silica concentration levels while
the RWST is required to be operable following upgrade of the RWP system
to satisfy reactor coolant chemistry limits.
    Date of issuance: January 29, 2001.
    Effective date: January 29, 2001, and shall be implemented in the
next periodic update to the FSAR Update, following upgrade of the
refueling water purification system, in accordance with 10 CFR
50.71(e).
    Amendment Nos.: Unit 1--144 ; Unit 2-143.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the FSAR Update.
    Date of initial notice in Federal Register: July 12, 2000 (65 FR
43050).
    The January 25, 2001, supplemental letter provided additional
clarifying information, did not expand the scope of the application as
originally noticed,

[[Page 11068]]

and did not change the staff's original proposed no significant hazards
consideration determination.
    The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 29, 2001.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California

    Date of application for amendments: September 6, 2000 (PCN-274,
Supplement 1).
    Brief description of amendments: The amendments revised the San
Onofre, Units 2 and 3 Technical Specification 3.3.11, ``Post Accident
Monitoring Instrumentation (PAMI),'' to extend the PAMI surveillance
frequency from 18 to 24 months to accommodate a 24-month fuel cycle.
    Date of issuance: January 30, 2001.
    Effective date: January 30, 2001, to be implemented within 30 days
of issuance.
    Amendment Nos.: Unit 2--176; Unit 3-167.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 18, 2000 (65 FR
62391).
    The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 30, 2001.
    No significant hazards consideration comments received: No.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California

    Date of application for amendments: October 6, 2000 (PCN-518).
    Brief description of amendments: The amendments revise TS 3.7.11,
``Control Room Emergency Air Cleanup System (CREACUS),'' to establish
actions to be taken for inoperable ventilation systems due to a
degraded control room pressure boundary. The amendments allow up to 24
hours to restore the pressure boundary to operable status when two
ventilation trains are inoperable due to an inoperable pressure
boundary in Modes 1, 2, 3, and 4. In addition, a limiting condition for
operation note is added to allow the pressure boundary to be opened
intermittently under administrative control without affecting CREACUS
operability.
    Date of issuance: January 30, 2001.
    Effective date: January 30, 2001, to be implemented within 30 days
of issuance.
    Amendment Nos.: Unit 2--177; Unit 3--168.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 15, 2000 (65
FR 69066).
    The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 30, 2001.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina

    Date of application for amendment: November 10, 2000.
    Brief description of amendment: This amendment will allow: (a) the
minimum fuel oil stored in the fuel oil storage tank (FOST) for each
emergency diesel generator (EDG) to be raised from 47,100 gallons to
48,500 gallons for Modes 1-4, and from 33,200 gallons to 42,500 gallons
for Modes 5 and 6; and (b) the minimum fuel oil maintained in the day
fuel tank for each EDG to be raised from 300 gallons to 360 gallons for
Modes 1-6.
    Date of issuance: February 2, 2001.
    Effective date: February 2, 2001.
    Amendment No.: 150.
    Facility Operating License No. NPF-12: Amendment revises the
Technical Specifications.
    Date of initial notice in Federal Register: November 20, 2000 (65
FR 69795).
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 2, 2001.
    No significant hazards consideration comments received: No.

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 amendments request: October 9, 2000, supplemented December
4, 2000.
    Brief Description of amendments: The amendments revise Technical
Specification 5.5.14, ``Technical Specification (TS) Bases Control
Program,'' to provide consistency with the changes in 10 CFR 50.59
which were published in the Federal Register on October 4, 2000.
    Date of issuance: January 31, 2001.
    Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
    Amendment Nos.: 148 and 140.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 2000 (65
FR 79907) The supplement dated December 4, 2000, provided clarifying
information that did not change the scope of the October 4, 2000,
application nor the initial proposed no significant hazards
consideration determination.
    The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 31, 2001.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia

    Date of application for amendments: June 14, 2000.
    Brief description of amendments: The amendments revised the
Technical Specifications (TSs) Surveillance Requirements (SR) 3.8.1.9
and 3.8.1.14 to reduce diesel generators loading requirements from
 6800 kW and  7000 kW to  6500 kW and
 7000 kW. These changes will make the above SRs consistent
with SRs 3.8.1.3 and 3.8.1.13, which are in the current TSs. In
addition, the proposed changes would correct a typographical error in
Section 5.6.7, ``EDG Failure Report,'' in the Vogtle TS. This editorial
change will correctly reference Regulatory Position C.4 of Regulatory
Guide 1.9, Revision 3 instead of Regulatory Position C.5.
    Date of issuance: January 31, 2001.
    Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--117; Unit 2--95.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 6, 2000 (65
FR 54087).
    The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 31, 2001.
    No significant hazards consideration comments received: No.

[[Page 11069]]

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2 and 3, Limestone County, Alabama

    Date of application for amendments: August 11, 2000 (TS-400) as
supplemented by letter dated October 20, 2000.
    Brief description of amendments: The amendments revised the
surveillance test requirements for excess flow check valves.
    Date of issuance: January 29, 2001.
    Effective date: January 29, 2001.
    Amendment Nos.: 268 and 228.
    Facility Operating License Nos. DPR-52 and DPR-68: Amendments
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 6, 2000 (65
FR 54088). The October 20, 2000, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 29, 2001.
    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: December 7, 2000 (ET 00-0041).
    Brief description of amendment: The amendment changes Table 3.3.2-
1, ``Engineered Safety Feature Actuation System Instrumentation,'' of
the Technical Specifications. The change adds Surveillance Requirement
(SR) 3.3.2.10 for the following two engineered safety feature actuation
system instrumentation in the table: item 6.f, loss of offsite power,
and item 6.h, auxiliary feedwater pump suction transfer on suction
pressure--low.
    Date of issuance: February 06, 2001.
    Effective date: February 06, 2001, and shall be implemented
including the changes to the Bases for the response times, within 60
days of the date of issuance.
    Amendment No.: 136.
    Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65
FR 81932).
    The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 06, 2001.
    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 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, located at One White Flint North, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852, 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 March 23, 2001, 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

[[Page 11070]]

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, located at One White Flint North, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852, 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 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 20555-001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, located at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852, 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 20555-001, 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 Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
    Date of amendment request: December 19, 2000.
    Description of amendment request: The amendment revises the
Technical Specifications to indicate that quadrant power tilt limits
apply only when reactor power is greater than 50 percent.
    Date of issuance: December 20, 2000.
    Effective Date: As of its date of issuance and shall be implemented
within 30 days.
    Amendment No.: 204.
    Facility Operating License No. DPR-64: Amendment revises the
Technical Specifications.
    Public comments requested as to proposed no significant hazards
consideration: No.
    The Commission's related evaluation of the amendment, finding of
emergency circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated
December 20, 2000.
    Attorney for licensee: Mr. John M. Fulton, Assistant General
Counsel Entergy Nuclear Generating Co. Pilgrim Station, 600 Rocky Hill
Road Plymouth, MA 02360.
    NRC Section Chief: Marsha Gamberoni.

    Dated at Rockville, Maryland, this 14th day of February 2001.
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 01-4228 Filed 2-20-01; 8:45 am]
BILLING CODE 7590-01-U 

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